Damage cross-section library DAMSIG84 (in a 640 group structure of the SAND-II type)
The damage cross-section library DAMSIG84 is an updated and extended version of the damage cross-section library DAMSIG81. The library contains energy dependent group cross-section data for a number of materials to facilitate the calculations of damage production (based on displacements of atoms), the calculations of probable zones and the calculation of gas production due to (n,α) and (n,p) reactions. The group cross-section data are given for a fine group structure of the SAND-II type with 640 groups. This library contains for some materials more than one cross-section set originating from different evaluations. Cross-section data sets for the activation reactions 54Fe(n,p), 58Ni(n,p), 59Co(n,γ), and 63Cu(n,α), which reactions are commonly used to determine thermal and fast neutron fluences, have been included also. Moreover also some artificial cross-sections are incorporated in this library which can be used to calculate values for some physical quantities characterizing neutron spectra, such as mean lethargy , and mean energy . Also cross-sections for B, Al and Cd are included; these are required to reach compatibility with other libraries in the SAND-II format
Cross-section library DOSCROS81 (in a 640 group structure of the SAND-II type)
The cross section library DOSCROS81 is an updated and extended version of the dosimetry cross section library DOSCROS77. The library contains energy dependent fine group cross section values for a number of reactions which are applied in neutron metrology and in neutron activation spectrometry. The library contains data from the ENDF/B-V file supplemented with information from the ENDF/B-IV and from the INDL/V. The total number of reaction cross section sets incorporated in this library is 70 (+ 3 cover cross section sets). A documentation of the library is presented. The library is written in the SAND-II format. The numerical data are available on microfiche upon request to ECN. The library will be available in a computer compatible form from the OECD NEA Data Bank and from the RSIC at Oak Ridge
1 - Description of problem or function: - ZZ-IRDF-82: ENDF-5 Format; 620 group (SAND II) Dosimetry Library. Nuclides: Li, B, F, Na, Mg, Al, P, S, Sc, Ti, Mn, Fe, Co, Ni, Cu, Zn, Zr, Nb, Rh, In, I, Au, Th, U, Np, Pu, Am. - ZZ-IRDF-90: ENDF-6 Format; 640 groups extended SAND II structure. Nuclides: Li, B, F, Mg, Al, P, S, Sc, Ti, Cr, Mn, Fe, Co, Ni, Cu, Zn, Zr, Nb, Rh, Cd, Ir, Gd, Au, Th, U, Np, Pu, V. Damage cross section for Fe, Cr, Ni. Weighting spectrum: Maxwell spectrum, 1/E spectrum and Watt fission spectrum. - ZZ-IRDF-2002: ENDF-6 Format (pointwise cross-section data). SAND II 640 energy group structure (multigroup data). Nuclides: Li, B, F, Na, Mg, Al, P, S, Sc, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zn, As, Y, Zr, Nb, Rh, Ag, In, I, La, Pr, Tm, Ta, W, Au, Hg, Pb, Th, U, Np, Pu, Am, Cd, Gd. Damage cross section for Fe, Cr, Ni, Si, GaAs displacement. Weighting spectrum: - Typical MTR spectrum used in the input of the cross-section uncertainty processing code. - Flat weighting spectrum used in converting the pointwise cross-section data to the extended SAND-II group structure. - ZZ-IRDF-2002-ACE: ACE Format (continuous energy cross-section data for Monte Carlo). Nuclides: Li, B, F, Na, Mg, Al, P, S, Sc, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zn, As, Y, Zr, Nb, Rh, Ag, In, I, La, Pr, Tm, Ta, W, Au, Hg, Pb, Th, U, Np, Pu, Am, Cd, Gd. Damage cross section for Fe, Cr, Ni, Si, GaAs displacement. - (A) ZZ-IRDF-82: The 1982 version of the International Reactor Dosimetry File is composed of two different parts. The first part is made up of a collection of dosimetry Cross sections and the second part contains a collection of benchmark spectra. For ease of use in dosimetry applications both Cross sections and spectra are distributed in multigroup form. Each of these two parts is in the ENDF/B-V Format as a separate computer file. I) The dosimetry cross section library contains the following data: (1) The entire ENDF/B-V Dosimetry Library (Mod. 1) in the form of 620 group averaged Cross
Damage cross section library (DAMSIG77)
The damage cross sections of various materials are converted to a data format, which can be used as library for the program SAND-II. The materials available in this library are graphite, stainless steel, aluminium, silicium, chromium, iron, nickel, copper, zirconium, molybdenum, tungsten, vanadium and niobium. A number of these materials have more than one cross section set, originating from different evaluations. Cross sections for some activation reactions, commonly used to determine thermal and fast neutron fluences have been included too. Moreover, also some artificial cross sections are introduced in this library which can be used to derive values for some physical quantities which may characterize neutron spectra
DAMSIG81. ECN Radiation Damage Cross Section Library. Contents and documentation
DAMSIG81, the Radiation Damage Cross-Section Library by ECN, Netherlands, includes neutron cross-sections for about 20 reactor structural materials for calculating radiation damage by atomic displacements and by gas production, together with some additional related data. The data are presented in a 640 group structure similar to SAND-II. The library can be obtained free of charge from the IAEA Nuclear Data Section. The present version (DAMSIG81A) has a minor correction compared to the original version. (author)
Multigroup cross section library; WIMS library
The WIMS library has been extensively used in thermal reactor calculations. This multigroup constants library was originally developed from the UKNDL in the late 60's and has been updated in 1986. This library has been distributed with the WIMS-D code by NEA data bank. The references to WIMS library in literature are the 'old' which is the original as developed by the AEA Winfrith and the 'new' which is the current 1986 WIMS library. IAEA has organised a CRP where a new and fully updated WIMS library will soon be available. This paper gives an overview of the definitions of the group constants that go into any basic nuclear data library used for reactor calculations. This paper also outlines the contents of the WIMS library and some of its shortcomings
Testing of cross section libraries for TRIGA criticality benchmark
Influence of various up-to-date cross section libraries on the multiplication factor of TRIGA benchmark as well as the influence of fuel composition on the multiplication factor of the system composed of various types of TRIGA fuel elements was investigated. It was observed that keff calculated by using the ENDF/B VII cross section library is systematically higher than using the ENDF/B-VI cross section library. The main contributions (∼220 pcm) are from 235U and Zr. (author)
Experience With the SCALE Criticality Safety Cross Section Libraries
Bowman, S.M.
2000-08-21
This report provides detailed information on the SCALE criticality safety cross-section libraries. Areas covered include the origins of the libraries, the data on which they are based, how they were generated, past experience and validations, and performance comparisons with measured critical experiments and numerical benchmarks. The performance of the SCALE criticality safety cross-section libraries on various types of fissile systems are examined in detail. Most of the performance areas are demonstrated by examining the performance of the libraries vs critical experiments to show general trends and weaknesses. In areas where directly applicable critical experiments do not exist, performance is examined based on the general knowledge of the strengths and weaknesses of the cross sections. In this case, the experience in the use of the cross sections and comparisons with the results of other libraries on the same systems are relied on for establishing acceptability of application of a particular SCALE library to a particular fissile system. This report should aid in establishing when a SCALE cross-section library would be expected to perform acceptably and where there are known or suspected deficiencies that would cause the calculations to be less reliable. To determine the acceptability of a library for a particular application, the calculational bias of the library should be established by directly applicable critical experiments.
Graphs of all neutron cross sections and photon production cross sections on the Alternate Monte Carlo Cross Section (AMCCS) library have been plotted along with local neutron heating numbers. The values of ν-bar, the average number of neutrons per fission, are also plotted for appropriate isotopes
Cross section library based discrepancies in MCNP criticality calculations
In nuclear engineering several reactor physics problems can be approached using Monte Carlo neutron transport techniques, which usually give reliable results when properly used. The quality of the results is largely determined by the accuracy of the geometry model and the statistical uncertainty of the Monte Carlo calculation. There is, however, another potential source of error, namely the cross section data used with the Monte Carlo codes. It has been shown in several studies that there may be significant discrepancies between results calculated using cross section libraries based on different evaluated nuclear data files. These discrepancies are well known to the evaluators of nuclear data but less acknowledged by reactor physicists, who often rely on a single cross section library in their calculations. In this study, discrepancies originating from base nuclear data were investigated in a systematic manner using the MCNP4C code. Calculations on simplified UOX and MOX fuelled LWR lattices were carried out using cross section libraries based on ENDF/B-VI.8, JEFF-3.0, JENDL-3.3, JEF-2.2 and JENDL-3.2 evaluated data files. The neutron spectrum of the system was varied over a wide range by changing the ratio of hydrogen to heavy metal atoms. The essential isotopes underlying the discrepancies were identified and the roles of fission and absorption cross sections of the most important nuclides assessed. The results confirm that there are large systematic differences up to a few per cent in the multiplication factors of LWR lattices. The discrepancies are strongly dependent on material compositions and neutron spectra, and largely originate from U-238 and the primary fissile isotopes. It is concluded that these discrepancies should be taken into account in all reactor physics calculations, and that reactor physicists should not rely on results based on a single cross section library. (author)
Accurate Development of Thermal Neutron Scattering Cross Section Libraries
Hawari, Ayman; Dunn, Michael
2014-06-10
The objective of this project is to develop a holistic (fundamental and accurate) approach for generating thermal neutron scattering cross section libraries for a collection of important enutron moderators and reflectors. The primary components of this approach are the physcial accuracy and completeness of the generated data libraries. Consequently, for the first time, thermal neutron scattering cross section data libraries will be generated that are based on accurate theoretical models, that are carefully benchmarked against experimental and computational data, and that contain complete covariance information that can be used in propagating the data uncertainties through the various components of the nuclear design and execution process. To achieve this objective, computational and experimental investigations will be performed on a carefully selected subset of materials that play a key role in all stages of the nuclear fuel cycle.
Nuclear cross section library for oil well logging analysis
As part of the IRTMBA (Improved Radiation Transport Modelling for Borehole Applications) Project of the EU Community's 5th Programme a special purpose multigroup cross section library to be used in the deterministic (as well as Monte Carlo) oil well logging particle transport calculations was prepared. This library is expected to improve the prediction of the neutron and gamma spectra at the detector positions of the logging tool, and their use for the interpretation of the neutron logging measurements was studied. Preparation and testing of this library is described. (author)
Testing of cross section libraries on zirconium benchmarks
Highlights: ► Calculations with ENDF/B-VII.0 nuclear data overpredict keff of Zr benchmarks. ► TRIGA criticality benchmark sensitive to Zr data. ► Zr scattering cross section responsible for differences in keff. ► Need for new experimental data on Zr cross sections. - Abstract: In this paper we investigate the influence of various up-to-date nuclear data libraries, such as ENDF/B-VI.6, ENDF/B-VII.0 and JEFF 3.1, on the multiplication factor of the TRIGA benchmark with fuel made of enriched uranium and zirconium hydride and SB light-water reactor benchmarks with fuel made of fissile material in zirconium matrix. The calculations are performed with the Monte Carlo computer code MCNP. Differences of ∼600 pcm in keff are observed for the benchmark model of the TRIGA reactor, while there are practically no differences in the kinf of the fuel. Therefore, an investigation is performed also for hypothetical homogeneous and heterogeneous systems with different leakage. The uncertainty analysis shows that the most important contributors to the difference in keff are the Zr isotopes (especially 90Zr and 91Zr) and thermal scattering data for H and Zr in ZrH. As the differences in keff due to the use of different cross section libraries are relatively large, there is certainly a need for a review of the evaluated cross section data of the zirconium isotopes.
Benchmark calculations of 150-group cross section library for LMR's
For the purpose of diversification of selection of cross section library for neutron calculation of LMR, the 150 multi-group cross section library was generated from ENDF-VI release. The set was then examined by analyzing measured reactivity quantities such as control rod worth, Doppler effect and sodium void effect for BFS critical assemblies that we obtained through the critical experiment plan for developing the KALIMER core design. The calculated results based on 9 group structure using the new set were also compared with those of JEF set based on the same group structure and compared with those of the same set based on 25 group structure to find the proper group structure. ENDF-VI-based set shows a small deviation in predicting measured integral quantities in comparison with the previous set and a small group effect
Status of standard cross section library and future plan
JSSTDL-300 multi-group cross section library with 300 neutron energy groups coupled with 104 group γ-ray cross sections was developed for general users in nuclear reactor physics and/or design, whose source data is the evaluated nuclear data library JENDL-3.2. For the purpose of a standard or common use, several famous cross section libraries worldwide used, i.e., ABBN-25, GAM-123, VITAMIN-C/J(E+C), MGCL-137, BERMUDA-12 and FNS-125 for neutron, and LANL-12, -24-, -48, and CSEWG-94 for γ-ray, are consulted about setting the common energy group structure. Furthermore, in order to expand the applicability, the top energy is set on 20 MeV and the lowest energy is 10-5 eV. In the thermal neutron energy region, the JSSTDL-300 has about 20 energy groups. Besides, many utility codes for group collapsing and for data format transformation are provided for general users. (author)
Status of standard cross section library and future plan
Zukeran, Atsushi [Hitachi Ltd., Power and Industrial System R and D Laboratory, Hitachi, Ibaraki (Japan)
2001-08-01
JSSTDL-300 multi-group cross section library with 300 neutron energy groups coupled with 104 group {gamma}-ray cross sections was developed for general users in nuclear reactor physics and/or design, whose source data is the evaluated nuclear data library JENDL-3.2. For the purpose of a standard or common use, several famous cross section libraries worldwide used, i.e., ABBN-25, GAM-123, VITAMIN-C/J(E+C), MGCL-137, BERMUDA-12 and FNS-125 for neutron, and LANL-12, -24-, -48, and CSEWG-94 for {gamma}-ray, are consulted about setting the common energy group structure. Furthermore, in order to expand the applicability, the top energy is set on 20 MeV and the lowest energy is 10{sup -5} eV. In the thermal neutron energy region, the JSSTDL-300 has about 20 energy groups. Besides, many utility codes for group collapsing and for data format transformation are provided for general users. (author)
MOX Cross-Section Libraries for ORIGEN-ARP
Gauld, I.C.
2003-07-01
The use of mixed-oxide (MOX) fuel in commercial nuclear power reactors operated in Europe has expanded rapidly over the past decade. The predicted characteristics of MOX fuel such as the nuclide inventories, thermal power from decay heat, and radiation sources are required for design and safety evaluations, and can provide valuable information for non-destructive safeguards verification activities. This report describes the development of computational methods and cross-section libraries suitable for the analysis of irradiated MOX fuel with the widely-used and recognized ORIGEN-ARP isotope generation and depletion code of the SCALE (Standardized Computer Analyses for Licensing Evaluation) code system. The MOX libraries are designed to be used with the Automatic Rapid Processing (ARP) module of SCALE that interpolates appropriate values of the cross sections from a database of parameterized cross-section libraries to create a problem-dependent library for the burnup analysis. The methods in ORIGEN-ARP, originally designed for uranium-based fuels only, have been significantly upgraded to handle the larger number of interpolation parameters associated with MOX fuels. The new methods have been incorporated in a new version of the ARP code that can generate libraries for low-enriched uranium (LEU) and MOX fuel types. The MOX data libraries and interpolation algorithms in ORIGEN-ARP have been verified using a database of declared isotopic concentrations for 1042 European MOX fuel assemblies. The methods and data are validated using a numerical MOX fuel benchmark established by the Organization for Economic Cooperation and Development (OECD) Working Group on burnup credit and nuclide assay measurements for irradiated MOX fuel performed as part of the Belgonucleaire ARIANE International Program.
AFCI-2.0 Library of Neutron Cross Section Covariances
Herman, M.; Herman,M.; Oblozinsky,P.; Mattoon,C.; Pigni,M.; Hoblit,S.; Mughabghab,S.F.; Sonzogni,A.; Talou,P.; Chadwick,M.B.; Hale.G.M.; Kahler,A.C.; Kawano,T.; Little,R.C.; Young,P.G.
2011-06-26
Neutron cross section covariance library has been under development by BNL-LANL collaborative effort over the last three years. The primary purpose of the library is to provide covariances for the Advanced Fuel Cycle Initiative (AFCI) data adjustment project, which is focusing on the needs of fast advanced burner reactors. The covariances refer to central values given in the 2006 release of the U.S. neutron evaluated library ENDF/B-VII. The preliminary version (AFCI-2.0beta) has been completed in October 2010 and made available to the users for comments. In the final 2.0 release, covariances for a few materials were updated, in particular new LANL evaluations for {sup 238,240}Pu and {sup 241}Am were adopted. BNL was responsible for covariances for structural materials and fission products, management of the library and coordination of the work, while LANL was in charge of covariances for light nuclei and for actinides.
The European Activation File. EAF-99 cross section library
In the European fusion programme, safety and environmental issues are of great importance in the continuing development of power plants. In support of this programme, a sound, complete and reliable neutron cross section data library is required. The European Activation File (EAF) project has been an ongoing process performed through European and world-wide co-operation that has led to the creation of succeeding EAF versions. The latest release, EAF-99, has benefited from the generation and maintenance of comprehensive activation files and the maturing of the processing code SYMPAL. Cross section validation exercises against both experimental data and systematics, which were started on the EAF-4 files, enable a comprehensive assessment of the data. Systematically, using SYMPAL, 3,168 reactions and branching ratios have been normalised to either experimental or systematic data, 26% of all the reactions. These are challenging tasks when the source contains threshold reactions with an energy dependent branching ratio. Although EAF-99 is the best-validated cross section library in the world, currently less than 16% of all EAF-99 reactions can be compared with experimental information, and sometimes then only for very limited, and not always relevant, energy ranges. Recently, for the first time, results of integral experiments have been used to adjust data. Validation of activation code predictions, and thereby of cross section and decay data, has been performed by means of direct comparison with measurements of sample structural material under fusion-relevant neutron spectra. Irradiations have been carried out at ENEA FNG, FZK Isochron-cyclotron, Sergiev Posad SNEG-13 and JAERI FNS and integral C/E comparisons made (C/E is the ratio of the library to the experimental value). The results of these benchmarking exercises have indicated, when correlated with other sources of information, corrective measures that have been taken on a selection of important reactions. The EAF
The European activation file: EAF-97 - cross section library
In the European Fusion programme, safety and environmental impact issues are of great importance in the continuing development of fusion plants. As part of this programme, a sound, complete and reliable neutron cross section data library is required. The European Activation File (EAF) project has been an ongoing process performed through European and worldwide cooperation that has led to the creation of succeeding EAF versions: from 1, in 1989, to 4.1 in 1995. The latest release, EAF-97, has benefited from the generation and maintenance of comprehensive activation files and the maturing of the processing code SYMPAL. Although the neutron cross-sections included in EAF come from sources having varying levels of quality, important new reliable sources (JENDL-3.2/A, IRK, FEI, JAERI, CRP) were available for EAF-97. Validation processes against either experimental data (compilation and EXFOR) or systematics, started on the EAF-4 files, enable a comprehensive assessment of the data. Automatically, in SYMPAL, 3,159 reactions and branching ratios were normalised to either experimental or systematic data, 25% of all the reactions. These are challenging tasks when the source contains threshold reactions with an energy dependent branching ratio. Although EAF-97 is the best-validated cross section library in the world, currently however, less than 16% of all EAF-97 reactions can be compared with experimental information, and sometimes only for very limited, not always relevant, energy ranges. The EAF-97 library contains about 12,500 excitation functions involving 766 different targets from 1H to 257Fm, atomic numbers 1 to 100, in the energy range 10-5 eV to 20 MeV. The 1,500,000 lines that make up the pointwise file are then processed into numerous groupwise files with different micro-flux weighting spectra to fit various user needs. Uniquely, an uncertainty file is also provided that quantifies the degree of confidence placed on the data for each reaction channel. (author)
The European Activation File. EAF-2001 cross section library
In the European fusion programme, safety and environmental issues are of great importance in the continuing development of power plants. In support of this programme, a sound, complete and reliable neutron cross section data library is required. The European Activation File (EAF) project has been an ongoing process performed through European and world-wide co-operation that has led to the creation of succeeding EAF versions. The latest release, EAF-2001, has benefited from the generation and maintenance of comprehensive activation files and the development of the new processing code SAFEPAQ-II. Cross section validation exercises against both experimental data and systematics, which were started on the EAF-4 files, enable a comprehensive assessment of the data. SAFEPAQ-II is used to apply a series of modifications to the original source data. A very important set of modifications concerns renormalisation and branching using experimental or systematic data. A total of 3,377 reactions have been so changed; 27% of all the reactions. These are challenging tasks when the source contains non-threshold reactions with an energy dependent branching ratio. Although EAF-2001 is the best-validated cross section library in the world, currently less than 16% of all the reactions can be compared with experimental information, and sometimes then only for very limited, and not always relevant, energy ranges. As with EAF-99, results of integral experiments have been used to adjust data. For a small number of reactions this can be done using SAFEPAQ-II: the remaining integral data was compared with activation predictions made using EAF-99 and adjustment factors found. Validation using integral data has been performed by means of direct comparison with measurements of sample structural material under fusion-relevant neutron spectra. Irradiations have been carried out at ENEA FNG, FZK Isochron- cyclotron, Sergiev Posad SNEG-13 and JAERI FNS and integral C/E comparisons made (C/E is the
Neutron Cross Section Libraries for Cryogenic Aromatic Moderator Materials
The dynamics of a set of aromatic hydrocarbons, such as benzene, toluene, mesitylene and a 3:2 mixture (by volume) of mesitylene and toluene, all of them in solid phase, was studied as potential moderator materials for cold neutron sources. Cross section libraries were generated for hydrogen bounded in those materials, at several temperatures in ACE format, and they were used in MCNP calculations to analyze their neutron production compared with traditional materials like solid methane and liquid hydrogen. In particular, cross section libraries were generated at 20 K, which is the operating temperature of the majority of the existing cold neutron sources. Although solid methane is the best moderator in terms of cold neutron production, it has very poor radiation resistance, causing spontaneous burping even at fairly low doses. Such effect is considerably reduced in the aromatic hydrocarbons. On the other hand, all of them show a similar and significant neutron production, with the exception of benzene. Even though those aromatic materials are very easy to handle, the solid phases that produce an enhanced flux of cold neutrons correspond to amorphous structures rich in low-energy excitations, and they can be created through lengthy cooling processes requiring in many cases additional annealing stages. The 3:2 mesitylene-toluene mixture, that forms in a simple and direct manner the appropriate disordered structure, constitutes an excellent cryogenic moderator material, as it is able to produce an intense flux of cold neutrons while presenting high resistance to radiation, thus conforming a new and advantageous alternative to traditional moderator materials. (authors)
LLL evaluated-nuclear-data library (ENDL): graphs of cross sections from the library
Graphs of neutron cross sections and related parameters for the LLL ENDL are presented. LLL and others use the evaluated nuclear data given here for neutron transport and related calculations. This document presents data contained in the library as of October 31, 1978. Evaluations of 79 nuclides/elements have been changed since the 1976 edition. Five tables discuss the information in the report, while the cross sections themselves are included on six microfiche
Neutron cross-section library for SAND-2 and its service program
The logical structure of the neutron cross-section library used in the SAND-2 program complex is considered. The organization of the DSIG01 program creating and servicing the neutron cross section library is described. The DSIG 01 program is written on FORTRAN and permits to create the neutron cross section library on the ES computer magnetic discs operating under the control of the ES operating system and to perform certain manipulations therewith
Revised transport cross-sections for the WIMS library
WIMS transport cross-sections above 4 eV are formed by a column-sum correction in which an assumed current spectrum is used to weight the P1 scattering data for a given isotope. Revised weighting spectra lead to improved transport cross-sections for the principal moderators: the effect on calculations of k-infinity is small but leakage calculations, for the homogenised cell, are now in close agreement with corresponding B1 calculations using explicit P1 data. Energy condensation of the B0 (transport corrected) equations appears to be more valid than the procedure used to condense the B1 equations. (author)
SHAMSI, 48 group cross-section library for fusion nucleonics analysis
A P3 48 group coupled neutron gamma-ray (34 N - 14 G) cross-section library is produced and validated for neutronic studies in fusion reactor blanket/shield. This report describes the library content, the procedure adopted and the results of the calculations performed for testing the cross sections
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
MCNP and MATXS cross section libraries based on JENDL-3.3
The continuous energy cross section library for the Monte Carlo transport code MCNP-4C, FSXLIB-J33, has been generated from the latest version of JENDL-3.3. The multigroup cross section library with the MATXS format, MATXS-J33, has been generated also from JENDL-3.3. Both libraries contain all nuclides in JENDL-3.3 and are processed at 300 K by the nuclear data processing system NJOY99. (author)
Expanded and applied sixteen-neutron-energy-group cross-section library
The purpose of the work reported in this paper was five-fold: (1) Develop an expanded neutron cross-section library containing ∼1,200 cross-section sets with the Hansen-Roach (H-R) 16-neutron-energy-group structure. (2) Provide an enhanced computational tool on a personal computer for criticality calculations. (3) Provide consistent values of the effective scattering cross sections (σs) for each set of the expanded H-R library for use in the selection of the resonance self-shielded cross sections (σp). (4) Develop a consistent technique for calculating σp in order to select and apply specific self-shielded cross-section sets. (5) Apply the cross sections and the selection technique to a wide variety of criticality calculational benchmarks
This document presents the description of a physical tape containing the basic evaluated nuclear data library of neutron-nucleus interaction cross sections, photon production cross sections and photon-atom interaction cross sections for fusion applications. It is part of the evaluated nuclear data library for fusion applications FENDL-2. The data are available cost-free from the Nuclear Data Section upon request. The data can also be retrieved by the user via online access through international computer networks. (author)
ACT-1000. Group activation cross-section library for WWER-1000 type reactors
The ACT-1000, a problem-oriented library of group-averaged activation cross-sections for WWER-1000 type reactors, is based on evaluated microscopic cross-section data files. The ACT-1000 data library was designed for calculating induced activity for the main dose-generated nuclides contained in WWER-1000 structural materials. In preparing the ACT-1000 library, 47 group-averaged cross-section data for the 10-9-17.33 MeV energy range were used to calculate the spatial-energy neutron flux distribution. (author)
Comparative analysis of the neutron cross-sections of iron from various evaluated data libraries
The comparative analysis of neutron cross-sections of iron from evaluated nuclear data libraries SOKRATOR, KEDAK, ENDL is done in energy interval from 0.025 eV to 20 MeV. Some of iron cross-sections from SOKRATOR library are revised and new data, which are obtained by using new experimental data and more comprehensive theoretical methods, are recommended. As a result the new version of the iron neutron cross-section file (BNF-2012) is produced for SOKRATOR library. (author)
Cai, Xiao-Xiao; Llamas-Jansa, Isabel; Mullet, Steven;
2013-01-01
Geant4 is an open source general purpose simulation toolkit for particle transportation in matter. Since the extension of the thermal scattering model in Geant4.9.5 and the availability of the IAEA HP model cross section libraries, it is now possible to extend the application area of Geant4 to re...... models and the G4NDL library. However, cross sections of those missing isotopes were made available recently through the IAEA project “new evaluated neutron cross section libraries for Geant4”....
ZZ XCOM, Photon Cross-Section Library for Personal Computer
1 - Description of program or function: Format: The input file FDAT produces the binary file UDAT (direct access un-formatted). This file is then used by the program XCOM1 to retrieve and display the photon cross-sections and attenuation coefficients. Number of groups: Photon cross-section data files (partial interaction coefficients and total attenuation coefficients) for 100 elements in the energy range 1 KeV to 100 GeV. Materials:H, He, Li, Be, B, C, N, O, F, Ne, Na, Mg, Al, Si, P, S, Cl, Ar, K, Ca, Sc, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zn, Ga, Ge, As, Se, Br, Kr, Rb, Sr, Y, Zr, Nb, Mo, Tc, Ru, Rh, Pd, Ag, Cd, In, Sn, Sb, Te, I, Xe, Cs, Ba, La, Ce, Pr, Nd, Pm, Sm, Eu, Gd, Tb, Dy, Ho, Er, Tm, Yb, Lu, Hf, Ta, W, Re, Os, Ir, Pt, Au, Hg, Tl, Pb, Bi, Po, At, Rn, Fr, Ra, Ac, Th, Pa, U, Np, Pu, Am, Cm, Bk, Cf, Es, Fm. Origin: Several sources. It is based on an experimental data base consisting of 21000 data points from 512 literature sources. Same sources as DLC-136/PHOTX. Weighting spectrum: The weighting factors, i.e., the fractions by weights of the atomic constituents, are calculated from the chemical formula entered by the user. The National Institute of Standards and Technology, through its Office of Standard Reference Data, has long maintained and published compilations of measured and evaluated photon cross sections. This compilation of XCOM Version 1.2, released on personal computer media, represents best values as determined in 1987. XCOM1 (Version 1.3, copyright 1991) is similar to XCOM but uses the direct-access un-formatted database file UDAT. 2 - Method of solution: The data from the National Institute of Standards and Technology are in binary files for 100 elements covering the energy range 1 keV to 100 GeV. The reactions considered are coherent and incoherent scattering, photoelectric absorption, and pair production. The XCOM data are derived from the same source as DLC-0136/ZZ-PHOTX
Pritychenko, B.; Mughabghab, S.F.
2012-01-01
We present calculations of neutron thermal cross sections, Westcott factors, resonance integrals, Maxwellian-averaged cross sections and astrophysical reaction rates for 843 ENDF materials using data from the major evaluated nuclear libraries and European activation file. Extensive analysis of newly-evaluated neutron reaction cross sections, neutron covariances, and improvements in data processing techniques motivated us to calculate nuclear industry and neutron physics quantities, produce s-...
Zaidi, H
1999-01-01
the many applications of Monte Carlo modelling in nuclear medicine imaging make it desirable to increase the accuracy and computational speed of Monte Carlo codes. The accuracy of Monte Carlo simulations strongly depends on the accuracy in the probability functions and thus on the cross section libraries used for photon transport calculations. A comparison between different photon cross section libraries and parametrizations implemented in Monte Carlo simulation packages developed for positron emission tomography and the most recent Evaluated Photon Data Library (EPDL97) developed by the Lawrence Livermore National Laboratory was performed for several human tissues and common detector materials for energies from 1 keV to 1 MeV. Different photon cross section libraries and parametrizations show quite large variations as compared to the EPDL97 coefficients. This latter library is more accurate and was carefully designed in the form of look-up tables providing efficient data storage, access, and management. Toge...
Development of a Multi-Group Neutron Cross Section Library Generation System for PWR
Kim, Kang Seog; Hong, Ser Gi; Song, Jae Seung; Lee, Kyung Hoon; Cho, Jin Young; Kim, Ha Yong; Koo, Bon Seung; Shim, Hyung Jin; Park, Sang Yoon
2008-10-15
This report describes a generation system of multi-group cross section library which is used in the KARMA lattice calculation code. In particular, the theoretical methodologies, program structures, and input preparations for the constituent programs of the system are described in detail. The library generation system consists of the following five programs : ANJOY, GREDIT, MERIT, SUBDATA, and LIBGEN. ANJOY generates automatically the NJOY input files and two batch files for automatic NJOY run for all the nuclides considered. The automatic NJOY run gives TAPE 23 (PENDF output file of BROADR module of NJOY) and TAPE24 (GENDF output file of GROUPR module of NJOY) files for each nuclide. GREDIT prepares a formatted multi-group cross section file in which the cross sections are tabulated versus temperature and background cross section after reading the TAPE24 file. MERIT generates the hydrogen equivalence factors and the resonance integral tables by solving the slowing down equation with ultra-fine group cross sections which are prepared with the TAPE 23 file. SUBDATA generates the subgroup data including subgroup levels and weights after reading the MERIT output file. Finally, LIBGEN generates the final multi-group library file by assembling the data prepared in the previous steps and by reading the other data such as fission product yield data and decay data.The multi-group cross section library includes general multi-group cross sections, resonance data, subgroup data, fission product yield data, kappa-values (energy release per fission), and all the data which are required in the depletion calculation. The addition or elimination of the cross sections for some nuclides can be easily done by changing the LIBGEN input file if the general multi-group cross section and the subgroup data files are prepared.
XNWLUP, Graphical user interface to plot WIMS-D library multigroup cross sections
1 - Description of program or function: XnWlup is a computer program with user-friendly graphical interface to help the users of WIMS-D library to enable quick visualisation of the plots of the energy dependence of the multigroup cross sections of any nuclide of interest. This software enables the user to generate and view the histogram of 69 multi-group cross sections as a function of neutron energy under Microsoft Windows environment. This software is designed using Microsoft Visual C++ and Microsoft Foundation Classes Library. IAEA1395/05: New features of version 3.0: - Plotting absorption and fission cross sections of resonant nuclide after applying the self-shielding cross section. - Plotting the data of Resonant Integral table, as a function of dilution cross section for a selected temperature and for a given energy group. - Plotting the data of Resonant Integral table, as a function of temperature for a selected background dilution cross section and for a given energy group. - Clearing all the graphs except one graph from the display screen is easily done by using a tool bar button. - Displaying the coordinate of the cursor point with appropriate units. 2 - Methods: XnWlup helps to obtain histogram plots of the values of cross section data of an element/isotope available as 69-group WIMS-D library as a function of energy bins. The software XnWlup is developed with this graphical user interface in order to help those users who frequently refer to the WIMS-D library cross section data of neutron-nuclear reactions. The software also helps to produce handbook of WIMS-D cross sections
Development of a Multi-Group Neutron Cross Section Library Generation System for PWR
This report describes a generation system of multi-group cross section library which is used in the KARMA lattice calculation code. In particular, the theoretical methodologies, program structures, and input preparations for the constituent programs of the system are described in detail. The library generation system consists of the following five programs : ANJOY, GREDIT, MERIT, SUBDATA, and LIBGEN. ANJOY generates automatically the NJOY input files and two batch files for automatic NJOY run for all the nuclides considered. The automatic NJOY run gives TAPE 23 (PENDF output file of BROADR module of NJOY) and TAPE24 (GENDF output file of GROUPR module of NJOY) files for each nuclide. GREDIT prepares a formatted multi-group cross section file in which the cross sections are tabulated versus temperature and background cross section after reading the TAPE24 file. MERIT generates the hydrogen equivalence factors and the resonance integral tables by solving the slowing down equation with ultra-fine group cross sections which are prepared with the TAPE 23 file. SUBDATA generates the subgroup data including subgroup levels and weights after reading the MERIT output file. Finally, LIBGEN generates the final multi-group library file by assembling the data prepared in the previous steps and by reading the other data such as fission product yield data and decay data.The multi-group cross section library includes general multi-group cross sections, resonance data, subgroup data, fission product yield data, kappa-values (energy release per fission), and all the data which are required in the depletion calculation. The addition or elimination of the cross sections for some nuclides can be easily done by changing the LIBGEN input file if the general multi-group cross section and the subgroup data files are prepared
A subject of activation cross section library for IRAC code system
The IRAC code system has a major activation cross section library, ACSELA94, which has been calculated using ALICE-F code on the basis of selection of geometry depended hybrid models with conditions of 9 incident particles and 136 target nuclides in the range of 1H - 209Bi. The incident energies of light ions (1H, 2H, 4He) and neutrons are from threshold energy to 150 MeV, and of heavy ions (12C, 14N, 16O, 20Ne, 40Ar) are from threshold energy to 500 MeV, respectively. To obtain the general tendency of the calculated cross section, the sum total of individual isotope production cross sections was compared with nonelastic cross section. It was found that they are in good agreement with the both in which the mass number of target nuclides is from 15 to 185. Furthermore some subjects of the cross section in ACSELA94 were found out. (author)
Neutron cross-section libraries in the AMPX master interface format for thermal and fast reactors
Neutron cross-section libraries in the AMPX master interface format have been created for three reactor types. Included are an 84-group library for use with light-water reactors, a 27-group library for use with heavy-water CANDU reactors and a 126-group library for use with liquid metal fast breeder reactors. In general, ENDF/B data were used in the creation of these libraries, and the nuclides included in each library should be sufficient for most neutronic analyses of reactors of that type. Each library has been used successfully in fuel depletion calculations
KOEBLIB1.0: Two group polynomial cross section library for Koeberg version 1
The mathematical models and engineering data used for the generation of version 1 of the 2-group polynomial cross section library for the two PWR units at the Koeberg Nuclear Power Station, are described. This library was prepared using the few-group coarse mesh cross section generation package of the Reactor Theory Programme at the Atomic Energy Corporation of South Africa Ltd. An overall description of the calculational scheme as well as descriptions of the various modules used for the generation of the cross section library is given. The fuel assembly model is described in detail and the values of the operational parameters used, are given. The methods used to generate the ex-core reflector data are described. Details of the generation of the polynomial library are given and the assembly and reflector engineering data are listed. 2 figs., 6 tabs., 19 refs
A code system has been developed to produce neutron cross section libraries for the MVP continuous energy Monte Carlo code from an evaluated nuclear data library in the ENDF format. The code system consists of 9 computer codes, and can process nuclear data in the latest ENDF-6 format. By using the present system, MVP neutron cross section libraries for important nuclides in reactor core analyses, shielding and fusion neutronics calculations have been prepared from JENDL-3.1, JENDL-3.2, JENDL-FUSION file and ENDF/B-VI data bases. This report describes the format of MVP neutron cross section library, the details of each code in the code system and how to use them. (author)
Library of neutron reaction cross-sections in the ABBN-93 constant system
The library of neutron reaction group cross-sections in the ABBN-93 constant set is described. The format used for data representation, the content and purpose of the sub-libraries and their practical application in the SCALE criticality safety estimation system are discussed. (author)
Up to date cross sections library for Thermos and Record codes
Reactor cell analysis is the first step in determining reactor core behavior and is required in the reload licensing process. For best results, reactor cell analysis should be carried out with libraries of up to date, accurate cross sections produced with well described methods from standard evaluated nuclear data. At first step in this work were determined the library structure for RECORD and THERMOS and were prepared the cross sections libraries using the NJOY nuclear data processing system and the ENDF-B/IV evaluated nuclear data. These libraries were used by the codes and some samples were perform, the result show some differences against the results obtained using the previous libraries. By other hand the libraries contain various adjustments to correct for deficiencies in nuclear data or analytical methods. These adjustments doesn't have any documentation, although some of them were identified in this work. (Author). 25 refs, 78 figs, 55 tabs
Basis calculation of phase cross section library in a low power fast reactor neutronic simulation
In order to implement the utilization of the efficient multidimensional cubic SPLINE interpolation, we determine the phase library bases for net like relevant state components. A generic cubic surface and a weighted plane pertinent alternative interpolating methods used capable to generate cross sections values for fixed coordinates from cell code calculated data points is used. It is verified that the phase library bases increases or decrease smoothly and monotonically with the spectrum asymmetry and total flux buckling. This justifies its use in cross section updating avoiding cell calculations. (author)
Description of the ENDF-NJOY system for the generation of cross sections libraries
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)
TEST, Sort, Delete, List ANISN and DOT Cross-Sections Library Data
1 - Description of program or function: Test is an auxiliary program for sorting, deleting and listing data contents of ANISN and DOT cross section libraries, generated with AMICO or any other program. 2 - Restrictions on the complexity of the problem: No restrictions on the number of energy groups or materials are noted because the program uses the variable dimension technique
ECNJEFI. A JEFI based 219-group neutron cross-section library: User's manual
This manual describes the contents of the ECNJEF1 library. The ECNJEF1 library is a JEF1.1 based 219-group AMPX-Master library for reactor calculations with the AMPX/SCALE-system, e.g. the PASC-3 system as implemented at the Netherlands Energy Research Foundation in Petten, Netherlands. The group cross-section data were generated with NJOY and NPTXS/XLACS-2 from the AMPX system. The data on the ECNJEF1 library allows resolved-resonance treatment by NITAWL and/or unresolved resonance self-shielding by BONAMI. These codes are based upon the Nordheim and Bondarenko methods, respectively. (author). 10 refs., 7 tabs
EJ1: a JEF1 based 219-group neutron cross-section library
This manual describes the contents of the EJ1 library. The EJ1 library is a JEF1.1 based 219-group AMPX-Master library for reactor calculations with the AMPX/SCALE-3 system, e.g. the PASC-3 system, as implemented at ECN-Petten. The group cross-section data were generated with NJOY. The data on the EJ1 library allow resolved-resonance treatment by NITAWL and unresolved resonance self-shielding by BONAMI. These codes are based upon the Nordheim and Bondarenko methods, respectively. (author). 24 refs., 8 tabs
How to Use Benchmark and Cross-section Studies to Improve Data Libraries and Models
Wagner, V.; Suchopár, M.; Vrzalová, J.; Chudoba, P.; Svoboda, O.; Tichý, P.; Krása, A.; Majerle, M.; Kugler, A.; Adam, J.; Baldin, A.; Furman, W.; Kadykov, M.; Solnyshkin, A.; Tsoupko-Sitnikov, S.; Tyutyunikov, S.; Vladimirovna, N.; Závorka, L.
2016-06-01
Improvements of the Monte Carlo transport codes and cross-section libraries are very important steps towards usage of the accelerator-driven transmutation systems. We have conducted a lot of benchmark experiments with different set-ups consisting of lead, natural uranium and moderator irradiated by relativistic protons and deuterons within framework of the collaboration “Energy and Transmutation of Radioactive Waste”. Unfortunately, the knowledge of the total or partial cross-sections of important reactions is insufficient. Due to this reason we have started extensive studies of different reaction cross-sections. We measure cross-sections of important neutron reactions by means of the quasi-monoenergetic neutron sources based on the cyclotrons at Nuclear Physics Institute in Řež and at The Svedberg Laboratory in Uppsala. Measurements of partial cross-sections of relativistic deuteron reactions were the second direction of our studies. The new results obtained during last years will be shown. Possible use of these data for improvement of libraries, models and benchmark studies will be discussed.
Effects of cross sections libraries parameters on the OECD/NEA Oskarshamn-2 benchmark solution
The OECD/NEA proposes a new international benchmark based on the data collected from an instability transient occurred at the Oskarshamn-2 NPP with the aim to test the coupled 3D Neutron Kinetic/Thermal Hydraulic codes on challenging situations. The ENEA 'Casaccia' Research Center is participating to this benchmark, developing a computational model using RELAP5-3D code. The 3DNK model was developed starting from the cross sections datasets calculated by OKG, the Oskarshamn-2 licensee, using the CASMO lattice code. Integration of neutron cross sections database in RELAP5-3D required data fitting by a n-dimensional polynomials, calculations of the various polynomial coefficients and of the base cross sections values. An ad-hoc tool named PROMETHEUS has been developed for automatically generate the RELAP5-3D-compatible cross sections libraries. Thanks to this software it has been easily possible to visualize the complex structure of the neutronic data sets and to derive different cross sections libraries for evaluating the effects of some neutronic parameters on the prediction of the reactor instability. Thus, the effects of the fuel temperature and control rod history, of the discontinuity factors (averaged/not averaged), and of the neutron poisons has been assessed. A ranking table has been produced, demonstrating the relevance of the not-averaged discontinuity factors and of the on-transient neutron poisons calculations for the correct prediction of the Oskarshamn-2 event. (author)
Effects of cross sections library parameters on the OECD/NEA Oskarshamn-2 benchmark solution
Highlights: • A 3D NK–TH model was developed using RELAP5-3D© for studying BWR instability events. • A cross section library was generated using the available CASMO format data. • To evaluate reactor stability parameters a tool was developed and validated. • The effect of some neutronic parameters on the reactor stability was investigated. • The Oskarshamn-2 1999 event stability parameters were properly reproduced. - Abstract: The OECD/NEA proposes a new international benchmark based on data collected during an instability transient occurred at the Oskarshamn-2 NPP. This benchmark is aimed at testing the coupled 3D Neutron Kinetic–Thermal Hydraulic (3D NK–TH) codes on challenging situations. The ENEA “Casaccia” Research Center, is participating to this benchmark, developing a computational model using the RELAP5-3D© code. The 3DNK model has been already developed from the cross sections dataset calculated by OKG, the Oskarshamn-2 licensee, through the CASMO lattice code. In order to use this neutron cross sections database in RELAP5-3D© a n-dimensional polynomials data fitting and base cross sections values calculations are required. An ad-hoc tool, named PROMETHEUS, has been developed for automatically generating RELAP5-3D©-compatible cross sections libraries. This tool allows at easily visualizing the complex structure of the neutronic datasets; moreover it is exploited for deriving different cross sections libraries needed to evaluate neutronic parameters effects on the reactor instability prediction. Thus, the effects of the fuel temperature and control rod histories, of the discontinuity factors (averaged/not averaged) and of the neutron poisons has been assessed. A ranking table has been produced, demonstrating the relevance of the not-averaged discontinuity factors and of the on-transient neutron poisons calculations for the correct prediction of the Oskarshamn-2 event
Development of an iterative lattice-core coupling method based on MICROX-2 cross section libraries
This paper describes an innovative online cross section generation method, developed based on Iterative Diffusion-Transport (IDT) calculation to minimize the inconsistency and inaccuracy in determining physics parameters by feeding actual reactor core conditions into the cross section generation process. A 2-dimensional (2-D) pin-by-pin lattice program, NEMA, was also developed to generate assembly lattice parameters using the embedded MICROX-2 cross section libraries and Nodal Expansion Method (NEM). The proposed methods were validated against a 2-D miniature core benchmark problem for both NEMA itself and its coupling to a reactor code by the IDT method (NEMA-DIF3D). The computational benchmark calculations have shown that the IDT improves the eigenvalue and power distribution predictions when compared with the conventional offline method. (author)
FENDL/A-2.0. Neutron activation cross section data library for fusion applications
This document describes the contents of a comprehensive neutron cross section data library for 13,006 neutron activation reactions with 739 target nuclides from H (A=1,Z=1) to Cm (A=248,Z=96), in the incident energy range up to 20 MeV. FENDL/A-2 is a sublibrary of FENDL-2, the second revision of the evaluated nuclear data library for fusion applications. It is supplemented by a decay data library FENDL/D-2 in ENDF-6 format for 1867 nuclides. The data are available from the IAEA Nuclear Data Section online via INTERNET by FTP command, or on magnetic tape upon request. (author)
Evaluation of neutron cross sections for the Pd isotopes (RCN-3 data library)
The evaluation procedure to obtain neutron cross sections of 102Pd, 104Pd, 105Pd, 106Pd, 107Pd, 108Pd and 110Pd for inclusion in the RCN-3 data library of fission-product cross sections is described. The new evaluation takes into account the results of recent differential and integral data. Most of the adopted resolved resonance parameters have been taken from the new CBNM measurements; for 107Pd the recent RPI data have been used. These resolved resonance parameters have been extensively analysed to obtain average values for the level spacing, capture width and neutron s- and p-wave strength functions. The systematics of the single-particle level-density parameter α and the capture width show significant odd-even effects. Optical-model and statistical-model calculations have been performed to obtain cross sections of reactions at energies from 1 meV to 15 MeV. The results for the capture cross sections based upon the analysed average resonance parameters turned out to be systematically lower than the ORELA average capture data and also lower than indicated by the most integral STEK data (except for 105Pd). As a compromise we have performed adjustments to increase the calculated fast capture cross sections for 104Pd, 106Pd and 108Pd
ZZ MONTAGE-400, Neutron Activation 100-Group Cross-Section Library of Fusion Reactor Materials
1 - Description of problem or function: Format: GAM-II group structure and ANISN; Number of groups: 100-group cross sections. Nuclides: H, He, Li, Be, B, N, O, F, Na, Al, Si, P, S, Cl, K, Ca, Sc, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Y, Zr, Nb, Mo, Tc, Ru, Ag, Sn, Cs, Hf, Ta, W, Re, Au, Pb. Origin: derived from ENDF/B, or calculated at Brookhaven National Laboratory. Weighting spectrum: 1/E except near 14 MeV where a thermally broadened fusion peak, assuming a temperature of 20 MeV, is employed. This data library contains 100- group cross sections, with GAM-II group structure, for 421 neutron activation reactions with fusion reactor structural and coolant materials. The weighting function is 1/E except near 14 MeV where a thermally broadened fusion peak, assuming a temperature of 20 MeV, is employed. The library also contains half life information for the activated nuclei. 2 - Method of solution: The thermal group cross sections were calculated from the 2200 m/s value, when available, otherwise from the group 99 value. The majority of the non-thermal cross sections were derived from pointwise data derived from ENDF/B, or calculated at Brookhaven National Laboratory using the nuclear systematics code THRESH. These were converted to multigroup from using the codes ETOG and NJOY
EJ2-MCNPlib. Contents of the JEF-2.2 based neutron cross-section library for MCNP4A
In this report a description is given of the EJ2-MCNPlib library. The EJ2-MCNPlib library is to be used for reactivity/critically calculations and general neutron/photon transport calculations with the Monte Carlo code MCNP4A. The library is based on the European JEF-2.2 nuclear data evaluation and contains data for all (i.e. 313) nuclides available on this evaluation.The cross-section data were generated using the NJOY cross-section processing code system, version 91.118. For easy reference cross-section plots are given in this report for the total, elastic and absorption cross sections for all nuclides on the EJ2-MCNPlib library. Furthermore, for verification purposes a graphical intercomparison is given of the results of standard benchmark calculations performed with JEF-2.2 cross-section data and with ENDF/B-V cross-section data (whenever available). 6 refs
On the basis of the library structure of the RECORD code a method to generate the neutron cross sections by means of the ENDF-B/IV database and the NJOY code has been developed. The obtained cross sections are compared with those of the current library which was processed using the ENDF-B/III version. (Author)
PWR ENDF/B-VII Cross-Section Libraries for ORIGEN-ARP
New pressurized water reactor (PWR) cross-section libraries were generated for use with the ORIGEN-ARP depletion sequence in the SCALE nuclear analysis code system. These libraries are based on ENDF/B-VII nuclear data and were generated using the two-dimensional depletion sequence, TRITON/NEWT, in SCALE 6.1. The libraries contain multiple burnup-dependent cross-sections for seven PWR fuel designs, with enrichments ranging from 1.5 to 6 wt% 235U. The burnup range has been extended from the 72 GWd/MTU used in previous versions of the libraries to 90 GWd/MTU. Validation of the libraries using radiochemical assay measurements and decay heat measurements for PWR spent fuel showed good agreement between calculated and experimental data. Verification against detailed TRITON simulations for the considered assembly designs showed that depletion calculations performed in ORIGEN-ARP with the pre-generated libraries provide similar results as obtained with direct TRITON depletion, while greatly reducing the computation time.
PWR ENDF/B-VII cross-section libraries for ORIGEN-ARP
New pressurized water reactor (PWR) cross-section libraries were generated for use with the ORIGEN-ARP depletion sequence in the SCALE nuclear analysis code system. These libraries are based on ENDF/B-VII nuclear data and were generated using the two-dimensional depletion sequence, TRITON/NEWT, in SCALE 6.1. The libraries contain multiple burnup-dependent cross-sections for seven PWR fuel designs, with enrichments ranging from 1.5 to 6 wt% 235U. The burnup range has been extended from the 72 GWd/MTU used in previous versions of the libraries to 90 GWd/MTU. Validation of the libraries using radiochemical assay measurements and decay heat measurements for PWR spent fuel showed good agreement between calculated and experimental data. Verification against detailed TRITON simulations for the considered assembly designs showed that depletion calculations performed in ORIGEN-ARP with the pre-generated libraries provide similar results as obtained with direct TRITON depletion, while greatly reducing the computation time. (authors)
Development of an iterative diffusion-transport method based on MICROX-2 cross section libraries
Highlights: • Innovative Iterative Diffusion Transport (IDT) method is developed. • A 2-dimensional (2-D) pin-by-pin lattice program, NEMA, is also developed. • The developed methods and codes are verified on benchmark problems. • Results show that the IDT method improves the global and local predictions. - Abstract: This paper introduces an innovative online cross section generation method, developed based on Iterative Diffusion-Transport (IDT) calculation to minimize the inconsistency and inaccuracy in determining physics parameters by feeding actual reactor core conditions into the cross section generation process. A two-dimensional (2-D) pin-by-pin lattice program, NEMA, was developed to generate assembly lattice parameters using the refined MICROX-2 cross section libraries and Nodal Expansion Method (NEM). The proposed method was verified against a 2-D miniature core (mini-core) benchmark problem. First, the few-group cross sections generated by NEMA were compared with those calculated by a Monte Carlo method code Serpent. Next, the analysis of a 2-D Light Water Reactor (LWR) mini-core benchmark problem was carried out by the nodal transport code DIF3D using few-group cross sections generated by NEMA, and the results were compared with those obtained from the Serpent full core calculation. Finally, the same benchmark problem was solved by the NEMA-DIF3D approach using the IDT coupling method. The computational benchmark calculations have shown that the homogenization technique implemented in NEMA is reliable when producing the few-group cross sections for the reactor core calculation. The IDT method also improves the eigenvalue and power distribution predictions
ARP: A PC-compatible scheme for generating ORIGEN-S cross section library
The SAS2H sequence of the SCALE code system has been widely used for treating problems related to the characterization of nuclear systems for disposal, storage, and shipment. The calculations, in general, consist of determining the isotope compositions of the different materials present in the problem as a function of time, which subsequently enable determination of the heat generation and radiation source terms. In the SAS2H scheme, time-dependent material concentrations are obtained using the ORIGEN-S code based on a point-depletion calculation that utilizes problem-dependent cross-section libraries generated by distinct codes of the SAS2H sequence. In this paper we will be concerned with the methodology utilized in the SAS2H control module to create cross-section libraries for point-depletion calculations with the ORIGEN-S code. A brief description of the SAS2H scheme will be given, and a new capability, the automatic rapid processing (ARP), for generating problem-dependent ORIGEN-S cross-section libraries will be presented. Use of ARP can enable execution of ORIGEN-S on a personal computer with identical accuracy to that obtained with SAS2H
ORIGEN-S cross section libraries for CANDU used-fuel characterization
A code system for producing burn-up dependent cross-section libraries for CANDU used-fuel characterization for use with the ORIGEN-S isotope generation and depletion code system is described. Benchmark results against experimental isotopic data for three CANDU-PHW reactor stations are presented. The code system couples the WIMS-AECL reactor physics analysis code with an ORIGEN-S depletion analysis to produce application-specific libraries that can be used in subsequent used-fuel analyses. 11 refs., 1 fig., 3 tabs
Consistent Generation and Verification of 190 Group Cross Section Library Data for Primary Nuclides
The multigroup cross section data used in the lattice transport or the direct whole core transport codes such as HELIOS and DeCART have a significant impact on the accuracy of the criticality prediction. If a large discrepancy is noted in the analysis of critical experiments, it is customary to adjust the resonance integral (RI) data of U-238 given in the cross section library in order to match the measurement. In case of HELIOS, the unadjusted library gives about 300∼550 pcm lower reactivity than the adjusted one. The sole adjustment of the U238 RI, however, is to blame only U238 for all the discrepancies that can originate from various sources. One of the sources of the error would be the inaccuracy of subgroup parameters used in the a group codes which employ the subgroup method for resonance treatment. The inconsistency problem noted in the subgroup parameter generation and usage steps which was reported in our previous work can be smeared out by the RI adjustment. Thus such blind adjustment of the resonance integral is to be avoided. In this work, we examine a new procedure for generating multigroup cross section data from the ENDF/B files, which would not require any forced adjustment. One of the distinct steps in the procedure is to employ a consistent method of generating subgroup parameters formulated by imposing a shielded cross section conservation principle rather than the resonance integral conservation. In order to check the validity of the procedure, multigroup data are generated only for a group of primary nuclides which appear in a fresh fuel UO2 pin cell, namely, U-235, U-238, H-1, O-16, and Zr. The accuracy of the new library is assessed by comparing the reactivity with those obtained from corresponding continuous energy Monte Carlo calculations. Since recently the ENDF/B-VII file was released which reflects improvements in the U238 resonance data, the difference between the multigroup cross section libraries generated from the new ENDF file
Consistent Generation and Verification of 190 Group Cross Section Library Data for Primary Nuclides
Kim, Gwan Young; Joo, Han Gyu [Seoul National University, Seoul (Korea, Republic of)
2008-05-15
The multigroup cross section data used in the lattice transport or the direct whole core transport codes such as HELIOS and DeCART have a significant impact on the accuracy of the criticality prediction. If a large discrepancy is noted in the analysis of critical experiments, it is customary to adjust the resonance integral (RI) data of U-238 given in the cross section library in order to match the measurement. In case of HELIOS, the unadjusted library gives about 300{approx}550 pcm lower reactivity than the adjusted one. The sole adjustment of the U238 RI, however, is to blame only U238 for all the discrepancies that can originate from various sources. One of the sources of the error would be the inaccuracy of subgroup parameters used in the a group codes which employ the subgroup method for resonance treatment. The inconsistency problem noted in the subgroup parameter generation and usage steps which was reported in our previous work can be smeared out by the RI adjustment. Thus such blind adjustment of the resonance integral is to be avoided. In this work, we examine a new procedure for generating multigroup cross section data from the ENDF/B files, which would not require any forced adjustment. One of the distinct steps in the procedure is to employ a consistent method of generating subgroup parameters formulated by imposing a shielded cross section conservation principle rather than the resonance integral conservation. In order to check the validity of the procedure, multigroup data are generated only for a group of primary nuclides which appear in a fresh fuel UO{sub 2} pin cell, namely, U-235, U-238, H-1, O-16, and Zr. The accuracy of the new library is assessed by comparing the reactivity with those obtained from corresponding continuous energy Monte Carlo calculations. Since recently the ENDF/B-VII file was released which reflects improvements in the U238 resonance data, the difference between the multigroup cross section libraries generated from the new
Benchmark Tests of the Multigroup Cross Section Libraries for Fast Reactors
In Korea, a design study for a fast breeder reactor named KALIMER (Korea Advanced LIquid MEtal Reactor) has been carried out. The simulations of the KALIMER core have been performed with the JEF-2.2- based 80-group neutron library KAFAX-F22 or the ENDF/B-VI.6-based 150-group neutron library KAFAXE66. Recently, newly evaluated nuclear data files such as ENDF/B-VII (beta 0 and 1), JEFF-3.1, and JENDL-3.3 have been released. And thus there is a need to update the libraries for the KALIMER by using the new data files. In this study, the fast cross section sets with 150 groups were prepared based on ENDF/B-VII beta 0, JEFF-3.1, and JENDL-3.3. The validations of the libraries have been carried out for 14 Cross Section Evaluation Working Group (CSEWG) fast benchmark problems through the 1-D and 2-D DANTSYS calculations. The effective multiplication factors (keff's) and central spectral indices have been compared with the experimental values and the results by the MCNPX calculations
Recent validation experience with multigroup cross-section libraries and scale
This paper will discuss the results obtained and lessons learned from an extensive validation of new ENDF/B-V and ENDF/B-VI multigroup cross-section libraries using analyses of critical experiments. The KENO V. a Monte Carlo code in version 4.3 of the SCALE computer code system was used to perform the critical benchmark calculations via the automated SCALE sequence CSAS25. The cross-section data were processed by the SCALE automated problem-dependent resonance-processing procedure included in this sequence. Prior to calling KENO V.a, CSAS25 accesses BONAMI to perform resonance self-shielding for nuclides with Bondarenko factors and NITAWL-II to process nuclides with resonance parameter data via the Nordheim Integral Treatment
The French 'CEA 86' multigroup cross-section library and its integral qualification
This paper describe the up-dated 99 groups library of the APOLLO French neutron computer code, the denominated 'CEA 86' library. The multigroup cross-section sets are based on the more recent nuclear data evaluations. The THEMIS code was generally used for the JEF-1 processing. In order to account for recent differential measurements and to improve the consistency between calculation and integral experiments, we produced our own CEA evaluations for the actinide nuclides: 235U, 238U, 239Pu, 240Pu, 241Am. This new APOLLO library was checked against critical experiments and PWR measurements: computed Conversion Factor, Reactivity Coefficients, Multiplication Factor, and Pu build-up are now in good agreement with LWR experimental results. PWR Pu recycling calculations, as does as HCLWR design studies, are also significantly improved. (author)
RCPL1 is a FORTRAN digital computer program designed and developed to prepare neutron and photon cross section libraries for the RCP01 Monte Carlo computer program for solving neutron and photon transport problems in three-dimensional geometry with detailed energy description. The neutron libraries prepared by RCPL1 contain detailed Doppler-broadened resonance cross sections from unresolved and either single-level or multilevel resonance parameters, for any number of nuclides, within an arbitrary energy structure, and the photon libraries contain tabulations of the interaction cross sections and gamma emission spectra. This report describes the various RCPL1 program options, calculational details, and input requirements. All data used for library construction are extracted from a multigroup cross section library system XAP, described in an appendix to the report, which contains Evaluated Nuclear Data File (ENDF) data. 5 figures, 6 tables
MICROX-2 cross section library based on ENDF/B-VII
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 238U-to-235U fission of ZEBRA lattice against the reference calculation done by MCNP5. (authors)
Full text: New releases of nuclear data files made available during the few recent years. The reference MCNP5 code (1) for Monte Carlo calculations is usually distributed with only one standard nuclear data library for neutron interactions based on ENDF/B-VI. The main goal of this work is to process new neutron cross sections libraries in ACE continuous format for MCNP code based on the most recent data files recently made available for the scientific community : ENDF/B-VII.b2, ENDF/B-VI (release 8), JEFF3.0, JEFF-3.1, JENDL-3.3 and JEF2.2. In our data treatment, we used the modular NJOY system (release 99.9) (2) in conjunction with its most recent upadates. Assessment of the processed point wise cross sections libraries performances was made by means of some criticality prediction and analysis of other integral parameters for a set of reactor benchmarks. Almost all the analyzed benchmarks were taken from the international handbook of Evaluated criticality safety benchmarks experiments from OECD (3). Some revised benchmarks were taken from references (4,5). These benchmarks use Pu-239 or U-235 as the main fissionable materiel in different forms, different enrichments and cover various geometries. Monte Carlo calculations were performed in 3D with maximum details of benchmark description and the S(α,β) cross section treatment was adopted in all thermal cases. The resulting one standard deviation confidence interval for the eigenvalue is typically +/-13% to +/-20 pcm
Testing of the IRDF-90 cross-section library in benchmark neutron spectra
The new version of the International Reactor Dosimetry File IRDF-90 (called ''Version April 1993'') has been tested by calculation of average cross-sections and their uncertainties in a coarse three energy group structure and by neutron spectrum adjustments in reference neutron spectra. This paper presents the results obtained and compares them with the corresponding ones of the old IRDF-85 and with the data of the Nuclear Data Guide for Reactor Neutron Metrology. The applicability of the new library in the field of neutron metrology is discussed. (orig.)
Comparative calculations of the experimental benchmark of iron sphere with Cf source have been performed in order to assess the sensibility of the calculations of neutron transmission through iron media to different multigroup libraries generated on the base of ENDF/B-6 and ENDF/B-4. Similar calculations and comparison of the neutron flux passed through media typical as geometry and material compositions for the WWER-1000 and WWER-440 vessels have been carried out. Except the already well-known problem dependent libraries, the new libraries BGL-440 and BGL-1000 generated on the base of ENDF/B-6 for the WWER-440 and WWER-1000 RPV neutron fluence calculations have been applied. The solving of neutron transport through iron media using ENDF/B-6 data gives better consistency with the experiment than using ENDF/B-4. The latter underestimate the experimental fluxes more substantially in the energy range above 2 MeV and the evaluations of the neutron flux responses for the WWER vessel surveillance is preferably to be carried out by the appropriate BGL library. Key words: neutron transport, multigroup neutron cross section libraries
VITAMIN E: a multipurpose ENDF/B-V coupled neutron-gamma cross section library
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
A 39 neutron group self-shielded cross section library for the Lotus fusion-fission test facility
A 39 neutron group cross section library for fusion fission blanket calculations and especially for the analysis of the LOTUS experiment has been processed using the NJOY system. The library has been generated mostly using the ENDF/B-IV basic files at 296 K. All cross sections were self-shielded using the Bondarenko method. 5 background cross sections, namely 1010, 104, 102, 10 and 1 barns respectively were considered. The tabulated dilution dependent cross sections have been interpolated with the code TRANSX-CTR which is adequate for fusion applications. The fission spectrum of the fissionable material thorium has been collapsed from the fission matrices using the Bondarenko weighting scheme. The correct geometry of the LOTUS blanket and the cell specifications were correctly considered in the interpolation scheme. Some reaction cross sections for dosimetry applications have been included into the library. These base on the more recent ENDF/B-V evaluation. Transport and response edit cross sections have been coupled in the usual way to form P0 - P3 card image tables. Furthermore they have been converted into a binary file suitable to our RSYST computational system. Moreover the cross section card image tables have been reformatted and fitted into a BXSLIB binary library for the LANL-ONEDANT transport module. (Auth.)
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
ZZ COV-15GROUP-2006, 15-group cross section covariance matrix library
Description: ZZ-COV-15GROUP is a 15-group cross section covariance matrix library presenting a general overview of the presently available data. Number of groups: 15 neutron. Nuclides: H-1, Li-6, Li-7, Be-9, B-10, C-12, N-14, O-16, F-19, Na-23, Al-27, Si-28, Si-nat, Cr-52, Mn-55, Fe-56, Fe-57, Ni-58, Zr-90, Pb-nat, Pb-206, Pb-207, Pb-208, Th-232, U-235, U-238, Np-237, Pu-239, Pu-240, Pu-241, Am-241. Origin: ENDF/B-V, /B-VI.8, JENDL-3.3, JEFF-3.0, IRDF-2002 and IAEA Version 02 differs from version 01 in the following features: The input files (original BOXER format covariance libraries and ANGELO inputs instructions) have been included thus allowing to convert the covariance matrices to a user-defined energy group structure. Examples of output for the 15 group structure are provided. The code LAMBDA for verification of mathematical properties of the matrices (e. g. eigenvalues) is also included. This verification is highly recommended before using any covariance matrices. Version 03 differs from version 02 in the following features: The library file covfils2.lib was corrected (energy group structure was provided only for one isotope), as well as the corresponding test case outputs
New Standard Evaluated Neutron Cross Section Libraries for the GEANT4 Code and First Verification
Mendoza, Emilio; Koi, Tatsumi; Guerrero, Carlos
2014-01-01
The Monte Carlo simulation of the interaction of neutrons with matter relies on evaluated nuclear data libraries and models. The evaluated libraries are compilations of measured physical parameters (such as cross sections) combined with predictions of nuclear model calculations which have been adjusted to reproduce the experimental data. The results obtained from the simulations depend largely on the accuracy of the underlying nuclear data used, and thus it is important to have access to the nuclear data libraries available, either of general use or compiled for specific applications, and to perform exhaustive validations which cover the wide scope of application of the simulation code. In this paper we describe the work performed in order to extend the capabilities of the GEANT4 toolkit for the simulation of the interaction of neutrons with matter at neutron energies up to 20 MeV and a first verification of the results obtained. Such a work is of relevance for applications as diverse as the simulation of a n...
ZZ RFL-2-DTF, Group Constant Library of Reaction Cross-Section, Gas Production, Kerma, DPA
1 - Description of program or function: Format: DTF format and the structure is adopted from the MACKLIB-IV library. Number of groups: group library of reaction cross sections, gas production, kerma and DPA. Materials: H-1, H-2, H-3, He-3, He-4, Li-6, Li-7, Be-9, B-10, B-11, C-Nat, N-14, N-15, O-16, F-19, Na-23, Mg-Nat, Al-27, Si-28, P-31, S-Nat, Cl-Nat, Ar-36, Ar-38, Ar-40, K-Nat, Ca-Nat, Ti-Nat, V-Nat, Cr-50, Cr-52, Cr-53, Cr-54, Mn-55, Fe-54, Fe-56, Fe-57, Fe-58, Co-59, Ni-58, Ni-60, Ni-61, Ni-62, Ni-64, Cu-63, Cu-65, Zr-Nat, Nb-93, Mo-92, Mo-94, Mo-95, Mo-96, Mo-97, Mo-98, Mo-100, Ag-107, Ag-109, Cd-Nat, In-113, In-115, Sn-112, Sn-114, Sn-115, Sn-116, Sn-117, Sn-118, Sn-119, Sn-120, Sn-12,2 Sn-124, Ba-130, Ba-134, Ba-135, Ba-136, Ba-137, Ba-138, Hf-174, Hf-176, Hf-177, Hf- 178, Hf-179, Hf-180, Ta-181, W-Nat, W-182, W-183, W-184, W-186, Re-185, Re-187, Pb-Nat, Bi-209. Temperatures: T=293.6 K. Origin: GEFF-2 and GEPDL. RFL-2 is a group library of reaction cross sections, gas production, kerma and DPA based upon GEFF-2 and GEPDL - which are included in the package ZZ-GEFF-2-GENDF - and upon DECNET - included in the ZZ-DECNET-GENDF package (see below the description of these libraries). RFL-2 has been derived from them by the GENTORFL code (GENdf To RFL). Its primary use is to complete the neutron transport libraries in ANISN or FIDO format with data normally not present in the traditional files. It includes all GEFF-2 materials at T=293.6 K and σ0 = infinity; as qualifying point it gives 'delayed' kerma and 'delayed' gamma-ray production matrices, i.e. the energy release and the photons, respectively, generated by the decay of radioactive nuclei produced in the primary reactions; decay events that occur within 10000 seconds from the primary reaction are taken into account. The library includes many isotopes, since for each natural element included in GEFF-2 the decay of all component isotopes have been traced out. The library is in DTF format and the structure is
A 75 group neutron-photon coupled cross-section library has been developed for 42 reactor nuclides utilizing the basic cross-section files - ENDF/B-IV for neutrons and DLC-7F for photons. 50 neutron energy groups and gamma energy groups are included in this library which should be well suited to carry out safety, shielding and core physics studies of nuclear reactors based on fission or fusion processes. This library is also adequate for oil logging and mineral exploration investigations. (author). 11 refs., 3 tabs
WLUP3.0, 69 and 172 Group Cross Section Libraries for WIMS
Description or function: WLUP contains validated WIMS-D formatted cross section libraries in 69 and 172 energy group structures for nuclear reactor calculations. Materials from recently released evaluated nuclear data libraries are included. The NJOY nuclear data processing system was applied for generating the cross section files following the models and conventions built into the WIMS-D lattice code. The relevant features for the WIMS users are: - Energy group structures: 69 and 172 energy groups. - List of materials: WIMS ID, general information, source of data. - Cross sections: 69 and 172 group plots. - Resonance data: WIMS ID, temperature, background cross sections. - Goldstein-Cohen factors: Goldstein-Cohen lambda values. - Thermal scattering data: thermal scattering laws and P1 matrixes. - Fission spectrum: fission spectrum data. - Burnup data: burnup chains. - Fission product yields: fission yield tables. - Pseudo lumped fission product: Description of pseudo fission product. - Energy release by fission: table of energy released by fission. - Dosimetry data: dosimetry reactions, source of data. - Averaging flux and current spectra: flux and current spectra plots (Numerical data on NJOY inputs). - WIMSD5B updates: WIMSD5B extensions and updates. - Processing methods: Brief description on processing methods. Moderators: 1-H-H2O, 1-H-ZrH, 1-D-D2O, 4-Be, 6-C, 8-O-16. Structural materials: 2-He-3, 2-He-4, 3-Li-6, 3-Li-7, 5-B-10, 5-B, 7-N, 9-F, 11-Na, 12-Mg, 13-Al, 14-Si, 15-P, 16-S, 17-Cl, 20-Ca, 22-Ti, 23-V, 24-Cr, 25-Mn, 26-Fe, 27-Co-59, 28-Ni, 29-Cu, 40-Zr, 41-Nb-93, 42-Mo, 47-Ag, 48-Cd, 49-In, 50-Sn, 51-Sb-121, 51-Sb-123, 63-Eu, 72-Hf, 73-Ta, 74-W, 82-Pb. Burnable materials: 5-B-10, 5-B-11, 72-Hf-176, 72-Hf-177, 72-Hf-178, 72-Hf-179, 72-Hf-180. Fission products: 36-Kr-83, 42-Mo-95, 43-Tc-99, 44-Ru-101, 44-Ru-103, 44-Ru-106, 45-Rh-103, 45-Rh-105, 46-Pd-105, 46-Pd-107, 46-Pd-108, 47-Ag-109, 48-Cd-113, 49-In-115, 51-Sb-125, 52-Te-127, 53-I-127, 53-I-135, 54-Xe
ZZ MCJEF22NEA.BOLIB, MCNP Cross Section Library Based on JEF-2.2
1 - Description or function: Continuous energy cross-section data library for the Monte Carlo program MCNP based on the JEF-2.2 evaluated nuclear data library (ACE Format). Format: ACE Number of groups: Continuous energy Nuclides (107): H-1, H-2, He-4, Li-6, Li-7, Be-9, B-10, B-11, C-nat, N-14, N-15, O-16, O-17, F-19, Na-23, Mg-nat, Al-27, Si-nat, Cl-nat, Ti-nat, Cr-50, Cr-52, Cr-53, Cr-54, Mn-55, Fe-54, Fe-56, Fe-57, Fe-58, Co-59, Ni-58, Ni-60, Ni-61, Ni-62, Ni-64, Zr-90, Zr-91, Zr-92, Zr-94, Zr-96, Zr-nat, Nb-93, Mo-92, Mo-94, Mo-95, Mo-96, Mo-97, Mo-98, Mo-100, Mo-nat, Tc-99, Ru-101, Ru-102, Ru-104, Rh-103, Pd-105, Pd-107, Ag-109, I-129, Xe-131, Cs-133, Pr-141, Nd-143, Nd-145, Pm-147, Sm-147, Sm-149, Sm-150, Sm-151, Sm-152, Eu-153, Gd-154, Gd-155, Gd-156, Gd-157, Gd-158, Gd-160, Hf-174, Hf-176, Hf-177, Hf-178, Hf-179, Hf-180, Pb-nat, Bi-209, Th-232, U-234, U-235, U-236, U-238, Np-237, Pu-238, Pu-239, Pu-239bis, Pu-240, Pu-241, Pu-242, Am-241, Am-242, Am-242m, Am-243, Cm-242, Cm-243, Cm-244, Cm-245, Cm-246, Cm-247, Cm-248. Temperatures: 300 deg. K, 500 deg. K, 560 deg. K, 760 deg. K, 800 deg. K, 1000 deg. K, 1500 deg. K and 2200 deg. K. Thermal scattering (for diverse Temperatures); H in CH2 (polyethylene), H in H2O (light water), D in D2O (heavy water), C (graphite), Be (beryllium metal). Dosimetry cross-section: 16-S-32, 48-Cd-0, 79-Au-197; Origin: JEF-2.2, IRDF-90 V2. 2 - Methods: This library was generated with the NJOY-94.66 nuclear data processing system
1 - Description of program or function: specified on ORNL-RSIC-25, shielding benchmark problems. - BP-3 (Neutron cross sections): Format: ANISN, DOT and MORSE; Number of groups: 22 neutron / 18 gamma-ray; Nuclides: air; Origin: ENDF/B; Weighting spectrum: 1/E; - BP-6 (neutron and gamma-ray cross sections): Format: ANISN, DOT and MORSE; Number of groups: 22 neutron / 18 gamma-ray; Nuclides: Borated Polyethylene (C-12, H, and B-10); Origin: ENDF/B-II. The cross section data can be used to repeat the Shielding Benchmark Problems 3.0 and 6.0 for testing against the results published in ORNL-RSIC-25. 2 - Method of solution: ZZ-BP-3 neutron cross sections from the CCC-17/05R library were processed into 104 neutron groups using the PSR-9/CSP code. The fine-group neutron cross sections were collapsed to 22 broad groups using CCC-254/ANISN with an equilibrium fission spectrum source. The resulting multigroup cross sections are P5 coefficients punched on cards in format suitable for input to ANISN, DOT, and MORSE. ZZ-BP-6 neutron and gamma-ray cross sections for 12C, H, and 10B were from ENDF/B-II data. The neutron multigroup cross sections were generated into 104 neutron groups using the PSR-13/SUPERTOG code. The fine-group neutron cross sections were collapsed to 22 broad groups using CCC-254/ANISN with an equilibrium fission spectrum source. The gamma-ray multigroup cross sections were generated using PSR-7/MUG. The neutron-gamma-ray coupling utilized yield data from the DLC-12/POPOP4 library (data sets 010101, 060101, 060301, and 05100201). The neutron-gamma-ray coupled multigroup cross-section set was generated using the SAMPLE COUPLING CODE (ASCC). The multigroup cross sections are in a 22-18 group structure with P3 coefficients punched on cards in format suitable for input to ANISN, DOT, and MORSE
ZZ CANDULIB-AECL, Burnup-Dependent ORIGEN-S Cross-Section Libraries for Candu Reactor Fuels
1 - Historical background and information: - 28-element fuel cross-section library: Format: Designed for use with the ORIGEN-S isotope generation and depletion code. Materials: Co, Ge, As, Se, Br, Kr, Rb, Sr, Y, Zr, Nb, Mo, Tc, Ru, Rh, Pd, Ag, Cd, In, Sn, Sb, Te, I, Xe, Cs, Ba, La, Ce, Pr, Nd, Pm, Sm, Eu, Gd, Tb, Dy, Ho, Er, Lu, Ta, W, Re, Au, Th, Pa, U, Np, Pu, Am, Cm. Origin: ENDSF, ENDF/B-IV, -V and -VI Weighting spectrum: determined using WIMS-AECL transport code. - 37-element fuel cross-section library: Format: Designed for use with the ORIGEN-S isotope generation and depletion code. Materials: Co, Ge, As, Se, Br, Kr, Rb, Sr, Y, Zr, Nb, Mo, Tc, Ru, Rh, Pd, Ag, Cd, In, Sn, Sb, Te, I, Xe, Cs, Ba, La, Ce, Pr, Nd, Pm, Sm, Eu, Gd, Tb, Dy, Ho, Er, Lu, Ta, W, Re, Au, Th, Pa, U, Np, Pu, Am, Cm. Origin: ENDSF, ENDF/B-IV, -V and -VI Weighting spectrum: determined using WIMS-AECL transport code. In 1995, updated ORIGEN-S cross-section libraries were created as part of a program to upgrade and standardize the computer codes and nuclear data employed for used fuel characterization. This effort was funded through collaboration between Atomic Energy of Canada Limited and the Canadian Nuclear Power Utilities, under the Candu Owners Group (COG). The updated cross sections were generated using the WIMS-AECL lattice code and ENDF/B-V and -VI based data to provide cross section consistency with reactor physics codes. 2 - Application of the data: The libraries in this data collection are designed for characterising used fuel from Candu pressurized heavy water reactors. Two libraries are provided: one for the standard 28-element fuel bundle design, the other for the 37-element fuel bundle design. The libraries were generated for typical reactor operating conditions. The libraries are designed for use with the ORIGEN-S isotope generation and depletion code. 3 - Source and scope of data: The Candu libraries are updated with cross sections from a variety of different sources. Capture
Leal, L.C.; Hermann, O.W.; Bowman, S.M.; Parks, C.V.
1998-04-01
In this report, a methodology is described which serves as an alternative to the SAS2H path of the SCALE system to generate cross sections for point-depletion calculations with the ORIGEN-S code. ARP, Automatic Rapid Processing, is an algorithm that allows the generation of cross-section libraries suitable to the ORIGEN-S code by interpolation over pregenerated SAS2H libraries. The interpolations are carried out on the following variables: burnup, enrichment, and water density. The adequacy of the methodology is evaluated by comparing measured and computed spent fuel isotopic compositions for PWR and BWR systems.
A methodology is described that serves as an alternative to the SAS2H path of the SCALE system to generate cross sections for point-depletion calculations with the ORIGEN-S code. Automatic Rapid Processing (ARP) is an algorithm that allows the generation of cross-section libraries suitable to the ORIGEN-S code by interpolation over pregenerated SAS2H libraries. The interpolations are carried out on the following variables: burnup, enrichment, and water density. The adequacy of the methodology is evaluated by comparing measured and computed spent-fuel isotopic compositions for pressurized water reactor and boiling water reactor systems
Accuracy of thorium cross section of JENDL-4.0 library in thorium based fuel core evaluation
Highlights: ► Critical experiments on Th core were conducted to verify the accuracy of Th232 cross section of JENDL-4.0 library. ► Calculations are found to overestimate effective multiplication factor about (0.90 ± 0.01–0.99 ± 0.01)%. ► Comparison between measured and calculated Th sample worth reassures Th232 capture underestimation of JENDL-4.0 library. ► Th capture cross section is needed to be adjusted at thermal energy range to provide more reliable evaluation. - Abstract: Considering the importance of thorium data and concerning about the accuracy of Th232 cross section library, a series of experiments on thorium critical core with different neutron spectra has been implemented at Kyoto University Critical Assembly (KUCA). Reactivity worth of control rod and thorium sample was measured after the cores experimentally achieved critical state. In order to verify the accuracy of thorium cross section library, calculations of effective multiplication factor, control rod worth, reactivity worth of Th plates for the same core configurations were done by MVP code (Nagaya et al., 2005) using JENDL-4.0 library (Shibata et al., 2011). From the comparison between the measured and calculated results, the calculations are found to overestimate effective multiplication factor about (0.90 ± 0.01–0.99 ± 0.01)%. By comparing the measured Th sample worth with the calculated one, Th capture underestimation is reassured. Sensitive study on reactivity worth evaluation was conducted and it suggests that Th capture cross section is needed to be adjusted at thermal energy range to provide more reliable evaluation for thorium based fuel core design and safety calculation
Preparation of lumped fission product (FP) cross sections for a multigroup library
A method for the calculation of lumped Fission Product (FP) cross sections has been developed. The group constants fo each nuclide are generated by NJOY code, based on ENDF/B-V data. In this first version, cross section of 28 nuclides are lumped for typical characteristics of Binary Breeder Reactor (BBR). One energy group calculations are made for a 1000 MWe fast reactor to verify the influence of burnup, number of FP and fuel composition on the lumped fission product cross sections. (Author)
Effect of three cross-section libraries on the calculated neutron flux in the cavity of a PWR
The objective of this study was to compare calculations of pressure vessel surveillance dosimetry foil reaction rates computed using the ENDF/B-VI cross-section libraries for all reactor core and in-vessel materials except the reactor pressure vessel for which the ENDF/B-V, ENDF/B-VI, and LANL T-2 iron cross sections were substituted. Reaction rates for dosimetry foils in the cavity surrounding the pressure vessel of a pressurized water reactor were determined using the MCNP4A code. These calculations were compared to measured reaction rates from dosimetry foil experiments conducted during cycle 10 of Arkansas Nuclear One unit 1 (ANO-1)
Tardelli, T.C.; Stecher, L.C.; Coelho, T.S.; Castro, V.A. De; Cavalieri, T.A.; Menzel, F.; Giarola, R.S.; Domingos, D.B.; Yoriyaz, H., E-mail: tiago.tardelli@gmail.com [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear
2013-08-15
Absorbed dose calculations depend on a consistent set of nuclear data used in simulations in computer codes. Nuclear data are stored in libraries, however, the information available about the differences in dose caused by different libraries are rare. The libraries are processed by a computer system to be able to be used by a radiation transport code. One of the systems capable of processing nuclear data is the NJOY system. The objective of this study is to evaluate the nuclear data libraries for neutrons available in the literature, and to quantify the differences in absorbed dose obtained using the libraries JENDL 4.0, JEFF 3.3.1 and ENDF/B.VII. The absorbed dose calculation was performed on a simple geometric model, as spheres, and in anthropomorphic model of the human body based on the ICRP-110 for neutron transport simulation using the MCNP5 code. The results were compared with literature data. The results obtained with cross sections from the libraries JEFF and ENDF/B.VII have shown to be identical in most cases, except for one case where the difference has exceeded 10%. The results obtained with JENDL library has shown to be considerably different in most cases comparing to other two libraries. Some differences were over 200%. The dose calculations showed differences between the libraries, which is justified by differences in the cross sections. It has been observed that the cross sections values of certain nuclides assume quite different values in different libraries. These differences in turn cause considerable differences in dose calculations. (author)
As a result of the IAEA Co-ordinated Research Programme entitled 'Final Stage of the WIMS Library Update Project', new and updated WIMS-D libraries based upon ENDF/B-VI.5, JENDL-3.2 and JEF-2.2 have become available. A project to prepare an exhaustive handbook of WIMS-D cross sections from old and new libraries has been taken up by the authors. As part of this project, we have developed a computer program XnWlup with user-friendly graphical interface to help the users of WIMS-D library to enable quick visualization of the plots of the energy dependence of the multigroup cross sections of any nuclide of interest. This software enables the user to generate and view the histogram of 69 multi-group cross sections as a function of neutron energy under Microsoft Windows environment. This software is designed using Microsoft Visual C++ and Microsoft Foundation Classes Library. The current features of the software, on-line help manual and future plans for further development are described in this paper
ZZ SIGMNA-A, Photon Interaction and Absorption Cross-Section Library
1 - Description of program or function: - Format: special format; - Number of groups: Photon interaction and absorption coefficients covering the energy range 1 KeV to 100 MeV. - Nuclides: Materials: A150TE PLAST (H, C, N, O, F, Ca); Ac; Air (N, O, Ar); Sb; Ar; As; At; Bakelite (C, H, O); Ba; BARSO4; Be; Bk; Bi; Bone (H, C, N, O, Mg, P, S, Ca); B; Br; C552SHONKA P (H, C, O, F, Si); Cd; Ca; Cf; CAPINTEC (H, C, O, F, Si); C; Ce; Cs; Cl; Cr; Co; Concrete (H, O, Na, Mg, Al, Si, S, K, Ca, Fe); Cu; Cm; Delrin (C, H, O); Dy; Er; Eu; Fat (H, C, N, O, S); F; Fr; FRICK8 (H, O, Na, S, Cl, Fe); Gd; Ga; Ge; Au; Hf; He; Ho; H; ICRP Cortical bone (H, C, N, O, Mg, P, S, Ca, Zn); ICRP Tissue (H, C, N, O, S, Mg, P, S, Cl, K, Ca, Fe, Zn); ICRU Tissue (H, C, N, O); In; I; Ir; Fe; Kr; Pb; LIFTLD (Li, F); Li; Lucite (C, H, O); Lu; Mg; Mn; Hg; Mo; Muscle (H, C, N, O, S, Mg, P, S, K, Ca); Nd; Ne; Np; Ni; Nb; N; Nylon (H, C, N, O); O; Pd; P; Pt; Pu; Po; Polyethylene (C, H); Polystyrene (C, H); K; Pr; Pm; Pa; Ra; Re; Rh; Rb; Ru; Sm; Sc; Se; Si; Ag; Sodium-iodide; Na; SOLWA1; SOLWA2; Sr; S; Ta; Te; Tb; Tl; Th; Tm; Sn; Ti; W; U; V; Water (H, O); Xe; Yb; Y; Zn; Zr. - Origin: Howerton, JRC. An extensive library of photon interaction coefficients has been developed by the Ontario Cancer Institute, Toronto, Ontario, Canada, based on the compilation of Plechaty, Cullen, and Howerton. In addition to partial cross section data, the following are given: mass attenuation coefficients, mass energy transfer coefficients, mass energy absorption coefficients, average energy transferred to electrons, average energy absorbed per interaction, and average stopping power of electrons. Partial interaction coefficients and absorption coefficients are useful in any radiation transport or other radiation analysis application. The data from the Ontario Cancer Institute are given for 94 elements and 25 composite materials covering the energy range 1 KeV to 100 MeV. The reactions considered are coherent and
VELM61 and VELM22: Multigroup cross-section libraries for sodium-cooled reactor shield analysis
Two coupled neutron and photon multigroup cross-section libraries, derived from ENDF/B-V nuclear data, are described. The energy group structures, 61n/23γ and 22n/10γ, are subsets of the Vitamin-E 174n/38γ group structure, and are tailored to the iron and sodium resonances, windows, and capture gamma-ray spectra. Each of the two libraries are available in two formats, the AMPX master format and the ANISN format. Cross sections for all materials in the Vitamin-E library were collapsed using a standard energy weighting function, and in addition, several cross-section sets for each of the major constituents of commercial grade sodium, stainless steel (types 304 and 316), and carbon steel were derived using several problem-dependent weighting functions for averaging the fine groups. Effects of various group structures and weighting functions on the accuracy of the broad group libraries are studied by ANISN analysis of a typical sodium-iron shield configuration
Evaluated cross-section libraries and kerma factors for neutrons up to 100 MeV on 12C
A program is being carried out at Lawrence Livermore National Laboratory to develop high-energy evaluated nuclear data libraries for use in Monte Carlo simulations of cancer radiation therapy. In this report we describe evaluated cross sections and kerma factors for neutrons with incident energies up to 100 MeV on 12C. The aim of this effort is to incorporate advanced nuclear physics modeling methods, with new experimental measurements, to generate cross section libraries needed for an accurate simulation of dose deposition in fast neutron therapy. The evaluated libraries are based mainly on nuclear model calculations, benchmarked to experimental measurements where they exist. We use the GNASH code system, which includes Hauser-Feshbach, preequilibrium, and direct reaction mechanisms. The libraries tabulate elastic and nonelastic cross sections, angle-energy correlated production spectra for light ejectiles with A≤and kinetic energies given to light ejectiles and heavy recoil fragments. The major steps involved in this effort are: (1) development and validation of nuclear models for incident energies up to 100 MeV; (2) collation of experimental measurements, including new results from Louvain-la-Nueve and Los Alamos; (3) extension of the Livermore ENDL formats for representing high-energy data; (4) calculation and evaluation of nuclear data; and (5) validation of the libraries. We describe the evaluations in detail, with particular emphasis on our new high-energy modeling developments. Our evaluations agree well with experimental measurements of integrated and differential cross sections. We compare our results with the recent ENDF/B-VI evaluation which extends up to 32 MeV
ZZ DLC-13B, Resonance Cross-Section Group Constant Library for Tungsten and Depleted Pu
Nature of physical problem solved: Format: GAM-II; Number of groups: 32-energy-group split (0.4 to 1234 eV). Nuclides: tungsten (W,) and depleted uranium (U,) slabs. Multigroup capture and scatter cross sections in the resolved resonance region were calculated for tungsten and depleted uranium slabs for use in shielding calculations of neutron transport and capture distributions. Slabs of thickness of 1 to 8 centimeters surrounded by hydrogen or lithium hydride were considered. GAROL was used to generate the cross sections, a method previously observed to preserve the total capture rate in a detailed multigroup neutron transport calculation for a thick resonance absorber. Average cross sections were calculated for a 32-energy-group split (0.4 to 1234 eV) compatible with that used by GAM-2. Group fluxes are also presented permitting further group collapsing either by hand calculations or with an included computer program
ORIGEN-ARP Cross-Section Libraries for Magnox, Advanced Gas-Cooled, and VVER Reactor Designs
Murphy, BD
2004-03-10
Cross-section libraries for the ORIGEN-ARP system were extended to include four non-U.S. reactor types: the Magnox reactor, the Advanced Gas-Cooled Reactor, the VVER-440, and the VVER-1000. Typical design and operational parameters for these four reactor types were determined by an examination of a variety of published information sources. Burnup simulation models of the reactors were then developed using the SAS2H sequence from the Oak Ridge National Laboratory SCALE code system. In turn, these models were used to prepare the burnup-dependent cross-section libraries suitable for use with ORIGEN-ARP. The reactor designs together with the development of the SAS2H models are described, and a small number of validation results using spent-fuel assay data are reported.
A 35 group cross-section set with P3-anisotropic scattering matrices and resonance self-shielding factors has been generated from the basic ENDF/B-IV cross-section Library for 57 reactor elements. This library, called BARC35, is considered to be well suited for the neutronics and safety analysis of fission, fusion and hybrid systems. (author)
The ADS (accelerator driven sub-critical system) has great energy spans, complex energy spectrum structures and strong physical effects. Hence, the existing nuclear data libraries can't fully meet the needs of nuclear analysis in ADS. In order to do nuclear analysis for ADS system, a point-wise data library HENDL-ADS/MC (hybrid evaluated nuclear data library) was produced by FDS team. Meanwhile, to test the availability and reliability of the HENDL-ADS/MC data library, a series of shielding and critical safety benchmarks were performed. To validate and qualify the reliability of the high-energy cross section for HENDL-ADS/MC library further, a series of high neutronics integral experiments have been performed. The testing results confirm the accuracy and reliability of HENDL-ADS/MC. (authors)
ZZ BARC-27GRP, 27-Group Infinitely Dilute and Bondarenko Cross-Section Library from ENDF/B
1 - Description of problem or function: - BARC-27GRP: Format: 1-DX; Number of groups: 27; Nuclides: U-235, U-238, Pu-239, Pu-240, Pu-241, C, O, H, Al, Si, Na, Mg, Cr, Fe, Ni, Mo; Origin: ENDF/B-IV; Weighting spectrum: flux weighting proportional to 1/ΣT(u); fission weighting plus 1/E spectrum. - BARC-35-A: Format: SPHINX, Fx2-TH; Number of groups: 35; Nuclides: Al, He, Si, H, Fe, O, C, Na, Li, B, Be, N, Ca, Mn, V, Mo, Pb, Pu, Gd, K, Sm, Dy, Lu, Nb, U, Cr, Ni, Th, Np, Am, Zr, Cd, Eu, Mg, Ta, Cm, F, Ti, W. Origin: ENDF/B-IV; Weighting spectrum: fission - 1/E - thermal Maxwellian. - IAEA0856/01: 27-group resonance self-shielding factors and infinite diluted Cross sections for U-235, U-238, Pu-239, Pu-240, Pu-241, C, O, H, Al, Si, Na, Mg, Cr, Fe, Ni, Mo, generated by using the basic cross section and resonance parameter data from the ENDF/B-4 library. 2 - Method of solution: The 27-group constants were obtained by integrating the microscopic data over group intervals using a flux weighting proportional to 1/ΣT(u) and a fission plus 1/E spectrum. The standard ABBN group structure is used. The self-shielding factors were calculated for the following temperatures: 300, 900, 2100 (degrees Kelvin) and for potential scattering Cross sections of 10000, 100, 10, 1 barns. A thermal group is also included. For the 35-group library, resonance self-shielding factors are given at 300, 900, and 2100 K for a variety of dilution constants. Group Cross sections cover the energy range from 15 MeV to 0.005 eV and have been derived using Bondarenko flux approximation with a fission-1/E-thermal Maxwellian spectrum. The scattering Cross sections have been represented by a P3 Legendre expansion
ZZ DLC-16 COBB, 123 Neutron-Group Cross-Section Library from ENDF/B for XSDRN Calculation
1 - Nature of physical problem solved: Format: XSDRN; Number of groups: 123; Nuclides: H, D, He, Be-9, B-10, C-12, O-16, Na-23, Mg, Al-27, Ti, Cr, Mn-55, Fe, Ni, Cu, Cu-63, Cu-65, Nb-93, Mo, Xe-135, Sm-149, Eu-151, Eu-153, Gd, Dy-164, Lu-175, Lu-176, W-182, W-183, W-184, W-186, Re-185, Re-187, Au-197, U-233, U-234, U-235, U-236, U-238, Np-237, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, Am-243, and Cm-244. Origin: Mainly ENDF/B; Weighting spectrum: Fast cross sections → 1/E (14 MeV to .414 eV) Thermal cross sections → 1/E (1.86 eV to 0.125 eV) → Maxwell-Boltzmann (0.125 eV to 0.0047 eV). The library is intended to be a source of evaluated data for the cross section preparation code XSDRN. It supplements, rather than replaces, the existing XSDRN master library which is distributed with the code package. The library contains data for H, D, He, 9-Be, 10-B, 12-C, 16-O, 23-Na, Mg, 27-Al, Ti, Cr, 55-Mn, Fe, Ni, Cu, 63-Cu, 65-Cu, 93-Nb, Mo, 135-Xe, 149-Sm, 151-Eu, 153-Eu, Gd, 164-Dy, 175-Lu, 176-Lu, 182-W, 183-W, 184-W, 186-W, 185-Re, 187-Re, 197-Au, 233-U, 234-U, 235-U, 236-U, 238-U, 237-Np, 238-Pu, 239-Pu, 240-Pu, 241-Pu, 242-Pu, 241-Am, 243-Am, and 244-Cm. 2 - Method of solution: The library contains ENDF/B version 2 cross sections processed through several steps (primarily by SUPERTOG) into the standard XSDRN 123-group energy structure. These steps are - (a) process fast cross sections with SUPERTOG into standard GAM-2 energy structure (14 MeV to 0.414 eV), using a 1/E weighting function, and produce a GAM-2 tape. (This step was performed by R. Q. Wright, Math Div., ORNL). (b) Process thermal cross sections with SUPERTOG into standard 30-group THERMOS energy group structure (1.86 eV to 0.0047 eV), using a Maxwell-Boltzmann distribution with temperature 293 deg.K as a weighting function for E < 0.125 eV coupled to a 1/E weighting function for E from 0.125 eV to 1.86 eV. (c) Compute room temperature free-gas kernels, using THERMOS tape-making program, and
1 - Description of program or function: - Format: ANISN; - Number of groups: 34 neutron groups - 14 gamma groups; - Nuclides: (2)H, D, (3)O, Li6, Li7, B10, B11, C, Al, Si, Ti, V, Cr, Mn-55, Fe, Ni, Cu, Nb, Mo, W, Pb, (4)SS. - Origin: ENDF/B (DLC-0037); Weighting spectrum: 1/E weighted for neutron energies exceeding 0.345 eV, below this energy a Maxwellian distribution peaked at 800 K is used. The photon interaction cross sections are flat weighted. A P3 48-group coupled neutron and gamma-ray (34 neutron groups - 14 gamma groups) cross section library for neutronic studies in fusion reactor blankets or shield for the following 28 elements: (2)H, D, (3)O, Li6, Li7, B10, B11, C, Al, Si, Ti, V, Cr, Mn55, Fe, Ni, Cu, Nb, Mo, W, Pb, (4)SS. The cross section data are given in ANISN card image format. 4. Method of solution: The library has been produced by collapsing DLC-37, 100 neutron and 21 gamma groups to 34 neutron and 14 gamma groups. A rather fine mesh is maintained in the higher energy range where gamma production, activation and heat deposition are relatively more important. One of the files contains in the first position of the Po the kerma factor instead of absorption. Kerma factors were obtained from MACKLIB-IV
Nuclear data, cross section libraries and their application in nuclear technology
These proceedings contain the articles presented at the named seminar. The articles deal with evaluated nuclear data libraries, computer codes for neutron transport and reactor calculations using nuclear data libraries, and the application of nuclear data libraries for the calculation of the interaction of neutron beams with materials. (HSI)
ZZ WM-NRSM, Neutron and Gamma Group Cross-Section Library for Nuclear Rocket Shielding Calculations
Description of problem or function: - Master Library 1: Format: ANISN-W, DOT-IIW and APPROPOS. Number of groups: 52; Nuclides: Al, Be, B, B-10, Cd, C, Cr, Co, Cu, Gd, Au, H, In-115, Fe, Pb, Li, Li-6, Li-7, Mg, Mn, Mo, Ni, Nb, N, O, Si, Ta, Ti, W, U-235, U-238, Zr. Origin: Westinghouse Astro-nuclear Laboratory. Weighting spectrum: 1/E, flux and current spectra. - Master Library 2: Format: ANISN-W, DOT-IIW and APPROPOS. Number of groups: 52; Nuclides: Al, Be, B, B-10, Cd, C, Cr, Co, Cu, Gd, Au, H, In-115, Fe, Pb, Li, Li-6, Li-7, Mg, Mn, Mo, Ni, Nb, N, O, Si, Ta, Ti, W, U-235, U-238, Zr. Origin: Westinghouse Astro-nuclear Laboratory. Weighting spectrum: 1/E, flux and current spectra. - Master Library 3: Format: APPROPOS. Number of groups: 52; Nuclides: Al, Be, B, B-10, Cd, C, Cr, Co, Cu, Gd, Au, H, In-115, Fe, Pb, Li, Li-6, Li-7, Mg, Mn, Mo, Ni, Nb, N, O, Si, Ta, Ti, W, U-235, U-238, Zr. Origin: Westinghouse Astro-nuclear Laboratory. Weighting spectrum: 1/E, flux and current spectra. - Master Library 5: Format: KAP-VI, GAMLEG-W, MAC and SCAP. Number of groups: energy points in the range of 0.01 MeV to 20.0 MeV; Nuclides: H, He, Li, Be, B, C, N, O, Na, Mg, Al, Si, P, S, K, Ca, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zn, Y, Zr, Nb, Mo, Ag, Cd, In, Sn, Cs, Ba, Sm, Gd, Dy, Y, Hf, Ta, W, Au, Hg, Pb, Po, Th, Pa, U, Np, Pu. Origin: Westinghouse Astro-nuclear Laboratory. - Basic set of nuclear data (Library 6): Format: ANISN-W and DOT-IIW. Number of groups: 52; Nuclides: H, Be, B, C, U-235, U-238, N, O, Mg, Al, Si, Cr, Mn, Fe Co, Ni, Cu, Zr, Mo, Ag, In, Cd, Gd, Pb, Nb, Ti, Ta, Li-6, Li-7, B-10 W, S. Origin: Master Libraries. Weighting spectrum: decided by user. WANL-MSFC Nuclear Rocket Shielding Data Generators GAMLEG-W, APPROPOS, NAGS, SATURN and Neutron and photon Cross Section Libraries 1-6. Applications of the Data: Transport codes which use the data are ANISN-W, KAP-VI, DOT-IIW, MAC and SCAP. The transport codes, also available from RSIC, the cross section processing codes, and
Voloschenko Andrey
2016-01-01
Full Text Available The broad-group library BGL1000_B7 for neutron and gamma transport calculations in VVER-1000 internals, RPV and shielding was carried out on a base of fine-group library v7-200n47g from SCALE-6 system. The comparison of the library BGL1000_B7 with the library v7-200n47g and the library BGL1000 (the latter is using for VVER-1000 calculations is demonstrated on several calculation and experimental tests.
1 - Description of problem or function: Format: 'data base' for subsequent collapsing into both fine and broad group data in various formats (working and/or weighted ANISN, CCCC, etc.). Number of groups: AMPX-2/123 → 123 group structure; AMPX-2/219 → 219 group structure. Nuclides: H, He, Li, Be, B, C, N, O, F, Na, Mg, Al, Si, Cl, K, Ca, Ti, V, Cr, Mn, Fe, Ni, Cu, Kr, Zirc, Mo, Tc, Rh, Ag, Cd, Xe, Sm, Eu, Gd, Dy, Cu, Ta, W, Re, Pb, Th, Pa, U, Np, Pu, Am, Cm. Origin: ENDF/B-IV. Weighting spectrum: Most data were generated using a standard flux over three energy ranges (fission - 1/E - Maxwellian) as point-to-fine-group cross sections weighing function. The AMPX-2 P3 123- and 219- Group Neutron Cross-Section Master Interface Libraries may be considered as 'data bases' for subsequent collapsing into both fine and broad group data in various formats (working and/or weighted ANISN, CCCC, etc.). The built-in 123 and 219 group structures have been used to process all available data of ENDF/B-IV. 2 - Method of solution: The program AMPX-2 has been used to generate the data. By various executions of the module XLACS-2 (XLACS for bound H-1 in some materials) a number of independent libraries were generated which then were combined using the AMPX-2 module AJAX. Most data were generated using a standard flux over three energy ranges (fission - 1/E - Maxwellian) as point-to-fine-group cross sections weighing function. For some structural materials (e.g. Fe, Cr,...) different master data sets were produced using a weighting function fission - 1/E sigma T(SS-304) - Maxwellian, and the three parts of the spectrum were joined at properly selected energies. For some nuclides (e.g. 238U and 240Pu) various master data sets have been produced which contain problem-dependent unresolved cross sections characterized by the associated potential scattering cross sections. Some data sets contain P3 thermal scattering matrices, for which ENDF/B File 7 S(alpha, beta) data were used, e
A Validated MCNP(X) Cross Section Library based on JEFF 3.1
Haeck, W.; Verboomen, B.
2006-10-15
ALEPH-LIB is a multi-temperature neutron transport library for standard use by MCNP(X) and ALEPH generated with ALEPH-DLG. This is an auxiliary computer code to ALEPH, the Monte Carlo burn-up code under development at SCK-CEN in collaboration with Ghent university. ALEPH-DLG automates the entire process of generating library files with NJOY and takes care of the first requirement of a validated application library: verify the processing. It produces tailor made NJOY input files using data from the original ENDF file (initial temperature, the fact if the nuclide is fissile or if it has unresolved resonances, etc.) When the library files have been generated, ALEPH-DLG will also process the output from NJOY by extracting all messages and warnings. If ALEPH-DLG finds anything out of the ordinary, it will either warn the user or perform corrective actions. The temperatures included in the ALEPH-LIB library are 300, 600, 900, 1200, 1500 and 1800 K. Library files were produced for the JEF 2.2, JEFF 3.0, JEFF 3.1, JENDL 3.3 and ENDF/B-VI.8 nuclear data libraries. This will be extended with ENDF/B-VII when it becomes available. This report deals with the JEFF 3.1 files included in ALEPH-LIB that are now released by the NEA-OECD.
ZZ UKCTR-1, Cross-Section Library for Neutron Flux and Neutron Reaction Rates in CTR Calculation
1 - Description of problem or function: Format: ANISN, DOT, MORSE, SWANLAKE; Number of Groups: 46 energy group structure from 14.2 MeV to 1 MeV; Nuclides: Li-6, Li-7, O, Be, Pb, Nb, Fe, Ni, Cr, Zr, V, Ti, H, D, T, C, Al, B-10, B-11, Cu-63, Cu-65, F, Na, K, Mo. Origin: UKNDL; Weighting Spectrum: 1/(Sigma t (E).E) weighting is used for groups 1 to 44 with Maxwellian weighting for the two thermal groups. UKCTR1 is a data library of neutron cross sections for 25 materials in a 46 energy group structure from 14.2 MeV to 1 MeV. It is designed for calculation of neutron fluxes and reaction rates in controlled thermonuclear reactors. The energy group structure is fine at 14 MeV and there are two thermal groups; the lethargy interval width per energy group for decreasing energy is as follows: 0.014, 0.036, 2 x 0.15, 15 x 0.3, 25 x 0.5, 2.935 and 3.091. Reaction cross sections including partial inelastic data are provided for the following materials: Li-6, Li-7, O, Be, Pb, Nb, Fe, Ni, Cr, Zr, V, Ti, H, D, T, C, Al, B-10, B-11, Cu-63, Cu-65, F, Na, K, Mo. 1/(Sigma t (E).E) weighting is used for groups 1 to 44 with Maxwellian weighting for the two thermal groups. Anisotropy of scattering is represented by a P order up to 4 (usually 0 to 4). Data for hydrogen and deuterium both in water and heavy water and in the gaseous state is available. As a supplement, neutron kerma factors are included for each of the nuclides in the library as well as 98 activation cross sections of importance in fusion reactor work. (These 98 activation cross sections have been extracted from the bulk of the UKCTR-I library to be in a more convenient form for programs such as ANISN.) The kerma factors were computed using the code ENBAL2, a revised version of ENBAL, which calculates multigroup kerma factors directly from multigroup cross sections together with reaction Q-values. This approach allows neutron heating calculations to be performed consistently with the flux calculation. 2 - Method of
Development of the adjusted nuclear cross-section library based on JENDL-3.2 for large FBR
JNC (and PNC) had developed the adjusted nuclear cross-section library in which the results of the JUPITER experiments were reflected. Using this adjusted library, the distinct improvement of the accuracy in nuclear design of FBR cores had been achieved. As a recent research, JNC develops a database of other integral data in addition to the JUPITER experiments, aiming at further improvement for accuracy and reliability. In 1991, the adjusted library based on JENDL-2, JFS-3-J2 (ADJ91R), was developed, and it has been used on the design research for FBR. As an evaluated nuclear library, however, JENDL-3.2 is recently used. Therefore, the authors developed an adjusted library based on JENDL-3.2 which is called JFS-3-J3.2(ADJ98). It is known that the adjusted library based on JENDL-2 overestimated the sodium void reactivity worth by 10-20%. It is expected that the adjusted library based on JENDL-3.2 solve the problem. The adjusted library JFS-3-J3.2(ADJ98) was produced with the same method as the adjusted library JFS-3-J2(ADJ91R) and used more integral parameters of JUPITER experiments than the adjusted library JFS-3-J2(ADJ91R). This report also describes the design accuracy estimation on a 600 MWe class FBR with the adjusted library JFS-3-J3.2(ADJ98). Its main nuclear design parameters (multiplication factor, burn-up reactivity loss, breeding ratio, etc.) except the sodium void reactivity worth which are calculated with the adjusted library JFS-3-J3.2(ADJ98) are almost the same as those predicted with JFS-3-J2(ADJ91R). As for the sodium void reactivity, the adjusted library JFS-3-J3.2(ADJ98) estimates about 4% smaller than the JFS-3-J2(ADJ91R) because of the change of the basic nuclear library from JENDL-2 to JENDL-3.2. (author)
Kulesza Joel A.
2016-01-01
Full Text Available This paper describes a comparison of contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a boiling water reactor calculational benchmark problem. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission. The benchmark problem was originally evaluated by Brookhaven National Laboratory using the Oak Ridge National Laboratory discrete ordinates code DORT and the BUGLE-93 cross-section library. In this paper, the Westinghouse RAPTOR-M3G three-dimensional discrete ordinates code was used. A variety of cross-section libraries were used with RAPTOR-M3G including the BUGLE93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory and ALPAN-VII.0 developed at Westinghouse. In comparing the calculated fast reaction rates using the four aforementioned cross-section libraries in the pressure vessel capsule, for six dosimetry reaction rates, a maximum relative difference of 8% was observed. As such, it is concluded that the results calculated by RAPTOR-M3G are consistent with the benchmark and further that the different vintage BUGLE cross-section libraries investigated are largely self-consistent.
Kulesza, Joel A.; Arzu Alpan, F.
2016-02-01
This paper describes a comparison of contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a boiling water reactor calculational benchmark problem. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission. The benchmark problem was originally evaluated by Brookhaven National Laboratory using the Oak Ridge National Laboratory discrete ordinates code DORT and the BUGLE-93 cross-section library. In this paper, the Westinghouse RAPTOR-M3G three-dimensional discrete ordinates code was used. A variety of cross-section libraries were used with RAPTOR-M3G including the BUGLE93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory and ALPAN-VII.0 developed at Westinghouse. In comparing the calculated fast reaction rates using the four aforementioned cross-section libraries in the pressure vessel capsule, for six dosimetry reaction rates, a maximum relative difference of 8% was observed. As such, it is concluded that the results calculated by RAPTOR-M3G are consistent with the benchmark and further that the different vintage BUGLE cross-section libraries investigated are largely self-consistent.
ZZ-IRAN-LIB, Multigroup Neutron Gamma Cross-Section Library for 33 Elements in ANISN Format
Description of program or function: - Format: ANISN/PC; - Number of groups: IRAN1.LIB (22 neutrons 18 gammas); IRAN2.LIB (17 neutrons, 18 gammas); IRAN3.LIB (7 neutrons, 18 gammas); IRAN4.LIB (7 neutrons, 6 gammas); IRAN5.LIB (5 neutrons, 4 gammas); IRAN6.LIB (2 neutrons, 4 gammas). - Nuclides: H-1, H-2, Li-6, Li-7, Be-9, B-10, C-12, N-14, O-16, Na, Mg, Al-27, Si, K, V, Cr, Mn-55, Fe, Ni, Nb-93, Pb, U-235, U-238, Pu-239, Ba-134, Ba-135, Ba-136, Ba-137, Ba-140, Bi-209, Ca-nat, Zr-nat, Cd-nat. - Origin: VITAMIN-4C; ENDF/B-IV and V, and JENDL-3. Weighting spectrum: IRAN.LIB's data (microscopic cross sections) is suitable for neutron, gamma and coupled neutron- gamma transport calculation (shielding). It is intended for use by the multigroup discrete ordinates code ANISN/PC (CCC-0514) using anisotropic scattering by Legendre expansion up to order P-3. IRAN.LIB is a collection of libraries for elements (H-1; H-2; Li-6; Li-7; Be-9; B-10; C-12; N-14; O-16; Na; Mg; Al-27; Si; K; V; Cr; Mn-55; Fe; Ni; Nb-93; Pb; U-235; U-238; Pu-239; Ba-134; Ba-135; Ba-136; Ba-137; Ba-140; Bi-209; Ca-nat; Zr-nat; Cd-nat) in ISOTXS format with a different group structure for each library, that is, IRAN1.LIB (22 neutrons, 18 gammas); IRAN2.LIB (17 neutrons, 18 gammas); IRAN3.LIB (7 neutrons, 18 gammas); IRAN4.LIB (7 neutrons, 6 gammas); IRAN5.LIB (5 neutrons, 4 gammas); IRAN6.LIB (2 neutrons, 4 gammas). 2 - Method of solution: The basic data sources were VITAMIN-4C; ENDF/B-IV and V and JENDL-3. Most of the data were taken from VITAMIN-4C (H-1, H-2, Li-6, Li-7, Be-9, B-10, C-12, N-14, O-16, Na, Mg, Al-27, Si, K, V, Cr, Mn-55, Fe, Ni, Nb-93, Pb, U-235, U-238, Pu-239) and collapsing them using AMPX-II modules. The AJAX module extracts the neutron cross sections of desired elements from VITAMIN-4C. CHOX module combines master neutron, gamma production and gamma interaction libraries into a coupled neutron-gamma library. MALOCS module collapses the cross sections into given energy groups and
Benchmarking of the FENDL-3 Neutron Cross-Section Data Library for Fusion Applications
This report summarizes the benchmark analyses performed in a joint effort of ENEA (Italy), JAEA (Japan), KIT (Germany), and the University of Wisconsin (USA) with the objective to test and qualify the neutron induced general purpose FENDL-3.0 data library for fusion applications. The benchmark approach consisted of two major steps including the analysis of a simple ITER-like computational benchmark, and a series of analyses of benchmark experiments conducted previously at the 14 MeV neutron generator facilities at ENEA Frascati, Italy (FNG) and JAEA, Tokai-mura, Japan (FNS). The computational benchmark revealed a modest increase of the neutron flux levels in the deep penetration regions and a substantial increase of the gas production in steel components. The comparison to experimental results showed good agreement with no substantial differences between FENDL-3.0 and FENDL-2.1 for most of the responses analysed. There is a slight trend, however, for an increase of the fast neutron flux in the shielding experiment and a decrease in the breeder mock-up experiments. The photon flux spectra measured in the bulk shield and the tungsten experiments are significantly better reproduced with FENDL-3.0 data. In general, FENDL-3, as compared to FENDL-2.1, shows an improved performance for fusion neutronics applications. It is thus recommended to ITER to replace FENDL-2.1 as reference data library for neutronics calculation by FENDL-3.0. (author)
Verification of a Multi-group Cross Section Library for Burnup Calculation
Daing, Aung Tharn; Kim, Myung Hyun [Kyung Hee Univ., Yongin (Korea, Republic of); Joo, Hang Yu [Seoul National Univ., Seoul (Korea, Republic of)
2013-05-15
Despite satisfying the estimation of the neutronic parameters without depletion to some extent, it still requires detailed investigation of the behavior of a fuel with strong neutron absorber over its operating life time by nTRACER, the direct whole core calculation code with the conventional semi Predictor-Corrector method. This study is mainly focused on the verification of the newly generated multi-group library for burnup calculation by nTRACER through the analysis of its performance of depletion calculation of UO{sub 2} fuel with strong neutron absorbers such as Gadolinium. Firstly, the depletion calculation results of nTRACER are presented by comparing the evolution of k-inf and the inventories of commonly found important isotopes as a function of burnup in the cases of gadolinia(GAD)-bearing fuel pin and fuel assembly (FA) with those of MCNPX-version.2.6.0. The newly generated multi-group library for burnup calculation by nTRACER was verified through GAD-bearing fuel after the new approach of resonance treatment had been employed. Though very good agreement in the overall effect reflected on the multiplication factor of FA at BOC, the evolution of k-inf along fuel irradiation history was systematically well underestimated by nTRACER when compared to Monte Carlo results.
Quantitative and quality test of cross section library ENDF/B-b2
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)
The one-dimension SN method code ANISN and specific cross section library ZPR-22 have been used to perform the design calculation of dose rate distribution along the radial and axial direction of HWZPR shielding. Through multi-case calculations and optimization analysis works, a double slab cover structure is adopted. It is combined with the feasibility of structure and the possibility of boron concentration to be merged in paraffin for design case. The calculation results of axial direction: the core lattice distance is 18 cm; core radius R = 113 cm; reflector saving of radial direction is 25 cm; transfer leakage Dy = Dz = 244.6 cm. The calculation results of radial direction; the core lattice distance is 18 cm; critical water level 138.5 cm; reflector saving of axial direction is 20 cm; transfer leakage correction parameter Dy = 160 cm
The PSIMECX library contains calculated nuclide production cross-sections from neutron-induced reactions in the energy range about 2 to 800 MeV in the following 72 stable isotopes of 24 elements: 12C, 13C, 16O, 17O, 18O, 23Na, 24Mg, 25Mg, 26Mg, 27Al, 28Si, 29Si, 30Si, 31P, 32S, 33S, 34S, 36S, 35Cl, 37Cl, 39K, 40K, 41K, 40Ca, 42Ca, 43Ca, 44Ca, 46Ca, 48Ca, 46Ti, 47Ti, 48Ti, 49Ti, 50Ti, 50V, 51V, 50Cr, 52Cr, 53Cr, 54Cr, 55Mn, 54Fe, 56Fe, 57Fe, 58Fe, 58Ni, 60Ni, 61Ni, 62Ni, 64Ni, 63Cu, 65Cu, 64Zn, 66Zn, 67Zn, 68Zn, 70Zn, 92Mo, 94Mo, 95Mo, 96Mo, 97Mo, 98Mo, 100Mo, 121Sb, 123Sb, 204Pb, 206Pb, 207Pb, 208Pb, 232Th and 238U. The energy range covers essentially all transmutation channels other than capture. The majority of the selected elements are principal constituents of normal materials of construction used in and around accelerator facilities and the library is, first and foremost, designed to be a tool for the estimation of their activation in wide-band neutron fields. This first report, from a set of three, describes the form and usage of the library; the other two reports document the calculational methods. The present organisation of the library is the author's first idea and adequate for the intended use (activation calculations); being machine readable, translation of the library into other formats is straightforward. (author)
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
The original ANSL-V cross-section libraries (ORNL-6618) were developed over a period of several years for the physics analysis of the ANS reactor, with little thought toward including the materials commonly needed for shielding applications. Materials commonly used for shielding applications include calcium barium, sulfur, phosphorous, and bismuth. These materials, as well as 6Li, 7Li, and the naturally occurring isotopes of hafnium, have been added to the ANSL-V libraries. The gamma-ray production and gamma-ray interaction cross sections were completely regenerated for the ANSL-V 99n/44g library which did not exist previously. The MALOCS module was used to collapse the 99n/44g coupled library to the 39n/44g broad- group library. COMET was used to renormalize the two-dimensional (2- D) neutron matrix sums to agree with the one-dimensional (1-D) averaged values. The FRESH module was used to adjust the thermal scattering matrices on the 99n/44g and 39n/44g ANSL-V libraries. PERFUME was used to correct the original XLACS Legendre polynomial fits to produce acceptable distributions. The final ANSL-V 99n/44g and 39n/44g cross-section libraries were both checked by running RADE. The AIM module was used to convert the master cross-section libraries from binary coded decimal to binary format (or vice versa)
A paper was published in 1979 containing a compilation of experimental data on the cross-sections of (n,p), (n,α) and (n,2n) threshold reactions and recommended excitation functions. A further paper considered the development of evaluation methods based on the use of theoretical model calculations, an increase in the number of recommended excitation functions, correction of the recommended cross-sections on the basis of integral experiments and allowance for recent experimental data. To satisfy the wide circle of users, BOSPOR-80 - a machine library of evaluated threshold reaction cross-sections - was set up
Newly produced multigroup cross-section libraries require detailed testing to ensure that they are suitable for the applications intended. This requires that the libraries be tested against approved experimental benchmarks and/or well-posed calculational benchmarks. Following this tradition, the recently produced fine-group VITAMIN-B6 library and its derivative BUGLE-96 broad-group library have been tested against calculational and experimental benchmarks that are sensitive to neutrons with energies in the moderate-energy range (10.0 to 20.0 MeV). Iron is prominent in each benchmark as it is in many shielding configurations, and iron cross-section data have posed significant problems in many shielding designs. These benchmarks provide stringent tests for the iron cross sections. Calculated results obtained using the new libraries were compared to measured results or results from other calculations. In some cases, results were in good agreement. In other cases, there were significant discrepancies between results due to deficient measurements in a few comparisons and to method or data deficiencies in other comparisons. It is concluded that there is still need for further measurements and evaluations of the iron cross-section data in the energy region below 6.0 MeV. While fluxes in the moderate-energy range and the associated downscatter sources may be calculated adequately, the inadequate low-energy cross sections can lead to rather large discrepancies in integral quantities such as dose or heating
ZZ MCB63NEA.BOLIB, MCNP Cross Section Library Based on ENDF/B-VI Release 3
1 - Description of program or function: Continuous energy cross-section data library for the Monte Carlo program MCNP based on the ENDF/B-VI Release 3 evaluated nuclear data library (ACE Format). Format: ACE; Number of groups: Continuous energy; Nuclides (107): H-1, H-2, He-4, Li-6, Li-7, Be-9, B-10, B-11, C-nat, N-14, N-15, O-16, O-17, Na-23, Mg-nat, Al-27, Si-nat, Cl-nat, Ti-nat, Cr-50, Cr-52, Cr-53, Cr-54, Mn-55, Fe-54, Fe-56, Fe-57, Fe-58, Co-59, Ni-58, Ni-60, Ni-61, Ni-62, Ni-64, Zr-90, Zr-91, Zr-92, Zr-94, Zr-96, Zr-nat, Nb-93, Mo-94, Mo-95, Mo-96, Mo-97, Mo-nat, Tc-99, Ru-101, Ru-102, Ru-104, Rh-103, Pd-105, Pd-107, Ag-109, I-129, Xe-131, Cs-133, Pr-141, Nd-143, Nd-145, Pm-147, Sm-147, Sm-149, Sm-150, Sm-151, Sm-152, Eu-153, Gd-154, Gd-155, Gd-156, Gd-157, Gd-158, Gd-160, Hf-174, Hf-176, Hf-177, Hf-178, Hf-179, Hf-180, Hf-nat, Pb-206, Pb-207, Pb-208, Bi-209, Th-232,U-233, U-234, U-235, U-236, U-238, Np-237, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, Am-242, Am-242m, Am-243, Cm-242, Cm-243, Cm-244, Cm-245, Cm-246, Cm-247, Cm-248. Temperatures: 300 deg. K, 500 deg. K, 560 deg. K, 760 deg. K, 800 deg. K, 1000 deg. K, 1500 deg. K and 2200 deg. K. Thermal scattering (for diverse Temperatures): H in CH2 (polyethylene), H in H2O (light water), D in D2O (heavy water), C (graphite), Be (beryllium metal). Dosimetry cross-section: 16-S-32, 48-Cd-0, 79-Au-197. Origin: ENDF/B-VI Release 3, IRDF-90 Version 2. 2 - Methods: This library was generated with the NJOY-94.66 nuclear data processing system
ZZ DLC-2D/100G, 100 Neutron-Group Cross-Section Library by SUPERTOG Calculation for ANISN, DOT
1 - Nature of physical problem solved: Format: ANISN, DOT or DTF-4; Number of groups: 100; Nuclides: H, D, He, He-3, Li-6, Li-7, Be-9, B-10, B-11, C-12, N-14, O-16, Na-23, Mg, Al-27, Si, Cl, K, Ca, V, Cr, Mn-55, Fe, Co-59, Ni, Cu, Cu-63, Cu-65, Nb, Mo, Ag-107, Xe-135, Cs-133, Sm-149, Eu-151, Eu-153, Gd, Dy-164, Lu-175, Lu-176, Ta-181, Ta-182, W-182, W-183, W-184, W-186, Re-185, Re-187, Au-197, Pb, Th-232, Pa-233, U-234, U-235, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, Am-243, and Cm-244. Origin: The nuclides in DLC-2 are those which have been released as category I ENDF/B by the National Neutron Cross Section Center, Brookhaven National Laboratory. Weighting spectrum: The explicit assumption was made that the flux has the shape of a fission spectrum joined at 0.0674 MeV by a 1/E tail. Neutron transport calculations can be performed with DLC-2 data. Since the data are intended for use in multigroup discrete-ordinates or Monte Carlo transport codes which treat anisotropic scattering, possible cross section angular expansion is limited only by the options available in the particular code used. Specifically, the retrieval program manipulates DLC-2 such that it conforms to input requirements of the ANISN, DOT, or DTF-4 codes, or any computer code using data in the ANISN or DTF-4 format. The nuclides in DLC-2 are those which have been released as category I ENDF/B by the National Neutron Cross Section Center, Brookhaven National Laboratory. The library contains data for H, D, He, 3-He, 6-Li, 7-Li, 9-Be, 10-B, 11-B, 12-C, 14-N, 16-O, 23-Na, Mg, 27-Al, Si, Cl, K, Ca, V, Cr, 55-Mn, Fe, 59-Co, Ni, Cu, 63-Cu, 65-Cu, Nb, Mo, 107-Ag, 135-Xe, 133-Cs, 149-Sm, 151-Eu, 153-Eu, Gd, 164-Dy, 175-Lu, 176-Lu, 181-Ta, 182-Ta, 182-W, 183-W, 184-W, 186-W, 185-Re, 187-Re, 197-Au, Pb, 232-Th, 233-Pa, 234-U, 235-U, 238-U, 238-Pu, 239-Pu, 240-Pu, 241-Pu, 242-Pu, 241-Am, 243-Am, and 244-Cm. 2 - Method of solution: DLC-2 was generated by SUPERTOG from nuclear data in either point
Arzu Alpan, F.; Kulesza, Joel A.
2016-02-01
This paper compares contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a pressurized water reactor calculational benchmark problem with a standard out-in core loading. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission and used the Oak Ridge National Laboratory two-dimensional discrete ordinates code DORT and the BUGLE-93 cross-section library for the calculations. In this paper, a Westinghouse three-dimensional discrete ordinates code with parallel processing, the RAPTOR-M3G code was used. A variety of cross section libraries were used with RAPTOR-M3G including the BUGLE-93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory, and the broad-group ALPAN-VII.0 cross-section library developed at Westinghouse. In comparing the calculation-to-calculation reaction rates using the BUGLE-93 cross-section library at the thermal shield, pressure vessel, and cavity capsules, for eleven dosimetry reaction rates, a maximum relative difference of 5% was observed, with the exception of 65Cu(n,2n) in the pressure vessel capsule that had a 90% relative difference with respect to the reference results. It is thought that the 65Cu(n,2n) reaction rate reported in the reference for the pressure vessel capsule is not correct. In considering the libraries developed after BUGLE-93, a maximum relative difference of 12% was observed in reaction rates, with respect to the reference results, for 237Np(n,f) in the cavity capsule using BUGLE-B7.
ZZ SINEX, 100 Neutron-Group Neutron Reaction Cross-Section Library from ENDF/B by SUPERTOG for ANISN
1 - Nature of physical problem solved: Format: ANISN; Number of groups: 100 group reaction cross sections for neutron interactions. Nuclides: H, D, He, He-3, Li-6, Li-7, Be-9, B-10, B-11, C-12, N-14, O-16, Na-23, Mg, Al-27, Si, Cl, K, Ca, V, Cr, Mn-55, Fe, Co-59, Ni, Cu, Cu-63, Cu-65, Nb, Mo, Ag-107, Ag-109, Xe-135, Cs-133, Sm-149, Eu-151, Eu-153, Gd, Dy-164, Lu-175, Lu-176, Ta-181, Ta-182, W-182, W-183, W-184, W-186, Re-185, Re-187, Au-197, Pb, Th-232, Pa-233, U-234, U-235, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, Am-243, and Cm-244. Origin: ENDF/B; Weighting spectrum: For the top 99 groups, the explicit assumption was made that the flux (weighting function) has the shape of a fission spectrum jointed at 0.0674 MeV by a 1/E tail. For the thermal group (group 100), values for all materials except hydrogen were taken from the Maxwellian average values derived from the ENDF/B data. The data can be used in combination with 100 group neutron transport calculations (using, e. g., the DLC-2 library) to determine the spatial distribution of individual reaction rates. In particular, the retrieval program allows the preparation of dummy materials based on DLC-24 which can be used in the activity calculation option in ANISN to calculate the desired reaction rates. The library consists of 100 group reaction cross sections for neutron interactions as follows - total, elastic, inelastic, (n,2n), fission, (n,n'α), (n,n'3α), (n,2nα), absorption, (n,n'p), capture, (n,γ), (n,p), (n,d), (n,t), (n,He3), (n,α), (n,2α), and ν-bar. The units are barns, except that ν-bar is the average number of neutrons per fission event. A table listing the reactions included for each material is found in ref.1. The nuclides in DLC-24 are those which have been released as category I ENDF/B by the National Neutron Cross Section Center, Brookhaven National Laboratory. The library contains data for H, D, He, 3-He, 6-Li, 7-Li, 9-Be, 10-B, 11-B, 12-C, 14-N, 16-O, 23-Na, Mg, 27-Al, Si
The PSIMECX library contains calculated nuclide production cross-sections from neutron-induced reactions in the energy range about 2 to 800 MeV in the following 72 stable isotopes of 24 elements: 12C, 13C, 16O, 17O, 18O, 23Na, 24Mg, 25Mg, 26Mg, 27Al, 28Si, 29Si, 30Si, 31P, 32S, 33S, 34S, 36S, 35Cl, 37Cl, 39K, 40K, 41K, 40Ca, 42Ca, 43Ca, 44Ca, 46Ca, 48Ca, 46Ti, 47Ti, 48Ti, 49Ti, 50Ti, 50V, 51V, 50Cr, 52Cr, 53Cr, 54Cr, 55Mn, 54Fe, 56Fe, 57Fe, 58Fe, 58Ni, 60Ni, 61Ni, 62Ni, 64Ni, 63Cu, 65Cu, 64Zn, 66Zn, 67Zn, 68Zn, 70Zn, 92Mo, 94Mo, 95Mo, 96Mo, 97Mo, 98Mo, 100Mo, 121Sb, 123Sb, 204Pb, 206Pb, 207Pb, 208Pb, 232Th and 238U. The energy range covers essentially all transmutation channels other than capture. The majority of the selected elements are main constituents of normal materials of construction used in and around accelerator facilities and the library is, first and foremost, designed to be a tool for the estimation of their activation in wide-band neutron fields. This third report describes and discusses the calculational methods used for the heavy nuclei. The library itself has been described in the first report of this series and the treatment for the medium and light mass nuclei is given in the second. (author)
This paper presents a synthesis of the ENEA-Bologna Nuclear Data Group programme dedicated to generate and validate group-wise cross section libraries for shielding and radiation damage deterministic calculations in nuclear fission reactors, following the data processing methodology recommended in the ANSI/ANS-6.1.2-1999 (R2009) American Standard. The VITJEFF311.BOLIB and VITENDF70.BOLIB fine group coupled n-γ (199 n + 42 γ - VITAMIN-B6 structure) multi-purpose cross section libraries, based on the Bondarenko method for neutron resonance self-shielding and respectively on JEFF-3.1.1 and ENDF/B-VII.0 evaluated nuclear data, were produced in AMPX format using the NJOY-99.259 and the ENEA-Bologna 2007 Revision of the SCAMPI nuclear data processing systems. Two derived broad-group coupled n-γ (47 n + 20 γ - BUGLE-96 structure) working cross section libraries in FIDO-ANISN format for LWR shielding and pressure vessel dosimetry calculations, named BUGJEFF311.BOLIB and BUGENDF70.BOLIB, were generated by the revised version of SCAMPI, through problem-dependent cross section collapsing and self-shielding from the cited fine-group libraries. The validation results on the criticality safety benchmark experiments for the fine-group libraries and the preliminary validation results for the broad-group working libraries on the PCA-Replica and VENUS-3 engineering neutron shielding benchmark experiments are reported in synthesis. (authors)
Updates to the ORIGEN-S Cross-Section Libraries Using ENDF-VI, EAF-99, and FENDL-2.0
Murphy, B.D.
2004-11-04
The standard cross-section library for light-water reactor (LWR) analyses used by the ORIGEN-S depletion and decay code has been extensively updated. This work entailed the development of broad multigroup neutron cross sections for ORIGEN-S from several sources of pointwise continuous-energy cross-section evaluations, including the U.S. Evaluated Nuclear Data Files ENDF/B-VI Release 7, the Fusion Evaluated Nuclear Data Library FENDL-2.0, and the European Activation File EAF-99. The pointwise cross sections were collapsed to a three-group structure using a continuous-energy neutron flux spectrum representative of the typical neutronic conditions of typical LWR fuel and formatted for use by ORIGEN-S. In addition, the fission-product library has been expanded to include ENDF/B-VI fission yield data for 30 fissionable actinides. The processing codes and procedures are explained. Preliminary verification studies using the updated libraries were performed using the modules of the SCALE (Standardized Computer Analyses for Licensing Evaluation) system. Comparisons between the previous basic ORIGEN-S libraries and the updated libraries developed in this work are presented.
Pescarini, M.; Sinitsa, V.; Orsi, R.; Frisoni, M.
2013-03-01
This paper presents a synthesis of the ENEA-Bologna Nuclear Data Group programme dedicated to generate and validate group-wise cross section libraries for shielding and radiation damage deterministic calculations in nuclear fission reactors, following the data processing methodology recommended in the ANSI/ANS-6.1.2-1999 (R2009) American Standard. The VITJEFF311.BOLIB and VITENDF70.BOLIB finegroup coupled n-γ (199 n + 42 γ - VITAMIN-B6 structure) multi-purpose cross section libraries, based on the Bondarenko method for neutron resonance self-shielding and respectively on JEFF-3.1.1 and ENDF/B-VII.0 evaluated nuclear data, were produced in AMPX format using the NJOY-99.259 and the ENEA-Bologna 2007 Revision of the SCAMPI nuclear data processing systems. Two derived broad-group coupled n-γ (47 n + 20 γ - BUGLE-96 structure) working cross section libraries in FIDO-ANISN format for LWR shielding and pressure vessel dosimetry calculations, named BUGJEFF311.BOLIB and BUGENDF70.BOLIB, were generated by the revised version of SCAMPI, through problem-dependent cross section collapsing and self-shielding from the cited fine-group libraries. The validation results on the criticality safety benchmark experiments for the fine-group libraries and the preliminary validation results for the broad-group working libraries on the PCA-Replica and VENUS-3 engineering neutron shielding benchmark experiments are reported in synthesis.
Orsi R.
2013-03-01
Full Text Available This paper presents a synthesis of the ENEA-Bologna Nuclear Data Group programme dedicated to generate and validate group-wise cross section libraries for shielding and radiation damage deterministic calculations in nuclear fission reactors, following the data processing methodology recommended in the ANSI/ANS-6.1.2-1999 (R2009 American Standard. The VITJEFF311.BOLIB and VITENDF70.BOLIB finegroup coupled n-γ (199 n + 42 γ – VITAMIN-B6 structure multi-purpose cross section libraries, based on the Bondarenko method for neutron resonance self-shielding and respectively on JEFF-3.1.1 and ENDF/B-VII.0 evaluated nuclear data, were produced in AMPX format using the NJOY-99.259 and the ENEA-Bologna 2007 Revision of the SCAMPI nuclear data processing systems. Two derived broad-group coupled n-γ (47 n + 20 γ – BUGLE-96 structure working cross section libraries in FIDO-ANISN format for LWR shielding and pressure vessel dosimetry calculations, named BUGJEFF311.BOLIB and BUGENDF70.BOLIB, were generated by the revised version of SCAMPI, through problem-dependent cross section collapsing and self-shielding from the cited fine-group libraries. The validation results on the criticality safety benchmark experiments for the fine-group libraries and the preliminary validation results for the broad-group working libraries on the PCA-Replica and VENUS-3 engineering neutron shielding benchmark experiments are reported in synthesis.
Orsi R.; Sinitsa V.; Pescarini M.; Frisoni M.
2013-01-01
This paper presents a synthesis of the ENEA-Bologna Nuclear Data Group programme dedicated to generate and validate group-wise cross section libraries for shielding and radiation damage deterministic calculations in nuclear fission reactors, following the data processing methodology recommended in the ANSI/ANS-6.1.2-1999 (R2009) American Standard. The VITJEFF311.BOLIB and VITENDF70.BOLIB finegroup coupled n-γ (199 n + 42 γ – VITAMIN-B6 structure) multi-purpose cross section libraries, based o...
The cell code WIMSD4 is used for the analysis of PROTEUS-LWHCR experiments. A library for this code which is based on the European evaluation JEF-1 was produced at EIR using the Los Alamos NJOY system with its module WIMSR and the Canadian management code WILMA. In general, this library delivered more accurate eigenvalues and reaction rates than the WIMS-Standard and WIMS81 libraries did in comparison to experimental values from PROTEUS-LWHCR Cores 1-3. However, large discrepancies (up to about 10%) occured between calculated migration areas (M2). Additional investigations have been undertaken to clarify this problem, since theoretical M2-values are needed for deducing k-infinity in the experiments. This has been done in the context of calculations for a reference LWHCR test lattice. The following major reasons for these deviations were found. First, the self-scattering term in non-moderators (P0 matrix) in the JEF-1 library was not transport corrected. Second, Standard and JEF-1 libraries use infinite dilute cross sections for 238U, whereas the WIMS81 library uses fully shielded cross sections. Third, the standard library uses the 'row' formula for the transport correction, whereas the 'inflow' formula is applied in the case of JEF-1 and WIMS81 libraries. Lastly, oxygen and 238U scattering cross sections in the fast energy range are smaller in the case of the WIMS81 library. Differences in calculated k-infinity values between the currently used library and WIMS81 (up to 3%) come (in order of importance for the reference LWHCR lattice) mainly from resonance cross sections for 240Pu capture, 238U capture and 239Pu fission. Recommendations have been made for generating a new JEF-1 library using updated versions of WIMSR and WILMA. (author)
The elimination of a large number of approximations that lead to numerous errors in the neutronic reactor calculations was the main purpose behind developing Monte Carlo codes. The MCNP series of codes (Monte Carlo Nuclear Particle) are developed and extensively used in neutronic core calculations. Although the neutronic data input to these codes is the pointwise cross section files as presented by ENDF libraries or similar ones, are comprehensive and detailed. Yet the major sources of errors in the core calculations stem from the uncertainty in the cross section data. In this paper the effect of estimated uncertainty in the values of cross sections in the ENDF/B-V1 library, on the neutronic parameters of the ETRR-2 reactor is studied. MCNP code is used to simulate a three dimensional model for the reactor core considering all the materials composition and geometrical details. Perturbation technique is used to determine the effect of uncertainty in cross sections for a number of isotopes in the reactor on the fission rates and a comparison is made between the fission rate values with and without the uncertainty values for the different cross section types and different energy ranges. It is shown that for all the considered isotopes the effect of uncertainty in the cross section data on the fission rate values is very small, where the differences in fission rates do not exceed 10% and this value is accepted
Evaluated cross section libraries and kerma factors for neutrons up to 100 MeV on 16O and 14N
We present evaluations of the interaction of 20 to 100 MeV neutrons with oxygen and nitrogen nuclei, which follows on from our previous work on carbon. Our aim is to accurately represent integrated cross sections, inclusive emission spectra, and kerma factors, in a data library which can be used in radiation transport calculations. We apply the FKK-GNASH nuclear model code, which includes Hauser-Feshbach, preequilibrium, and direct reaction mechanisms, and use experimental measurements to optimize the calculations. We determine total, elastic, and nonelastic cross sections, angle-energy correlated emission spectra, for light ejectiles with A≤4 and gamma-rays, and average energy depositions. Our results for charged-particle emission spectra agree well with the measurements of Subramanian et al.. We compare kerma factors derived from our evaluated cross sections with experimental data, providing an integral benchmarking of our work. The evaluated data libraries are available as electronic files
For both type of reactors, WWER-440 and WWER-1000, two different libraries have been created: BGL440 and BGL1000 respectively. The libraries have been produced by collapsing the American fine-group library VITAMIN-B6 (199 neutron and 42 gamma groups) to 67 group structure (47 neutron and 20 gamma groups). The libraries consider the features (detailed 1D geometry and material compositions) of the appropriate reactor and contain upscattering data for the five thermal energy groups. The order of scattering of the Legendre expansion is P5. Each library consists of 2 parts. The first part consists of neutron/gamma cross section data for all reactor materials: BGL441 consists of neutron/gamma cross section data for 150 isotopes (17 chemical elements which appear with different densities and temperatures in the different reactor materials that comprise the WWER-440 reactor); BGL1001 consists of cross sections for 140 nuclides (22 chemical elements which comprise the materials in the WWER-1000). For collapsing cross-sections (previously energy self-shielded) from the 241 group structure (VITAMIN-B6) to the 67 group structure the appropriate average neutron flux in each reactor zone has been used. These datasets can be used for detailed computations of neutron transport. The second parts of each library, BGL442 and BGL1002, consist of cross sections for all 120 nuclides in the VITAMIN-B6 based on the infinitely dilute values only without energy self-shielding. The neutron spectrum just beyond the Reactor Pressure Vessel (RPV) was used for this collapsing. These second datasets can be used for describing non-reactor materials such as dosimeters, capsules, specimens, etc., which may be inserted in the region behind the RPV. (author). 3 refs, 2 figs, 9 tabs
The ENEA-Bologna Nuclear Data Group produced the JEFF-3.1 VITJEFF31.BOLIB and MATJEFF31. BOLIB fine-group coupled neutron and photon (199 n + 42 γ) cross section libraries for nuclear fission applications, respectively in AMPX and MATXS format, with the same specifications and energy group structure of the Endf/B-VI-3 VITAMIN-B6 American library. Each library, containing 181 nuclide cross section files, was generated from the same set of cross section data files in GENDF format, obtained through the Bondarenko (f-factor) method, with an ENEA-Bologna revised version of the GROUPR module of the NJOY-99.160 system. Collapsed working libraries of self-shielded cross sections in FIDO-ANISN format, used by the deterministic transport codes of the DANTSYS and DOORS systems, can be generated from VITJEFF31.BOLIB and MATJEFF31.BOLIB through, respectively, further data processing with an ENEA-Bologna revised version of the SCAMPI system and with the TRANSX code. This paper describes the methodology and specifications of the data processing performed and presents some results of the VITJEFF31.BOLIB validation. (authors)
EJ2-XMAS. Contents of the JEF2.2 based neutron cross-section library in the XMAS group structure
This report describes the contents of the EJ2-XMAS library. The EJ2-XMAS library is a JEF2.2 based 172-group AMPX-Master library in the XMAS group structure for reactor calculations with the SCALE-4 system, as implemented at ECN-Petten. The group cross section data were generated with NJOY89/NSLINK4 and NJOY91/NSLINK4. The data on the EJ2-XMAS library allow resolved-resonance treatment by NITAWL and unresolved resonance self-shielding by BONAMI. These codes are based upon the Nordheim and Bondarenko methods, respectively. (orig.)
and ENDF/B-V data. The group constants for minor actinides such as Np, Am, and Cm have been produced on the basis of the JENDL-2 data, to be used for TRU-transmutation calculations. This library is designed for JAERI fast reactor analysis and design code system. This library contains the 70 group constants with quarter-lethargy width for the following 59 nuclides and 12 lumped fission products: H-1, He-4, Be-9, B-10, B-11, C-12, N-14, O-16, Na-23, Al-27, Si, Ar, Ti, V, Cr, Mn-55, Fe, Ni, Cu, Zr, Nb-93, Mo, Eu-151, Eu-153, Gd, Gd-155, Gd-156, Gd-157, Gd-158, Gd-160, Ta-181, W, Th-228, Th-230, Th-232, Th-233, Th-234, Pa-233, U-233, U-234, U-235, U-236, U-238, Np-237, Np-239, Pu-236, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, Am-242m, Am-242g, Am-243, Cm-242, Cm-243, Cm-244, Cm-245 and 12 LFPs for 4 mother nuclides (U-235, U-238, Pu-239 and Pu-241) and 3 burnup days (180, 1080 and 1800). ZZ-JFS-V2: 25-group constants in ABBN energy structure and 70-group constants in JFS energy structure for the following elements: Be, B-10, B-11, C, O, Na, Al, Si, Cr, Mn, Fe, Ni, Cu, Mo, Th-232, U-233, U-234, U-235, U-236, U-238, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, fission products of U-235, and fission products of Pu-239. 2 - Method of solution: ZZ-JFS-3/J2: Group constants are generated with the data processing code system TIMS-PGG. In this code, the collision density spectrum for a typical large LMFBR core spectrum is used as the weighting function. The group constants in the unresolved resonance region are produced on the basis of the random sampling resonance generation method. The ultra-fine group calculation method is used for the resonance region. The resonance shielding factors are tabulated for 8 background potential scattering cross sections (0, 1, 10, 100, 1000, 10000, 100000 and 1000000 barns), 4 temperatures (300, 800, 2100, and 4500) and 4 resonance interface parameters. ZZ-JFS-V2: The cross-section adjustment has been made by using an auxiliary equation for
A new pointwise energy neutron cross section library named ENDFb7–r in ACE format for Reactor Monte Carlo code RMC has been generated by Reactor Cross Section Processing code RXSP using ENDF/B-VII.0. The pointwise energy cross section library called ENDFb7–n generated by NJOY has also been constructed for inter-comparison of results. Benchmark tests for series of criticality reactor cores and assemblies including both uranium and plutonium fuels with thermal, intermediate and fast neutron spectrum have been performed with the code RMC using these two libraries ENDFb7–r and ENDFb7–n. The k-effective and neutron flux calculated with two libraries show very good agreement with each other. Moreover, another practical PWR fuel assembly depletion model is further constructed and simulated by RMC. The calculated results of k-effective and isotopic concentration swings with burnup agree very well with each other. It has been proved that the self-processed library named ENDFb7–r is accurate enough to be used for both criticality and depletion calculations. (author)
The ENEA-Bologna Nuclear Data Group produced the VITJEF22.BOLIB (NEA-1699/01 ZZ VITJEF22.BOLIB) and MATJEF22.BOLIB (NEA-1740/01 ZZ MATJEF22.BOLIB) fine-group coupled neutron and photon (199 n + 42 γ) cross section libraries for nuclear fission applications, respectively in AMPX and MATXS format and based on the JEF-2.2 European nuclear data file. Both the libraries were produced from the same set of cross section files in GENDF format, generated with the NJOY-94.66 nuclear data processing system. The present libraries can be considered as European counterparts of the VITAMIN-B6 (DLC-0184 ZZ VITAMIN-B6) American library in AMPX format, based on the ENDF/B-VI Release 3 American nuclear data file. In fact they have the same general features and the same neutron and photon energy group structures as VITAMIN-B6. In particular, all these libraries are pseudo-problem-independent and based on the Bondarenko (f-factor) method for the treatment of neutron resonance self-shielding and temperature effects. Each ENEA-Bologna library contains a set of 133 nuclide cross section files processed at 4 temperatures (300 K, 600 K, 1000 K and 2100 K) and obtained for the most part with 6 to 8 values of the background cross section σ0. Thermal scattering cross sections were processed at all the temperatures available in the JEF-2.2 thermal scattering law data file for 5 additional bound nuclides: H-1 in light water, H-1 in polyethylene, H-2 in heavy water, C in graphite and Be in beryllium metal. Collapsed working libraries of self-shielded cross sections in the formats used by the deterministic transport codes of the DANTSYS and DOORS systems can be generated from VITJEF22.BOLIB and MATJEF22.BOLIB through, respectively, further problem-dependent data processing with the AMPX or SCAMPI nuclear data processing systems and with the TRANSX code. (authors)
Pescarini, M.; Orsi, R.; Martinelli, T.; Sinitsa, V. [ENEA - Centro Ricerche - Ezio Clementel - Bologna (Italy); Blokhin, A.I. [Institute of Physics and Power Engineering (IPPE), Kaluga Region (Russian Federation)
2005-07-01
The ENEA-Bologna Nuclear Data Group produced the VITJEF22.BOLIB (NEA-1699/01 ZZ VITJEF22.BOLIB) and MATJEF22.BOLIB (NEA-1740/01 ZZ MATJEF22.BOLIB) fine-group coupled neutron and photon (199 n + 42 {gamma}) cross section libraries for nuclear fission applications, respectively in AMPX and MATXS format and based on the JEF-2.2 European nuclear data file. Both the libraries were produced from the same set of cross section files in GENDF format, generated with the NJOY-94.66 nuclear data processing system. The present libraries can be considered as European counterparts of the VITAMIN-B6 (DLC-0184 ZZ VITAMIN-B6) American library in AMPX format, based on the ENDF/B-VI Release 3 American nuclear data file. In fact they have the same general features and the same neutron and photon energy group structures as VITAMIN-B6. In particular, all these libraries are pseudo-problem-independent and based on the Bondarenko (f-factor) method for the treatment of neutron resonance self-shielding and temperature effects. Each ENEA-Bologna library contains a set of 133 nuclide cross section files processed at 4 temperatures (300 K, 600 K, 1000 K and 2100 K) and obtained for the most part with 6 to 8 values of the background cross section {sigma}{sub 0}. Thermal scattering cross sections were processed at all the temperatures available in the JEF-2.2 thermal scattering law data file for 5 additional bound nuclides: H-1 in light water, H-1 in polyethylene, H-2 in heavy water, C in graphite and Be in beryllium metal. Collapsed working libraries of self-shielded cross sections in the formats used by the deterministic transport codes of the DANTSYS and DOORS systems can be generated from VITJEF22.BOLIB and MATJEF22.BOLIB through, respectively, further problem-dependent data processing with the AMPX or SCAMPI nuclear data processing systems and with the TRANSX code. (authors)
Evaluated cross section libraries and kerma factors for neutrons up to 100 MeV on 40Ca and 31P
The authors present evaluations of the interaction of 20 to 100 MeV neutrons with calcium and phosphorus, which follows on from the previous work on carbon, nitrogen, and oxygen. The aim is to accurately represent integrated cross sections, inclusive emission spectra, and kerma factors, in a data library which can be used in radiation transport calculations. They apply the GNASH nuclear model code, which includes Hauser-Feshbach, preequilibrium, and direct reaction mechanisms, and use experimental measurements to optimize the calculations. Total, elastic, and nonelastic cross sections, angle-energy correlated emission spectra for light ejectiles with A ≤ 4 and gamma-rays, and average energy depositions, are determined. The expected accuracy of the calculated cross sections and kerma factors is discussed
This report discusses specifications which have been developed for a new multigroup cross section library based on ENDF/B-VI data for light water reactor shielding and reactor pressure vessel dosimetry applications. The resulting broad-group library and an intermediate fine-group library are defined by the specifications provided in this report. Processing ENDF/B-VI into multigroup format for use in radiation transport codes will provide radiation shielding analysts with the most currently available nuclear data. it is expected that the general nature of the specifications will be useful in other applications such as reactor physics
Selected neutron reaction nuclear data evaluations and photon-atomic interaction cross section libraries for elements of interest to the IAEA's program on Fusion Evaluated Nuclear Data Library (FENDL) have been processed into GENDF and MATXS format using the NJOY system by R.E. MacFarlane, in VITAMIN-J group structure with VITAMIN-E weighting spectrum. This document summarizes the resulting multigroup data library FENDL/MG version 1.1. The data are available costfree, upon request from the IAEA Nuclear Data Section, online or on magnetic tape. (author). 7 refs, 1 tab
White, J.E.; Wright, R.Q.; Roussin, R.W.; Ingersoll, D.T.
1992-11-01
This report discusses specifications which have been developed for a new multigroup cross section library based on ENDF/B-VI data for light water reactor shielding and reactor pressure vessel dosimetry applications. The resulting broad-group library and an intermediate fine-group library are defined by the specifications provided in this report. Processing ENDF/B-VI into multigroup format for use in radiation transport codes will provide radiation shielding analysts with the most currently available nuclear data. it is expected that the general nature of the specifications will be useful in other applications such as reactor physics.
This study deals with the neutronic and thermal hydraulic analysis of the 3MW TRIGA MARK II research reactor to upgrade it to a higher flux. The upgrading will need a major reshuffling and reconfiguration of the current core. To reshuffle the current core configuration, the chain of NJOY94.10 - WIMSD-5A - CITATION - PARET - MCNP4B2 codes has been used for the overall analysis. The computational methods, tools and techniques, customisation of cross section libraries, various models for cells and super cells, and a lot of associated utilities have been standardised and established/validated for the overall core analysis. Analyses using the 4-group and 7-group libraries of macroscopic cross sections generated from the 69-group WIMSD-5 library showed that a 7-group structure is more suitable for TRIGA calculations considering its LEU fuel composition. The MCNP calculations established that the CITATION calculations and the generated cross section library are reasonably good for neutronic analysis of TRIGA reactors. Results obtained from PARET demonstrated that the flux upgrade will not cause the temperature limit on the fuel to be exceeded. Also, the maximum power density remains, by a substantial margin below the level at which the departure from nucleate boiling could occur. A possible core with two additional irradiation channels around the CT is projected where almost identical thermal fluxes as in the CT are obtained. The reconfigured core also shows 7.25% thermal flux increase in the Lazy Susan. (author)
A new executable, identified as NJOY99.0 has been created to generate the 69-group cross-section library for the reactor lattice transport code WIMS. The new code incorporates modifications in the WIMSR module of NJOY to generate the 69-group library, which will be used for TRIGA reactor calculations. The basic evaluated nuclear data file JEF-2.2 was used to generate the 69-group cross-section library in WIMS format. The results for TRX-1, TRX-2, BAPL-1, BAPL-2, and BAPL-3 benchmarks obtained by using the generated 69-group cross-section library from JEF-2.2 were analyzed. The following integral parameters were considered for the validation of the 69-group library: finite medium effective multiplication factor (keff), Ratio of epithermal to thermal 238U captures (ρ28), Ratio of epithermal to thermal 235U fission (δ25), Ratio of 238U fission to 235U fission (δ28) and Ratio of 238U captures to 235U fissions (C*). The TRX and BAPL benchmark lattices were modeled with optimized inputs, which were suggested in the final report of the WIMS Library Update Project (WLUP) Stage-I by Ravnik. The calculated results of the integral parameters of TRX and BAPL Benchmark Lattices obtained by using the new version of code WIMSD-5B were found to be in good agreement with the experimental values. Besides, The TRX and BAPL calculation results showed that JEF-2.2 is reliable for thermal reactor calculations and validated the 69-group library, which will be used for the neutronic calculation of the TRIGA Mark-II research reactor at AERE, Savar, Dhaka, Bangladesh. (authors)
Alonso V, G.; Hernandez L, H. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)
1991-11-15
On the basis of the library structure of the RECORD code a method to generate the neutron cross sections by means of the ENDF-B/IV database and the NJOY code has been developed. The obtained cross sections are compared with those of the current library which was processed using the ENDF-B/III version. (Author)
The new cross-section covariance matrix library ZZ-VITAMIN-J/COVA/EFF3 intended to simplify and encourage sensitivity and uncertainty analysis was prepared and is available from the NEA Data Bank. The library is organised in a ready-to-use form including both the covariance matrix data as well as processing tools:-Cross-section covariance matrices from the EFF-3 evaluation for five materials: 9Be, 28Si, 56Fe, 58Ni and 60Ni. Other data will be included when available. -FORTRAN program ANGELO-2 to extrapolate/interpolate the covariance matrices to a users' defined energy group structure. -FORTRAN program LAMBDA to verify the mathematical properties of the covariance matrices, like symmetry, positive definiteness, etc. The preparation, testing and use of the covariance matrix library are presented. The uncertainties based on the cross-section covariance data were compared with those based on other evaluations, like ENDF/B-VI. The collapsing procedure used in the ANGELO-2 code was compared and validated with the one used in the NJOY system
Kodeli, Ivan-Alexander [OECD NEA-DB, 12 Bd des Iles, 92130 Issy-les-Moulineaux (France)]. E-mail: ivo.kodeli@oecd.org
2005-11-15
The new cross-section covariance matrix library ZZ-VITAMIN-J/COVA/EFF3 intended to simplify and encourage sensitivity and uncertainty analysis was prepared and is available from the NEA Data Bank. The library is organised in a ready-to-use form including both the covariance matrix data as well as processing tools:-Cross-section covariance matrices from the EFF-3 evaluation for five materials: {sup 9}Be, {sup 28}Si, {sup 56}Fe, {sup 58}Ni and {sup 60}Ni. Other data will be included when available. -FORTRAN program ANGELO-2 to extrapolate/interpolate the covariance matrices to a users' defined energy group structure. -FORTRAN program LAMBDA to verify the mathematical properties of the covariance matrices, like symmetry, positive definiteness, etc. The preparation, testing and use of the covariance matrix library are presented. The uncertainties based on the cross-section covariance data were compared with those based on other evaluations, like ENDF/B-VI. The collapsing procedure used in the ANGELO-2 code was compared and validated with the one used in the NJOY system.
SNL RML recommended dosimetry cross section compendium
Griffin, P.J.; Kelly, J.G.; Luera, T.F. [Sandia National Labs., Albuquerque, NM (United States); VanDenburg, J. [Science and Engineering Associates, Inc., Albuquerque, NM (United States)
1993-11-01
A compendium of dosimetry cross sections is presented for use in the characterization of fission reactor spectrum and fluence. The contents of this cross section library are based upon the ENDF/B-VI and IRDF-90 cross section libraries and are recommended as a replacement for the DOSCROS84 multigroup library that is widely used by the dosimetry community. Documentation is provided on the rationale for the choice of the cross sections selected for inclusion in this library and on the uncertainty and variation in cross sections presented by state-of-the-art evaluations.
A project to prepare an exhaustive handbook of WIMS-D cross section libraries for thermal reactor applications comparing different WIMS-D compatible nuclear data libraries originating from various countries has been successfully designed. To meet the objectives of this project, a computer software package with graphical user interface for MS Windows has been developed at BARC, India. This article summarizes the salient features of this new software and presents significant improvements and extensions in relation to its first version [Ann Nucl Energ 29 (2002) 1735
To improve the accuracy of the neutron analyses for subcritical systems with thermal fission blanket, a coupled neutron and photon (315 n + 42γ) fine-group cross section library HENDL3.0/FG based on ENDF/B-Ⅶ. 0 has been produced by FDS team. In order to test the availability and reliability of the HENDL3.0/FG data library, shielding and critical safety benchmarks were performed with VisualBUS code. The testing results indicated that the discrepancy between calculation and experimental values of nuclear parameters fell in a reasonable range. (authors)
New WWER-1000 fuel libraries with cross-sections were created, they are intended to work with the ORIGEN-S module of the SCALE4.4a system. The used model is described and main input data about geometry and material composition of WWER-1000 fuel assembly, densities, temperatures, masses, and others are given too. Comparison by nuclide concentrations, between SCALE4.4a with the 17x17 library for PWR and tvsm1000 library for WWER-1000, and the HELIOS-1.5 code is realized. Comparison by radioactivity and decay heat between the libraries 17x17 for PWR and tvsm1000 for WWER-1000 is realized for different nuclides and total (Authors)
This study deals with the analysis of some neutronics and safety parameters of the current core of a 3-MW TRIGA MARK-II research reactor and validation of the generated macroscopic cross-section library and calculational techniques by benchmarking with experimental, operational, and available Safety Analysis Report (SAR) values. The overall strategy is: (a) generation of the problem-dependent cross-section library from basic Evaluated Nuclear Data Files such as ENDF/B-VI and JENDL-3.2 with NJOY94.10+, (b) use of the WIMSD-5 package to generate a few-group neutron macroscopic cross section for all of the materials in the core and its immediate neighborhood, (c) use the three-dimensional CITATION code to perform the global analysis of the core, and (d) checking of the validity of the CITATION diffusion code with the MCNP4B2 Monte Carlo code. The ultimate objective is to establish methods for reshuffling the current core configuration to upgrade the thermal flux at irradiation locations for increased isotope production. The computational methods, tools and techniques, customization of cross-section libraries, various models for cells and supercells, and many associated utilities are standardized and established/validated for the overall neutronic analysis. The excess reactivity, neutron flux, power distribution, power peaking factors, determination of the hot spot, and fuel temperature reactivity coefficients αf in the temperature range of 45 to 1000 deg. C are studied. All the analyses are performed using the 4- and 7-group libraries of the macroscopic cross sections generated from the 69-group WIMSD-5 library. The 7-group calculations yield comparatively better agreement with the experimental value of keff and the other core parameters. The CITATION test runs using different cross-section sets based on the different models applied in the WIMSD-5 calculations show a strong influence of those models on the final integral parameter. Some of the cells are specially
The D-T neutron-induced radioactivity constitutes one of the foremost issues in fusion reactor design. The validation of activation cross-sections and decay data libraries is one of the important requirements for validating ITER design from safety and waste disposal viewpoints. An elaborate, experimental program was initiated in 1988, under USDOE-JAERI collaborative program, to validate the radioactivity codes/libraries. The measurements of decay-γ spectra from irradiated, high purity samples of Al, Si, Ti, V, Cr, Mn-Cu alloy, Fe, Co, Ni, Cu, stainless steel 316 (AISI 316), Zn, Zr, Nb, Mo, In, Sn, Ta, W, and Pb, among others, were conducted under D-T neutron fluences varying from 1.6 x 1010ncm-2 to 6.1 x 1013ncm-2. As many as 14 neutron energy spectra were covered for a number of materials. The analysis of isotopic activities of the irradiated materials using activation cross-section libraries of four leading radioactivity codes, i.e. ACT4/THIDA-2, REAC-3, DKR-ICF, and RACC, has shown large discrepancies among the calculations, on the one hand, and between the calculations and the measurements, on the other. A discussion is also presented on definition and obtention of safety cum quality factors for various activation libraries. (orig.)
Goluoglu, S.
2003-12-01
A review of the degree of applicability of benchmarks containing gadolinium using the computer code KENO V.a and the gadolinium cross sections from the 238-group SCALE cross-section library has been performed for a system that contains {sup 239}Pu, H{sub 2}O, and Gd{sub 2}O{sub 3}. The system (practical problem) is a water-reflected spherical mixture that represents a dry-out condition on the bottom of a sludge receipt and adjustment tank around steam coils. Due to variability of the mixture volume and the H/{sup 239}Pu ratio, approximations to the practical problem, referred to as applications, have been made to envelop possible ranges of mixture volumes and H/{sup 239}Pu ratios. A newly developed methodology has been applied to determine the degree of applicability of benchmarks as well as the penalty that should be added to the safety margin due to insufficient benchmarks.
A patch for the NJOY data processing system was prepared to enable the correct processing of covariances stored in ENDF File 40, as used in IRDF-2002. This patch has now been included in the official NJOY processing system, release NJOY99.336. A number of minor corrections were also required to the File 40 covariances in nine evaluations included in IRDF-2002. These corrected pointwise cross section files, designated as IRDF-2002.1, are available from the IAEA website: http://www-nds.iaea.org/irdf2002/. (author)
For research reactor applications of neutron activation analysis, the evaluated neutron reaction cross sections and resonance integrals in some different libraries available were analyzed comparatively. In order to check these data, the thermal neutron capture cross section (σ0) and the resonance integral (I0) of 23Na(n, γ )24Na, 58Fe(n, γ) 59Fe, 59Co(n, γ )60Co, 27Al(n, γ )28Al, 109Ag(n, γ) 110mAg, 197Au(n, γ)198Au and 238U(n, γ )239U reactions from different libraries were used for comparative analysis with experimental measurements based on fundamental neutron activation equation. The targets were irradiated with neutrons in a research nuclear reactor 100 kW power, Triga Mark I. A high purity Ge detector was used for the gamma ray measurements of the irradiated samples. The evaluated results have been in general agreement with the current data according to different library sources. (author)
Pritychenko, B.; Mughabghab, S. F.
2012-12-01
We present calculations of neutron thermal cross sections, Westcott factors, resonance integrals, Maxwellian-averaged cross sections and astrophysical reaction rates for 843 ENDF materials using data from the major evaluated nuclear libraries and European activation file. Extensive analysis of newly-evaluated neutron reaction cross sections, neutron covariances, and improvements in data processing techniques motivated us to calculate nuclear industry and neutron physics quantities, produce s-process Maxwellian-averaged cross sections and astrophysical reaction rates, systematically calculate uncertainties, and provide additional insights on currently available neutron-induced reaction data. Nuclear reaction calculations are discussed and new results are presented. Due to space limitations, the present paper contains only calculated Maxwellian-averaged cross sections and their uncertainties. The complete data sets for all results are published in the Brookhaven National Laboratory report.
The present report contains the Summary of the Second IAEA Research Co-ordination Meeting of the Co-ordinated Research Programme on ''Establishment of an International Reference Data Library of Nuclear Activation Cross Sections''. The meeting was organized by the IAEA Nuclear Data Section with co-operation and assistance of local organizers from the Instituto de Fusion Nuclear de la Universidad Politecnica de Madrid, Spain, from 13 to 16 May 1996. Summarized are the conclusions and recommendations of the meeting together with a list of actions and deadlines. Attached are the detailed agenda and list of participants. (author). 4 refs, 1 tab
New activation cross section data
New nuclear cross section libraries (known as USACT92) have been created for activation calculations. A point-wise file was created from merging the previous version of the activation library, the U.S. Nuclear Data Library (ENDF/B-VI), and the European Activation File (EAF-2). 175 and 99 multi-group versions were also created. All the data are available at the National Energy Research Supercomputer Center
The neutronic and thermal hydraulic analysis of the 3 MW TRIGA MARK II research reactor to upgrade it is presented. The upgrading will need a major reshuffling and reconfiguration of the current core. To realize this objective, the overall strategy followed is: 1.) generation of problem dependent cross section library from basic Evaluated Nuclear Data Files such as ENDF/B-VI, JENDL3.2 with NJOY94.10+, 2.) use WIMSD-5 package to generate cell constants for all of the materials in the core and its immediate neighborhood, 3.) use CITATION to perform 3-D global analysis of the core to study multiplication factor, neutron flux and power distributions, power peaking factors, temperature reactivity coefficients, etc., 4.) couple output of CITATION with PARET to study thermal hydraulic behavior to predict safety margins, 5.) check the validity of the deterministic codes with the Monte Carlo code MCNP4B2 , and 6.) reshuffle the current core configuration to achieve the desired objectives. The computational methods, tools and techniques, customization of cross section libraries, various models for cells and super cells, and a lot of associated utilities have been standardized and established/validated for the overall core analysis
1 - Description: Format: MATXS, 142 nuclides processed with NJOY99.245. Number of groups: 150 neutron-, 12 photon-groups. 142 nuclides: H-1, H-2, He-3, He-4, Li-6, Li-7, Be-9, B-10, B-11, C-nat, N-14, N-15, O-16, F-19, Na-23, Mg-24, Mg-25, Mg-26, Al-27, Si-28, Si-29, Si-30, P-31, Cl-35, Cl-37, Ar-40, K-39, K-40, K-41, Ca-40, Ca-42, Ca-43, Ca-44, Ca-46, Ca-48, Ti-46, Ti-47, Ti-48, Ti-49, Ti-50, V-nat, Cr-50, Cr-52, Cr-53, Cr-54, Mn-55, Fe-54, Fe-56, Fe-57, Fe-58, Co-59, Ni-58, Ni-60, Ni-61, Ni-62, Ni-64, Cu-63, Cu-65, Ga-nat, Y-89, Zr-90, Zr-91, Zr-92, Zr-93, Zr-94, Zr-95, Zr-96, Nb-93, Mo-92, Mo-94, Mo-95, Mo-96, Mo-97, Mo-98, Mo-99, Mo-100, Ag-107, Ag-109, Cd-106, Cd-108, Cd-110, Cd-111, Cd-112, Cd-113, Cd-114, Cd-115m, Cd-116, Sn-112, Sn-114, Sn-115, Sn-116, Sn-117, Sn-118, Sn-119, Sn-120, Sn-122, Sn-123, Sn-124, Sn-125, Sn-126, Eu-151, Eu-153, Gd-152, Gd-154, Gd-155, Gd-156, Gd-157, Gd-158, Gd-160, W-182, W-183, W-184, W-186, Re-185, Re-187, Au-197, Pb-206, Pb-207, Pb-208, Bi-209, Th-232, Pa-233, U-233, U-234, U-235, U-236, U-238, Np-237, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, Am-242, Am-242m, Am-243, Cm-242, Cm-243, Cm-244, Cm-245, Cm-246. Origin: JEFF-3.1. Weighting spectrum: 300, 600, 900, 1200 K. The KAFAX-F31 is a MATXS-format, 150-group neutron and 12-group photon cross section library for fast reactors based on JEFF-3.1. This library was originally generated for the KALIMER (Korea Advanced LIquid Metal Reactor) core analyses. It includes 142 nuclide data (Table 1) processed by the NJOY99.245 code patched with NEA020. The library can be utilized to generate the problem-dependent group constants for neutron and/or photon transport calculations through the DANTSYS, DOORS, or PARTISN code systems. 2 - Methods: The KAFAX-F31 was generated at 300, 600, 900, and 1200 K. It contains the self-shielded cross sections for 5 to 10 background cross sections depending on the nuclides. The neutron group structure consists of one-eighth lethargy widths in almost
1 - Description: Format: MATXS, 144 nuclides processed with NJOY99.245. Number of groups: 150 neutron-, 12 photon-groups. 144 nuclides: H-1, H-2, He-3, He-4, Li-6, Li-7, Be-9, B-10, B-11, C-nat, N-14, N-15, O-16, F-19, Na-23, Mg-24, Mg-25, Mg-26, Al-27, Si-28, Si-29, Si-30, P-31, Cl-35, Cl-37, Ar-40, K-39, K-40, K-41, Ca-40, Ca-42, Ca-43, Ca-44, Ca-46, Ca-48, Ti-46, Ti-47, Ti-48, Ti-49, Ti-50, V-nat, Cr-50, Cr-52, Cr-53, Cr-54, Mn-55, Fe-54, Fe-56, Fe-57, Fe-58, Co-59, Ni-58, Ni-60, Ni-61, Ni-62, Ni-64, Cu-63, Cu-65, Ga-69, Ga-71, Y-89, Zr-90, Zr-91, Zr-92, Zr-93, Zr-94, Zr-95, Zr-96, Nb-93, Mo-92, Mo-94, Mo-95, Mo-96, Mo-97, Mo-98, Mo-99, Mo-100, Ag-107, Ag-109, Cd-106, Cd-108, Cd-110, Cd-111, Cd-112, Cd-113, Cd-114, Cd-115m, Cd-116, Sn-112, Sn-113, Sn-114, Sn-115, Sn-116, Sn-117, Sn-118, Sn-119, Sn-120, Sn-122, Sn-123, Sn-124, Sn-125, Sn-126, Eu-151, Eu-153, Gd-152, Gd-154, Gd-155, Gd-156, Gd-157, Gd-158, Gd-160, W-182, W-183, W-184, W-186, Re-185, Re-187, Au-197, Pb-206, Pb-207, Pb-208, Bi-209, Th-232, Pa-233, U-233, U-234, U-235, U-236, U-238, Np-237, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, Am-242, Am-242m, Am-243, Cm-242, Cm-243, Cm-244, Cm-245, Cm-246. Origin: ENDF/B-VII.0. Weighting spectrum: 300, 600, 900, 1200 k. The ZZ-KAFAX-E70 is a MATXS-format, 150-group neutron and 12-group photon cross section library for fast reactors based on ENDF/B-VII.0. This library was originally generated for the KALIMER (Korea Advanced Liquid Metal Reactor) core analyses. It includes 144 nuclide data processed with the NJOY99.245 code patched with NEA020. The library can be used to generate the problem-dependent group constants for neutron and/or photon transport calculations through the DANTSYS, DOORS, or PARTISN code systems. 2 - Methods: The KAFAX-E70 was generated at 300, 600, 900, and 1200 K. It contains the self-shielded cross sections for 5 to 10 background cross sections depending on the nuclides. The neutron group structure consists of one-eighth lethargy
1 - Description: Format: MATXS, 136 nuclides processed with NJOY99.245. Number of groups: 150 neutron-, 12 photon-groups. 136 Nuclides: H-1, H-2, He-3, He-4, Li-6, Li-7, Be-9, B-10, B-11, C-nat, N-14, N-15, O-16, F-19, Na-23, Mg-24, Mg-25, Mg-26, Al-27, Si-28, Si-29, Si-30, P-31, Cl-35, Cl-37, Ar-40, K-39, K-40, K-41, Ca-40, Ca-42, Ca-43, Ca-44, Ca-46, Ca-48, Ti-46, Ti-47, Ti-48, Ti-49, Ti-50, V-nat, Cr-50, Cr-52, Cr-53, Cr-54, Mn-55, Fe-54, Fe-56, Fe-57, Fe-58, Co-59, Ni-58, Ni-60, Ni-61, Ni-62, Ni-64, Cu-63, Cu-65, Ga-69, Ga-71, Y-89, Zr-90, Zr-91, Zr-92, Zr-93, Zr-94, Zr-95, Zr-96, Mo-92, Mo-94, Mo-95, Mo-96, Mo-97, Mo-98, Mo-99, Mo-100, Ag-107, Ag-109, Cd-106, Cd-108, Cd-110, Cd-111, Cd-112, Cd-113, Cd-114, Cd-116, Sn-112, Sn-114, Sn-115, Sn-116, Sn-117, Sn-118, Sn-119, Sn-120, Sn-122, Sn-123, Sn-124, Sn-126, Eu-151, Eu-153, Gd-152, Gd-154, Gd-155, Gd-156, Gd-157, Gd-158, Gd-160, W-182, W-183, W-184, W-186, Pb-206, Pb-208, Bi-209, Th-232, Pa-233, U-233, U-234, U-235, U-236, U-238, Np-237, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, Am-242, Am-242m, Am-243, Cm-242, Cm-243, Cm-244, Cm-245, Cm-246. Origin: JENDL-3.3. Weighting spectrum: 300, 600, 900, 1200 K. The KAFAX-J33 is a MATXS-format, 150-group neutron and 12-group photon cross section library for fast reactors based on JENDL-3.3. This library was originally generated for the KALIMER (Korea Advanced LIquid Metal Reactor) core analyses. It includes 136 nuclide data processed by the NJOY99.245 code patched with NEA020. The library can be utilized to generate the problem-dependent group constants for neutron and/or photon transport calculations through the DANTSYS, DOORS, or PARTISN code systems. 2 - Methods: The KAFAX-J33 was generated at 300, 600, 900, and 1200 K. It contains the self-shielded cross sections for 5 to 10 background cross sections depending on the nuclides. The neutron group structure consists of one-eighth lethargy widths in almost all the energy ranges, except between 1 and 10 keV in
Minnesota Department of Natural Resources — FEMA Cross Sections are required for any Digital Flood Insurance Rate Map database where cross sections are shown on the Flood Insurance Rate Map (FIRM). Normally...
The Generation IV [1] International forum identified six advanced reactor concepts and related fuel cycles along with the R and D programs necessary to achieve the four key goals: (1) sustainability, (2) safety and reliability, (3) economics, (4) proliferation resistance and physical protection. Among these six promising reactor concepts, the lead-cooled fast reactor (LFR) has been selected for development by EURATOM, which in 2006 decided to finance the European Lead Cooled System (ELSY) project. The aim of the project is to demonstrate the possibility to design a safe and competitive lead-cooled fast power reactor using simple engineering solutions. This paper demonstrates the use of the code package SCALE5.1 and its NEWT/TRITON modules [3] for preliminary neutronic core analysis of a LFR within Generation IV Nuclear Energy systems program. More specifically, the analysis of the reference design of the ELSY-600 open square fuel assembly is presented. In particular, the use of ENDF/B-V and ENDF/B-VI.7 and multigroup energy structure was investigated. The homogenized cross sections calculated for the ELSY fuel assembly 2D model have been evaluated and compared to the results obtained with calculations performed with the deterministic code ERANOS/ECCO using JEFF2.2 cross section library. A good agreement has been observed in the energy range of interests, and generally for energy above 1 eV. (authors)
1 - Description: Format: MATXS. Number of groups: 80 neutron-, 24 photon-groups. 97 Nuclides: 1-H-1, 1-H-2, 2-He-3, 2-He-4, 3-Li-6, 3-Li-7, 4-Be-9, 5-B-10, 5-B-11, 6-C- nat., 7-N-14, 7-N-15, 8-O-16, 9-F-19, 11-Na-23, 12-Mg-nat., 13-Al-27, 14-Si-nat., 15-P-31, 17-Cl-nat., 18-Ar-40, 19-K-nat., 20-Ca-nat., 22-Ti-nat., 23-V-nat., 24-Cr-50, 24-Cr-52, 24-Cr-53, 24-Cr-54, 25-Mn-25, 26-Fe-54, 26-Fe-56, 26-Fe-57, 26-Fe-58, 27-Co-59, 28-Ni-58, 28-Ni-60, 28-Ni-61, 28-Ni-62, 28-Ni-64, 29-Cu-nat., 31-Ga-nat., 39-Y-89, 40-Zr-nat., 41-Nb-93, 42-Mo-nat., 47-Ag-107, 47-Ag-109, 48-Cd-nat., 50-Sn-nat., 63-Eu-151, 63-Eu-153, 64-Gd-152, 64-Gd-154, 64-Gd-155, 64-Gd-156, 64-Gd-157, 64-Gd-158, 64-Gd-160, 73-Ta-181, 74-W-182, 74-W-183, 74-W-184, 74-W-186, 75-Re-185, 75-Re-187, 79-Au-197, 82-Pb-nat., 83-Bi-209, 90-Th-232, 91-Pa-233, 92-U-232, 92-U-233, 92-U-234, 92-U-235, 92-U-236, 92-U-237, 92-U-238, 93-Np-237, 93-Np-238, 94-Pu-238, 94-Pu-239, 94-Pu-240, 94-Pu-241, 94-Pu-242, 95-Am-241, 95-Am-242, 95-Am-242m, 95-Am-243, 96-Cm-242, 96-Cm-243, 96-Cm-244, 96-Cm-245, 96-Cm-246, 96-Cm-247, 96-Cm-248, 98-Cf-252 Origin: JEF-2.2; Weighting spectrum: Thermal + 1/E + fast reactor + fusion. The library is focused on the fast reactor analyses. It has 80 and 24 energy group structures for neutron and photon, respectively. It includes 97 nuclide data based on JEF-2.2 and has a Format of MATXS processed by the NJOY94 code. It can be used to calculate the problem dependant group constants with the TRANSX code for neutron and gamma transport. 2 - Methods: The data were generated at 300 ∼ 2500 Kelvin degrees and at 4∼7 background cross sections for the self shielding considerations. The weighting function used for group averaged neutron cross sections from the pointwise data is 'thermal + 1/E + fast reactor + fusion'. The library has been validated through the CSEWG benchmark analyses such as VERA-11A, ZPR-3-12, SNEAK-7B, ZPPR-2, ZPR-6-7, etc. 3 - Related or auxiliary programs: - BBC: Program to convert
Perkins, S.T.; Cullen, D.E. (Lawrence Livermore National Lab., CA (United States)); Seltzer, S.M. (National Inst. of Standards and Technology (NML), Gaithersburg, MD (United States). Center for Radiation Research)
1991-11-12
Energy-dependent evaluated electron interaction cross sections and related parameters are presented for elements H through Fm (Z = 1 to 100). Data are given over the energy range from 10 eV to 100 GeV. Cross sections and average energy deposits are presented in tabulated and graphic form. In addition, ionization cross sections and average energy deposits for each shell are presented in graphic form. This information is derived from the Livermore Evaluated Electron Data Library (EEDL) as of July, 1991.
Bretscher, M.M.
1993-12-31
The WIMS-D4 code has been modified (WIMS-D4M) to produce microscopic isotopic cross sections in ISOTXS format for use in diffusion and transport calculations. Beginning with 69-group libraries based on ENDF/B-V data, numerous cell calculations have been made to prepare a set of broad group cross sections for use in diffusion calculations. Global calculations have been made for two control rod states of the Romanian steady state TRIGA reactor with 29 fresh HEU fuel clusters. Detailed Monte Carlo calculations also have been performed for the same reactor configurations using data based on ENDF/B-V. Results from these global calculations are compared with each other and with the measured excess reactivities. Although region-averaged macroscopic principal cross sections obtained from WIMS-D4M are in good agreement with the corresponding Monte Carlo values, problems exist with the high energy (E > 10 keV) microscopic hydrogen transport cross sections.
1 - Description: Efforts devoted to developing or improving thermal scattering data (S(alpha,beta)) for very cold and cryogenic temperatures have recently been carried out. Here several evaluations carried out at the Institut fuer Kernenergetik und Energiesysteme (IKE), University of Stuttgart are made available. They are listed in the following Table. Liquid hydrogen and deuterium for the two modifications: ortho and para. para Hydrogen: MAT*: 2, Temperatures (K): 14, 16 and 20.38, ID ACE* files: pH.00t, pH.01t, pH.03t; ortho Hydrogen: MAT*: 3, Temperatures (K): 14, 16 and 20.38, ID ACE* files: oH.00t, oH.01t, oH.03t; para Deuterium: MAT*: 12, Temperatures (K): 19 and 23.65, ID ACE* files: pD.00t, pD.01t; ortho Deuterium: MAT*: 13, Temperatures (K): 19 and 23.65, ID ACE* files: oD.00t, oD.01t; H in polyethylene (CH2): H in CH2: MAT: 37, Temperatures (K): 87, 293.6 and 350, ID ACE files: poly.01t, poly.03t, poly.04t, poly.11t, poly.13t, poly.14t; Liquid argon: 18-Ar: MAT: 18, Temperature (K): 87, ID ACE file: argon.01t, argon.11t; Aluminium face centred cubic lattice: 13-Al-27: MAT: 61, Temperatures (K): 20, 77, 87, 100, 293.6, 400, ID ACE files: al.00t, al.01t, al.02t, al.03t, al.04t, al.05t, al.10t, al.11t, al.12t,al.13t, al.14t, al.15t (* MAT numbers for the ENDF files and ID's for ACE (MCNP continuous energy data libraries)). The datasets are provided in the standard ENDF-6 format and in the ACE format, used for continuous energy Monte Carlo applications. Cross section libraries can be produced also for deterministic approaches through the use of the NJOY computer code. It should be noted that for very cold temperatures special care must be taken in processing the data and occasionally patches need to be applied to the processing code. Processing of the S(alpha,beta) data to energy dependent differential and integral cross sections as well as data sets for neutron transport calculations has been carried out e.g. MCNP(X). This was done with the following
This work presents a theoretical re-evaluation of a set of original experiments included in the 2009 issue of the International Handbook of Evaluated Criticality Safety Benchmark Experiments, as “Concrete Reflected Cylinders of Highly Enriched Solutions of Uranyl Nitrate” (identification number: HEU-SOL-THERM- 002) [4]. The present evaluation has been made according to benchmark specifications [4], and added data taken out of the original published report [3], but applying a different approach, resulting in a more realistic calculation model. In addition, calculations have been made using the latest version of MCNPX Monte Carlo code, combined with an updated set of cross section data, the continuous-energy ENDF/B-VI library. This has resulted in a comprehensive model for the given experimental situation. Uncertainties analysis has been made based on the evaluation of experimental data presented in the HEU-SOLTHERM-002 report. Resulting calculations with the present improved physical model have been able to reproduce the criticality of configurations within 0.5%, in good agreement with experimental data. Results obtained in the analysis of uncertainties are in general agreement with those at HEU-SOL-THERM-002 benchmark document. Qualitative results from analyses made in the present work can be extended to similar fissile systems: well moderated units of 235U solutions, reflected with concrete from all directions. Results have confirmed that neutron absorbers, even as impurities, must be taken into account in calculations if at least approximate proportions were known. (authors)
Selected neutron reaction nuclear data evaluations for elements of interest to the IAEA's program on Fusion Evaluated Nuclear Data Library (FENDL) have been processed into ACE format using the NJOY system by R.E. MacFarlane. This document summarizes the resulting continuous energy cross-section data library FENDL/MC version 1.1. The data are available cost free, upon request from the IAEA Nuclear Data Section, online or on magnetic tape. (author). 1 tab
The work performed to revise the REAC data file of Mann et al., containing cross-sections for neutron activation and transmutation reactions for use in fusion-reactor technology, is described. The revisions were made by means of renormalizations of the cross-sections to experimental data at 14.5 MeV or to data from 14.5 MeV systematics. Uncertainty estimates are given for the systematics. Furthermore, a number of reactions have been added. The file essentially contains cross-sections for almost all stable and unstable nuclides with half lives exceeding 1 day. If a reaction can produce one or two isomers the cross-sections for producing the ground and isomeric states are given separately. In most cases these cross-sections were obtained by a simple scaling using isomer ratios at 14.5 MeV, based upon experimental data or recently developed systematics. For about 50 reactions leading to long-lived states a special treatment was followed including a detailed uncertainty analysis. The revised file is called REAC-ECN-3. A version with multi-group cross-sections has also been generated (GREAC-ECN-3). The report contains an Appendix with 14.5 MeV cross-sections for all isotopes considered in the data file. 7 figs.; 29 refs.; 5 tabs.; 1 appendix
A new interactive program called CAPSIZE has been written for the IBM-PC to rapidly determine the likely impact that proposed design objectives might have on the size and capacity of spent fuel shipping casks designed to meet those objectives. Given the burnup of the spent fuel, its cooling time, the thickness of the internal basket walls, the desired external dose rate, and the nominal weight limit of the loaded cask, the CAPSIZE program will determine the maximum number of PWR fuel assemblies that may be shipped in a lead-, steel-, or uranium-shielded cask meeting those objectives. The necessary neutron and gamma shield thicknesses are determined by the program in such a way as to meet the specified external dose rate while simultaneously minimizing the overall weight of the loaded cask. The one-group cross-section library used in the CAPSIZE program has been distilled from the intermediate results of several hundred 1-D multigroaup discrete ordinates calculations for different types of casks. Neutron and gamma source terms, as well as the decay heat terms, are based on ORIGEN-S analyses of PWR fuel assemblies having exposures of 10, 20, 30, 40, 50, and 60 gigawatt days per metric tonne of initial heavy metal (GWD/MTIHM). In each case, values have been tabulated at 17 different decay times between 120 days and 25 years. Other features of the CAPSIZE program include a steady-state heat transfer calculation which will minimize the size and weight of external cooling fins, if and when such fins are required. Comparisons with previously reported results show that the CAPSIZE program can generally estimate the necessary neutron and gamma shield thicknesses to within 0.16 in. and 0.08 in., respectively. The corresponding cask weights have generally been found to be within 1000 lbs of previously reported results. 13 refs., 20 figs., 54 tabs
The neutron activation technique is routinely used in fusion experiments to measure the neutron yields. This paper investigates the uncertainty on these measurements as due to the uncertainties on dosimetry and activation reactions. For this purpose, activation cross-sections were taken from the International Reactor Dosimetry and Fusion File (IRDFF-v1.05) in 640 groups ENDF-6 format for several reactions of interest for both 2.5 and 14 MeV neutrons. Activation coefficients (reaction rates) have been calculated using the neutron flux spectra at JET vacuum vessel, both for DD and DT plasmas, calculated by MCNP in the required 640-energy group format. The related uncertainties for the JET neutron spectra are evaluated as well using the covariance data available in the library. These uncertainties are in general small, but not negligible when high accuracy is required in the determination of the fusion neutron yields
In May 2010, JENDL-4.0 was released from Japan Atomic Energy Agency as the updated Japanese Nuclear Data Library. It was processed by the nuclear data processing system LICEM and an arbitrary-temperature neutron cross section library MVPlib-nJ40 was produced for the neutron and photon transport calculation code MVP based on the continuous-energy Monte Carlo method. The library contains neutron cross sections for 406 nuclides on the free gas model, thermal scattering cross sections, and cross sections of pseudo fission products for burn-up calculations with MVP. Criticality benchmark calculations were carried out with MVP and MVPlib-nJ40 for about 1,000 cases of critical experiments stored in the hand book of International Criticality Safety Benchmark Evaluation Project (ICSBEP), which covers a wide variety of fuel materials, fuel forms, and neutron spectra. We report all comparison results (C/E values) of effective neutron multiplication factors between calculations and experiments to give a validation data for the prediction accuracy of JENDL-4.0 for criticalities. (author)
ZZ ETOG-1-DATA, Cross-Section Library for Programs MUFT3, MUFT5, GAM1, GAM2 Generated from ENDF/B
; 1030 Gd; 1051 Pu-239; 1067 U-233NFP; 1015 Al-27; 1031 Dy-164; 1052 Pu-239FP; 1068 U-235SFP; 1016 Ti; 1032 Lu-175; 1053 Pu-240; 1069 U-235NFP; 1017 V51; 1033 Lu-176; 1054 Pu-241; 1070 Pu-239SFP; 1018 Cr; 1035 Ta-181; 1055 Pu-242; 1071 Pu-239NFP. 3 - Restrictions on the complexity of the problem: Library/No. of groups: GAM1/68, GAM2/99, MUFT4,5/54. A 1/E weighting function joined to the fission spectrum was used. A 1.0*107 value was used as the non-resonance potential scattering cross section per absorber atom. For MUFT libraries the 26. group was the lowest group number in the resonance region, the 25. group was the highest in the inelastic region
G. GiacomelliBologna University and INFN
2014-01-01
The measurements of the hadron-hadron total cross sections are the first measurements performed when a new hadron accelerator opens up a new energy region; the measurements were made as function of the incoming beam momentum or c.m. energy and have often been repeated with improved accuracy and finer energy spacing.
Recommended evaluation procedure for photonuclear cross section
Lee, Young-Ouk; Chang, Jonghwa; Fukahori, Tokio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1998-03-01
In order to generate photonuclear cross section library for the necessary applications, data evaluation is combined with theoretical evaluation, since photonuclear cross sections measured cannot provide all necessary data. This report recommends a procedure consisting of four steps: (1) analysis of experimental data, (2) data evaluation, (3) theoretical evaluation and, if necessary, (4) modification of results. In the stage of analysis, data obtained by different measurements are reprocessed through the analysis of their discrepancies to a representative data set. In the data evaluation, photonuclear absorption cross sections are evaluated via giant dipole resonance and quasi-deutron mechanism. With photoabsorption cross sections from the data evaluation, theoretical evaluation is applied to determine various decay channel cross sections and emission spectra using equilibrium and preequilibrium mechanism. After this, the calculated results are compared with measured data, and in some cases the results are modified to better describe measurements. (author)
ZZ-SCALE5.1/COVA-44G, 44-group cross section covariance matrix library extracted from SCALE5.1
1 - Description: ZZ-SCALE5.1/COVA-44G is a 44-group cross section covariance matrix library retrieved from the SCALE-5.1 package. The package includes the following 4 covariance libraries in COVERX format: - 44GROUPV5COV, Basic ENDF/B-V Covariance Library - 44GROUPV5REC, Recommended ENDF/B-V Covariance Library - 44GROUPV6COV, Basic ENDF/B-VI Covariance Library - 44GROUPV6REC, Recommended ENDF/B-VI Covariance Library The files contain the covariance data for the following reactions or parameters: total, elastic, inelastic, (n,2n), fission, chi, (n,gamma), (n,p), (n,d), (n,t), (n,3He), (n,α), and ν-bar. The nuclides or materials (in ZA order) for which covariance data are provided. In parentheses the total number of the different relative covariance matrices in the four libraries for each nuclide is specified. H-1(10),H-2(3),H-3(2),He-3(2),He-4,Li-6(2),Li-7(3),Be-9(2), B-10(3),B-11(2),C-0(6),N-14(2),N-15,O-16(3),O-17,F-19(3), Na-23(3),Mg-0,Al-27(2),Si-0(3),Si-28,Si-29,Si-29,Si-30, P-31,S-0,S-32,Cl-0,K-0,Ca-0,Sc-45(2),Ti-0, V-0(2),Cr-0(2),Cr-50,Cr-52,Cr-53,Cr-54,Mn-55(3),Fe-0(2), Fe-54,Fe-56,Fe-57,Fe-58,Co-59(3),Ni-0(2),Ni-58,Ni-60, Ni-61,Ni-62,Ni-64,Cu-0,Cu-63,Cu-65,Ga-0,Ge-72, Ge-73,Ge-74,Ge-76,As-75,Se-74,Se-76,Se-77,Se-78, Se-80,Se-82,Br-79,Br-81,Kr-78,Kr-80,Kr-82,Kr-83, Kr-84,Kr-85,Kr-86,Rb-85,Rb-87,Sr-84,Sr-86,Sr-87, Sr-88,Sr-89,Sr-90,Y-89,Y-89,Y-90,Y-91,Zr-0, Zr-90,Zr-91,Zr-92,Zr-93,Zr-94,Zr-96,Nb-93,Nb-93, Nb-94,Nb-95,Mo-0,Mo-94,Mo-95,Mo-96,Mo-97,Tc-99, Ru-96,Ru-99,Ru-100,Ru-101,Ru-102,Ru-104,Ru-105,Ru-106, Rh-103,Rh-105,Pd-102,Pd-104,Pd-105,Pd-106,Pd-107,Pd-108, Pd-110,Ag-107,Ag-109,Ag-111,Cd-0,Cd-106,Cd-108,Cd-110, Cd-111,Cd-112,Cd-113,Cd-114,Cd-116,In-0,In-113,In-115, Sn-112,Sn-114,Sn-115,Sn-116,Sn-117,Sn-118,Sn-119,Sn-120, Sn-122,Sn-124,Sb-121,Sb-123,Sb-124,Te-120,Te-122,Te-123, Te-124,Te-125,Te-126,Te-127(m),Te-128,Te-130,I-127,I-129, I-130,I-131,Xe-124,Xe-126,Xe-128,Xe-129,Xe-130,Xe-131, Xe-132,Xe-133,Xe-134,Xe-135,Xe-136,Cs-133,Cs-134,Cs-135, Cs-137
Pescarini, Massimo; Sinitsa, Valentin; Orsi, Roberto; Frisoni, Manuela
2016-02-01
Two broad-group coupled neutron/photon working cross section libraries in FIDO-ANISN format, dedicated to LWR shielding and pressure vessel dosimetry applications, were generated following the methodology recommended by the US ANSI/ANS-6.1.2-1999 (R2009) standard. These libraries, named BUGJEFF311.BOLIB and BUGENDF70.BOLIB, are respectively based on JEFF-3.1.1 and ENDF/B-VII.0 nuclear data and adopt the same broad-group energy structure (47 n + 20 γ) of the ORNL BUGLE-96 similar library. They were respectively obtained from the ENEA-Bologna VITJEFF311.BOLIB and VITENDF70.BOLIB libraries in AMPX format for nuclear fission applications through problem-dependent cross section collapsing with the ENEA-Bologna 2007 revision of the ORNL SCAMPI nuclear data processing system. Both previous libraries are based on the Bondarenko self-shielding factor method and have the same AMPX format and fine-group energy structure (199 n + 42 γ) as the ORNL VITAMIN-B6 similar library from which BUGLE-96 was obtained at ORNL. A synthesis of a preliminary validation of the cited BUGLE-type libraries, performed through 3D fixed source transport calculations with the ORNL TORT-3.2 SN code, is included. The calculations were dedicated to the PCA-Replica 12/13 and VENUS-3 engineering neutron shielding benchmark experiments, specifically conceived to test the accuracy of nuclear data and transport codes in LWR shielding and radiation damage analyses.
Zizin, M. N.; Zimin, V. G.; Zizina, S. N.; Kryakvin, L. V.; Pitilimov, V. A.; Tereshonok, V. A.
2010-12-01
The ShIPR intellectual code system for mathematical simulation of nuclear reactors includes a set of computing modules implementing the preparation of macro cross sections on the basis of the two-group library of neutron-physics cross sections obtained for the SKETCH-N nodal code. This library is created by using the UNK code for 3D diffusion computation of first VVER-1000 fuel loadings. Computation of neutron fields in the ShIPR system is performed using the DP3 code in the two-group diffusion approximation in 3D triangular geometry. The efficiency of all groups of control rods for the first fuel loading of the third unit of the Kalinin Nuclear Power Plant is computed. The temperature, barometric, and density effects of reactivity as well as the reactivity coefficient due to the concentration of boric acid in the reactor were computed additionally. Results of computations are compared with the experiment.
The effect of angular anisotropy in inelastic secondary neutron scattering on neutron leakage spectra from the surface of spherical specimens is investigated. It is shown how inadequate representation of the cross-section structure in the neutron energy resonance region can affect the neutron leakage spectrum. (author). 19 refs, 5 figs, 6 tabs
Pritychenko, B.; Mughaghab, S. F.; Sonzogni, A. A.
2009-01-01
We calculated the Maxwellian-averaged cross sections (MACS) and astrophysical reaction rates of the stellar nucleosynthesis reactions (n,$\\gamma$), (n,fission), (n,p), (n,$\\alpha$) and (n,2n) using the ENDF/B-VII.0-, JEFF-3.1-, JENDL-3.3-, and ENDF/B-VI.8-evaluated nuclear-data libraries. Four major nuclear reaction libraries were processed under the same conditions for Maxwellian temperatures ({\\it kT}) ranging from 1 keV to 1 MeV. We compare our current calculations of the {\\it s}-process n...
Evaluated neutron reaction data and photon-atom interaction cross sections for materials contained in the general purpose Fusion Evaluated Nuclear Data Library (FENDL/E2.0) have been processed with the NJOY code system into VITAMIN-J multigroup structure, for use in discrete-ordinates transport codes, and into continuous energy ACE format, for use in the Monte Carlo transport code MCNP. This document summarizes the resulting data libraries FENDL/MG-2.0 version 1 and FENDL/MC-2.0 version 1. The data are available costfree from the IAEA Nuclear Data Section online or on magnetic tape. (author)
1 - Description of program or function: MCB-JEF2.2 is a continuous-energy cross section libraries in ACE Format suitable for the MCB-1C and MCNP codes. Libraries for various materials were generated at six different Temperatures, and cover the energy range up to 20 MeV. Format: ACE. Number of groups: Continuous energy. Nuclides: H-1, H-2, H-3, He-3, He-4, Li-6, Li-7, Be-9, B-10, B-11, C-nat., N-14, N-15, O-16, O-17, Na-23, F-19, Mg-nat., Al-27, Si-nat., P-31, S-32, S-33, S-34, S-36, Cl-nat, K-nat, Ca-nat., Ti-nat, V-nat, Cr-50, Cr-52, Cr-53, Cr-54, Mn-55, Fe-54, Fe-56, Fe-57, Fe-58, Co-59, Ni-58, Ni-59, Ni-60, Ni-61, Ni-62, Ni-64, Cu-nat, Ga-nat, Ge-72, Ge-73, Ge-74, Ge-76, As-75, Se-74, Se-76, Se-77, Se-78, Se-80, Se-82, Br-79, Br-81, Kr-78, Kr-80, Kr-82, Kr-83, Kr-84, Kr-85, Kr-86, Rb-85, Rb-86, Rb-87, Sr-84, Sr-86, Sr-87, Sr-88, Sr-89, Sr-90, Y-89, Y-90, Y-91, Zr-nat, Zr-90, Zr-91, Zr-92, Zr-93, Zr-94, Zr-95, Zr-96, Nb-93, Nb-94, Nb-95, Mo-nat, Mo-92, Mo-94, Mo-95, Mo-96, Mo-97, Mo-98, Mo-99, Mo-100, Tc-99, Ru-96, Ru-98, Ru-99, Ru-100, Ru-101, Ru-102, Ru-103, Ru-104, Ru-105, Ru-106, Rh-103, Rh-105, Pd-102, Pd-104, Pd-105, Pd-106, Pd-107, Pd-108, Pd-110, Ag-107, Ag-109, Ag-111, Cd-nat., Cd-106, Cd-110, Cd-111, Cd-112, Cd-113, Cd-114, Cd-115, Cd-116, In-113, In-115, Sn-114, Sn-115, Sn-116, Sn-117, Sn-118, Sn-119, Sn-120, Sn-122, Sn-123, Sn-24, Sn-125, Sn-126, Sb-121, Sb-123, Sb-124, Sb-125, Sb-126, Te-120, Te-122, Te-123, Te-124, Te-125, Te-126, Te-127, Te-128, Te-129, Te-130, Te-132, I-127, I-129, I-130, I-131, I-135, Xe-124, Xe-126, Xe-128, Xe-129, Xe-130, Xe-131, Xe-132, Xe-133, Xe-134, Xe-135, Xe-136, Cs-133, Cs-134, Cs-135, Cs-136, Cs-137, Ba-134, Ba-135, Ba-136, Ba-137, Ba-138, Ba-140, La-139, La-140, Ce-140, Ce-141, Ce-142, Ce-143, Ce-144, Pr-141, Pr-142, Pr-143, Nd-142, Nd-143, Nd-144, Nd-145, Nd-146, Nd-147, Nd-148, Nd-150, Pm-147, Pm-148, Pm-149, Pm-151, Sm-144, Sm-147, Sm-148, Sm-149, Sm-150, Sm-151, Sm-152, Sm-153, Sm-154, Eu-151, Eu-152, Eu-153, Eu
Gollapinni, Sowjanya
2016-01-01
The study of neutrino-nucleus interactions has recently received renewed attention due to their importance in interpreting the neutrino oscillation data. Over the past few years, there has been continuous disagreement between neutrino cross section data and predictions due to lack of accurate nuclear models suitable for modern experiments which use heavier nuclear targets. Also, the current short and long-baseline neutrino oscillation experiments focus in the few GeV region where several distinct neutrino processes come into play resulting in complex nuclear effects. Despite recent efforts, more experimental input is needed to improve nuclear models and reduce neutrino interaction systematics which are currently dominating oscillation searches together with neutrino flux uncertainties. A number of new detector concepts with diverse neutrino beams and nuclear targets are currently being developed to provide necessary inputs required for next generation oscillation experiments. This paper summarizes these effor...
Group cross sections calculations
Just a few methods have been developped to compute multigroup cross-sections from ENDF data. We have developped an original method in order to get accuracy and to reduce the number of discretization points in the same time; this is why we have tried to use polynomial integration. In this paper, we describe this method: in the first part, we recall some physical hypothesis generally used to solve the linear Boltzmann equation: that is the frame in which the numerical method has been developped. Polynomial methods are really powerfull only if discretization points are suitably chosen. This choice is explained in the next part of this paper. In conclusion, some numerical results are given to illustrate our method
White, J.E.
2001-04-19
A revised multigroup cross-section library based on Release 3 of ENDF/B-VI data has been produced and tested 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 data library released in February 1994 and replaces the data package for BUGLE-93 in the Radiation Safety Information Computational Center (formerly RSIC). The processing methodology is the same as that used for producing BUGLE-93 and is consistent with ANSI/ANS 6.1.2. The ENDF data were first processed into a fine-group, pseudo-problem-independent format and then collapsed into the final broad-group format. The fine-group library, which is designated VITAMIN-B6, contains 120 nuclides. The BUGLE-96 47-neutron-group/20-gamma-ray-group library contains the same 120 nuclides processed as infinitely dilute and collapsed using a weighting spectrum typical of a concrete shield. Additionally, nuclides processed with resonance self-shielding and weighted using spectra specific to BWR and PWR material compositions and reactor models are available. As an added feature of BUGLE-96, cross-section sets having upscatter data for four thermal neutron groups are included. The upscattering data should improve the application of BUGLE-96 to the calculation of more accurate thermal fluences, although more computer time will be required. Several new dosimetry response functions and kerma factors for all 120 nuclides are also included in the library. The incorporation of feedback from users has resulted in a data library that addresses a wider spectrum of user needs.
LINX-1: a code for linking polynomial cross section files
The capabilities of the LINX-1 code are described. It was developed for the purpose of linking seperate fuel assembly and reflector node polynomial cross section files, obtained by the POLX-1 code, together into a single reactor polynomial cross section library. The output of the polynomial cross section library can be in either binary or fixed (BCD) format. Input data requirements and the format of the output file generated by LINX-1 are also described. 2 refs
Diffractive and rising cross sections
The energy dependence of the diffractive component of the proton-proton cross section is discussed and its contribution to the rise of the total cross section at high energies is examined. 17 refs., 9 figs
[Fast neutron cross section measurements
This paper discusses the following topics: 14 MeV pulsed neutron facility; detection and measurement system; 238U capture cross sections at 23 and 964 keV using photon neutron sources; capture cross sections of Au-197 at 23 and 964 keV; and yttrium nuclear cross section measurement
Pescarini, Massimo; Orsi, Roberto; Frisoni, Manuela
2016-02-01
The PCA-Replica 12/13 (H2O/Fe) neutron shielding benchmark experiment was analysed using the ORNL TORT-3.2 3D SN code. PCA-Replica, specifically conceived to test the accuracy of nuclear data and transport codes employed in LWR shielding and radiation damage calculations, reproduces a PWR ex-core radial geometry with alternate layers of water and steel including a PWR pressure vessel simulator. Three broad-group coupled neutron/photon working cross section libraries in FIDO-ANISN format with the same energy group structure (47 n + 20 γ) and based on different nuclear data were alternatively used: the ENEA BUGJEFF311.BOLIB (JEFF-3.1.1) and BUGENDF70.BOLIB (ENDF/B-VII.0) libraries and the ORNL BUGLE-96 (ENDF/B-VI.3) library. Dosimeter cross sections derived from the IAEA IRDF-2002 dosimetry file were employed. The calculated reaction rates for the Rh-103(n,n')Rh-103 m, In-115(n,n')In-115m and S-32(n,p)P-32 threshold activation dosimeters and the calculated neutron spectra are compared with the corresponding experimental results.
Pescarini Massimo
2016-01-01
Full Text Available The PCA-Replica 12/13 (H2O/Fe neutron shielding benchmark experiment was analysed using the ORNL TORT-3.2 3D SN code. PCA-Replica, specifically conceived to test the accuracy of nuclear data and transport codes employed in LWR shielding and radiation damage calculations, reproduces a PWR ex-core radial geometry with alternate layers of water and steel including a PWR pressure vessel simulator. Three broad-group coupled neutron/photon working cross section libraries in FIDO-ANISN format with the same energy group structure (47 n + 20 γ and based on different nuclear data were alternatively used: the ENEA BUGJEFF311.BOLIB (JEFF-3.1.1 and BUGENDF70.BOLIB (ENDF/B-VII.0 libraries and the ORNL BUGLE-96 (ENDF/B-VI.3 library. Dosimeter cross sections derived from the IAEA IRDF-2002 dosimetry file were employed. The calculated reaction rates for the Rh-103(n,n′Rh-103 m, In-115(n,n′In-115m and S-32(n,pP-32 threshold activation dosimeters and the calculated neutron spectra are compared with the corresponding experimental results.
Pescarini Massimo; Orsi Roberto; Frisoni Manuela
2016-01-01
The PCA-Replica 12/13 (H2O/Fe) neutron shielding benchmark experiment was analysed using the ORNL TORT-3.2 3D SN code. PCA-Replica, specifically conceived to test the accuracy of nuclear data and transport codes employed in LWR shielding and radiation damage calculations, reproduces a PWR ex-core radial geometry with alternate layers of water and steel including a PWR pressure vessel simulator. Three broad-group coupled neutron/photon working cross section libraries in FIDO-ANISN format with ...
Atlas of neutron capture cross sections
This report describes neutron capture cross sections in the range 10-5 eV - 20 MeV as evaluated and compiled in recent activation libraries. The selected subset comprise the (n,γ) cross sections for a total of 739 targets for the elements H (Z = 1, Z = 1) to Cm (Z = 96, A = 238) totaling 972 reactions. Plots of the point-wise data are shown and comparisons are made with the available experimental values at thermal energy, 30 keV and 14.5 MeV. 10 refs, 7 tabs
Jordan, W.C.
1993-02-01
A version of KENO V.a and the 27-group library in SCALE-4.0 were validated for use in evaluating the nuclear criticality safety of low-enriched uranium systems. A total of 59 critical systems were analyzed. A statistical analysis of the results was performed, and subcritical acceptanced criteria are established.
Jordan, W.C.
1993-02-01
A version of KENO V.a and the 27-group library in SCALE-4.0 were validated for use in evaluating the nuclear criticality safety of low-enriched uranium systems. A total of 59 critical systems were analyzed. A statistical analysis of the results was performed, and subcritical acceptanced criteria are established.
Verification of important cross section data
Full text: Continuing efforts in nuclear data development have made the design of a fusion power system less uncertain. The fusion evaluated nuclear data library (FENDL) development effort since 1987 under the leadership of the IAEA Nuclear Data Section has provided a credible international library for the investigation and design of the International Thermonuclear Engineering Reactor (ITER). Integral neutronics experiments are being carried out for ITER and fusion power plant blanket and shield assemblies to validate the available nuclear database and to identify deficiencies for further improvement. Important cross section data need experimental verifications if these data are evaluated based on physics model calculations and there are no measured data points available. A particular reaction cross section is Si28(n,x)Al27, which is the important cross section to determine whether the low activation SiC composite structure can be qualified as low level nuclear waste after life time exposure in the first wall neutron environment in a fusion power plant. Measurements of helium production data for candidate fusion materials are also needed, particularly at energies above 14 MeV for the assessment of materials damage in the IFMIF neutron spectrum. To a less extent, it appears that V51(n,x)Ti50 reaction cross section also needs to be measured to further confirm a recent new evaluation of vanadium for ENDF/B-VII. (author)
Most of the fission products and a few of the actinides in ENDF/B-V do not have (n,2n) cross sections. A complete set of these cross sections is presented in the multigroup structure defined. These were constructed for future use in the DANDE Code System
XCOM: Photon Cross Sections Database
SRD 8 XCOM: Photon Cross Sections Database (Web, free access) A web database is provided which can be used to calculate photon cross sections for scattering, photoelectric absorption and pair production, as well as total attenuation coefficients, for any element, compound or mixture (Z <= 100) at energies from 1 keV to 100 GeV.
Cross Sections and Lorentz Violation
Colladay, Don; Kostelecky, Alan
2001-01-01
The derivation of cross sections and decay rates in the Lorentz-violating standard-model extension is discussed. General features of the physics are described, and some conceptual and calculational issues are addressed. As an illustrative example, the cross section for the specific process of electron-positron pair annihilation into two photons is obtained.
Simplified polynomial representation of cross sections for reactor calculation
It is shown a simplified representation of a cross section library generated by transport theory using the cell model of Wigner-Seitz for typical PWR fuel elements. The effect of burnup evolution through tables of reference cross sections and the effect of the variation of the reactor operation parameters considered by adjusted polynomials are presented. (M.C.K.)
Nuclear characteristics of Pu fueled LWR and cross section sensitivities
Takeda, Toshikazu [Osaka Univ., Suita (Japan). Faculty of Engineering
1998-03-01
The present status of Pu utilization to thermal reactors in Japan, nuclear characteristics and topics and cross section sensitivities for analysis of Pu fueled thermal reactors are described. As topics we will discuss the spatial self-shielding effect on the Doppler reactivity effect and the cross section sensitivities with the JENDL-3.1 and 3.2 libraries. (author)
The report contains the Summary of the First IAEA Research Co-ordination Meeting (RCM) of the new Co-ordinated Research Programme (CRP) on ''Establishment of an International Reference Data Library of Nuclear Activation Cross Sections''. The meeting was organized by the IAEA Nuclear Data Section with co-operation and assistance of local organizers from the Institute of Experimental Physics and held in Debrecen, Hungary, from 4 to 7 October 1994. The purpose of the RCM was to discuss the scope and goals of the CRP, to report and evaluate the first results of the research carried out by each participating laboratory, to review the current tasks, identify further actions of participants and agree on the coordination of work under this CRP. The detailed agenda, the list of participants, conclusions and recommendations of the meeting are presented in the summary report. (author)
Energy-dependent evaluated photon interaction cross sections and related parameters are presented for elements H through Cf(Z = 1 to 98). Data are given over the energy range from 0.1 keV to 100 MeV. The related parameters include form factors and average energy deposits per collision (with and without fluorescence). Fluorescence information is given for all atomic shells that can emit a photon with a kinetic energy of 0.1 keV or more. In addition, the following macroscopic properties are given: total mean free path and energy deposit per centimeter. This information is derived from the Livermore Evaluated-Nuclear-Data Library (ENDL) as of October 1978
Positive Scattering Cross Sections using Constrained Least Squares
A method which creates a positive Legendre expansion from truncated Legendre cross section libraries is presented. The cross section moments of order two and greater are modified by a constrained least squares algorithm, subject to the constraints that the zeroth and first moments remain constant, and that the standard discrete ordinate scattering matrix is positive. A method using the maximum entropy representation of the cross section which reduces the error of these modified moments is also presented. These methods are implemented in PARTISN, and numerical results from a transport calculation using highly anisotropic scattering cross sections with the exponential discontinuous spatial scheme is presented
Positive Scattering Cross Sections using Constrained Least Squares
Dahl, J.A.; Ganapol, B.D.; Morel, J.E.
1999-09-27
A method which creates a positive Legendre expansion from truncated Legendre cross section libraries is presented. The cross section moments of order two and greater are modified by a constrained least squares algorithm, subject to the constraints that the zeroth and first moments remain constant, and that the standard discrete ordinate scattering matrix is positive. A method using the maximum entropy representation of the cross section which reduces the error of these modified moments is also presented. These methods are implemented in PARTISN, and numerical results from a transport calculation using highly anisotropic scattering cross sections with the exponential discontinuous spatial scheme is presented.
Development of automatic cross section compilation system for MCNP
A development of a code system to automatically convert cross-sections for MCNP is in progress. The NJOY code is, in general, used to convert the data compiled in the ENDF format (Evaluated Nuclear Data Files by BNL) into the cross-section libraries required by various reactor physics codes. While the cross-section library: FSXLIB-J3R2 was already converted from the JENDL-3.2 version of Japanese Evaluated Nuclear Data Library for a continuous energy Monte Carlo code MCNP, the library keeps only the cross-sections at room temperature (300 K). According to the users requirements which want to have cross-sections at higher temperature, say 600 K or 900 K, a code system named 'autonj' is under development to provide a set of cross-section library of arbitrary temperature for the MCNP code. This system can accept any of data formats adopted JENDL that may not be treated by NJOY code. The input preparation that is repeatedly required at every nuclide on NJOY execution is greatly reduced by permitting the conversion process of as many nuclides as the user wants in one execution. A few MCNP runs were achieved for verification purpose by using two libraries FSXLIB-J3R2 and the output of autonj'. The almost identical MCNP results within the statistical errors show the 'autonj' output library is correct. In FY 1998, the system will be completed, and in FY 1999, the user's manual will be published. (K. Tsuchihashi)
A detailed three-dimensional, continuous-energy MCNP4B model of the LWR-PROTEUS critical facility has been developed for the analysis of whole-reactor characteristics using ENDF/B-V, ENDF/B-VI and JEF-2.2 cross-section sets. The model has been applied to the determination of the critical loading, as well as the evaluation of reactivity worths for safety/shutdown rods, control rods, and individual driver-region fuel rods. The initially obtained results for the first configuration investigated (Core 1B) indicated that, for the same geometrical and materials specifications, the ENDF/B-V data library yields the closest critical prediction (discrepancy of 640±40 pcm), followed by ENDF/B-VI (980±40 pcm) and JEF-2.2 (1340±40 pcm). The differences in results between the three data libraries were studied by considering the contributions of individual materials to the neutron balance. 235U and 238U cross-sections from JEF-2.2, for example, explain an effect of ∼400 pcm. Refinement of the materials specifications in the MCNP4B whole-reactor model, in particular the impurities assumed for the graphite driver of the driver and reflector regions, has been shown to reduce the final discrepancy of the ENDF/B-V based keff result to ∼0.2%. The MCNP4B results for relative reactivity effects, such as control rod worths, are found to agree within experimental errors with the measured values
Measurement of fission cross sections
A review is presented on the recent progress in the experiment of fission cross section measurement, including recent activity in Japan being carried out under the project of nuclear data measurement. (author)
R. Vogt
2007-01-01
We assess the theoretical uncertainties on the total charm cross section. We discuss the importance of the quark mass, the scale choice and the parton densities on the estimate of the uncertainty. We conclude that due to the small charm quark mass, which amplifies the effect of the other parameters in the calculation, the uncertainty on the total charm cross section is difficult to quantify.
Thanh Mai Vu; Takanori Kitada
2014-01-01
Recently, the researches on fast neutron spectrum system utilized thorium fuel are widely conducted. However, the recent thorium cross section libraries are limited compared to uranium cross section libraries. The impact of thorium cross section uncertainty on thorium fuel utilized accelerator driven system (ADS) reactivity calculation is estimated in this study. The uncertainty of the keff caused by 232Th capture cross section of JENDL-4.0 is about 1.3%. The uncertainty of JENDL-4.0 is neede...
Revolutionizing Cross-sectional Imaging
Fan, Yifang; Luo, Liangping; Lin, Wentao; Li, Zhiyu; Zhong, Xin; Shi, Changzheng; Newman, Tony; Zhou, Yi; Lv, Changsheng; Fan, Yuzhou
2014-01-01
Cross-sectional imaging is so important that, six Nobel Prizes have been awarded to the field of nuclear magnetic resonance alone because it revolutionized clinical diagnosis. The BigBrain project supported by up to 1 billion euro each over a time period of 10 years predicts to "revolutionize our ability to understand internal brain organization" (Evan 2013). If we claim that cross-sectional imaging diagnosis is only semi-quantitative, some may believe because no doctor would ever tell their patient that we can observe the changes of this cross-sectional image next time. If we claim that BigBrain will make no difference in clinical medicine, then few would believe because no doctor would ever tell their patient to scan this part of the image and compare it with that from the BigBrain. If we claim that the BigBrain Project and the Human Brain Project have defects in their key method, one might believe it. But this is true. The key lies in the reconstruction of any cross-sectional image along any axis. Using Ga...
Terahertz radar cross section measurements
Iwaszczuk, Krzysztof; Heiselberg, Henning; Jepsen, Peter Uhd
2010-01-01
We perform angle- and frequency-resolved radar cross section (RCS) measurements on objects at terahertz frequencies. Our RCS measurements are performed on a scale model aircraft of size 5-10 cm in polar and azimuthal configurations, and correspond closely to RCS measurements with conventional radar...
Cross sections for nuclear astrophysics
General properties of low-energy cross sections and of reaction rates are presented. We describe different models used in nuclear astrophysics: microscopic models, the potential model, and the R-matrix method. Two important reactions, 7Be(p,γ)8B and 12C(α,γ)16O, are then briefly discussed. (author)
International evaluation cooperation Subgroup 7: Multigroup cross section processing
Roussin, R.W.; White, J.E. (Oak Ridge National Lab., TN (USA)); Sartori, E. (NEA Data Bank, 91 - Gif-sur-Yvette (France)); Panini, G. (ENEA, Bologna (Italy)); MacFarlane, R. (Los Alamos National Lab., NM (USA)); Muir, D. (International Atomic Energy Agency, Vienna (Austria). Nuclear Data Section); Mattes, M. (Stuttgart Univ. (Germany, F.R.). Inst. fuer Kernenergetik und Energiesysteme); Hasegawa, I
1991-01-01
The chairmen of the ENDF/B, JEF, EFF, and JENDL evaluated data files adopted a proposal to develop a fine-group processed cross section library based on the VITAMIN'' concept. The authors listed above, with support from others, are participating in this project. The end result will be a pseudo-problem-independent fine-group cross section library generated from the latest evaluated data in ENDF/B-VI, JEF-2, EFF-2, and JENDL-3. Initial applications of the library will be for shielding, fast reactor physics, and fusion neutronics. Progress made to date will be discussed. 8 refs.
Cross Sections for Inner-Shell Ionization by Electron Impact
An analysis is presented of measured and calculated cross sections for inner-shell ionization by electron impact. We describe the essentials of classical and semiclassical models and of quantum approximations for computing ionization cross sections. The emphasis is on the recent formulation of the distorted-wave Born approximation by Bote and Salvat [Phys. Rev. A 77, 042701 (2008)] that has been used to generate an extensive database of cross sections for the ionization of the K shell and the L and M subshells of all elements from hydrogen to einsteinium (Z = 1 to Z = 99) by electrons and positrons with kinetic energies up to 1 GeV. We describe a systematic method for evaluating cross sections for emission of x rays and Auger electrons based on atomic transition probabilities from the Evaluated Atomic Data Library of Perkins et al. [Lawrence Livermore National Laboratory, UCRL-ID-50400, 1991]. We made an extensive comparison of measured K-shell, L-subshell, and M-subshell ionization cross sections and of Lα x-ray production cross sections with the corresponding calculated cross sections. We identified elements for which there were at least three (for K shells) or two (for L and M subshells) mutually consistent sets of cross-section measurements and for which the cross sections varied with energy as expected by theory. The overall average root-mean-square deviation between the measured and calculated cross sections was 10.9% and the overall average deviation was −2.5%. This degree of agreement between measured and calculated ionization and x-ray production cross sections was considered to be very satisfactory given the difficulties of these measurements
Evaluation of neutron reaction cross sections for astrophysics
We have developed a code system to evaluate nuclear reaction cross sections for the nucleosynthesis. The system includes an interface to Reference Input Parameter Library (RIPL), as well as some systematics to extrapolate the parameters into unstable regions. We are focusing on neutron capture processes important for s- and r-processes. The structure of the system is reviewed, and calculated capture cross sections in the fission product mass region are compared with experimental data available. (author)
Parametric equations for calculation of macroscopic cross sections
Botelho, Mario Hugo; Carvalho, Fernando, E-mail: mariobotelho@poli.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear
2015-07-01
Neutronic calculations of the core of a nuclear reactor is one thing necessary and important for the design and management of a nuclear reactor in order to prevent accidents and control the reactor efficiently as possible. To perform these calculations a library of nuclear data, including cross sections is required. Currently, to obtain a cross section computer codes are used, which require a large amount of processing time and computer memory. This paper proposes the calculation of macroscopic cross section through the development of parametric equations. The paper illustrates the proposal for the case of macroscopic cross sections of absorption (Σa), which was chosen due to its greater complexity among other cross sections. Parametric equations created enable, quick and dynamic way, the determination of absorption cross sections, enabling the use of them in calculations of reactors. The results show efficient when compared with the absorption cross sections obtained by the ALPHA 8.8.1 code. The differences between the cross sections are less than 2% for group 2 and less than 0.60% for group 1. (author)
Metonymy and Cross Section Demand
Evstigneev, Igor V.; Hildenbrand, Werner; Jerison, Michael
1996-01-01
Cross section consumer expenditure data are frequently used to make conclusions about consumer demand behavior. Such conclusions, however, can only be justified under certain assumptions, which are often left unstated in the empirical demand literature. An assumption of this type, the metonymy hypothesis, was stated rigorously and then exploited by Hardle, Hildenbrand and Jerison when analyzing the monotonicity property of aggregate demand functions. The purpose of the present paper is to exa...
Wind Turbine Radar Cross Section
David Jenn; Cuong Ton
2012-01-01
The radar cross section (RCS) of a wind turbine is a figure of merit for assessing its effect on the performance of electronic systems. In this paper, the fundamental equations for estimating the wind turbine clutter signal in radar and communication systems are presented. Methods of RCS prediction are summarized, citing their advantages and disadvantages. Bistatic and monostatic RCS patterns for two wind turbine configurations, a horizontal axis three-blade design and a vertical axi...
Microscopic cross sections: An utopia?
Hilaire, S. [CEA Bruyeres-le-Chatel, DIF 91 (France); Koning, A.J. [Nuclear Research and Consultancy Group, PO Box 25, 1755 ZG Petten (Netherlands); Goriely, S. [Institut d' Astronomie et d' Astrophysique, Universite Libre de Bruxelles, Campus de la Plaine, CP 226, 1050 Brussels (Belgium)
2010-07-01
The increasing need for cross sections far from the valley of stability poses a challenge for nuclear reaction models. So far, predictions of cross sections have relied on more or less phenomenological approaches, depending on parameters adjusted to available experimental data or deduced from systematical relations. While such predictions are expected to be reliable for nuclei not too far from the experimentally known regions, it is clearly preferable to use more fundamental approaches, based on sound physical bases, when dealing with very exotic nuclei. Thanks to the high computer power available today, all major ingredients required to model a nuclear reaction can now be (and have been) microscopically (or semi-microscopically) determined starting from the information provided by a nucleon-nucleon effective interaction. We have implemented all these microscopic ingredients in the TALYS nuclear reaction code, and we are now almost able to perform fully microscopic cross section calculations. The quality of these ingredients and the impact of using them instead of the usually adopted phenomenological parameters will be discussed. (authors)
[Fast neutron cross section measurements
In this report, we outline the progress achieved in two distinct under the DOE-sponsored cross section project: the initial results obtained from the pulsed 14 MeV neutron facility, and a cooperative effort with Argonne National Laboratory in the measurement of fast neutron cross sections in yttrium. In the 14 MeV neutron laboratory, this year has seen the maturation of the project into one in which initial scattering measurements are now underway. We have improved the accelerator and ion source in several significant ways, so that neutron intensities have now been proven to be adequate for our series of elastic scattering angular distribution measurements outlined in our initial proposal of two years ago. We have successfully tested all components of the time-of-flight spectrometer and recorded initial neutron spectra from the ring targets that we have obtained for our first angular distribution measurements. Examples of the time-of-flight spectra that have been obtained are given later in this report. At the present time, the accelerator is operating with the highest degree of reliability that we have experienced since installing the pulsing system. Improvements made over the past year have not only increased the available neutron intensity, but also increased our capability to deal with inevitable component failures that require repair or replacement. The measurements carried out in conjunction with Argonne have contributed significantly to the available database on fast neutron interactions in yttrium. Results indicate that the cross section for the 89 Y(n,p)89Sr reaction is substantially higher than represented in ENDF/B-VI
Wind Turbine Radar Cross Section
David Jenn
2012-01-01
Full Text Available The radar cross section (RCS of a wind turbine is a figure of merit for assessing its effect on the performance of electronic systems. In this paper, the fundamental equations for estimating the wind turbine clutter signal in radar and communication systems are presented. Methods of RCS prediction are summarized, citing their advantages and disadvantages. Bistatic and monostatic RCS patterns for two wind turbine configurations, a horizontal axis three-blade design and a vertical axis helical design, are shown. The unique electromagnetic scattering features, the effect of materials, and methods of mitigating wind turbine clutter are also discussed.
Porosity effects in the neutron total cross section of graphite
Graphite has been used in nuclear reactors since the birth of the nuclear industry due to its good performance as a neutron moderator material. Graphite is still an option as moderator for generation IV reactors due to its good mechanical and thermal properties at high operation temperatures. So, there has been renewed interest in a revision of the computer libraries used to describe the neutron cross section of graphite. For sub-thermal neutron energies, polycrystalline graphite shows a larger total cross section (between 4 and 8 barns) than predicted by existing theoretical models (0.2 barns). In order to investigate the origin of this discrepancy we measured the total cross section of graphite samples of three different origins, in the energy range from 0.001 eV to 10 eV. Different experimental arrangements and sample treatments were explored, to identify the effect of various experimental parameters on the total cross section measurement. The experiments showed that the increase in total cross section is due to neutrons scattered around the forward direction. We associate these small-angle scattered neutrons (SANS) to the porous structure of graphite, and formulate a very simple model to compute its contribution to the total cross section of the material. This results in an analytic expression that explicitly depends on the density and mean size of the pores, which can be easily incorporated in nuclear library codes.
Neutron standard cross sections in reactor physics - Need and status
The design and improvement of nuclear reactors require detailed neutronics calculations. These calculations depend on comprehensive libraries of evaluated nuclear cross sections. Most of the cross sections that form the data base for these evaluations have been measured relative to neutron cross-section standards. The use of these standards can often simplify the measurement process by eliminating the need for a direct measurement of the neutron fluence. The standards are not known perfectly, however; thus the accuracy of a cross-section measurement is limited by the uncertainty in the standard cross section relative to which it is measured. Improvements in a standard cause all cross sections measured relative to that standard to be improved. This is the reason for the emphasis on improving the neutron cross-section standards. The continual process of measurement and evaluation has led to improvements in the accuracy and range of applicability of the standards. Though these improvements have been substantial, this process must continue in order to obtain the high-quality standards needed by the user community
Windowed multipole for cross section Doppler broadening
Josey, C.; Ducru, P.; Forget, B.; Smith, K.
2016-02-01
This paper presents an in-depth analysis on the accuracy and performance of the windowed multipole Doppler broadening method. The basic theory behind cross section data is described, along with the basic multipole formalism followed by the approximations leading to windowed multipole method and the algorithm used to efficiently evaluate Doppler broadened cross sections. The method is tested by simulating the BEAVRS benchmark with a windowed multipole library composed of 70 nuclides. Accuracy of the method is demonstrated on a single assembly case where total neutron production rates and 238U capture rates compare within 0.1% to ACE format files at the same temperature. With regards to performance, clock cycle counts and cache misses were measured for single temperature ACE table lookup and for windowed multipole. The windowed multipole method was found to require 39.6% more clock cycles to evaluate, translating to a 7.9% performance loss overall. However, the algorithm has significantly better last-level cache performance, with 3 fewer misses per evaluation, or a 65% reduction in last-level misses. This is due to the small memory footprint of the windowed multipole method and better memory access pattern of the algorithm.
Electron-Impact Ionization Cross Section Database
SRD 107 Electron-Impact Ionization Cross Section Database (Web, free access) This is a database primarily of total ionization cross sections of molecules by electron impact. The database also includes cross sections for a small number of atoms and energy distributions of ejected electrons for H, He, and H2. The cross sections were calculated using the Binary-Encounter-Bethe (BEB) model, which combines the Mott cross section with the high-incident energy behavior of the Bethe cross section. Selected experimental data are included.
[Fast neutron cross section measurements
In the 14 MeV Neutron Laboratory, we have continued the development of a facility that is now the only one of its kind in operation in the United States. We have refined the klystron bunching system described in last year's report to the point that 1.2 nanosecond pulses have been directly measured. We have tested the pulse shape discrimination capability of our primary NE 213 neutron detector. We have converted the RF sweeper section of the beamline to a frequency of 1 MHz to replace the function of the high voltage pulser described in last year's report which proved to be difficult to maintain and unreliable in its operation. We have also overcome several other significant experimental difficulties, including a major problem with a vacuum leak in the main accelerator column. We have completed additional testing to prove the remainder of the generation and measurement systems, but overcoming some of these experimental difficulties has delayed the start of actual data taking. We are now in a position to begin our first series of ring geometry elastic scattering measurements, and these will be underway before the end of the current contract year. As part of our longer term planning, we are continuing the conceptual analysis of several schemes to improve the intensity of our current pulsed beam. These include the provision of a duoplasmatron ion source and/or the provision of preacceleration bunching. Additional details are given later in this report. A series of measurements were carried out at the Tandem Dynamatron Facility involving the irradiation of a series of yttrium foils and the determination of activation cross sections using absolute counting techniques. The experimental work has been completed, and final analysis of the cross section data will be completed within several months
Reference solution for cross section parametrization
Core calculations of nuclear reactors are usually performed by core physics codes (e.g. with NEM or FDM solvers) in diffusion or SP3 approximation of the transport equation. For each fuel type parameterized data libraries are prepared by means of a lattice code. The data libraries are burnup dependent, and the parameterization covers the hyperspace of admissible values of all operational parameters (fuel temperature, moderator density, boron concentration etc.) This approach has two weak spots. The first is, that it is difficult to make perfect parameterization of the data library because of relatively broad range of the parameter values and the fact that the parameters' effect on the macroscopic cross-sections are not mutually independent. The second is that even for perfect parameterizations with precise approximations of the data changes with respect to the feedback parameters the so-called history effects are neglected. It is generally difficult to assess the cumulative errors arising due to the approximative parameterization of the data libraries and due to the history effects. It is as well difficult to assess the efficiency of techniques developed in order to incorporate the history effect in the data library (such as time integration). In this paper we present a tool for reference core calculations in which the above stated approximations are eliminated. This paper presents the solution method, its implementation, as well as the results of a demonstration calculation showing the improvement of the calculation results over the traditional approach, assessing the magnitude of history and parameterization effects importance. The most important feature of the presented method is that it provides the perfect parameterization of macroscopic data, allowing the core physics code developers to understand sources of modeling uncertainties by completely removing the parameterization error (including, unlike other approaches, a complete representation of the
Impact of the ENDF/B-VI Cross Section on the RPV Fluence Determination
The calculations with the broad-group cross-section library Bugle-96, and atom displacement (dpa) cross sections for iron, both derived from ENDF/B-VI data, result in higher calculated fast neutron fluxes, better agreement of calculations with radiometric dosimeter measurements, and significantly slower dpa rate attenuation through pressure vessel walls relative to the results with their predecessors: the Sailor library and ASTM iron dpa cross sections
Evaluation of cross section for 103Rh
A completely new evaluation for the neutron cross sections is presented. The experimental data mainly referred to EXFOR, and the recommended cross sections are compared with ENDF/B-6, BROND-2, JENDL-3.2 and JEF-2
Cross section generation for LWR pin lattices simulations
Velasquez, Carlos E.; Macedo, Anderson A.P.; Cardoso, Fabiano S.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Costa, Antonella L. [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil); Instituto Nacional de Ciencia e Tecnologia de Reatores Nucleares Inovadores/CNPq, Brasilia, DF (Brazil); Barros, Graiciany de P. [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)
2015-07-01
The majority of the neutron data library provided with the MCNP code is set at room temperature. Therefore, it is important to generate continuous energy cross section library for MCNP code for different temperatures. To evaluate the methodology used, the criticality calculations obtained using MCNP with the cross section generated at DEN/UFMG, are compared with the criticality data from the International Handbook of Evaluated Reactor Physics Benchmarks Experiments about the PIN lattices for light water reactors. It was used nuclear data from the ENDF-VII.1, JEFF-3.1 and TENDL-2014, which were processed using the NJOY99 code for different energies and temperatures. This code provides the nuclear data in ACE libraries, which then are added to MCNP libraries to perform the simulations. The results indicate the methodology efficiency developed by DEN/UFMG. (author)
Photoproduction total cross section and shower development
Cornet, F.; García Canal, C. A.; Grau, A.; Pancheri, G.; Sciutto, S. J.
2015-12-01
The total photoproduction cross section at ultrahigh energies is obtained using a model based on QCD minijets and soft-gluon resummation and the ansatz that infrared gluons limit the rise of total cross sections. This cross section is introduced into the Monte Carlo system AIRES to simulate extended air showers initiated by cosmic ray photons. The impact of the new photoproduction cross section on common shower observables, especially those related to muon production, is compared with previous results.
Photoproduction total cross section and shower development
Cornet, F; Grau, A; Pancheri, G; Sciutto, S J
2015-01-01
The total photoproduction cross section at ultra-high energies is obtained using a model based on QCD minijets and soft-gluon resummation and the ansatz that infrared gluons limit the rise of total cross sections. This cross section is introduced into the Monte Carlo system AIRES to simulate extended air-showers initiated by cosmic ray photons. The impact of the new photoproduction cross section on common shower observables, especially those related to muon production, is compared with previous results.
Plots of the experimental and evaluated photoneutron cross-sections
Graphical plots of experimental data of photon induced nuclear reaction cross-sections are given for many elements and isotopes. The numerical data were taken from the international EXFOR data library which is available from the nuclear data centers. For selected nuclides evaluated data have been included in the plots. (author). Refs, 3 tabs
Generation of neutron scattering cross sections for silicon dioxide
A set of neutron scattering cross sections for silicon and oxygen bound in silicon dioxide were generated and validated. The cross sections were generated in the ACE format for MCNP using the nuclear data processing system NJOY, and the validation was done with published experimental data. This cross section library was applied to the calculation of five critical configurations published in the benchmark Critical Experiments with Heterogeneous Compositions of Highly Enriched Uranium, Silicon Dioxide and Polyethylene. The original calculations did not use the thermal scattering libraries generated in this work and presented significant differences with the experimental results. For this reason, the newly generated library was added to the input and the multiplication factor for each configuration was recomputed. The utilization of the thermal scattering libraries did not result in an improvement of the computational results. Based on this we conclude that integral experiments to validate this type of thermal cross sections need to be designed with a higher influence of thermal scattering in the measured result, and the experiments have to be performed under more controlled conditions.
JENDL gas-production cross section file
The JENDL gas-production cross section file was compiled by taking cross-section data from JENDL-3 and by using the ENDF-5 format. The data were given to 23 nuclei or elements in light nuclei and structural materials. Graphs of the cross sections and brief description on their evaluation methods are given in this report. (author)
Status of multigroup cross-section data for shielding applications
Multigroup cross-section libraries for shielding applications in formats for direct use in discrete ordinates or Monte Carlo codes have long been a part of the Data Library Collection (DLC) of the Radiation Shielding Information Center (RSIC). In recent years libraries in more flexible and comprehensive formats, which allow the user to derive his own problem-dependent sets, have been added to the collection. The current status of both types is described, as well as projections for adding data libraries based on ENDF/B-V
[Fast neutron cross section measurements
From its inception, the Nuclear Data Project at the University of Michigan has concentrated on two major objectives: (1) to carry out carefully controlled nuclear measurements of the highest possible reliability in support of the national nuclear data program, and (2) to provide an educational opportunity for students with interests in experimental nuclear science. The project has undergone a successful transition from a primary dependence on our photoneutron laboratory to one in which our current research is entirely based on a unique pulsed 14 MeV fast neutron facility. The new experimental facility is unique in its ability to provide nanosecond bursts of 14 MeV neutrons under conditions that are ''clean'' and as scatter-free as possible, and is the only one of its type currently in operation in the United States. It has been designed and put into operation primarily by graduate students, and has met or exceeded all of its important initial performance goals. We have reached the point of its routine operation, and most of the data are now in hand that will serve as the basis for the first two doctoral dissertations to be written by participating graduate students. Our initial results on double differential neutron cross sections will be presented at the May 1993 Fusion Reactor Technology Workshop. We are pleased to report that, after investing several years in equipment assembly and optimization, the project has now entered its ''data production'' phase
Thanh Mai Vu
2014-01-01
Full Text Available Recently, the researches on fast neutron spectrum system utilized thorium fuel are widely conducted. However, the recent thorium cross section libraries are limited compared to uranium cross section libraries. The impact of thorium cross section uncertainty on thorium fuel utilized accelerator driven system (ADS reactivity calculation is estimated in this study. The uncertainty of the keff caused by 232Th capture cross section of JENDL-4.0 is about 1.3%. The uncertainty of JENDL-4.0 is needed to be enhanced to provide more reliable results on reactivity calculation for the fast system. The impact of uncertainty of 232Th capture cross section of ENDF/B-VII is small (0.1%. Therefore, it will cause no significant impact of the thorium cross section library on the thorium utilized ADS design calculation.
Recent fission cross section standards measurements
Wasson, O.A.
1985-01-01
The /sup 235/U(n,f) reaction is the standard by which most neutron induced fission cross sections are determined. Most of these cross sections are derived from relatively easy ratio measurements to /sup 235/U. However, the more difficult /sup 235/U(n,f) cross section measurements require the use of advanced neutron detectors for the determination of the incident neutron fluence. Examples of recent standard cross section measurements are discussed, various neutron detectors are described, and the status of the /sup 235/U(n,f) cross section standard is assessed. 23 refs., 8 figs., 4 tabs.
Recent fission cross section standards measurements
The 235U(n,f) reaction is the standard by which most neutron induced fission cross sections are determined. Most of these cross sections are derived from relatively easy ratio measurements to 235U. However, the more difficult 235U(n,f) cross section measurements require the use of advanced neutron detectors for the determination of the incident neutron fluence. Examples of recent standard cross section measurements are discussed, various neutron detectors are described, and the status of the 235U(n,f) cross section standard is assessed. 23 refs., 8 figs., 4 tabs
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.
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
Integral test of fission-product cross sections
A test of more than 50 nuclides of the fission-product file of the JEF-1 data library has been performed, using integral data measured in Dutch, French and US facilities. Some results are given for the capture cross sections of the 40 most important fission products in a fast reactor. The inelastic scattering cross sections of many even-mass nuclides are systematically too low due to neglect of direct-collective effects. In lumped fission-product cross sections the uncertainties due to the release of gaseous products have been reduced by means of a new burn-up model with parameters tuned to leakage data of irradiated PHENIX fuel pins
Ionization cross sections for low energy electron transport
Seo, Hee; Saracco, Paolo; Kim, Chan Hyeong
2011-01-01
Two models for the calculation of ionization cross sections by electron impact on atoms, the Binary-Encouter-Bethe and the Deutsch-Maerk models, have been implemented; they are intended to extend and improve Geant4 simulation capabilities in the energy range below 1 keV. The physics features of the implementation of the models are described, and their differences with respect to the original formulations are discussed. Results of the verification with respect to the original theoretical sources and of extensive validation with respect to experimental data are reported. The validation process also concerns the ionization cross sections included in the Evaluated Electron Data Library used by Geant4 for low energy electron transport. Among the three cross section options, the Deutsch-Maerk model is identified as the most accurate at reproducing experimental data over the energy range subject to test.
Background-cross-section-dependent subgroup parameters
A new set of subgroup parameters was derived that can reproduce the self-shielded cross section against a wide range of background cross sections. The subgroup parameters are expressed with a rational equation which numerator and denominator are expressed as the expansion series of background cross section, so that the background cross section dependence is exactly taken into account in the parameters. The advantage of the new subgroup parameters is that they can reproduce the self-shielded effect not only by group basis but also by subgroup basis. Then an adaptive method is also proposed which uses fitting procedure to evaluate the background-cross-section-dependence of the parameters. One of the simple fitting formula was able to reproduce the self-shielded subgroup cross section by less than 1% error from the precise evaluation. (author)
Vertically stabilized elongated cross-section tokamak
Sheffield, George V.
1977-01-01
This invention provides a vertically stabilized, non-circular (minor) cross-section, toroidal plasma column characterized by an external separatrix. To this end, a specific poloidal coil means is added outside a toroidal plasma column containing an endless plasma current in a tokamak to produce a rectangular cross-section plasma column along the equilibrium axis of the plasma column. By elongating the spacing between the poloidal coil means the plasma cross-section is vertically elongated, while maintaining vertical stability, efficiently to increase the poloidal flux in linear proportion to the plasma cross-section height to achieve a much greater plasma volume than could be achieved with the heretofore known round cross-section plasma columns. Also, vertical stability is enhanced over an elliptical cross-section plasma column, and poloidal magnetic divertors are achieved.
SCAMPI: A code package for cross-section processing
Parks, C.V.; Petrie, L.M.; Bowman, S.M.; Broadhead, B.L.; Greene, N.M.; White, J.E.
1996-04-01
The SCAMPI code package consists of a set of SCALE and AMPX modules that have been assembled to facilitate user needs for preparation of problem-specific, multigroup cross-section libraries. The function of each module contained in the SCANTI code package is discussed, along with illustrations of their use in practical analyses. Ideas are presented for future work that can enable one-step processing from a fine-group, problem-independent library to a broad-group, problem-specific library ready for a shielding analysis.
MPI version of NJOY and its application to multigroup cross-section generation
Multigroup cross-section libraries are needed in performing neutronics calculations. These libraries are referred to as broad-group libraries. The number of energy groups and group structure are highly dependent on the application and/or user's objectives. For example, for shielding calculations, broad-group libraries such as SAILOR and BUGLE with 47-neutron and 20-gamma energy groups are used. The common procedure to obtain a broad-group library is a three-step process: (1) processing pointwise ENDF (PENDF) format cross sections; (2) generating fine-group cross sections; and (3) collapsing fine-group cross sections to broad-group. The NJOY code is used to prepare fine-group cross sections by processing pointwise ENDF data. The code has several modules, each one performing a specific task. For instance, the module RECONR performs linearization and reconstruction of the cross sections, and the module GROUPR generates multigroup self-shielded cross sections. After fine-group, i.e., groupwise ENDF (GENDF), cross sections are produced, cross sections are self-shielded, and a one-dimensional transport calculation is performed to obtain flux spectra at specific regions in the model. These fluxes are then used as weighting functions to collapse the fine-group cross sections to obtain a broad-group cross-section library. The third step described is commonly performed by the AMPX code system. SMILER converts NJOY GENDF filed to AMPX master libraries, AJAX collects the master libraries. BONAMI performs self-shielding calculations, NITAWL converts the AMPX master library to a working library, XSDRNPM performs one-dimensional transport calculations, and MALOCS collapses fine-group cross sections to broad-group. Finally, ALPO is used to generate ANISN format libraries. In this three-step procedure, generally NJOY requires the largest amount of CPU time. This time varies depending on the user's specified parameters for each module, such as reconstruction tolerances, temperatures
The 233Pa fission cross-section measurement and evaluation
233Pa is a conspicuous example of strongly discrepant data in the accepted nuclear data evaluations. The precise knowledge of the neutron-induced reaction cross-section of this highly β-active nuclide (T1/2 = 27.0 d) is essential for the successful implementation of the thorium-based fuel cycle in advanced nuclear applications. The reactions involving 233Pa are responsible for the balance of nuclei as well as the average number of prompt fission neutrons in a contemplated reactor scenario. In an IAEA report, it is stated that there is a need to know the 233Pa(n, f) cross-section with an accuracy of 20%. The different evaluated neutron data libraries show, however, a difference of a factor of two for this cross-section. It has previously been deemed not feasible to measure this reaction directly due to its short half-life, high radioactivity and the in-growth of the daughter product 233U. Hence, the entries in the neutron libraries are based on theoretical predictions, which explains the large discrepancies. As reported recently the neutron-induced fission cross-section of 233Pa has been measured for the first time directly with mono-energetic neutrons from 1.0 to 3.0 MeV at the Van-de-Graaff facility of the IRMM. In the meantime, during two further measurement campaigns, the energy range has been extended up to 8.5 MeV. The experimental results will be presented together with recent model calculations of the fission cross-section applying the statistical model code STATIS, which improve the cross-section evaluation up to the second chance fission threshold. (authors)
Measurements of neutron capture cross sections
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 238U 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)
Compilation of cross-sections. Pt. 2
A compilation of integrated cross-sections for hadronic reactions is presented. This is an updated version of CERN/HERA 79-1, 79-2, 79-3. It contains all data published up to the beginning of 1982, but some more recent data have also been included. Plots of the cross sections versus incident laboratory momentum are also given. This volume II contains cross-sections for K+ and K- induced reactions. (orig.)
Cross Sections for Electron Collisions with Methane
Song, Mi-Young, E-mail: mysong@nfri.re.kr; Yoon, Jung-Sik [Plasma Technology Research Center, National Fusion Research Institute, 814-2 Osikdo-dong, Gunsan, Jeollabuk-do 573-540 (Korea, Republic of); Cho, Hyuck [Department of Physics, Chungnam National University, Daejeon 305-764 (Korea, Republic of); Itikawa, Yukikazu [Institute of Space and Astronautical Science, Sagamihara 252-5210 (Japan); Karwasz, Grzegorz P. [Faculty of Physics, Astronomy and Applied Informatics, University Nicolaus Copernicus, Grudziadzka 5, 87100 Toruń (Poland); Kokoouline, Viatcheslav [Department of Physics, University of Central Florida, Orlando, Florida 32816 (United States); Nakamura, Yoshiharu [6-1-5-201 Miyazaki, Miyamae, Kawasaki 216-0033 (Japan); Tennyson, Jonathan [Department of Physics and Astronomy, University College London, Gower Street, London WC1E 6BT (United Kingdom)
2015-06-15
Cross section data are compiled from the literature for electron collisions with methane (CH{sub 4}) molecules. Cross sections are collected and reviewed for total scattering, elastic scattering, momentum transfer, excitations of rotational and vibrational states, dissociation, ionization, and dissociative attachment. The data derived from swarm experiments are also considered. For each of these processes, the recommended values of the cross sections are presented. The literature has been surveyed through early 2014.
Ion and electron impact ionization cross sections
Several current projects are described in which cross sections of interest to radiation physics are being measured. These include total and multiple ionization cross sections for protons on several gases covering a wide energy range, the measurement of cross sections differential in the angle and energy of ejected electrons for several gases including water vapor, and a review of proton ionization data. The work on water vapor has also been extended to electron and neutral hydrogen impact. A brief discussion is also given of some systematics of ionization cross sections. 13 references
Improved Empirical Parametrization of Fragmentation Cross Sections
Sümmerer, Klaus
2012-01-01
A new version is proposed for the universal empirical formula, EPAX, which describes fragmentation cross sections in high-energy heavy-ion reactions. The new version, EPAX 3, can be shown to yield cross sections that are in better agreement with experimental data for the most neutron-rich fragments than the previous version. At the same time, the very good agreement of EPAX 2 with data on the neutron-deficient side has been largely maintained. Comparison with measured cross sections show that the bulk of the data is reproduced within a factor of about 2, for cross sections down to the pico-barn range.
Integral test on activation cross section of tag gas nuclides using fast neutron spectrum fields
Activation cross sections of xenon and krypton isotopes were evaluated as tag gases to identify capsule rupture in liquid metal cooled fast reactors. The accuracy of the activation cross sections was investigated with several tag gas samples irradiated in the standard neutron field of JOYO and YAYOI reactors. Comparing the measured radioactivities and calculated values using YAYOI neutron fluence and activation cross section processed from the JENDL(Japanese Evaluated Nuclear Data Library)-3.2 cross section library, yielded C/E values of approximately 0.86 to 2.6 for tag gas nuclides. The discrepancy between calculation and measurement appears due to the cross section uncertainty. This study confirmed the present accuracy of tag gas activation cross sections. (author)
Cross sections for fuel depletion and radioisotope production calculations in TRIGA reactors
For TRIGA Reactors, the fuel depletion and isotopic inventory calculations, depends on the computer code and in the cross sections of some important actinides used. Among these we have U-235, U-238, Pu-239, Pu-240 and Pu-241. We choose ORIGEN2, a code with a good reputation in this kind of calculations, we observed the cross sections for these actinides in the libraries that we have (PWR's and BWR), the fission cross section for U-235 was about 50 barns. We used a PWR library and our results were not satisfactory, specially for standard elements. We decided to calculate cross sections more suitable for our reactor, for that purpose we simulate the standard and FLIP TRIGA cells with the transport code WIMS. We used the fuel average flux and COLAPS (a home made program), to generate suitable cross sections for ORIGEN2, by collapsing the WIMS library cross sections of these nuclides. For the radioisotope production studies using the Central Thimble, we simulate the A and B rings and used the A average flux to collapse cross sections. For these studies, the required nuclides sometimes are not present in WIMS library, for them we are planning to process the ENDF/B data, with NJOY system, and include the cross sections to WIMS library or to collapse them using the appropriate average-flux and the program COLAPS. (author)
Validation of proton ionization cross section generators for Monte Carlo particle transport
Batic, Matej; Saracco, Paolo
2011-01-01
Three software systems, ERCS08, ISICS 2011 and \\v{S}mit's code, that implement theoretical calculations of inner shell ionization cross sections by proton impact, are validated with respect to experimental data. The accuracy of the cross sections they generate is quantitatively estimated and inter-compared through statistical methods. Updates and extensions of a cross section data library relevant to PIXE simulation with Geant4 are discussed.
Compilation of cross-sections. Pt. 4
This is the fourth volume in our series of data compilations on integrated cross-sections for weak, electromagnetic, and strong interaction processes. This volume covers data on reactions induced by photons, neutrinos, hyperons, and KL0. It contains all data published up to June 1986. Plots of the cross-sections versus incident laboratory momentum are also given. (orig.)
Compilation of cross-sections. Pt. 1
A compilation of integral cross-sections for hadronic reactions is presented. This is an updated version of CERN/HERA 79-1, 79-2, 79-3. It contains all data published up to the beginning of 1982, but some more recent data have also been included. Plots of the cross-sections versus incident laboratory momentum are also given. (orig.)
Nucleon-XcJ Dissociation Cross Sections
冯又层; 许晓明; 周代翠
2002-01-01
Nucleon-XcJ dissociation cross sections are calculated in a constituent interexchange model in which quark-quark potential is derived from the Buchmüller-Tye quark-anti-quark potential. These new cross sections for dominant reaction channels depend on the centre-of-mass energy of the nucleon and the charmonium.
Fission cross section calculations for Pa isotopes
Based on the recently measured cross-section values for the neutron-induced fission of 231Pa and our experience gained with other isotopes, new self consistent neutron cross section calculations for n+231Pa have been performed up to 30 MeV. The results are quite different to the existing evaluations, especially above the first chance fission threshold. (authors)
Comparative analysis among several cross section sets
Critical parameters were calculated using the one dimensional multigroup transport theory for several cross section sets. Calculations have been performed for water mixtures of uranium metal, plutonium metal and uranium-thorium oxide, and for metallics systems, to determine the critical dimensions of geometries (sphere and cylinder). For this aim, the following cross section sets were employed: 1) multigroup cross section sets obtained from the GAMTEC-II code; 2) the HANSEN-ROACH cross section sets; 3) cross section sets from the ENDF/B-IV, processed by the NJOY code. Finally, we have also calculated the corresponding critical radius using the one dimensional multigroup transport DTF-IV code. The numerical results agree within a few percent with the critical values obtained in the literature (where the greatest discrepancy occured in the critical dimensions of water mixtures calculated with the values generated by the NJOY code), a very good results in comparison with similar works. (Author)
Photoproton cross section for 17O
The measurement of the 17O(γ,p)16N reaction from threshold to an excitation energy of 44 MeV is presented. These results have been summed with the previously measured total photoneutron cross section to provide an approximation to the total photoabsorption cross section of 17O. The magnitude of the 17O photoabsorption cross section at the peak of the Giant Dipole Resonance is considerably less than the equivalent value for the photoabsorption cross sections of 16O and 18O. In addition, the integrated total photoabsorption cross section for 17O (up to 40 MeV) exhausts only about 58% of the sum rule; the values for the cases of 16O and 18O are significantly larger than this. The present data along with results from other reaction channels of this nucleus, were used to make spin, parity, and isospin assignments for several states in 17O. 48 refs., 4 tabs., 7 figs
Validation of the Dosimetry Cross Sections by Integral Experiments
Full text: A new version of the dosimetry library IRDFF has been released from the IAEA, featuring more reactions (compared to the older IRDF-2002), an extended energy range to 60 MeV, re-evaluation of several reaction channels with greatly improved covariance information, etc. The aim of the present CRP is the validation of the IRDFF library in order to enhance its reliability for the purposes it serves. Reaction rate measurements in well-defined neutron fields are a common method of validating the cross sections. This technique has been implemented to some extent in the preparation of the IRDFF library from the published data, namely the use of the published average cross sections in the 252Cf spontaneous fission neutron spectrum, thermal cross sections and resonance integrals. There is synergy with the community performing neutron activation analysis by the k0 standardisation technique. To some extent the k0 database has been used for a preliminary validation of the capture reactions in the IRDFF library, but one has to be very careful about the definition of the constants since the interpretation and the derivation of the commonly-used constants is often misleading. The proposed contribution of the Jozef Stefan Institute (JSI) is on the consistent definitions of the constants that allow unique interpretation of the measured reaction rate ratios (particularly for the capture reactions), together with advanced analysis techniques including Monte Carlo simulations of the experiments. The contribution can include the codes that are used at (JSI) for the purpose, namely the GRUPINT code for the analysis of measured reaction rate rations and spectrum unfolding and the RRUNC code for the calculation of the uncertainties in the calculated reaction rates due to the uncertainties in the cross sections and in the neutron spectrum. (author)
Neutron-induced fission cross sections of short-lived actinides via the surrogate reaction method
A brief discussion of surrogate reaction methods has been made and some of the recent results on neutron induced fission cross section measurements have been presented. The validation of the EMPIRE-3.1. predictions on neutron induced cross sections corresponding to fission barriers used from Barrier Formula (BF) and RIPL-1 libraries have been discussed
JSD1000: multi-group cross section sets for shielding materials
A multi-group cross section library for shielding safety analysis has been produced by using ENDF/B-IV. The library consists of ultra-fine group cross sections, fine-group cross sections, secondary gamma-ray production cross sections and effective macroscopic cross sections for typical shielding materials. Temperature dependent data at 300, 560 and 900 K have been also provided. Angular distributions of the group to group transfer cross section are defined by a new method of ''Direct Angular Representation'' (DAR) instead of the method of finite Legendre expansion. The library designated JSD1000 are stored in a direct access data base named DATA-POOL and data manipulations are available by using the DATA-POOL access package. The 3824 neutron group data of the ultra-fine group cross sections and the 100 neutron, 20 photon group cross sections are applicable to shielding safety analyses of nuclear facilities. This report provides detailed specifications and the access method for the JSD1000 library. (author)
Photoneutron cross sections for the silicon isotopes
The photoneutron cross sections for 28Si, 29Si, and 30Si have been measured up to 33 MeV with monoenergetic photons from the annihilation in flight of fast positrons, using neutron multiplicity counting. Average neutron energies were obtained simultaneously with the cross-section data by the ring-ratio technique. The giant dipole resonance for 28Si and 30Si exhibit appreciable fragmentation; that for 29Si does not. The (γ,2n) cross section for 30Si is large; that for 29Si is consistent with zero. The (γ,1n) cross section for 30Si decreases sharply with energy to values near zero as the (γ,2n) cross section grows, then increases to appreciable values as the (γ,2n) cross section diminishes; this extreme behavior, although never seen before, is attributable to the competition between the (γ,n), (γ,2n), and (γ,pn) decay channels. Some properties of the isospin components of the giant resonance are inferred. Other features of the data, including the integrated cross sections, are found to be similar in many respects to corresponding results for the oxygen and magnesium isotopes. The 28Si nucleus is found to be a better core for 29Si and 30Si than might have been expected from previous descriptions of its open-shell character
The 42Ca photoneutron cross section
The measurement of the 42Ca(γ,nsub(t)) is reported here over the energy range 10.5 - 28 MeV. Bremsstrahlung radiation from the 35 MeV Betatron at this University was used to measure a yield curve of photoneutrons, from which the (γ,nsub(t)) cross section was derived. Since proton and neutron emission are the major decay modes of the giant dipole resonance, summing these cross sections approximates the photo-absorption cross section. With this information the theoretical predictions can be checked
Compilation of cross-sections. Pt. 3
A compilation of integrated cross-sections for hadronic reactions is presented. This is an updated version of CERN/HERA 79-1, 79-2, 79-3. It contains all data published up to the beginning of 1982, but some more recent data, particularly those from the CERN Collider, have also been included. Plots of the cross-sections versus incident laboratory momentum are also given. This volume III contains cross-sections for p and anti p induced reactions. (orig.)
Screening corrections to the Rutherford cross section
Differential cross sections for elastic p-Au scattering were measured in the energy range between 0.2 and 0.8 MeV for scattering angles from 300 to 1500 in order to determine corrections to the Rutherford cross section due to the screening of the nuclear charge by the atomic electrons. Furthermore, differential cross sections have been calculated in the weakly screening region using various screening functions. A simple analytical expression has been derived for the representation of both experimental and theoretical results. (orig.)
Measurement of MA fission cross sections at YAYOI
Ohkawachi, Yasushi; Ohki, Shigeo; Wakabayashi, Toshio [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center
1998-03-01
Fission cross section ratios of minor actinide nuclides (Am-241, Am-243) relative to U-235 in the fast neutron energy region have been measured using a back-to-back (BTB) fission chamber at YAYOI fast neutron source reactor. A small BTB fission chamber was developed to measure the fission cross section ratios in the center of the core at YAYOI reactor. Dependence of the fission cross section ratios on neutron spectra was investigated by changing the position of the detector in the reactor core. The measurement results were compared with the fission cross sections in the JENDL-3.2, ENDF/B-VI and JEF-2.2 libraries. It was found that calculated values of Am-241 using the JENDL-3.2, ENDF/B-VI and JEF-2.2 data are lower by about 15% than the measured value in the center of the core (the neutron average energy is 1.44E+6(eV)). And, good agreement can be seen the measured value and calculated value of Am-243 using the JENDL-3.2 data in the center of the core (the neutron average energy is 1.44E+6)(eV), but calculated values of Am-243 using the ENDF/B-VI and JEF-2.2 data are lower by 11% and 13% than the measured value. (author)
A nuclear cross section data handbook
Fisher, H.O.M.
1989-12-01
Isotopic information, reaction data, data availability, heating numbers, and evaluation information are given for 129 neutron cross-section evaluations, which are the source of the default cross sections for the Monte Carlo code MCNP. Additionally, pie diagrams for each nuclide displaying the percent contribution of a given reaction to the total cross section are given at 14 MeV, 1 MeV, and thermal energy. Other information about the evaluations and their availability in continuous-energy, discrete-reaction, and multigroup forms is provided. The evaluations come from ENDF/B-V, ENDL85, and the Los Alamos Applied Nuclear Science Group T-2. Graphs of all neutron and photon production cross-section reactions for these nuclides have been categorized and plotted. 21 refs., 5 tabs.
Differential cross sections of positron hydrogen collisions
于荣梅; 濮春英; 黄晓玉; 殷复荣; 刘旭焱; 焦利光; 周雅君
2016-01-01
We make a detailed study on the angular differential cross sections of positron–hydrogen collisions by using the momentum-space coupled-channels optical (CCO) method for incident energies below the H ionization threshold. The target continuum and the positronium (Ps) formation channels are included in the coupled-channels calculations via a complex equivalent-local optical potential. The critical points, which show minima in the differential cross sections, as a function of the scattering angle and the incident energy are investigated. The resonances in the angular differential cross sections are reported for the first time in this energy range. The effects of the target continuum and the Ps formation channels on the different cross sections are discussed.
Systematics of (n,2n) Cross Sections
2008-01-01
<正>The experimental data of (n, 2n) cross sections were collected and evaluated as complete as possible. There are 640 sets of experimental data for 130 nuclei. The data were fitted to the expressions that describe the
Photoneutron cross section of 34S
Using an enriched 34S target, the reaction 34S(γ,sn)33S has been measured from below threshold (10.4 MeV) to 28 MeV by directly counting the photoneutrons as a function of bremsstrahlung energy. The resultant cross section shows gross splitting in the GDR region. The integrated cross section is discussed in the light of the systematics of similar nuclei having two neutrons outside a doubly closed shell/sub-shell core
Photoneutron cross section of 34S
Using an enriched 34S target, the reaction 34S(γ, sn) has been measured from below threshold (10.4 MeV) to 28 MeV by directly counting the photoneutrons as a function of bremsstrahlung energy. The resultant cross section shows gross splitting in the GDR region. The integrated cross section is discussed in the light of the systematics of similar nuclei having two neutrons outside a doubly closed shell/sub-shell core. (orig.)
Neutron capture cross sections from Surrogate measurements
Scielzo N.D.; Dietrich F.S.; Escher J.E.
2010-01-01
The prospects for determining cross sections for compound-nuclear neutron-capture reactions from Surrogate measurements are investigated. Calculations as well as experimental results are presented that test the Weisskopf-Ewing approximation, which is employed in most analyses of Surrogate data. It is concluded that, in general, one has to go beyond this approximation in order to obtain (n,γ) cross sections of sufficient accuracy for most astrophysical and nuclear-energy applications.
Neutron capture cross sections from Surrogate measurements
Scielzo N.D.
2010-03-01
Full Text Available The prospects for determining cross sections for compound-nuclear neutron-capture reactions from Surrogate measurements are investigated. Calculations as well as experimental results are presented that test the Weisskopf-Ewing approximation, which is employed in most analyses of Surrogate data. It is concluded that, in general, one has to go beyond this approximation in order to obtain (n,γ cross sections of sufficient accuracy for most astrophysical and nuclear-energy applications.
Evaluation methods for neutron cross section standards
Methods used to evaluate the neutron cross section standards are reviewed and their relative merits, assessed. These include phase-shift analysis, R-matrix fit, and a number of other methods by Poenitz, Bhat, Kon'shin and the Bayesian or generalized least-squares procedures. The problems involved in adopting these methods for future cross section standards evaluations are considered, and the prospects for their use, discussed. 115 references, 5 figures, 3 tables
Methods for calculating anisotropic transfer cross sections
The Legendre moments of the group transfer cross section, which are widely used in the numerical solution of the transport calculation can be efficiently and accurately constructed from low-order (K = 1--2) successive partial range moments. This is convenient for the generation of group constants. In addition, a technique to obtain group-angle correlation transfer cross section without Legendre expansion is presented. (author)
Photoproton cross section for 14C
Using bremsstrahlung, the 14C(γ,p) reaction cross section has been measured from threshold to 29 MeV. The integrated cross section up to 30 MeV is 18±3 MeV mb. Above 23.5 MeV, the reported cross section includes a contribution, estimated at 3.5 MeV mb, due to the 14C(γ,d) and 14Cγ,pn) reactions. Essentially the entire 14C(γ,p) cross section results from decay of T> dipole states. From knowledge of other decay channels estimates of the cross section, integrated to 30 MeV for the T and T> components of the giant resonance (GDR) of 81 MeV mb and 43 MeV mb are obtained. The splitting of the mean energies of the GDR isospin components is 8.5 MeV. Comparisons with several shell-model calculations are made with the data, and general agreement is found. A comparison of photonuclear absorption cross sections for 12,1314C and 16,17,18 O shows dramatic redistribution of dipole strength as neutrons are added to the core nuclei. 41 refs., 1 tab., 7 figs
Cross sections, benchmarks, etc.: What is data testing all about
In order to determine the consistency of two distinct measurements of a physical quantity, the discrepancy d between the two should be compared with its own standard deviation, σ = √(σ/sub 1//sup 2/+σ/sub 2//sup 2/). To properly test a given cross-section library by a set of benchmark (integral) measurements, the quantity corresponding to (d/σ)/sup 2/ is the quadratic d/sup dagger/C/sup -1/d. Here d is the vector of which the components are the discrepancies between the calculated values of the integral parameters and their corresponding measured values, and C is the uncertainty matrix of these discrepancies. This quadratic form is the only true measure of the joint consistency of the library and benchmarks. On the other hand, the very matrix C is essentially all one needs to adjust the library by the benchmarks. Therefore, any argument against adjustment simultaneously disqualifies all serious attempts to test cross-section libraries against integral benchmarks
A Pebble Bed Reactor cross section methodology
A method is presented for the evaluation of microscopic cross sections for the Pebble Bed Reactor (PBR) neutron diffusion computational models during convergence to an equilibrium (asymptotic) fuel cycle. This method considers the isotopics within a core spectral zone and the leakages from such a zone as they arise during reactor operation. The randomness of the spatial distribution of fuel grains within the fuel pebbles and that of the fuel and moderator pebbles within the core, the double heterogeneity of the fuel, and the indeterminate burnup of the spectral zones all pose a unique challenge for the computation of the local microscopic cross sections. As prior knowledge of the equilibrium composition and leakage is not available, it is necessary to repeatedly re-compute the group constants with updated zone information. A method is presented to account for local spectral zone composition and leakage effects without resorting to frequent spectrum code calls. Fine group data are pre-computed for a range of isotopic states. Microscopic cross sections and zone nuclide number densities are used to construct fine group macroscopic cross sections, which, together with fission spectra, flux modulation factors, and zone buckling, are used in the solution of the slowing down balance to generate a new or updated spectrum. The microscopic cross-sections are then re-collapsed with the new spectrum for the local spectral zone. This technique is named the Spectral History Correction (SHC) method. It is found that this method accurately recalculates local broad group microscopic cross sections. Significant improvement in the core eigenvalue, flux, and power peaking factor is observed when the local cross sections are corrected for the effects of the spectral zone composition and leakage in two-dimensional PBR test problems.
Multigroup cross sections of resonant nuclei considering moderator mass differences
The multigroup constants library MGCL in the nuclear criticality safety evaluation code system JACS has been produced by the Bondarenko method to treat self-shielding effects. For estimating errors of this treatment, the multigroup cross sections of MGCL are compared with those obtained by precise treatment, i.e. with the weighted cross sections by ultra-fine spectra of neutron. The precise calculations are made for homogeneous mixtures of a resonant nucleus (235U, 238U, 239Pu, 240Pu, 242Pu or 56Fe) and a fictitious moderator nucleus with mass number 1, 12 or 200. The ultra-fine spectrum is calculated by the RABBLE code. Distinct differences are found in the self-shielding factors by comparisons between both treatments. Moreover, as the mass number increases, depressions of the self-shielding factor at the resonance peaks and its enhancements at the window of resonances are observed. (author)
Reduction Methods for Total Reaction Cross Sections
Gomes, P. R. S.; Mendes Junior, D. R.; Canto, L. F.; Lubian, J.; de Faria, P. N.
2016-03-01
The most frequently used methods to reduce fusion and total reaction excitation functions were investigated in a very recent paper Canto et al. (Phys Rev C 92:014626, 2015). These methods are widely used to eliminate the influence of masses and charges in comparisons of cross sections for weakly bound and tightly bound systems. This study reached two main conclusions. The first is that the fusion function method is the most successful procedure to reduce fusion cross sections. Applying this method to theoretical cross sections of single channel calculations, one obtains a system independent curve (the fusion function), that can be used as a benchmark to fusion data. The second conclusion was that none of the reduction methods available in the literature is able to provide a universal curve for total reaction cross sections. The reduced single channel cross sections keep a strong dependence of the atomic and mass numbers of the collision partners, except for systems in the same mass range. In the present work we pursue this problem further, applying the reduction methods to systems within a limited mass range. We show that, under these circumstances, the reduction of reaction data may be very useful.
Neutron cross section of methane hydrate
Kiyanagi, Y.; Date, S.; Horikawa, T.; Takamine, J.; Iwasa, H.; Kamiyama, T. [Graduate School of Eng., Hokkaido Univ., Sapporo (Japan); Uchida, T.; Ebinuma, T.; Narrita, H. [National Inst. of Advanced Industrial Science, Tsukisamu, Sapporo (Japan); Bennington, S.M. [ISIS Dept., Rutherford Appleton, Chilton, Didcot, Oxon (United Kingdom)
2004-03-01
To estimate the neutronic characteristics of methane hydrate and also to synthesize cross section data for simulation we need neutron scattering data ranging wide energy and momentum region. We performed inelastic neutron scattering experiments to get information about the neutron cross section on methane hydrate. It was found that at high momentum transfer region rotational mode as well as vibration mode showed recoil like behavior. On the other hand, at low momentum region, as well known, free rotation like energy levels were observed. The energy level of ice in methane hydrate was very similar to normal ice. The results suggest that the rough expression of the cross section of the methane hydrate is presented by linear combination of the methane and ice. (orig.)
Prospects for Precision Neutrino Cross Section Measurements
Harris, Deborah A. [Fermilab
2016-01-28
The need for precision cross section measurements is more urgent now than ever before, given the central role neutrino oscillation measurements play in the field of particle physics. The definition of precision is something worth considering, however. In order to build the best model for an oscillation experiment, cross section measurements should span a broad range of energies, neutrino interaction channels, and target nuclei. Precision might better be defined not in the final uncertainty associated with any one measurement but rather with the breadth of measurements that are available to constrain models. Current experience shows that models are better constrained by 10 measurements across different processes and energies with 10% uncertainties than by one measurement of one process on one nucleus with a 1% uncertainty. This article describes the current status of and future prospects for the field of precision cross section measurements considering the metric of how many processes, energies, and nuclei have been studied.
Radiation pressure cross section for fluffy aggregates
We apply the discrete dipole approximation (DDA) to estimate the radiation pressure cross section for fluffy aggregates by computing the asymmetry parameter and the cross sections for extinction and scattering. The ballistic particle-cluster aggregate and the ballistic cluster-cluster aggregate consisting of either dielectric or absorbing material are considered to represent naturally existing aggregates. We show that the asymmetry parameter perpendicular to the direction of wave propagation is maximized where the wavelength is comparable to the aggregate size, which may be characterized by the area-equivalent radius or the radius of gyration rather than the volume-equivalent radius. The asymmetry parameter for the aggregate depends on the morphology of the particle, but not on the constituent material. Therefore, the dependence of the radiation pressure cross section on the material composition arises mainly from that of the extinction and scattering cross sections, in other words, the single-scattering albedo. We find that aggregates consisting of high-albedo material show a large deviation of radiation pressure from the direction of incident radiation. When the aggregates are illuminated by blackbody radiation, the deviation of the radiation pressure increases with increasing temperature of the blackbody. Since the parallel component of the radiation pressure cross section for the aggregates is smaller than that for the volume-equivalent spheres at the size parameter close to unity, the Planck-mean radiation pressure cross section for the aggregates having radius comparable to the effective wavelength of radiation shows a lower value, compared with the volume-equivalent sphere. Consequently, the slope of the radiation pressure force per mass of the particle as a function of particle mass shows a lower maximum for the aggregates than for compact spherical particles. (Copyright (c) 1998 Elsevier Science B.V., Amsterdam. All rights reserved.)
Group neutron fission and radiative-capture cross-sections for transactinides
A comparison is made between evaluations of radiative-capture and fission cross-sections for the isotopes 236U, 237Np, 238Pu, 241Am, 243Am, 242Cm and 244Cm, and group cross-sections for use in fast-reactor calculations are recommended. Group cross-sections obtained from the HEDL graphical data (evaluation for ENDF/B-V) are shown for 234U, 236Pu, 237Pu, 242Pu, 244Pu, sup(242m)Am, 241Cm, 243Cm and 248Cm. Group cross-sections for 32 isotopes from the ENDL-76 library files are also given. In choosing recommended cross-sections, account was taken of the extent of agreement with experimental data where these are available, the extent to which the cross-sections are documented and the extent to which they have been calculated from a theoretical model. The reliability of evaluations is discussed. An attempt is made to evaluate the error in single-group cross-sections averaged over a typical fast-reactor spectrum. Conclusions are drawn from a study of the literature on the current status of experimental and theoretical research on transactinide cross-sections, and from the spread of the different evaluation data. Finally, the situation with respect to the integral experiments which can be used for correcting transactinide cross-sections is discussed. (author)
Charged particle reaction cross sections and nucleosynthesis
The role of proton and α-particle induced reactions in carbon, neon, oxygen and silicon burning in massive stars is surveyed. The problems associated with determining thermonuclear reaction rates for reactions with widely spaced resonances and with closely spaced or overlapping resonances are discussed and the associated experimental approaches are reviewed. Experimental techniques which have been used in the measurement of reaction cross sections are discussed and their strengths and weaknesses are identified. Recent developments in attempts to establish reliable statistical-model codes for calculation of reaction cross sections are presented and discussed. The results of experimental tests of statistical model codes are summarised and evaluated
Neutron capture cross sections from surrogate measurements
The prospects for determining cross sections for compound-nuclear neutron-capture reactions from Surrogate measurements are investigated. Calculations as well as experimental results are presented that test the Weisskopf-Ewing approximation, which is employed in most analyses of Surrogate data. The method is applied to the 155Gd(n,γ) reaction. It is concluded that, in general, one has to go beyond this approximation in order to obtain (n,γ) cross sections of sufficient accuracy for most astrophysical and nuclear-energy applications. (authors)
Precise neutron inelastic cross section measurements
Negret, Alexandru
2012-11-01
The design of a new generation of nuclear reactors requires the development of a very precise neutron cross section database. Ongoing experiments performed at dedicated facilities aim to the measurement of such cross sections with an unprecedented uncertainty of the order of 5% or even smaller. We give an overview of such a facility: the Gamma Array for Inelastic Neutron Scattering (GAINS) installed at the GELINA neutron source of IRMM, Belgium. Some of the most challenging difficulties of the experimental approach are emphasized and recent results are shown.
Optical Model and Cross Section Uncertainties
Herman,M.W.; Pigni, M.T.; Dietrich, F.S.; Oblozinsky, P.
2009-10-05
Distinct minima and maxima in the neutron total cross section uncertainties were observed in model calculations using spherical optical potential. We found this oscillating structure to be a general feature of quantum mechanical wave scattering. Specifically, we analyzed neutron interaction with 56Fe from 1 keV up to 65 MeV, and investigated physical origin of the minima.We discuss their potential importance for practical applications as well as the implications for the uncertainties in total and absorption cross sections.
Saturation Effects in Hadronic Cross Sections
Shoshi, Arif I.; Steffen, Frank D.
2002-01-01
We compute total and differential elastic cross sections of high-energy hadronic collisions in the loop-loop correlation model that provides a unified description of hadron-hadron, photon-hadron, and photon-photon reactions. The impact parameter profiles of pp and gamma*p collisions are calculated. For ultra-high energies the hadron opacity saturates at the black disc limit which tames the growth of the hadronic cross sections in agreement with the Froissart bound. We compute the impact param...
Covariance Evaluation Methodology for Neutron Cross Sections
Herman,M.; Arcilla, R.; Mattoon, C.M.; Mughabghab, S.F.; Oblozinsky, P.; Pigni, M.; Pritychenko, b.; Songzoni, A.A.
2008-09-01
We present the NNDC-BNL methodology for estimating neutron cross section covariances in thermal, resolved resonance, unresolved resonance and fast neutron regions. The three key elements of the methodology are Atlas of Neutron Resonances, nuclear reaction code EMPIRE, and the Bayesian code implementing Kalman filter concept. The covariance data processing, visualization and distribution capabilities are integral components of the NNDC methodology. We illustrate its application on examples including relatively detailed evaluation of covariances for two individual nuclei and massive production of simple covariance estimates for 307 materials. Certain peculiarities regarding evaluation of covariances for resolved resonances and the consistency between resonance parameter uncertainties and thermal cross section uncertainties are also discussed.
CROSS SECTION EVALUATIONS FOR ENDF/B-VII.
HERMAN, M.; ROCHMAN, D.; OBLOZINSKY, P.
2006-06-05
This is the final report of the work performed under the LANL contract on neutron cross section evaluations for ENDF/B-VII (April 2005-May 2006). The purpose of the contract was to ensure seamless integration of the LANL neutron cross section evaluations in the new ENDF/B-VII library. The following work was performed: (1) LANL evaluated data files submitted for inclusion in ENDF/B-VII were checked and, when necessary, formal formatting errors were corrected. As a consequence, ENDF checking codes, run on all LANL files, do not report any errors that would rise concern. (2) LANL dosimetry evaluations for {sup 191}Ir and {sup 193}Ir were completed to match ENDF requirements for the general purpose library suitable for transport calculations. A set of covariances for both isotopes is included in the ENDF files. (3) Library of fission products was assembled and successfully tested with ENDF checking codes, processed with NJOY-99.125 and simple MCNP calculations. (4) KALMAN code has been integrated with the EMPIRE system to allow estimation of covariances based on the combination of measurements and model calculations. Covariances were produced for 155,157-Gd and also for 6 remaining isotopes of Gd.
Coupled neutron and photon cross sections for transport calculations
A compact set of multigroup cross sections and transfer tables for use in neutron and photon transport calculations was prepared from ENDF/B-IV using the NJOY processing system. The library includes prompt and steady-state coupled sets for neutrons and photons in FIDO format, prompt and steady-state fission spectra (chi vectors) for the fissionable isotopes, and a table of useful response functions including heating and gas production. These multigroup constants should be useful for a wide variety of problems where self-shielding is not important. 15 references
Removing the Artifacts from Artwork Cross-section Images
Beneš, Miroslav; Zitová, Barbara; Blažek, Jan; Hradilová, J.; Hradil, David
Brussels : IEEE, 2011, s. 3598-3601. ISBN 978-1-4577-1302-6. [ICIP 18th IEEE International Conference on Image Processing. Brussels (BE), 11.09.2011-14.09.2011] R&D Projects: GA MŠk 1M0572 Institutional research plan: CEZ:AV0Z10750506; CEZ:AV0Z40320502 Keywords : image processing * image enhancement * cultural heritage Subject RIV: JD - Computer Applications, Robotics http://library.utia.cas.cz/separaty/2011/ZOI/benes-removing the artifacts from artwork cross-section images.pdf
Fusion cross sections and the new dynamics
The prediction of the need for an extra push over the interaction barrier in order to make the heavier nuclei fuse is made the basis of a simple algebraic theory for the energy-dependence of the fusion cross-section. A comparison with recent experiments promises to provide a quantitative test of the New Dynamics
LSP-Nucleus Elastic Scattering Cross Sections
Vergados, J. D.; Kosmas, T. S.
1997-01-01
We calculate LSP-nucleus elastic scattering cross sections using some representative input in the restricted SUSY parameter space. The coherent matrix elements are computed throughout the periodic table while the spin matrix elements for the proposed $^{207}Pb$ target which has a rather simple nuclear structure. The results are compared to those given from other cold dark matter detection targets.
Electron impact excitation cross sections for carbon
Ganas, P. S.
1981-04-01
A realistic analytic atomic independent particle model is used to generate wave functions for the valence and excited states of carbon. Using these wave functions in conjunction with the Born approximation and the Russell-Saunders LS-coupling scheme, we calculate generalized oscillator strengths and integrated cross sections for various excitations from the 2p 2( 3P O) valence state.
Electron impact excitation cross sections for carbon
A realistic analytic atomic independent particle model is used to generate wave functions for the valence and excited states of carbon. Using these wave functions in conjunction with the Born approximation and the Russell-Saunders LS-coupling scheme, we calculate generalized oscillator strengths and integrated cross sections for various excitations from the 2p2(3P0) valence state. (orig.)
Top quark cross sections and differential distributions
Kidonakis, Nikolaos
2011-01-01
I present results for the top quark pair total cross section and the top quark transverse momentum distribution at Tevatron and LHC energies. I also present results for single top quark production. All calculations include NNLO corrections from NNLL threshold resummation.
Neutron cross sections of importance to astrophysics
Neutron reactions of importance to the various stellar burning cycles are discussed. The role of isomeric states in the branched s-process is considered for particular cases. Neutron cross section needs for the 187Re-187Os, 87Rb-87Sr clocks for nuclear cosmochronology are discussed. Other reactions of interest to astrophysical processes are presented. 35 references
Modelisation of the fission cross section
The neutron cross sections of four nuclear systems (n+235U, n+233U, n+241Am and n+237Np) are studied in the present document. The target nuclei of the first case, like 235U and 239Pu, have a large fission cross section after the absorption of thermal neutrons. These nuclei are called 'fissile' nuclei. The other type of nuclei, like 237Np and 241Am, fission mostly with fast neutrons, which exceed the fission threshold energy. These types of nuclei are called 'fertile'. The compound nuclei of the fertile nuclei have a binding energy higher than the fission barrier, while for the fissile nuclei the binding energy is lower than the fission barrier. In this work, the neutron induced cross sections for both types of nuclei are evaluated in the fast energy range. The total, reaction and shape-elastic cross sections are calculated by the coupled channel method of the optical model code ECIS, while the compound nucleus mechanism are treated by the statistical models implemented in the codes STATIS, GNASH and TALYS. The STATIS code includes a refined model of the fission process. Results from the theoretical calculations are compared with data retrieved from the experimental data base EXFOR. (author)
Neutron Capture Cross Sections for Radioactive Nuclei
Tonchev, Anton; Bedrossian, Peter; Escher, Jutta; Scielzo, Nicholas
2015-10-01
Accurate neutron-capture cross sections for radioactive nuclei near or far away from the line of beta stability are crucial for understanding the nucleosynthesis of heavy elements. However, neutron-capture cross sections for short-lived radionuclides are difficult to measure due to the fact that the measurements require both highly radioactive samples and intense neutron sources. Essential ingredients for describing the γ decays following neutron capture are the γ-ray strength function and level densities. We will compare different indirect approaches for obtaining observables that can constrain Hauser-Feshbach statistical model calculations of capture cross sections. Specifically, we will consider photon scattering, transfer reactions, and beta-delayed neutron emission. Challenges that exist on the path to obtaining neutron-capture cross sections for reactions on isotopes far from stability will be discussed. This work was performed under the auspices of US DOE by LLNL under contract DE-AC52-07NA27344. Funding was provided via the LDRD-ERD-069 project.
Symmetric charge transfer cross section of uranium
Symmetric charge transfer cross section of uranium was calculated under consideration of reaction paths. In the charge transfer reaction a d3/2 electron in the U atom transfers into the d-electron site of U+(4I9/2) ion. The J value of the U atom produced after the reaction is 6, 5, 4 or 3, at impact energy below several tens eV, only resonant charge transfer in which the product atom is ground state (J=6) takes place. Therefore, the cross section is very small (4-5 x 10-15 cm2) compared with that considered so far. In the energy range of 100-1000eV the cross section increases with the impact energy because near resonant charge transfer in which an s-electron in the U atom transfers into the d-electron site of U+ ion. Charge transfer cross section between U+ in the first excited state (289 cm-1) and U in the ground state was also obtained. (author)
Measurement cross sections for radioisotopes production
New radioactive isotopes for nuclear medicine can be produced using particle accelerators. This is one goal of Arronax, a high energy - 70 MeV - high intensity - 2*350 μA - cyclotron set up in Nantes. A priority list was established containing β- - 47Sc, 67Cu - β+ - 44Sc, 64Cu, 82Sr/82Rb, 68Ge/68Ga - and α emitters - 211At. Among these radioisotopes, the Scandium 47 and the Copper 67 have a strong interest in targeted therapy. The optimization of their productions required a good knowledge of their cross-sections but also of all the contaminants created during irradiation. We launched on Arronax a program to measure these production cross-sections using the Stacked-Foils' technique. It consists in irradiating several groups of foils - target, monitor and degrader foils - and in measuring the produced isotopes by γ-spectrometry. The monitor - natCu or natNi - is used to correct beam loss whereas degrader foils are used to lower beam energy. We chose to study the natTi(p,X)47Sc and 68Zn(p,2p)67Cu reactions. Targets are respectively natural Titanium foil - bought from Goodfellow - and enriched Zinc 68 deposited on Silver. In the latter case, Zn targets were prepared in-house - electroplating of 68Zn - and a chemical separation between Copper and Gallium isotopes has to be made before γ counting. Cross-section values for more than 40 different reactions cross-sections have been obtained from 18 MeV to 68 MeV. A comparison with the Talys code is systematically done. Several parameters of theoretical models have been studied and we found that is not possible to reproduce faithfully all the cross-sections with a given set of parameters. (author)
Importance of neutron cross-sections for transmutation
Accurate neutron cross-section data is fundamental to the reliable design of any transmutation device, and, in particular, of an accelerator-driven system (ADS). Calculations of the behaviour of the core depend strongly on the cross-section data: parameters such as the multiplication coefficient, power densities or reactivity may vary significantly depending on the nuclear-data (ND) library used. These potential discrepancies justify the need to improve the present data for several isotopes and reaction channels, for a wide range of neutron energies from thermal to high-energy. This paper follows on from work performed in the context of the nTOF-ND-ADS project of the EURATOM 5th framework program, where a preliminary analysis of the effects of different cross-section data was carried out using the Monte Carlo code package FLUKA-EAMC. That study was based on the Pb-Bi cooled 80 MWth energy-amplifier prototype, and included comparison of parameters such as source multiplication coefficient ksrc, neutron spectra, neutron balance and one-group cross-sections for different isotopes using different nuclear-data evaluations. The present work expands this analysis to other isotopes of interest such as 233U, 243Am, 244,245Cm and the long-lived fission fragments (LLFFs) 99Tc and 129I. A direct comparison of nuclear-data libraries to indicate the spread between values was performed. The paper also extends the sensitivity analysis of the parameters mentioned above to moderated systems, such as TRADE (triga accelerator-driven experiment): a 1 MW triga reactor coupled with a 110-140 MeV-2 mA proton cyclotron. Study of the discrepancies in the thermal and epithermal regions is essential for the design of systems for the transmutation of LLFF (transmutation by adiabatic resonance crossing, TARC) and also important for minor actinides (MAs) for which sub-threshold fission should not be neglected. These studies highlight the relative importance of different isotopes and assess the
Evaluated cross section data from Russian reactor dosimetry file
New updates to the RRDF-98 library were presented. Problems with the covariance matrix in previously submitted evaluations have been eliminated by using extended precision, which resolved the issue of negative Eigenvalues in some of the covariance matrices. A new 237Np(n, f) evaluation was also provided; this file does not yet have a File 1 comments section, but includes the data required to finalize the contents of the IRDF-2002 library. Several example cases were presented where excessive scatter in the experimental data had been resolved by carefully tracing the standards used by the experimenter and re-normalizing the data using the current best estimates of the reference standard cross sections. This renormalization required that Zolotarev track down the experimental details, often contacting the actual experimenters since details were not provided in their written documentation. Cases were shown where the resulting re-normalization procedure dramatically collapsed the spread in the experimental data. Some of Zolotorev's tables have the latest RRDF-98 contributions labelled as IRDF-2002 evaluations. Since the final IRDF-2002 library contents will not be decided until the end of this meeting, these contributions will be relabeled Updated RRDF-98. Zsolnay expressed the views of the whole meeting when she thanked Zolotarev for his extensive contributions to this DDP, and for his quick response to requests to resolve the issues that arose from the reviews. The issue of adding cumulative fission yields to the IRDF-2002 library was discussed since these data are part of the database required in using the new dosimetry cross sections. Trkov noted that the IAEA has an on-going project that addresses this need, but the results will not be available within the timeframe of the IRDF-2002 release. The issue of the addition of fission yields to the IRDF-2002 library was set aside for consideration as part of any future revision
Absolute average capture cross sections of gold, thorium, tantalum, molybdenum, copper and strontium in 252Cf spontaneous fission neutron spectrum were simulated for two types of experiment setups preformed by Z. Dezso and J. Csikai and by L. Green. The experiments were simulated with MCNP5 using cross section data from the ENDF/B-VII.0 library. The determination of neutron backscattering was calculated with the use of neutron flagging. Correction factors to experimentally measured values were determined to obtain average cross sections in a pure 252Cf spontaneous fission spectrum. Influence of concrete wall thickness, air moisture and room size on the average cross section was analyzed. Correction factors amounted to about 30%. Corrected values corresponding to average cross sections in a pure 252Cf spectrum were calculated for 197Au, 232Th, 181Ta, 98Mo, 65Cu and 84Sr. Average cross sections were also calculated with the RR-UNC software using IRDFF-v.1.05 and ENDF/B-VII.0 libraries. The revised average radiative capture cross sections are 75.5±0.1 mb for 197Au, 87.0±1.6 mb for 232Th , 98.0±4.5 mb for 181Ta, 21.2±0.5 mb for 98Mo, 10.3±0.3 mb for 63Cu, and 34.9±6.5 mb for 84Sr. - Highlights: • Average capture cross sections in 252Cf spontaneous fission spectrum were simulated. • Calculations were done using MCNP5 code and ENDF/B-VII.0 library. • Correction factors for self-shielding and room return effects were taken into account. • The revised average radiative capture cross sections for different materials are published
Graphs of neutron cross sections in JSD1000 for radiation shielding safety analysis
Graphs of neutron cross sections and self-shielding factors in the JSD1000 library are presented for radiation shielding safety analysis. The compilation contains various reaction cross sections for 42 nuclides from 1H to 241Am in the energy range from 3.51 x 10-4 eV to 16.5 MeV. The Bondarenko-type self-shielding factors of each reaction are given by the background cross sections from σ0 = 0 to σ0 = 10000. (author)
Effects of cross sections tables generation and optimization on rod ejection transient analyses
Sánchez-Cervera Huerta, Santiago; García Herranz, Nuria; Herrero Carrascosa, José Javier; Cuervo Gómez, Diana
2014-01-01
Best estimate analysis of rod ejection transients requires 3D kinetics core simulators. If they use cross sections libraries compiled in multidimensional tables,interpolation errors – originated when the core simulator computes the cross sections from the table values – are a source of uncertainty in k-effective calculations that should be accounted for. Those errors depend on the grid covering the domain of state variables and can be easily reduced, in contrast with other sources of uncertai...
Cabellos de Francisco, Oscar Luis; Klix, A.; Fischer, U.; García Herranz, Nuria; Sanz Gonzalo, Javier; SIMAKOV S.
2011-01-01
The prediction of the tritium production is required for handling procedures of samples, safety & maintenance and licensing of the International Fusion Materials Irradiation Facility (IFMIF). A comparison of the evaluated tritium production cross-sections with available experimental data from the EXFOR data base has shown insufficient validation. And significant discrepancies in evaluated cross-section libraries, including lack of tritium production reactions for some important elements, were...
The derivation and evaluation of resonance cross sections applicable to light-water reactors
A rigorous approach to resonance cross section theory is adopted. The R-matrix theory of Wigner and Eisenbud is reviewed and presented in a form applicable to nuclear reactor technology. The single-level Breit-Wigner, multi-level Breit-Wigner, Vogt, and Reich-Moore cross section formulas are derived. The application of these formulas in the construction of nuclear data libraries is discussed
Systematics of the (n, np) rection cross sections at 14,5 MeV
A new six-parameter formula for the estimation of the sum of the (n, np) + (n, pn) + (n, d) reaction cross section is proposed at the incident neutron energy of 14.5 MeV. On the base of experimental data available new cross section data library was worked out for these reactions. The systematics provides better agreement with the experimental data in comparison with the systematics proposed earlier by other authors
Hart, S. W. D. [University of Tennessee, Knoxville (UTK); Maldonado, G. Ivan [University of Tennessee, Knoxville (UTK); Celik, Cihangir [ORNL; Leal, Luiz C [ORNL
2014-01-01
For many Monte Carlo codes cross sections are generally only created at a set of predetermined temperatures. This causes an increase in error as one moves further and further away from these temperatures in the Monte Carlo model. This paper discusses recent progress in the Scale Monte Carlo module KENO to create problem dependent, Doppler broadened, cross sections. Currently only broadening the 1D cross sections and probability tables is addressed. The approach uses a finite difference method to calculate the temperature dependent cross-sections for the 1D data, and a simple linear-logarithmic interpolation in the square root of temperature for the probability tables. Work is also ongoing to address broadening theS (alpha , beta) tables. With the current approach the temperature dependent cross sections are Doppler broadened before transport starts, and, for all but a few isotopes, the impact on cross section loading is negligible. Results can be compared with those obtained by using multigroup libraries, as KENO currently does interpolation on the multigroup cross sections to determine temperature dependent cross-sections. Current results compare favorably with these expected results.
We are developing a method of (n,α) cross section measurement using gaseous samples in a gridded ionization chamber (GIC). This method enables cross section measurements in large solid angle without the distortion by the energy loss in a sample, but requires a method to estimate the detection efficiency. We solve this problem by using GIC signals and a tight neutron collimation. The validity of this method was confirmed through the 12C(n,α0)9Be measurement. We applied this method to the 16O(n,α)13C cross section around 14.1 MeV. (author)
Sanami, Toshiya; Baba, Mamoru; Saito, Keiichiro; Ibara, Yasutaka; Hirakawa, Naohiro [Tohoku Univ., Sendai (Japan). Faculty of Engineering
1997-03-01
We are developing a method of (n,{alpha}) cross section measurement using gaseous samples in a gridded ionization chamber (GIC). This method enables cross section measurements in large solid angle without the distortion by the energy loss in a sample, but requires a method to estimate the detection efficiency. We solve this problem by using GIC signals and a tight neutron collimation. The validity of this method was confirmed through the {sup 12}C(n,{alpha}{sub 0}){sup 9}Be measurement. We applied this method to the {sup 16}O(n,{alpha}){sup 13}C cross section around 14.1 MeV. (author)
Structured ion impact: Doubly differential cross sections
The electron emission in coincidence with a projectile that has been ionized has been measured, thus making it possible to separate and identify electrons resulting from these various mechanisms. In 1985, coincidence doubly differential cross sections were measured for 400 to 750 keV/atomic mass unit (amu) He+ impact on He, Ne, Ar, Kr, and H2O. Cross sections were measured for selected angles and for electron energies ranging from 10 to 1000 eV. Because of the coincidence mode of measurement, the total electron emission was subdivided into its target emission and its projectile emission components. The most interesting findings were that target ionization does not account for the electron emission spectrum at lower electron energies. A sizable percentage of these low-energy electrons were shown to originate as a result of simultaneous projectile/target ionizations. Similar features were observed for all targets and impact energies that were studied
Elliptical cross section fuel rod study II
In this paper it is continued the behavior analysis and comparison between cylindrical fuel rods of circular and elliptical cross sections. Taking into account the accepted models in the literature, the fission gas swelling and release were studied. An analytical comparison between both kinds of rod reveals a sensible gas release reduction in the elliptical case, a 50% swelling reduction due to intragranular bubble coalescence mechanism and an important swelling increase due to migration bubble mechanism. From the safety operation point of view, for the same linear power, an elliptical cross section rod is favored by lower central temperatures, lower gas release rates, greater gas store in ceramic matrix and lower stored energy rates. (author). 6 refs., 8 figs., 1 tab
Jet cross sections and PDF constraints
CMS Collaboration
2012-01-01
A measurement of inclusive jet and dijet production cross sections is presented. Data from LHC proton-proton collisions at $\\sqrt{s}=7\\TeV$, corresponding to $4.67\\fbinv$ of integrated luminosity, have been collected with the CMS detector. Jets are reconstructed with the anti-$k_T$ clustering algorithm of size parameter $R=0.7$, extending to rapidity $|y|=2.5$, transverse momentum $\\pt=2\\TeV$, and dijet invariant mass $M_{JJ}=5\\TeV$. The measured cross sections are corrected for detector effects and compared to perturbative QCD predictions at next-to-leading order, using various sets of parton distribution functions.
The photoneutron cross section of 20Ne
The photoneutron cross section of 20Ne has been measured over a photon energy range 16 to 29 MeV in steps of 100 keV. The giant dipole resonance is resolved into three strong peaks below 21 MeV and at least two broader resonances at higher excitations. This structure is consistent with earlier measurements of poorer resolution and shows a correlation with the recent calculations of Schmid and Do Dang. Comparisons with high resolution neutron time-of-flight and electron scattering data indicate that there appear to exist in the giant resonance of 20Ne, regions of structure roughly 2-3 MeV wide which exhibit localised characteristics related to the excitation mechanisms. The role of deformation and configuration splitting effects in the cross section are discussed and possible directions of further study are noted which might clarify the situation more fully
Electron capture cross sections for stellar nucleosynthesis
Giannaka, P G
2015-01-01
In the first stage of this work, we perform detailed calculations for the cross sections of the electron capture on nuclei under laboratory conditions. Towards this aim we exploit the advantages of a refined version of the proton-neutron quasi-particle random-phase approximation (pn-QRPA) and carry out state-by-state evaluations of the rates of exclusive processes that lead to any of the accessible transitions within the chosen model space. In the second stage of our present study, we translate the above mentioned $e^-$-capture cross sections to the stellar environment ones by inserting the temperature dependence through a Maxwell-Boltzmann distribution describing the stellar electron gas. As a concrete nuclear target we use the $^{66}Zn$ isotope, which belongs to the iron group nuclei and plays prominent role in stellar nucleosynthesis at core collapse supernovae environment.
Cross-section analysis for TRADE fuel
The TRIGA core includes bounded hydrogen in Zirconium hydride in its fuel meat allowing for fast reactivity transients. The inherent safety mechanism is based on the immediate increase of neutron up-scattering by the hydrogen as a result of a fuel temperature increase. The temperature dependent resonance absorption is the second safety feature. The special fuel type together with the introduction of an external source within it for the TRADE project necessitates an accurate evaluation of the bounded hydrogen cross section generation technique as well as of the resonance treatment. By comparing deterministic tools and Monte Carlo solution methods the generated bounded isotopes cross sections are analysed. Further, the importance of the Doppler and the thermal up-scattering effects are quantified and the sensitivities to the solution method are discussed. (authors)
Measurements of neutron spallation cross section. 2
Kim, E.; Nakamura, T. [Tohoku Univ., Sendai (Japan). Cyclotron and Radioisotope Center; Imamura, M.; Nakao, N.; Shibata, S.; Uwamino, Y.; Nakanishi, N.; Tanaka, Su.
1997-03-01
Neutron spallation cross section of {sup 59}Co(n,xn){sup 60-x}Co, {sup nat}Cu(n,sp){sup 56}Mn, {sup nat}Cu(n,sp){sup 58}Co, {sup nat}Cu(n,xn){sup 60}Cu, {sup nat}Cu(n,xn){sup 61}Cu and {sup nat}Cu(n,sp){sup 65}Ni was measured in the quasi-monoenergetic p-Li neutron fields in the energy range above 40 MeV which have been established at three AVF cyclotron facilities of (1) INS of Univ. of Tokyo, (2) TIARA of JAERI and (3) RIKEN. Our experimental data were compared with the ENDF/B-VI high energy file data by Fukahori and the calculated cross section data by Odano. (author)
Neutron capture cross section measurement techniques
A review of currently-used techniques to measure neutron capture cross sections is presented. Measurements involving use of total absorption and Moxon-Rae detectors are based on low-resolution detection of the prompt γ-ray cascades following neutron captures. In certain energy ranges activation methods are convenient and useful. High resolution γ-ray measurements with germanium detectors can give information on the parameters of resonance capture states. The use of these techniques is described. (U.S.)
Fusion cross sections at deep subbarrier energies
Hagino, K.; Rowley, N.; Dasgupta, M
2003-01-01
A recent publication reports that heavy-ion fusion cross sections at extreme subbarrier energies show a continuous change of their logarithmic slope with decreasing energy, resulting in a much steeper excitation function compared with theoretical predictions. We show that the energy dependence of this slope is partly due to the asymmetric shape of the Coulomb barrier, that is its deviation from a harmonic shape. We also point out that the large low-energy slope is consistent with the surprisi...
How to calculate colourful cross sections efficiently
Gleisberg, Tanju; Krauss, Frank
2008-01-01
Different methods for the calculation of cross sections with many QCD particles are compared. To this end, CSW vertex rules, Berends-Giele recursion and Feynman-diagram based techniques are implemented as well as various methods for the treatment of colours and phase space integration. We find that typically there is only a small window of jet multiplicities, where the CSW technique has efficiencies comparable or better than both of the other two methods.
Jet cross sections in leptoproduction from QCD
We have calculated the longitudinal and other polarization dependent cross sections for jet production in deep inelastic ep, νp and anti νp scattering up to order αsub(s) of the quark-gluon coupling constant. Fragmentation of final state partons into hadrons is taken into account. Distributions in thrust, p2sub(Tin) and p2sub(Tout) are predicted for all three reactions and various values of W and Q. (orig.)
Measurements of Fission Cross Sections of Actinides
Wiescher, M; Cox, J; Dahlfors, M
2002-01-01
A measurement of the neutron induced fission cross sections of $^{237}$Np, $^{241},{243}$Am and of $^{245}$Cm is proposed for the n_TOF neutron beam. Two sets of fission detectors will be used: one based on PPAC counters and another based on a fast ionization chamber (FIC). A total of 5x10$^{18}$ protons are requested for the entire fission measurement campaign.
Neutron cross section standards and instrumentation
1992-09-01
This report from the National Institute of Standards and Technology contains a summary of the accomplishments of the Neutron Cross Section Standards and Instrumentation Project during the second year of a three-year interagency agreement. This program includes a broad range of data measurements and evaluations. An emphasis has been focused on the (sup 10)B cross sections where serious discrepancies in the nuclear data base remain. In particular, there are important problems with the interpretation of the helium gas production associated with diagnostic measurements of interest in nuclear technology. The enhanced use of this isotope for medical treatment is also of significance. New measurements of neutron reaction cross sections for (sup 10)B are in progress in collaboration with scientists at the Oak Ridge National Laboratory. New experiments are in progress on the important dosimetry standards (sup 237)Np(n,f) and (sup 239)Pu(n,f) below 1 MeV neutron energy. In addition, new measurements of charged-particle production in basic biological elements for medical applications are underway. Further measurements are planned or in progress in collaborations which include fission fragment energy and angular distributions, and neutron energy spectra and angular distributions from neutron-induced fission. Also measurements of angular distributions of neutrons from scattering on protons, and determinations of capture cross section of gold are planned for a later time. Data evaluation will shift to include a unified international effort to motivate new measurements and evaluations. In response to the requests of the measurement community, NIST is beginning the formation of a national depository for fissionable isotope mass standards. This action will preserve for future measurements the valuable and irreplaceable critical samples whose masses and composition have been carefully determined and documented over the past 30 years of the nuclear program.
Electron collision cross sections and radiation chemistry
A survey is given of the cross section data needs in radiation chemistry, and of the recent progress in electron impact studies on dissociative excitation of molecules. In the former some of the important target species, processes, and collision energies are presented, while in the latter it is demonstrated that radiation chemistry is a source of new ideas and information in atomic collision research. 37 references, 4 figures
Atomic-process cross section data, 1
Compiled by the Data Study Group, the data are intended for fusion plasma physics research. Cross sections of the latest experimental and theoretic studies cover the processes involving H,D,T as principal plasma materials as well as photons and electrons: emission and absorption of electromagnetic wave, electron collision, ion collision, recombination, neutral atom mutual collision, etc. Edition is so made to enable the future renewal by users. (J.P.N.)
Cross section of the CMS solenoid
Tejinder S. Virdee, CERN
2005-01-01
The pictures show a cross section of the CMS solenoid. One can see four layers of the superconducting coil, each of which contains the superconductor (central part, copper coloured - niobium-titanium strands in a copper coating, made into a "Rutherford cable"), surrounded by an ultra-pure aluminium as a magnetic stabilizer, then an aluminium alloy as a mechanical stabilizer. Besides the four layers there is an aluminium mechanical piece that includes pipes that transport the liquid helium.
Neutron cross section standards and instrumentation
This report from the National Institute of Standards and Technology contains a summary of the accomplishments of the Neutron Cross Section Standards and Instrumentation Project during the second year of a three-year interagency agreement. This program includes a broad range of data measurements and evaluations. An emphasis has been focused on the 10B cross sections where serious discrepancies in the nuclear data base remain. In particular, there are important problems with the interpretation of the helium gas production associated with diagnostic measurements of interest in nuclear technology. The enhanced use of this isotope for medical treatment is also of significance. New measurements of neutron reaction cross sections for 10B are in progress in collaboration with scientists at the Oak Ridge National Laboratory. New experiments are in progress on the important dosimetry standards 237Np(n,f) and 239Pu(n,f) below 1 MeV neutron energy. In addition, new measurements of charged-particle production in basic biological elements for medical applications are underway. Further measurements are planned or in progress in collaborations which include fission fragment energy and angular distributions, and neutron energy spectra and angular distributions from neutron-induced fission. Also measurements of angular distributions of neutrons from scattering on protons, and determinations of capture cross section of gold are planned for a later time. Data evaluation will shift to include a unified international effort to motivate new measurements and evaluations. In response to the requests of the measurement community, NIST is beginning the formation of a national depository for fissionable isotope mass standards. This action will preserve for future measurements the valuable and irreplaceable critical samples whose masses and composition have been carefully determined and documented over the past 30 years of the nuclear program
The Pa-233 fission cross section
The energy dependent neutron-induced fission cross section of 233Pa has for the first time been measured directly with mono-energetic neutrons. This isotope is produced in the thorium fuel cycle and serves as an intermediate step between the 232Th source material and the 233U fuel material. Four neutron energies between 1.0 and 3.0 MeV have been measured in a first campaign. Some preliminary results are presented and compared to literature. (author)
Fusion cross sections measurements with MUSIC
Carnelli, P. F. F.; Fernández Niello, J. O.; Almaraz-Calderon, S.; Rehm, K. E.; Albers, M.; Digiovine, B.; Esbensen, H.; Henderson, D.; Jiang, C. L.; Nusair, O.; Palchan-Hazan, T.; Pardo, R. C.; Ugalde, C.; Paul, M.; Alcorta, M.; Bertone, P. F.; Lai, J.; Marley, S. T.
2014-09-01
The interaction between exotic nuclei plays an important role for understanding the reaction mechanism of the fusion processes as well as for the energy production in stars. With the advent of radioactive beams new frontiers for fusion reaction studies have become accessible. We have performed the first measurements of the total fusion cross sections in the systems 10 , 14 , 15C + 12C using a newly developed active target-detector system (MUSIC). Comparison of the obtained cross sections with theoretical predictions show a good agreement in the energy region accessible with existing radioactive beams. This type of comparison allows us to calibrate the calculations for cases that cannot be studied in the laboratory with the current experimental capabilities. The high efficiency of this active detector system will allow future measurements with even more neutron-rich isotopes. The interaction between exotic nuclei plays an important role for understanding the reaction mechanism of the fusion processes as well as for the energy production in stars. With the advent of radioactive beams new frontiers for fusion reaction studies have become accessible. We have performed the first measurements of the total fusion cross sections in the systems 10 , 14 , 15C + 12C using a newly developed active target-detector system (MUSIC). Comparison of the obtained cross sections with theoretical predictions show a good agreement in the energy region accessible with existing radioactive beams. This type of comparison allows us to calibrate the calculations for cases that cannot be studied in the laboratory with the current experimental capabilities. The high efficiency of this active detector system will allow future measurements with even more neutron-rich isotopes. This work is supported by the U.S. DOE Office of Nuclear Physics under Contract No. DE-AC02-06CH11357 and the Universidad Nacional de San Martin, Argentina, Grant SJ10/39.
Inclusive jet cross section at D0
Bhattacharjee, M. [Delhi Univ. (India). Dept. of Physics and Astrophysics
1996-09-01
Preliminary measurement of the central ({vert_bar}{eta}{vert_bar} {<=} 0.5) inclusive jet cross sections for jet cone sizes of 1.0, 0.7, and 0.5 at D{null} based on the 1992-1993 (13.7 {ital pb}{sup -1}) and 1994-1995 (90 {ital pb}{sup -1}) data samples are presented. Comparisons to Next-to-Leading Order (NLO) Quantum Chromodynamics (QCD) calculations are made.
Cross-section reconstruction during uniaxial loading
The inelastic response of materials to applied uniaxial loading is typically measured using tensile or compressive specimens of an initially circular cross-section. Under deformation, this cross-section may become elliptical due to anisotropic material behaviour. An optical technique for measuring the elliptical deformation of anisotropic, homogeneous cylindrical specimens undergoing uniaxial deformation is presented. It enables the quantification of anisotropic deformation in situ and provides data for material characterization. Three or more silhouette views of a specimen are obtained using multiple cameras or mirrored views. The positions of the edges are computed using a sub-pixel edge detection method, and 3D tangent rays from the camera through these positions are calculated. These bounding tangents are used as the basis for an elliptical fit by least squares at cross-sections along the length of the specimen. Stochastic error estimates are performed by simulation of the experiment. Error estimates, for the experimental set-up used, are also calculated by reconstructing elliptical prisms of precisely measured dimensions. Example reconstructions from specimens of rolled titanium deformed plastically in tension at quasi-static (7 × 10−4 s−1) and high strain rates (3 × 103 s−1) are presented
Effects of cross sections tables generation and optimization on rod ejection transient analyses
Highlights: • Different cross-section libraries are applied to a rod ejection transient benchmark. • Effects of the optimization of the library grid-point distribution are assessed. • Effects of the library generation are assessed by comparison with other solutions. • Interpolation errors contribute to neutronics uncertainties in modeling transients. - Abstract: Best estimate analysis of rod ejection transients requires 3D kinetics core simulators. If they use cross sections libraries compiled in multidimensional tables, interpolation errors – originated when the core simulator computes the cross sections from the table values – are a source of uncertainty in k-effective calculations that should be accounted for. Those errors depend on the grid covering the domain of state variables and can be easily reduced, in contrast with other sources of uncertainties such as the ones due to nuclear data, by choosing an optimized grid distribution. The present paper assesses the impact of the grid structure on a PWR rod ejection transient analysis using the coupled neutron-kinetics/thermal-hydraulics COBAYA3/COBRA-TF system. For this purpose, the OECD/NEA PWR MOX/UO2 core transient benchmark has been chosen, as material compositions and geometries are available, allowing the use of lattice codes to generate libraries with different grid structures. Since a complete nodal cross-section library is also provided as part of the benchmark specifications, the effects of the library generation on transient behavior are also analyzed. Results showed large discrepancies when using the benchmark library and own-generated libraries when compared with benchmark participants’ solutions. The origin of the discrepancies was found to lie in the nodal cross sections provided in the benchmark
POLIDENT: A Module for Generating Continuous-Energy Cross Sections from ENDF Resonance Data
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
POLIDENT: A Module for Generating Continuous-Energy Cross Sections from ENDF Resonance Data
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.
POLIDENT: A Module for Generating Continuous-Energy Cross Sections from ENDF Resonance Data
Dunn, M.E.
2000-10-20
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.
Evaluation of {sup 235}U, {sup 238}U, {sup 6}Li, and {sup 27}Al Cross Sections
Chandler, J.R.
2001-05-17
Good nuclear data are essential for accurate prediction of reactor parameters. Several cross section libraries are currently available for use with GLASS physics calculations. In recent Mark 15 and Mark 22 studies, cross section data were developed to provide more accurate buckling calculations for Mark 15 and Mark 22 charges. This report documents evaluation of these new data for universal application.
Averaging cross section data so we can fit it
Brown, D. [Brookhaven National Lab. (BNL), Upton, NY (United States). NNDC
2014-10-23
The ^{56}Fe cross section we are interested in have a lot of fluctuations. We would like to fit the average of the cross section with cross sections calculated within EMPIRE. EMPIRE is a Hauser-Feshbach theory based nuclear reaction code, requires cross sections to be smoothed using a Lorentzian profile. The plan is to fit EMPIRE to these cross sections in the fast region (say above 500 keV).
Averaging cross section data so we can fit it
The 56Fe cross section we are interested in have a lot of fluctuations. We would like to fit the average of the cross section with cross sections calculated within EMPIRE. EMPIRE is a Hauser-Feshbach theory based nuclear reaction code, requires cross sections to be smoothed using a Lorentzian profile. The plan is to fit EMPIRE to these cross sections in the fast region (say above 500 keV).
Reinforced concrete columns of variable cross section
Brant, N.F.A.
1984-01-01
The results of a series of 19 full scale tests carried out on pin-ended reinforced concrete columns are reported. The columns tested had either tapered rectangular sections along the length or octagonal cross sections. All columns, except the last 6, were subjected to uniaxial eccentricities at one of the ends (the stronger end), and a nominally concentric load at the other end. For the case of the last six columns the loading applied at the stronger end was biaxially eccentric. For each of t...
Nuclear interaction cross sections for proton radiotherapy
Chadwick, M B; Arendse, G J; Cowley, A A; Richter, W A; Lawrie, J J; Newman, R T; Pilcher, J V; Smit, F D; Steyn, G F; Koen, J W; Stander, J A
1999-01-01
Model calculations of proton-induced nuclear reaction cross sections are described for biologically-important targets. Measurements made at the National Accelerator Centre are presented for double-differential proton, deuteron, triton, helium-3 and alpha particle spectra, for 150 and 200 MeV protons incident on C, N, and O. These data are needed for Monte Carlo simulations of radiation transport and absorbed dose in proton therapy. Data relevant to the use of positron emission tomography to locate the Bragg peak are also described.
Neutron capture cross section of $^{93}$Zr
We propose to measure the neutron capture cross section of the radioactive isotope $^{93}$Zr. This project aims at the substantial improvement of existing results for applications in nuclear astrophysics and emerging nuclear technologies. In particular, the superior quality of the data that can be obtained at n_TOF will allow on one side a better characterization of s-process nucleosynthesis and on the other side a more accurate material balance in systems for transmutation of nuclear waste, given that this radioactive isotope is widely present in fission products.
Charge changing cross sections of relativistic uranium
We report equilibrium charge state distributions of uranium at energies of 962 MeV/nucleon, 437 MeV/nucleon and 200 MeV/nucleon in low Z and high Z targets and the cross sections for U92+ reversible U91+ and U91+ reversible U90+ at 962 MeV/nucleon and 437 MeV/nucleon. Equilibrium thickness Cu targets produce approx. = 5% bare U92+ at 200 MeV/nucleon and 85% U92+ at 962 MeV/nucleon. 7 references, 5 figures
Fission cross section measurements for minor actinides
Fursov, B. [IPPE, Obninsk (Russian Federation)
1997-03-01
The main task of this work is the measurement of fast neutron induced fission cross section for minor actinides of {sup 238}Pu, {sup 242m}Am, {sup 243,244,245,246,247,248}Cm. The task of the work is to increase the accuracy of data in MeV energy region. Basic experimental method, fissile samples, fission detectors and electronics, track detectors, alpha counting, neutron generation, fission rate measurement, corrections to the data and error analysis are presented in this paper. (author)
A New Neutrino Cross Section Data Ressource
Whalley, M R
2005-01-01
We describe a new web based data resource being developed to provide access to accurate and validated cross sections of low energy neutrino and antineutrino interactions. The proposed content of this database are outlined which cover total and differential cross from inclusive, quasi-elastic and exclusive pion production processes from charged and neutral current interactions. Efforts to obtain these data, which come mainly from old bubble chamber experiments, are described as well as the implementation of an embryonic web site to make the resource generally accessible.
LEP vacuum chamber, cross-section
1983-01-01
Cross-section of the final prototype for the LEP vacuum chamber. The elliptic main-opening is for the beam. The small channel to the left is for the cooling water, to carry away the heat deposited by the synchrotron radiation. The square channel to the right houses the Non-Evaporable Getter (NEG) pump. The chamber is made from extruded aluminium. Its outside is clad with lead, to stop the synchrotron radiation emitted by the beam. For good adherence between Pb and Al, the Al chamber was coated with a thin layer of Ni. Ni being slightly magnetic, some resulting problems had to be overcome. See also 8301153.
Critical behavior of cross sections at LHC
Dremin, I M
2016-01-01
Recent experimental data on elastic scattering of high energy protons show that the critical regime has been reached at LHC energies. The approach to criticality is demonstrated by increase of the ratio of elastic to total cross sections from ISR to LHC energies. At LHC it reaches the value which can result in principal change of the character of proton interactions. The treatment of new physics of hollowed toroid-like hadrons requires usage of another branch of the unitarity condition. Its further fate is speculated and interpreted with the help of the unitarity condition in combination with present experimental data. The gedanken experiments to distinguish between different possibilities are proposed.
Neutron absorption cross section of uranium-236
U-236 neutron absorption was measured as a function of neutron time-of-flight from 20 eV to 1 MeV. The neutron flux was monitored with a 6Li glass scintillator. Average cross sections from 3 keV to 1 MeV were derived. Estimated uncertainties were less than 5% below 600 keV and increased to 9.5% at 1 MeV. Resonance parametrization from 20 eV to a few keV remains to be done. 17 refs., 5 figs., 3 tabs
New numerical methods for nuclear cross section processing
Nuclear data allow to describe how a particle interacts with matter. These data are therefore at the basis of neutron transport and reactor physics calculations. Once measured and evaluated, they are given in libraries as a list of parameters. Before they can be used in neutron transport calculations, processing is required which includes taking into account several physical phenomena. This can be done by several softwares, such as NJOY, which all have the drawback to use old numerical methods derived from the same algorithms. For nuclear safety applications, it is important to rely on independent methods, to have a comparison point and to isolate the effects of the treatment on the final results. Moreover, it is important to properly master processing accuracy during its different steps. The objective of this PhD is then to develop independent numerical methods that can guarantee nuclear data processing within a given precision and to implement them practically, with the creation of the GAIA software. Our first step was the reconstruction of cross sections from the parameters given in libraries, with different approximations of the R-matrix theory. Reconstruction using the general formalism, without any approximation, has also been implemented, which has required the development of a new method to calculate the R-matrix. Tests have been performed on all existing formalisms, including the newest one. They have shown a good agreement between GAIA and NJOY. Reconstruction of angular differential cross sections directly from R-matrix parameters, using the Blatt-Biedenharn formula, has also been implemented and tested. The cross sections we have obtained at this point correspond to a target nucleus at absolute zero temperature. Because of thermal agitation, these cross sections are subject to a Doppler effect that is taken into account by integrating them with Solbrig's kernel. Our second step was then to calculate this integral. First, we have elaborated and
Use of Neutron Benchmark Fields for the Validation of Dosimetry Cross Sections
Griffin, Patrick
2016-02-01
The evolution of validation metrics for dosimetry cross sections in neutron benchmark fields is explored. The strength of some of the metrics in providing validation evidence is examined by applying them to the 252Cf spontaneous fission standard neutron benchmark field, the 235U thermal neutron fission reference benchmark field, the ACRR pool-type reactor central cavity reference benchmark fields, and the SPR-III fast burst reactor central cavity. The IRDFF dosimetry cross section library is used in the validation study and observations are made on the amount of coverage provided to the library contents by validation data available in these benchmark fields.
Computation of Resonance-Screened Cross Section by the Dorix-Speng System
The report describes a scheme for computation of group cross sections for fast reactors in energy regions where the resonance structure of the cross sections may be dense. A combination of the programmes Dorix and Speng is then used. Dorix calculates group cross sections for each resonance absorber separately. The interaction between resolved resonances in the same isotope is treated using a method described in a separate report. The interaction between correlated and non-correlated resonances in the unresolved region is also considered. By a Dorix calculation we obtain effective microscopic cross sections which are then read in on a library tape. This library contains both point-by-point data and group cross sections and is used in the Speng programme for computation of spectrum and/or macroscopic cross sections. The resonance interaction between different isotopes is computed in Speng by the same method as was used in the Dorix programme for non-correlated unresolved resonances. Consideration is also given to the width of the resonances compared to the energy loss by a neutron colliding with some of the scattering elements
A novel hybrid weighting scheme for multi-group cross section collapsing
Multi-group cross section library generation plays an important role in deterministic transport simulations. In this paper, a new fine-group to broad-group cross section collapsing method is introduced. Rather than a traditional flux weighting, the new method uses a hybrid weighing scheme to collapse the scattering cross section matrix. Based upon a matrix analysis approach, we generalize different weighting schemes and derive the new hybrid weighting scheme, which mathematically shows that it is rational for the scattering cross section to be weighted by the (1) forward fluxes of the incoming/in-bound neutron groups and (2) the adjoint functions of the outgoing/out-bound neutron energy groups. This approach also makes physical sense, since it conserves the “importance flow” of particles through scattering while collapsing cross sections. To conserve the reaction rates at the same time, we re-normalize the hybrid weighted scattering cross section to the original library total scattering reaction rate. We demonstrate that the hybrid weighting scheme is more accurate, especially for the detector response simulation problem in a Dual-Range Coincidence Counter (DRCC) 3-D SN transport model. (author)
Measurement of 233U fission spectrum-averaged cross sections for some threshold reactions
The 233U fission spectrum-averaged cross sections for twelve threshold reactions were measured relative to the average cross section of 0.688 ± 0.040 mb for the 27Al(n,α)24Na reaction. The reference value was obtained by calculation using the energy dependent cross section in the Japanese Evaluated Nuclear Data Library (JENDL) Dosimetry File and the Watt-type fission spectrum in ENDF/B-VI. General agreement was seen between the measured and the calculated fission-spectrum averaged cross sections. However, there exist discrepancies of more than 10% between the measured and the calculated average cross sections for the 24Mg(n,p)24Na, 47Ti(n,p)47Sc, and 64Zn(n,p)64Cu reactions. The tendencies in the calculated-to-measured ratios are similar to those for 235U fission spectrum-averaged cross sections the authors previously measured. The measured average cross sections were also applied for the spectrum adjustment of the 233U fission neutrons using the Neutron Unfolding Package Code (NEUPAC). The adjusted spectrum is close to the Watt-type fission spectrum of 233U within the uncertainties of the obtained spectrum, although there exist some fluctuations in the ratio spectrum of the adjusted to the Watt-type
The Elusive p-air Cross Section
Block, Martin M
2006-01-01
For the $\\pbar p$ and $pp$ systems, we have used all of the extensive data of the Particle Data Group[K. Hagiwara {\\em et al.} (Particle Data Group), Phys. Rev. D 66, 010001 (2002).]. We then subject these data to a screening process, the ``Sieve'' algorithm[M. M. Block, physics/0506010.], in order to eliminate ``outliers'' that can skew a $\\chi^2$ fit. With the ``Sieve'' algorithm, a robust fit using a Lorentzian distribution is first made to all of the data to sieve out abnormally high $\\delchi$, the individual i$^{\\rm th}$ point's contribution to the total $\\chi^2$. The $\\chi^2$ fits are then made to the sieved data. We demonstrate that we cleanly discriminate between asymptotic $\\ln s$ and $\\ln^2s$ behavior of total hadronic cross sections when we require that these amplitudes {\\em also} describe, on average, low energy data dominated by resonances. We simultaneously fit real analytic amplitudes to the ``sieved'' high energy measurements of $\\bar p p$ and $pp$ total cross sections and $\\rho$-values for $\\...
Total neutron cross section for 181Ta
Schilling K.-D.
2010-10-01
Full Text Available The neutron time of flight facility nELBE, produces fast neutrons in the energy range from 0.1 MeV to 10 MeV by impinging a pulsed relativistic electron beam on a liquid lead circuit [1]. The short beam pulses (∼10 ps and a small radiator volume give an energy resolution better than 1% at 1 MeV using a short flight path of about 6 m, for neutron TOF measurements. The present neutron source provides 2 ⋅ 104 n/cm2s at the target position using an electron charge of 77 pC and 100 kHz pulse repetition rate. This neutron intensity enables to measure neutron total cross section with a 2%–5% statistical uncertainty within a few days. In February 2008, neutron radiator, plastic detector [2] and data acquisition system were tested by measurements of the neutron total cross section for 181Ta and 27Al. Measurement of 181Ta was chosen because lack of high quality data in an anergy region below 700 keV. The total neutron cross – section for 27Al was measured as a control target, since there exists data for 27Al with high resolution and low statistical error [3].
Cross-section measurements for radioactive samples
The measurement of (n,p), (n,α) and (n,γ) cross sections for radioactive nuclei is of interest to both nuclear physics and astrophysics. For example, using these reactions, properties of levels in nuclei at high excitation energies, which are difficult or impossible to study using other reactions, can be investigated. Also, reaction rates for both big-bang and stellar nucleosynthesis can be obtained from these measurements. In the past, the large background associated with the sample activity limited these types of measurements to radioisotopes with very long half-lives. The advent of the low-energy, high-intensity neutron source at the Los Alamos Neutron Scattering CEnter (LANSCE) has greatly increased the number of nuclei which can be studied. Examples of (n,p) measurements on samples with half lives as short as fifty-three days will be given. The nuclear physics and astrophysics to be learned from these data will be discussed. Additional difficulties are encountered when making (n,γ) rather than (n,p) or (n,α) measurements. However, with a properly-designed detector, and the high peak neutron intensities now available, (n,γ) measurements can be made for nuclei with half lives as short as several months. Progress on the Los Alamos (n,γ) cross-section measurement program for radioactive samples will be discussed. 39 refs., 7 figs
Calculation of cross sections for heavy isotopes
In the present work an integrated system of codes for basic neutron data evaluation were assembled and built. Complete evaluations for the isotopes 240Pu, 241Pu, 242Pu and 238Pu were performed. The following cross sections: total, elastic, radiative capture, fission, total inelastic, partial inelastic, (n,2n), (n,3n) and differential elastic were evaluated as well as the average number of neutrons per neutron-induced fission and the average elastic scattering cosine in the lab system.The data for the plutonium isotopes were incorporated into the German KEDAK file. A method was developed for calculating the energy distributions of the second and third secondary neutrons from the A(n,2n) and (n,3n) reactions in the framework of the compound nucleus theory, and utilizing the nuclear data of the nuclei A, A-1, A-2. This method was used to generate the 238U secondary neutron energy distributions in the incident neutron energy range of 6 to 15 MeV. A nuclear data evaluation for 237U in the resolved inelastic scattering range (10-700 keV) was performed. The compound elastic and partial inelastic scattering cross sections were used in the 238U secondary neutron energy distribution calculations. (B.G.)
The prediction of the tritium production is required for handling procedures of samples, safety and maintenance and licensing of the International Fusion Materials Irradiation Facility (IFMIF). A comparison of the evaluated tritium production cross-sections with available experimental data from the EXFOR data base has shown insufficient validation. And significant discrepancies in evaluated cross-section libraries, including lack of tritium production reactions for some important elements, were found. Here, we have addressed an uncertainty analysis to draw conclusions on the reliability of the tritium prediction under the potential impact of activation cross-section uncertainties. We conclude that there is not sufficient experimental validation of the evaluated tritium production cross-sections, especially for iron and sodium. Therefore a dedicated experimental validation program for those elements should be desirable.
Validation of cross sections for Monte Carlo simulation of the photoelectric effect
Han, Min Cheol; Pia, Maria Grazia; Basaglia, Tullio; Batic, Matej; Hoff, Gabriela; Kim, Chan Hyeong; Saracco, Paolo
2016-01-01
Several total and partial photoionization cross section calculations, based on both theoretical and empirical approaches, are quantitatively evaluated with statistical analyses using a large collection of experimental data retrieved from the literature to identify the state of the art for modeling the photoelectric effect in Monte Carlo particle transport. Some of the examined cross section models are available in general purpose Monte Carlo systems, while others have been implemented and subjected to validation tests for the first time to estimate whether they could improve the accuracy of particle transport codes. The validation process identifies Scofield's 1973 non-relativistic calculations, tabulated in the Evaluated Photon Data Library(EPDL), as the one best reproducing experimental measurements of total cross sections. Specialized total cross section models, some of which derive from more recent calculations, do not provide significant improvements. Scofield's non-relativistic calculations are not surp...
Search for Optimum Subgroup Levels for Minimizing Errors in Resonance Shielded Cross Sections
The subgroup method is one of the most employed methods for resonance treatment in the lattice transport or the direct whole core transport codes such as HELIOS and DeCART. It requires the subgroup parameters which consist of subgroup levels and subgroup weights. Subgroup weights are produced from the given subgroup levels by solving an error minimization problem for the resonance shielded effective cross sections. The subgroup parameters have a significant impact on the accuracy of the effective cross section which is estimated by the subgroup method. The available subgroup levels for each resonance group of the existing libraries were not been thoroughly optimized. The purpose of this work is to devise a way to determine proper subgroup levels which can further reduce the error in the effective cross section errors. Needless to say, more correct resonance effective cross sections would improve the accuracy of the lattice transport or the direct whole transport calculation.
Search for Optimum Subgroup Levels for Minimizing Errors in Resonance Shielded Cross Sections
Lee, Kyong Seop; Kim, Gwan Young; Joo, Han Gyu [Seoul National University, Seoul (Korea, Republic of)
2010-05-15
The subgroup method is one of the most employed methods for resonance treatment in the lattice transport or the direct whole core transport codes such as HELIOS and DeCART. It requires the subgroup parameters which consist of subgroup levels and subgroup weights. Subgroup weights are produced from the given subgroup levels by solving an error minimization problem for the resonance shielded effective cross sections. The subgroup parameters have a significant impact on the accuracy of the effective cross section which is estimated by the subgroup method. The available subgroup levels for each resonance group of the existing libraries were not been thoroughly optimized. The purpose of this work is to devise a way to determine proper subgroup levels which can further reduce the error in the effective cross section errors. Needless to say, more correct resonance effective cross sections would improve the accuracy of the lattice transport or the direct whole transport calculation.
Correction of multigroup cross sections for resolved resonance interference in mixed absorbers
The effect that interference between resolved resonances has on averaging multigroup cross sections is examined for thermal reactor-type problems. A simple and efficient numerical scheme is presented to correct a preprocessed multigroup library for interference effects. The procedure is implemented in a design oriented lattice physics computer code and compared with rigorous numerical calculations. The approximate method for computing resonance interference correction factors is applied to obtaining fine-group cross sections for a homogeneous uranium-plutonium mixture and a uranium oxide lattice. It was found that some fine group cross sections are changed by more than 40% due to resonance interference. The change in resonance interference correction factors due to burnup of a PWR fuel pin is examined and found to be small. The effect of resolved resonance interference on collapsed broad-group cross sections for thermal reactor calculations is discussed
This paper discusses recent developments and future plans for the SENSIBL code (the successor to the SENSIT[6] and SENSIT-2D[7] codes), along with associated covariance data and cross section libraries. 34 refs
The evaluation of cross sections for n + 6Li reaction
Neutron nuclear data of 6Li are important for fusion neutronics calculation. Therefore, the cross sections for n + 6Li reaction are evaluated in the energy range from 10-5 eV to 20 MeV. In the evaluation, 6Li(n, d+n)4He and 6Li(n, n+d)4He reactions are included. It is concluded that there is really only the second excited level (3.562 MeV) in the inelastic scattering, no assumed levels were taken into account. The evaluated data describe the real process of n + 6Li reactions and improve the existing evaluated libraries such as ENDF/B-6 and JENDL-3
Elastic cross sections in an RSIIp scenario
The elastic differential cross section is calculated at low energies (below 100 MeV) for the elements 3He, 20Ne, 40Ar, 14N, 12C, and for the 208Pb using a finite electromagnetic potential, which is obtained by considering a Randall–Sundrum II scenario modified by the inclusion of p compact extra-dimensions. The length scale is adjusted in the potential to compare with known experimental data and to set bounds for the parameter of the model. The effective four-dimensional (4D) electromagnetic potential is produced by a point charge, as seen from the three-brane that contains it, in uniform motion in an RSIIp scenario. (paper)
Lunar Radar Cross Section at Low Frequency
Rodriguez, P.; Kennedy, E. J.; Kossey, P.; McCarrick, M.; Kaiser, M. L.; Bougeret, J.-L.; Tokarev, Y. V.
2002-01-01
Recent bistatic measurements of the lunar radar cross-section have extended the spectrum to long radio wavelength. We have utilized the HF Active Auroral Research Program (HAARP) radar facility near Gakona, Alaska to transmit high power pulses at 8.075 MHz to the Moon; the echo pulses were received onboard the NASA/WIND spacecraft by the WAVES HF receiver. This lunar radar experiment follows our previous use of earth-based HF radar with satellites to conduct space experiments. The spacecraft was approaching the Moon for a scheduled orbit perturbation when our experiment of 13 September 2001 was conducted. During the two-hour experiment, the radial distance of the satellite from the Moon varied from 28 to 24 Rm, where Rm is in lunar radii.
Plasma-based radar cross section reduction
Singh, Hema; Jha, Rakesh Mohan
2016-01-01
This book presents a comprehensive review of plasma-based stealth, covering the basics, methods, parametric analysis, and challenges towards the realization of the idea. The concealment of aircraft from radar sources, or stealth, is achieved through shaping, radar absorbing coatings, engineered materials, or plasma, etc. Plasma-based stealth is a radar cross section (RCS) reduction technique associated with the reflection and absorption of incident electromagnetic (EM) waves by the plasma layer surrounding the structure. A plasma cloud covering the aircraft may give rise to other signatures such as thermal, acoustic, infrared, or visual. Thus it is a matter of concern that the RCS reduction by plasma enhances its detectability due to other signatures. This needs a careful approach towards the plasma generation and its EM wave interaction. The book starts with the basics of EM wave interactions with plasma, briefly discuss the methods used to analyze the propagation characteristics of plasma, and its generatio...
Calculated medium energy fission cross sections
An analysis has been made of medium-energy nucleon induced fission of 238U and 237Np using detailed models of fission, based upon the Bohr-Wheeler formalism. Two principal motivations were associated with these calculations. The first was determination of barrier parameters for proton-rich uranium and neptunium isotopes normally not accessible in lower energy reactions. The second was examination of the consistency between (p,f) experimental data versus new (n,f) data that has recently become available. Additionally, preliminary investigations were also made concerning the effect of fission dynamics on calculated fission cross sections at higher energies where neutron emission times may be significantly less than those associated with fission
Partial cross sections in H- photodetachment
This dissertation reports experimental measurements of partial decay cross sections in the H- photodetachment spectrum. Observed decays of the 1P0 H-**(n) doubly-excitedresonances to the H(N=2) continuum are reported for n=2,3, and 4 from 1990 runs in which the author participated. A recent analysis of 1989 data revealing effects of static electric fields on the partial decay spectrum above 13.5 eV is also presented. The experiments were performed at the High Resolution Atomic Beam Facility. the Los Alamos Meson Physics Facility, with a relativistic H-beam (β=0.842)intersecting a ND:YAG laser. Variation of the intersection angle amounts to Doppler-shifting the photon energy, allowing continuous tuning of the laser energy as viewed from the moving ions' frame
Radar Cross Section of Moving Objects
Gholizade, H
2013-01-01
I investigate the effects of movement on radar cross section calculations. The results show that relativistic effects (the constant velocity case) can change the RCS of moving targets by changing the incident plane wave field vectors. As in the Doppler effect, the changes in the fields are proportional to $\\frac{v}{c}$. For accelerated objects, using the Newtonian equations of motion yields an effective electric field (or effective current density) on the object due to the finite mass of the conducting electrons. The results indicate that the magnetic moment of an accelerated object is different from that of an un-accelerated object, and this difference can change the RCS of the object. Results for moving sphere and non-uniformly rotating sphere are given and compared with static (\\textbf{v}=0) case.
Kalpakkam multigroup cross section set for fast reactor applications - status and performance
This report documents the status of the presently created set of multigroup constants at Kalpakkam. The list of nuclides processed and the details of multigroup structure are given. Also included are the particulars of dilutions and temperatures for each nuclide in the multigroup cross section set for which self shielding factors have been calculated. Using this new multigroup cross section set, measured integral quantities such as K-eff, central reaction rate ratios, central reactivity worths etc. were calculated for a few fast critical benchmark assemblies and the calculated values of neutronic parameters obtained were compared with those obtained using the available Cadarache cross section library and those published in literature for ENDF/B-IV based set and Japanese evaluated nuclear data library (JENDL). The details of analyses are documented along with the conclusions. (author). 17 refs., 12 tabs
Optimization of multidimensional cross-section tables for few-group core calculations
Highlights: • Optimization of tabulated cross-sections libraries for multi-group diffusion codes. • Sensitivity coefficients using perturbation theory are determined. • A non-uniform grid satisfying a given target accuracy in k-effective is built. • Satisfactory results are obtained using libraries with different accuracy level. - Abstract: Multigroup diffusion codes for three dimensional LWR core analysis use as input data pre-generated homogenized few group cross sections and discontinuity factors for certain combinations of state variables, such as temperatures or densities. The simplest way of compiling those data are tabulated libraries, where a grid covering the domain of state variables is defined and the homogenized cross sections are computed at the grid points. Then, during the core calculation, an interpolation algorithm is used to compute the cross sections from the table values. Since interpolation errors depend on the distance between the grid points, a determined refinement of the mesh is required to reach a target accuracy, which could lead to large data storage volume and a large number of lattice transport calculations. In this paper, a simple and effective procedure to optimize the distribution of grid points for tabulated libraries is presented. Optimality is considered in the sense of building a non-uniform point distribution with the minimum number of grid points for each state variable satisfying a given target accuracy in k-effective. The procedure consists of determining the sensitivity coefficients of k-effective to cross sections using perturbation theory; and estimating the interpolation errors committed with different mesh steps for each state variable. These results allow evaluating the influence of interpolation errors of each cross section on k-effective for any combination of state variables, and estimating the optimal distance between grid points
Several new WIMSD libraries based on recent cross-section evaluations such as IAEA, ENDFB-VI, JENDL, and JEF have been made available by IAEA. These libraries were used for the computation of multiplication factor and energy spectrum for Pakistan Research Reactor-1 (PARR-1). Methodology was validated for benchmark problems made available by IAEA and comparison with reference results. The value of effective multiplication factors for all newly released libraries are 1.8-3.2% less than that of 1981 WIMSD library. The effect of various cross-section libraries on neutron energy spectrum was also studied. Differences of about -10% to 12.5% were found in thermal flux using the newly released libraries as compared with that obtained using 1981 WIMSD library. From the analysis, it was found that the main source of the difference is the cross-sections of hydrogen bound in water. When these cross-sections of hydrogen (bound in water) from new libraries were used along with all other data in 1981 WIMSD library, the keff obtained in this way has a difference of only 0.02-0.8% with that obtained from new libraries, while the flux spectrum agreed within 1% below 1 MeV with new libraries
Ahmad, Siraj-ul-Islam [Pakistan Institute of Engineering and Applied Sciences (PIEAS), PO Nilore, Islamabad 45650 (Pakistan)]. E-mail: siraj1@hotpop.com; Ahmad, Nasir [Pakistan Institute of Engineering and Applied Sciences (PIEAS), PO Nilore, Islamabad 45650 (Pakistan); Aslam [Pakistan Institute of Engineering and Applied Sciences (PIEAS), PO Nilore, Islamabad 45650 (Pakistan)
2004-11-01
Several new WIMSD libraries based on recent cross-section evaluations such as IAEA, ENDFB-VI, JENDL, and JEF have been made available by IAEA. These libraries were used for the computation of multiplication factor and energy spectrum for Pakistan Research Reactor-1 (PARR-1). Methodology was validated for benchmark problems made available by IAEA and comparison with reference results. The value of effective multiplication factors for all newly released libraries are 1.8-3.2% less than that of 1981 WIMSD library. The effect of various cross-section libraries on neutron energy spectrum was also studied. Differences of about -10% to 12.5% were found in thermal flux using the newly released libraries as compared with that obtained using 1981 WIMSD library. From the analysis, it was found that the main source of the difference is the cross-sections of hydrogen bound in water. When these cross-sections of hydrogen (bound in water) from new libraries were used along with all other data in 1981 WIMSD library, the k{sub eff} obtained in this way has a difference of only 0.02-0.8% with that obtained from new libraries, while the flux spectrum agreed within 1% below 1 MeV with new libraries.
Single-level resonance parameters fit nuclear cross-sections
Drawbaugh, D. W.; Gibson, G.; Miller, M.; Page, S. L.
1970-01-01
Least squares analyses of experimental differential cross-section data for the U-235 nucleus have yielded single level Breit-Wigner resonance parameters that fit, simultaneously, three nuclear cross sections of capture, fission, and total.
Residual diagnostics for cross-section time series regression models
Baum, Christopher F
2001-01-01
These routines support the diagnosis of groupwise heteroskedasticity and cross-sectional correlation in the context of a regression model fit to pooled cross-section time series (xt) data. Copyright 2001 by Stata Corporation.
Photoneutron cross sections measured by Saclay and Livermore
The differences between the Saclay and Livermore photoneutron cross sections are discussed. It is shown that the differences between Saclay and Livermore (γ,n) and (γ,2n) cross sections arise from the neutron multiplicity sorting. (Author)
There are numerous uncertainties in the analyses of innovative reactor designs, arising from approximations used in the solution of the transport equation, and in nuclear data processing and cross section libraries generation. This paper describes: the problems encountered in the analysis of the lead cooled and reflected reactors; the new cross section data libraries developed to overcome these problems; and applications of these new data libraries to the Encapsulated Nuclear Heat Source (ENHS) core benchmark analysis. The ENHS is a new lead-bismuth or lead cooled novel reactor concept that is fuelled with metallic alloy of Pu, U and Zr, and is designed to operate for 20 effective full power years without refuelling and with very small burnup reactivity swing. The computational tool benchmarked include MOCUP, a coupled MCNP-4C and ORIGEN2.1 utility codes with MCNP data libraries based on the newest evaluations. (author)
Electron Elastic-Scattering Cross-Section Database
SRD 64 NIST Electron Elastic-Scattering Cross-Section Database (PC database, no charge) This database provides values of differential elastic-scattering cross sections, corresponding total elastic-scattering cross sections, phase shifts, and transport cross sections for elements with atomic numbers from 1 to 96 and for electron energies between 50 eV and 20,000 eV (in steps of 1 eV).
Total cross sections for neutron-nucleus scattering
Suryanarayana, S. V.; H. Naik; Ganesan, S; Kailas, S; Choudhury, R. K.; Kim, Guinyum
2010-01-01
Systematics of neutron scattering cross sections on various materials for neutron energies up to several hundred MeV are important for ADSS applications. Ramsauer model is well known and widely applied to understand systematics of neutron nucleus total cross sections. In this work, we examined the role of nuclear effective radius parameter (r$_0$) on Ramsauer model fits of neutron total cross sections. We performed Ramsauer model global analysis of the experimental neutron total cross section...
Finite sum expressions for elastic and reaction cross sections
Nuclear cross section calculations are often performed by using the partial wave method or the Eikonal method through Glauber theory. The expressions for the total cross section, total elastic cross section, and total reaction cross section in the partial wave method involve infinite sums and do not utilize simplifying approximations. Conversely, the Eikonal method gives these expressions in terms of integrals but utilizes the high energy and small angle approximations. In this paper, by using the fact that the lth partial wave component of the T-matrix can be very accurately approximated by its Born term, the infinite sums in each of the expressions for the differential cross section, total elastic cross section, total cross section, and total reaction cross section are re-written in terms of finite sums plus closed form expressions. The differential cross sections are compared to the Eikonal results for 16O+16O,12C+12C, and p+12C elastic scattering. Total cross sections, total reaction cross sections, and total elastic cross sections are compared to the Eikonal results for 12C+12C scattering
Cross sections for electron impact excitation of molecules
The discussion in this chapter is restricted to elastic scattering, rotational, vibrational, and electronic excitation and total scattering cross sections in electron molecule collisions. Experimental data on differential, integral and momentum transfer cross sections are surveyed and short remarks are made on experimental techniques and theoretical approaches used for generating cross section data. 11 references, 3 figures
30 CFR 779.25 - Cross sections, maps, and plans.
2010-07-01
... 30 Mineral Resources 3 2010-07-01 2010-07-01 false Cross sections, maps, and plans. 779.25 Section... RESOURCES § 779.25 Cross sections, maps, and plans. (a) The application shall include cross sections, maps..., maps and plans included in a permit application as required by this section shall be prepared by,...
30 CFR 783.25 - Cross sections, maps, and plans.
2010-07-01
... 30 Mineral Resources 3 2010-07-01 2010-07-01 false Cross sections, maps, and plans. 783.25 Section... ENVIRONMENTAL RESOURCES § 783.25 Cross sections, maps, and plans. (a) The application shall include cross sections, maps, and plans showing— (1) Elevations and locations of test borings and core samplings;...
DDXPLOT: a program to plot the energy angle double-differential cross sections
A program ''DDXPLOT'' has been developed to plot the energy-angle double differential cross sections (DDX) calculated from the current evaluated nuclear data library with the experimental data. The calculated DDX data of each reaction type are summed up and smoothed out according to the experimental conditions. Furthermore, the DDX data from various evaluated libraries can be compared with one another, by specifying the identification names in the data pool. This report is the manual of DDXPLOT. (author)
Radar Cross-section Measurement Techniques
V.G. Borkar
2010-03-01
Full Text Available Radar cross-section (RCS is an important study parameter for defence applications specially dealing with airborne weapon system. The RCS parameter guides the detection range for a target and is therefore studied to understand the effectiveness of a weapon system. It is not only important to understand the RCS characteristics of a target but also to look into the diagnostic mode of study where factors contributing to a particular RCS values are studied. This further opens up subject like RCS suppression and stealth. The paper discusses the RCS principle, control, and need of measurements. Classification of RCS in terms of popular usage is explained with detailed theory of RF imaging and inverse synthetic aperture radar (ISAR. The various types of RCS measurement ranges are explained with brief discussion on outdoor RCS measurement range. The RCS calibration plays a critical role in referencing the measurement to absolute values and has been described.The RCS facility at Reseach Centre Imarat, Hyderabad, is explained with some details of different activities that are carried out including RAM evaluation, scale model testing, and diagnostic imaging.Defence Science Journal, 2010, 60(2, pp.204-212, DOI:http://dx.doi.org/10.14429/dsj.60.341
Resonance capture cross section of 207Pb
Domingo-Pardo, C; Aerts, G; Alvarez-Pol, H; Alvarez-Velarde, F; Andrzejewski, J; Andriamonje, Samuel A; Assimakopoulos, P A; Audouin, L; Badurek, G; Baumann, P; Becvar, F; Berthoumieux, E; Bisterzo, S; Calviño, F; Cano-Ott, D; Capote, R; Carrapico, C; Chepel, V; Cennini, P; Chiaveri, Enrico; Colonna, N; Cortés, G; Couture, A; Cox, J; Dahlfors, M; David, S; Dillman, I; Dolfini, R; Dridi, W; Durán, I; Eleftheriadis, C; Embid-Segura, M; Ferrant, L; Ferrari, A; Ferreira-Marques, R; Fitzpatrick, L; Frais-Kölbl, H; Fujii, K; Furman, W; Gallino, R; Gonçalves, I; González-Romero, E M; Goverdovski, A; Gramegna, F; Griesmayer, E; Guerrero, C; Gunsing, F; Haas, B; Haight, R; Heil, M; Herrera-Martínez, A; Igashira, M; Isaev, S; Jericha, E; Kadi, Y; Käppeler, F K; Karamanis, D; Karadimos, D; Kerveno, M; Ketlerov, V; Köhler, P; Konovalov, V; Kossionides, E; Krticka, M; Lamboudis, C; Leeb, H; Lindote, A; Lopes, I; Lozano, M; Lukic, S; Marganiec, J; Marrone, S; Mastinu, P; Mengoni, A; Milazzo, P M; Moreau, C; Mosconi, M; Neves, F; Oberhummer, Heinz; Oshima, M; O'Brien, S; Pancin, J; Papachristodoulou, C; Papadopoulos, C; Paradela, C; Patronis, N; Pavlik, A; Pavlopoulos, P; Perrot, L; Plag, R; Plompen, A; Plukis, A; Poch, A; Pretel, C; Quesada, J; Rauscher, T; Reifarth, R; Rosetti, M; Rubbia, Carlo; Rudolf, G; Rullhusen, P; Salgado, J; Sarchiapone, L; Savvidis, I; Stéphan, C; Tagliente, G; Taín, J L; Tassan-Got, L; Tavora, L; Terlizzi, R; Vannini, G; Vaz, P; Ventura, A; Villamarín, D; Vincente6, M C; Vlachoudis, V; Vlastou, R; Voss, F; Walter, S; Wendler, H; Wiescher, M; Wisshak, K
2006-01-01
The radiative neutron capture cross section of 207Pb has been measured at the CERN neutron time of flight installation n_TOF using the pulse height weighting technique in the resolved energy region. The measurement has been performed with an optimized setup of two C6D6 scintillation detectors, which allowed us to reduce scattered neutron backgrounds down to a negligible level. Resonance parameters and radiative kernels have been determined for 16 resonances by means of an R-matrix analysis in the neutron energy range from 3 keV to 320 keV. Good agreement with previous measurements was found at low neutron energies, whereas substantial discrepancies appear beyond 45 keV. With the present results, we obtain an s-process contribution of 77(8)% to the solar abundance of 207Pb. This corresponds to an r-process component of 23(8)%, which is important for deriving the U/Th ages of metal poor halo stars.
Production cross sections from phenomenological constraints
Hadronic production cross sections ν sub(n) (s) satisfying exactly the high energy empirical laws known for the first, second and third multiplicity moments are determined. The result is obtained in the form of a second order linear differential equation for ν sub(n) (s) which allows one to calculate explicitly all successive moments. In particular, the fourth moment is in excellent agreement with the data. The asymptotic solution of the equation for ν sub(n) (s) is given analytically. KNO scaling turns out to be an asymptotic property of the solution. The full solution for ν sub(n) (s) is studied numerically and the KNO plot is compared with the data. No free parameters are left to be adjusted except for an overall normalization constant. As expected, KNO scaling sets in rather quickly with increasing n and the agreement with the data is progressively good. This agreement becomes excellent for the whole interval of n/ for which data exist (O) approximately equal to 2. It turns out that the asymptotic solution, given in analytic terms, is an excellent approximation to the data and can thus be used for practical purposes instead of the full solution for calculating ν sub(n) (s). (author)
Differential cross section and related integrals for the Moliere potential
The Moliere potential is widely used in radiation damage simulation studies. It is not much used in analytical transport theory calculations because of the awkward expression for the differential cross section corresponding to the potential. A two step process is followed to obtain a useful cross section: adopting the Lindhard, Nielsen and Scharff (LNS) approximations in order to generate a simpler form of the Moliere cross section and then creating a simple, easy-to-use, fit to that approximate form. Within the framework of the LNS treatment of atomic cross sections, our fit is accurate to 6%. Simple forms for the total cross section and several related quantities are presented. (author)
Cross section weighting spectrum for fast reactor analysis
Preparation of a nuclear data library is the first task that a reactor analyst needs to perform a neutronic analysis of a reactor type. Today, in fast reactor area, the scheme used to generate this library includes the processing of an evaluated nuclear data file to obtain cross sections, in thousands of groups. Sequentially, the nuclear data are processed by a cell code to obtain neutron flux that is used to condense the large amount of energy groups to a practical calculation number of groups that can be used in reactor analysis. In the first step of the scheme it is necessary a weighting spectrum to generate the nuclear data. Here, it is proposed to use the flux estimated by Monte Carlo code using cell or the exact geometries and actual composition of the problem to obtain the main portion of the weighting spectrum instead of a code built-in function. As an example, it is presented the differences between selected pins spectrums obtained with MCNP5 calculation of a hexagonal fast reactor fuel assembly. Also, it is showed a comparison between these spectra and the one obtained in the representative unit-cell model of this fuel assembly. The comparisons support that the proposed procedure, problem dependent, may be more accurate and a good choice to generate weighting spectrum in ultra-fine energy structure for fast reactor analysis. The proposed method will be used in space reactor neutronic analysis. (author)
Optimization of multi-group cross sections for fast reactor analysis
The selection of the number of broad energy groups, collapsed broad energy group boundaries, and their associated evaluation into collapsed macroscopic cross sections from a general 238-group ENDF/B-VII library dramatically impacted the k eigenvalue for fast reactor analysis. An analysis was undertaken to assess the minimum number of energy groups that would preserve problem physics; this involved studies using the 3D deterministic transport parallel code PENTRAN, the 2D deterministic transport code SCALE6.1, the Monte Carlo based MCNP5 code, and the YGROUP cross section collapsing tool on a spatially discretized MOX fuel pin comprised of 21% PUO2-UO2 with sodium coolant. The various cases resulted in a few hundred pcm difference between cross section libraries that included the 238 multi-group reference, and cross sections rendered using various reaction and adjoint weighted cross sections rendered by the YGROUP tool, and a reference continuous energy MCNP case. Particular emphasis was placed on the higher energies characteristic of fission neutrons in a fast spectrum; adjoint computations were performed to determine the average per-group adjoint fission importance for the MOX fuel pin. This study concluded that at least 10 energy groups for neutron transport calculations are required to accurately predict the eigenvalue for a fast reactor system to within 250 pcm of the 238 group case. In addition, the cross section collapsing/weighting schemes within YGROUP that provided a collapsed library rendering eigenvalues closest to the reference were the contribution collapsed, reaction rate weighted scheme. A brief analysis on homogenization of the MOX fuel pin is also provided, although more work is in progress in this area. (authors)
Comparison of ENDF/B-V and VI cross sections for dosimetry foil reaction rates
The objective of this study was to compare calculations of pressure vessel surveillance dosimetry foil reaction rates computed using the ENDF/B-V and ENDF/B-VI cross-section libraries. Reaction rates for dosimetry foils irradiated in the cavity surrounding the pressure vessel of a pressurized water reactor (PWR) were determined using the MCNP4A code with ENDF/B-V and -VI cross sections. The computed reaction rates were compared to measured ones obtained during three fuel cycles of the Arkansas Nuclear One Unit 1 (ANO-1)
Assessment of some optical model potentials in predicting neutron cross sections
Kumar, A.; Young, P.G.; Chadwick, M.B.
1998-03-01
Optical model potential parameters play an important role in the evaluation of nuclear data for applied purposes. The IAEA Coordinated Research Program on {open_quotes}Reference Input Parameter Library for Evaluation of Nuclear Data for Application in Nuclear Technology{close_quotes} aims to release a reference input file of various types of parameters for the evaluation of nuclear cross sections using nuclear model codes. Included in the parameter files are a collection of optical model potentials that are available in the literature to evaluate these cross sections. As part of this research program we assess the applicability of these potentials over a range of target mass and projectile energy.
Atlas of giant dipole resonances. Parameters and graphs of photonuclear reaction cross sections
Parameters of giant dipole resonances (GDR) observed in photonuclear reaction cross sections using various beams of incident photons are presented. Data, given for 200 stable isotopes from 2H to 243Am including their natural compositions, were collected from papers published over the years 1951-1996. GDR parameters, such as energy positions, amplitudes and widths, are included into the table and organized by element, isotope and reaction. Graphs of the majority of the photonuclear reaction cross sections, included in the international nuclear data library EXFOR by the end of 1998, are presented. The graphs are provided for 182 stable isotopes and natural compositions. (author)
TRANSX-2.15, Neutron Gamma Particle Transport Tables from MATXS Format Cross-Sections
1 - Description of program or function: TRANSX is a computer code that reads nuclear data from a library in MATXS format and produces transport tables compatible with many discrete-ordinates (SN) and diffusion codes. Tables can be produced for neutron, photon, charged-particle, or coupled transport. Options include adjoint tables, mixtures, homogeneous or heterogeneous self-shielding, group collapse, homogenization, thermal up-scatter, prompt or steady-state fission, transport corrections, elastic removal corrections, and flexible response function edits. 2 - Method of solution: TRANSX reads through the materials in a MATXS library and accumulates the cross sections into a transport table using the user's mix instructions. At the same time, response function edit cross sections are accumulated using the user's edit instructions. They can thus be any linear combination of the cross sections available in the library. When the table is complete, it is written out in the desired format. Output options include DTF-style card images, FIDO, ISOTXS, and the binary group-ordered GOXS format. Self-shielding is handled using the background cross section method. Heterogeneity options include homogeneous mixtures, escape using mean chord, lattices of cylinders by the Bell or Sauer approximations, and reflected or periodic slab cell by the bell or E3 approximations. 3 - Restrictions on the complexity of the problem: Only narrow- resonance self-shielding is available in this version. This may affect accuracy for thermal problems
Determination of the 241Pu(n,f) reaction cross sections by surrogate ratio method
In a recent work, direct measurement of neutron induced fission cross section of 241Pu(T1/2 = 14.1 yrs), has been carried out using neutron time of flight. Differences between the reported experimental measurement and evaluation by standard libraries exceeds by 30%, thus it is suggested to carry out further confirmatory measurements on 241Pu(n,f) cross sections. In this paper, we report the use of 238U(6Li,d)242Pu and 232Th(6Li,d)236U transfer reactions as surrogate of 241Pu(n,f) and 235U(n,f) compound nuclear reactions respectively. By employing the surrogate ratio method (SRM) and taking 235U(n,f) as reference, the 241Pu(n,f) cross sections have been determined in the equivalent neutron energy range of 11.0 MeV-16.0 MeV
Thermal neutron capture cross sections resonance integrals and g-factors
The thermal radiative capture cross sections and resonance integrals of elements and isotopes with atomic numbers from 1 to 83 (as well as 232Th and 238U) 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 113Cd, 124Xe and 157Gd at temperatures of 150, 294 and 400K. Comparisons have also been made of the newly evaluated capture cross sections of 6Li, 7Li, 12C and 207Pb with those determined by the k0 method. (author)
Activation cross sections of proton induced nuclear reactions on palladium up to 80MeV.
Tárkányi, F; Ditrói, F; Takács, S; Csikai, J; Hermanne, A; Uddin, M S; Baba, M
2016-08-01
Activation cross sections of proton induced nuclear reactions on palladium were measured up to 80MeV by using the stacked foil irradiation technique and gamma ray spectrometry. The beam intensity, the incident energy and the energy degradation were controlled by a method based on flux constancy via normalization to the excitation functions of monitor reactions measured in parallel. Excitation functions for direct and cumulative cross-sections were measured for the production of (104m,104g,105g,106m,110m)Ag, (100,101)Pd, (99m,99g,100,101m,101g,102m,102g,105)Rh and (103,97)Ru radioisotopes. The cross section data were compared with the theoretical predictions of TENDL-2014 and -2015 libraries. For practical applications thick target yields were derived from the measured excitation functions. Application in the field of medical radionuclide production is shortly discussed. PMID:27235887
Updated multi-group cross sections of minor actinides with improved resonance treatment
The study of minor actinide in transmutation reactors and other future applications makes resonance self-shielding treatment a significant issue for criticality and isotope depletion. Resonance treatment for minor actinides has been carried out by subgroup method with improved interference effect through interference correction. Subgroup data was generated using RMET21 and GENP codes along with multi-group cross section data by NJOY nuclear data processing system. Updated multi-group cross section data library for a neutron transport code nTRACER was compared with solutions from MCNPX. The resonance interaction of uranium with minor actinides has been included by modified interference treatment of interference correction in subgroup methodology. The comparison of cross sections and multiplication factor in pin and assembly problems showed significant improvement from systematic resonance treatment especially for 237Np and 243Am. (author)
Color dipole cross section and inelastic structure function
Jeong, Yu Seon; Reno, Mary Hall
2014-01-01
Instead of starting from a theoretically motivated form of the color dipole cross section in the dipole picture of deep inelastic scattering, we start with a parametrization of the deep inelastic structure function for electromagnetic scattering with protons, and then extract the color dipole cross section. Using the Donnachie-Landshoff parametrization of $F_2(x,Q^2)$, we find the dipole cross section from an approximate form of the presumed dipole cross section convoluted with the perturbative photon wave function for virtual photon splitting into a color dipole with massless quarks. The color dipole cross section determined this way works quite well in the massive case, reproducing the original Donnachie-Landshoff structure function for $0.1$ GeV$^2\\leq Q^2\\leq 10$ GeV$^2$. We discuss the large and small form of the dipole cross section and compare with other parameterizations.
Polynomial parameterized representation of macroscopic cross section for PWR reactor
Fiel, Joao Claudio B., E-mail: fiel@ime.eb.br [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Departamento de Engenharia Nuclear
2015-07-01
The purpose of this work is to describe, by means of Tchebychev polynomial, a parameterized representation of the homogenized macroscopic cross section for PWR fuel element as a function of soluble boron concentration, moderator temperature, fuel temperature, moderator density and {sup 235} U {sub 92} enrichment. Analyzed cross sections are: fission, scattering, total, transport, absorption and capture. This parameterization enables a quick and easy determination of the problem-dependent cross-sections to be used in few groups calculations. The methodology presented here will enable to provide cross-sections values to perform PWR core calculations without the need to generate them based on computer code calculations using standard steps. The results obtained by parameterized cross-sections functions, when compared with the cross-section generated by SCALE code calculations, or when compared with K{sub inf}, generated by MCNPX code calculations, show a difference of less than 0.7 percent. (author)
Resonance Averaged Photoionization Cross Sections for Astrophysical Models
Bautista, M A; Pradhan, A K
1997-01-01
We present ground state photoionization cross sections of atoms and ions averaged over resonance structures for photoionization modeling of astrophysical sources. The detailed cross sections calculated in the close-coupling approximation using the R-matrix method, with resonances delineated at thousands of energies, are taken from the Opacity Project database TOPbase and the Iron Project, including new data for the low ionization stages of iron Fe I--V. The resonance-averaged cross sections are obtained by convolving the detailed cross sections with a Gaussian distribution over the autoionizing resonances. This procedure is expected to minimize errors in the derived ionization rates that could result from small uncertainties in computed positions of resonances, while preserving the overall resonant contribution to the cross sections in the important near threshold regions. The detailed photoionization cross sections at low photon energies are complemented by new relativistic distorted-wave calculations for Z1...
Simulation of cross sections for practical ALCHEMI
Full text: Precisely known atomic scattering factors are essential for accurate atom location by channeling enhanced microanalysis (ALCHEMI) based on inner-shell ionization. For ALCHEMI using energy dispersive x-ray analysis (EDX), first principles calculations of ionization cross sections, realistically modelling the 'delocalization' of the ionization interaction, give excellent agreement with experiment. Such calculations are complex and computationally intensive. Hence, simple analytic forms are often assumed to describe the ionization potential. Such an approach assumes that the precise shape of the ionization potential is not important but that at least the half width at half maximum (HWHM) should be accurately estimated, for example using estimates of the HWHM from root-mean-square impact parameters for ionization. However this is generally not a good approximation and we have provided more realistic estimates. These are based on accurate atomic scattering form factors for ionization that have been calculated from first principles using relativistic Hartree-Fock wave functions for bound states and Hartree-Slater wave functions for the continuum states. The effective ionization interaction may be approximated by an equivalent local potential. The scattering factors have been calculated for K-shell ionization for elements in the range Z= 6 (carbon) to Z = 50 (tin) and for Z-shell ionization in the range Z = 20 (calcium) to Z = 60 (neodymium). Accurate values of the scattering factors can be obtained by interpolation for incident electron energies between 50 and 400 keV. The utility of these form factors is illustrated, using some data obtained by Matsumura and coworkers during their project to investigate radiation-induced disordering in magnesium aluminate spinel. High angular resolution electron channeling x-ray spectroscopy was employed to investigate ion displacements in MgOnAl2O3 (n = 1.0 and 2.4) irradiated with 1 MeV Ne+ ions or 900 keV electrons at 873
Projectile and Lab Frame Differential Cross Sections for Electromagnetic Dissociation
Norbury, John W.; Adamczyk, Anne; Dick, Frank
2008-01-01
Differential cross sections for electromagnetic dissociation in nuclear collisions are calculated for the first time. In order to be useful for three - dimensional transport codes, these cross sections have been calculated in both the projectile and lab frames. The formulas for these cross sections are such that they can be immediately used in space radiation transport codes. Only a limited amount of data exists, but the comparison between theory and experiment is good.
Proton-nucleus cross section at high energies
Wibig, Tadeusz; Sobczynska, Dorota
1998-01-01
Cross sections for proton inelastic collision with different nuclei are described within the Glauber and multiple scattering approximations. A significant difference between approximate `Glauber' formula and exact calculations with a geometrical scaling assumption for very high-energy cross section is shown. Experimental values of proton-proton cross sections obtained using extensive air shower data are based on the relationship of proton-proton and respective proton-air absorption cross sect...
Theoretical estimates of cross sections for neutron-nucleus collisions
Mukhopadhyay, Tapan; Lahiri, Joydev; Basu, D. N.
2010-01-01
We construct an analytical model derived from nuclear reaction theory and having a simple functional form to demonstrate the quantitative agreement with the measured cross sections for neutron induced reactions. The neutron-nucleus total, reaction and scattering cross sections, for energies ranging from 5 to 700 MeV and for several nuclei spanning a wide mass range are estimated. Systematics of neutron scattering cross sections on various materials for neutron energies upto several hundred Me...
CREST : a computer program for the calculation of composition dependent self-shielded cross-sections
A computer program CREST for the calculation of the composition and temperature dependent self-shielded cross-sections using the shielding factor approach has been described. The code includes the editing and formation of the data library, calculation of the effective shielding factors and cross-sections, a fundamental mode calculation to generate the neutron spectrum for the system which is further used to calculate the effective elastic removal cross-sections. Studies to explore the sensitivity of reactor parameters to changes in group cross-sections can also be carried out by using the facility available in the code to temporarily change the desired constants. The final self-shielded and transport corrected group cross-sections can be dumped on cards or magnetic tape in a suitable form for their direct use in a transport or diffusion theory code for detailed reactor calculations. The program is written in FORTRAN and can be accommodated in a computer with 32 K work memory. The input preparation details, sample problem and the listing of the program are given. (author)
Measurements of neutron capture cross section of 237Np for fast neutrons
The neutron capture cross section of 237Np has been measured for fast neutrons supplied at the center of the core in the Yayoi reactor. The activation method was used for the measurement, in which the amount of the product 238Np was determined by γ-ray spectroscopy using a Ge detector. The neutron flux at the center of the core calculated by the Monte Carlo simulation code MCNP was renormalized by using the activity of a gold activation foil irradiated simultaneously. The new convention is proposed in this paper to make possible a definite comparison of the integral measurement by the activation method using fast reactor neutrons with differential measurements using accelerator-based neutrons. 'Representative neutron energy' is defined in the convention at which the cross section deduced by the activation measurement has a high sensitivity. The capture cross section of 237Np corresponding to the representative neutron energy was deduced as 0.80±0.04 b at 214±9 keV from the measured reaction rate and the energy dependence of the cross section in the nuclear data library ENDF/B-VII.0. The deduced cross section of 237Np at the representative neutron energy agrees with the evaluated data of ENDF/B-VII.0, but is 15% higher than that of JENDL-3.3 and 13% higher than that of JENDL/AC-2008. (author)
Unresolved resonance range cross section, probability tables and self shielding factor
The performance and methodology of 4 processing codes have been compared in the unresolved resonance range of a selected set of isotopes. Those isotopes have been chosen to encompass most cases encountered in the unresolved energy range contained in major libraries like Endf/B-7 or Jeff-3.1.1. The code results comparison is accompanied by data format and formalism examinations and processing code fine-interpretation study. After some improvements, the results showed generally good agreement, although not perfect with infinite dilute cross-sections. However, much larger differences occur when shelf-shielded effective cross-sections are compared. The infinitely dilute cross-section are often plot checked but it is the probability table derived and shelf-shielded cross sections that are used and interpreted in criticality and transport calculations. This suggests that the current evaluation data format and formalism, in the unresolved resonance range should be tightened up, ambiguities removed. In addition production of the shelf shielded cross-sections should be converged to a much greater accuracy. (author)
Neutron-capture Cross Sections from Indirect Measurements
Escher, J E; Burke, J T; Dietrich, F S; Ressler, J J; Scielzo, N D; Thompson, I J
2011-10-18
Cross sections for compound-nuclear reactions play an important role in models of astrophysical environments and simulations of the nuclear fuel cycle. Providing reliable cross section data remains a formidable task, and direct measurements have to be complemented by theoretical predictions and indirect methods. The surrogate nuclear reactions method provides an indirect approach for determining cross sections for reactions on unstable isotopes, which are difficult or impossible to measure otherwise. Current implementations of the method provide useful cross sections for (n,f) reactions, but need to be improved upon for applications to capture reactions.
Fano interference and cross-section fluctuations in molecular photodissociation
We derive an expression for the total photodissociation cross section of a molecule incorporating both direct and indirect processes that proceed through excited resonances, and show that it exhibits generalized Beutler-Fano line shapes. Assuming that the closed system can be modeled by random-matrix theory, we derive the statistical properties of the photodissociation cross section and find that they are significantly affected by the direct processes. In the limit of isolated resonances, we find that direct processes suppress the correlation hole of the cross-section autocorrelation function and lead to a maximum in the cross-section distribution
Systematics of fission cross sections at the intermediate energy region
Fukahori, Tokio; Chiba, Satoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1997-03-01
The systematics was obtained with fitting experimental data for proton induced fission cross sections of Ag, {sup 181}Ta, {sup 197}Au, {sup 206,207,208}Pb, {sup 209}Bi, {sup 232}Th, {sup 233,235,238}U, {sup 237}Np and {sup 239}Pu above 20 MeV. The low energy cross section of actinoid nuclei is omitted from systematics study, since the cross section has a complicated shape and strongly depends on characteristic of nucleus. The fission cross sections calculated by the systematics are in good agreement with experimental data. (author)
Large discrepancies have been found among the various nuclear libraries, mainly in the production cross sections and angular distributions of gamma-rays following neutron interactions. These data are required for the design and application of contraband inspection systems based on neutron irradiation techniques. As a result of these discrepancies experiments were conducted to correct and complement the required data. Cross section evaluations of carbon, oxygen and chlorine are reviewed and compared with existing and the new experimental data. ((orig.))
Cross section model and scattering law of liquid water for design of a cold neutron source
A cross section model for cold neutron scattering in light water is developed, which describes various molecular motions inherent to hydrogen-bonded water molecules especially in terms of jump- and rotational-diffusion processes. Inter- and intra-molecular vibrations are also included. A systematic analysis is performed of a velocity autocorrelation function, a generalized frequency distribution and double-differential and total cross sections. Good agreement with the results of computer molecular dynamics and neutron scattering experiments is found. A wide range of cross section evaluation for neutron energies from 0.1 μeV to 10 eV and liquid temperatures between the melting and boiling points is performed. This permits us to generate such low-energy neutron cross section libraries as group constants set and scattering law for ultra-cold, very-cold, cold and thermal neutrons. Together with the libraries for liquid 4He, H2, D2 and solid and liquid CH4, a powerful tool for design of an advanced low-energy neutron source is now ready for use. (author)
Cross sections for the reaction 197Au(γ, chin)(chi<=12) have been measured for bremsstrahlung end-point energies in the range 60-340 MeV. From these dominant cross sections, the total photon absorption cross section is determined using a cascade-evaporation calculation to account for the missing reaction channels. The enhancement factor for the classical E1 sum rule is found to be 0.93+-0.10. (orig.)
Modeling and analysis of ground target radiation cross section
SHI Xiang; LOU GuoWei; LI XingGuo
2008-01-01
Based on the analysis of the passive millimeter wave (MMW) radiometer detection, the ground target radiation cross section is modeled as the new token for the target MMW radiant characteristics. Its ap-plication and actual testing are discussed and analyzed. The essence of passive MMW stealth is target radiation cross section reduction.
Analysis of cross sections using various nuclear potential
The relevant astrophysical reaction rates which are derived from the reaction cross sections are necessary input to the reaction network. In this work, we analyse several theoretical models of the nuclear potential which give better prediction of the cross sections for some selected reactions
Total Cross Sections at High Energies - An Update
Fazal-e-Aleem; Sohail Afzal Tahir; M. Alam Saeed; M. Qadeer Afzal
2002-01-01
Current and future measurements for the total cross sections at E-811, PP2PP, CSM, FELIX, and TOTEMhave been analyzed using various models. In the light of this study an attempt has been made to focus on the behaviorof total cross section at very high energies.
Surrogate reaction methods for neutron induced cross-sections
A brief discussion on surrogate reaction methods and some of the recent results on neutron induced fission cross-section measurements carried out by our group and the possibility of extending the measurements for determining (n,g), (n,2n) and (n,p) reaction cross-sections by surrogate reaction method are presented
Ni elemental neutron induced reaction cross-section evaluation
A completely new evaluation of the nickel neutron induced reaction cross sections was undertaken as a part of the ENDF/B-V effort. (n,xy) reactions and capture reaction time from threshold to 20 MeV were considered for 5860616264Ni isotopes to construct the corresponding reaction cross section for natural nickel. Both experimental and theoretical calculated results were used in evaluating different partial cross sections. Precompound effects were included in calculating (n,xy) reaction cross sections. Experimentally measured total section data extending from 0.7 MeV to 20 MeV were used to generate smooth cross section. Below 0.7 to MeV elastic and capture cross sections are represented by resonance parameters. Inelastic angular distributions to the discrete isotopic levels and elemental elastic angular distributions are included in the evaluated data file. Gamma production cross sections and energy distribution due to capture and the (n,xy) reactions were evaluated from experimental data. Finally, error files are constructed for all partial cross sections
Cross Sections for Electron Collisions with Carbon Monoxide
Cross section data are collected and reviewed for electron collisions with carbon monoxide. Collision processes included are total scattering, elastic scattering, momentum transfer, excitations of rotational, vibrational and electronic states, ionization, and dissociation. For each process, recommended values of the cross sections are presented, when possible. The literature has been surveyed through to the end of 2013
Applications of the BEam Cross section Analysis Software (BECAS)
Blasques, José Pedro Albergaria Amaral; Bitsche, Robert; Fedorov, Vladimir;
2013-01-01
A newly developed framework is presented for structural design and analysis of long slender beam-like structures, e.g., wind turbine blades. The framework is based on the BEam Cross section Analysis Software – BECAS – a finite element based cross section analysis tool. BECAS is used for the...
Learning of Cross-Sectional Anatomy Using Clay Models
Oh, Chang-Seok; Kim, Ji-Young; Choe, Yeon Hyeon
2009-01-01
We incorporated clay modeling into gross anatomy and neuro-anatomy courses to help students understand cross-sectional anatomy. By making clay models, cutting them and comparing cut surfaces to CT and MR images, students learned how cross-sectional two-dimensional images were created from three-dimensional structure of human organs. Most students…
On the scattering cross section of passive linear arrays
Solymar, L.
1973-01-01
A general formula is derived for the scattering cross section of a passiven-element linear array consisting of isotropic radiators. When all the reactances are tuned out and scattering in the mirror direction is investigated, it is found thatA_{sr}, the relative scattering cross section is equal to...
Cross section probability tables in multi-group transport calculations
The use of cross section probability tables in multigroup transport calculations is presented. Emphasis is placed on how probability table parameters are generated in a multigroup cross section processor and how existing transport codes must be modifed to use them. In order to illustrate the accuracy obtained by using probability tables, results are presented for a variety of neutron and photon transport problems
Possibility of spin mechanism of total cross section growth
The possibility of existence of the spin mechanism of total cross section growth is considered. A nucleon-nucleon scattering is studied. The energy dependence of scattering amplitude and possible effects related with the spin mechanism of total cross section growth are studied. It is shown that the considered mechanism can play a great role at high energies
Temperature dependence of the HNO3 UV absorption cross sections
Burkholder, James B.; Talukdar, Ranajit K.; Ravishankara, A. R.; Solomon, Susan
1993-01-01
The temperature dependence of the HNO3 absorption cross sections between 240 and 360 K over the wavelength range 195 to 350 nm has been measured using a diode array spectrometer. Absorption cross sections were determined using both (1) absolute pressure measurements at 298 K and (2) a dual absorption cell arrangement in which the absorption spectrum at various temperatures is measured relative to the room temperature absorption spectrum. The HNO3 absorption spectrum showed a temperature dependence which is weak at short wavelengths but stronger at longer wavelengths which are important for photolysis in the lower stratosphere. The 298 K absorption cross sections were found to be larger than the values currently recommended for atmospheric modeling (DeMore et al., 1992). Our absorption cross section data are critically compared with the previous measurements of both room temperature and temperature-dependent absorption cross sections. Temperature-dependent absorption cross sections of HNO3 are recommended for use in atmospheric modeling. These temperature dependent HNO3 absorption cross sections were used in a two-dimensional dynamical-photochemical model to demonstrate the effects of the revised absorption cross sections on loss rate of HNO3 and the abundance of NO2 in the stratosphere.
The effect of the decay data on activation cross section
The effect of the decay data on evaluation of activation cross section is investigated. Present work shows that these effects must be considered carefully when activation cross section is evaluated. Sometime they are main reason for causing the discrepancies among the experimental data
Minijets, soft gluon resummation and photon cross-sections
Godbole, R. M.; Grau, A.; Pancheri, G.; Srivastava, Y. N.
2008-01-01
We compare the high energy behaviour of hadronic photon-photon cross-sections in different models. We find that the photon-photon cross-section appears to rise faster than the purely hadronic ones (proton-proton and proton-antiproton).
ETOGM: epithermal cross section generation code using ENDF/B data
ETOGM processes ENDF/B FORMAT data to produce a master epithermal library containing multigroup cross sections, resolved and unresolved resonance parameters, and a scattering matrix for up to 100 materials of interest in reactor design calculations. The epithermal energy range may be divided into as many as 127 groups, and a weighting function may either be input or calculated by the code for use in calculating average group coefficients. Resonance contributions from thermal and negative energy resolved resonances, as well as infinite-dilute corrections in the epithermal range, are added to the appropriate smooth cross sections. Resolved and unresolved resonance parameters are tabulated when applicable. A combined inelastic-(n,2n) scattering matrix is calculated from secondary neutron energy distribution data. A fission spectrum is computed for each fissionable material. The master epithermal library is generated, updated, and edited by the ETOGM program
Measurement of the fission cross section of 238Pu
The fission cross sections of 238Pu have been measured from 0.1 eV to 80 keV energy range using the Rensselaer Intense Neutron Spectrometer. The cross sections were normalized to the 235U ENDF/B-V data broadened to the resolution of the Rensselaer Intense Neutron Spectrometer system. The fission areas and widths were determined for the resolved low-energy resonances. The ENDF/B-V fission cross sections for the 238Pu isotope are, in general, not in good agreement with the measured cross sections and a new evaluation is recommended. The observations of structure in the unresolved fission cross sections is suggestive of the existence of intermediate structure. 18 refs., 1 fig., 1 tab
Capture cross-section of threading dislocations in thin films
Highlights: ► We study the effect of film stress on capture cross-section of interacting threads. ► Capture cross-section area diverges near film channeling stress. ► Thread interactions are much more likely when local stress is near critical stress. - Abstract: The capture cross section for annihilation of two threads with opposite Burgers vectors moving on orthogonal slip planes in a thin film is examined using a numerical model. The initial configurations of threads that lead to annihilation are mapped out for a range of applied film stresses. The area of the region of initial configurations that lead to annihilation at a given stress and thickness is the capture cross-section. The size of the capture cross-section is shown to be highly sensitive to the applied stress relative to the critical stress for dislocation motion imposed by the film thickness.
Anomalously large neutron capture cross sections: a random phenomenon?
Carlson, B V; Kerman, A K
2015-01-01
We discuss the existence of huge thermal neutron capture cross sections in several nuclei. The values of the cross sections are several orders of magnitude bigger than expected at these very low energies. We lend support to the idea that this phenomenon is random in nature and is similar to what we have learned from the study of parity violation in the actinide region. The idea of statistical doorways is advanced as a unified concept in the delineation of large numbers in the nuclear world. The average number of maxima per unit mass, $$ in the capture cross section is calculated and related to the underlying cross section correlation function and found to be $ = \\frac{3}{\\pi \\sqrt{2}\\gamma_{A}}$, where $\\gamma_{A}$ is a characteristic mass correlation width which designates the degree of remnant coherence in the system. We trace this coherence to nucleosynthesis which produced the nuclei whose neutron capture cross sections are considered here.
Database for Proton Induced Residual Production Cross Sections up to 2.6 GeV
Cross sections for production of nuclides by protons up to 2.6 GeV from the target elements of C, N, O, F, Mg, Al, Si, Ca, Ti, V, Mn, Fe, Co, Ni, Cu, Sr, Y, Zr, Nb, Ba and Au are tabulated. All data sets are compiled in the EXFOR library, and available from the International Network of Nuclear Reaction Data Centres (NRDC). (author)
Trial publication after registration in ClinicalTrials.Gov: a cross-sectional analysis.
Ross, Joseph S.; Mulvey, Gregory K.; Hines, Elizabeth M.; Nissen, Steven E.; Krumholz, Harlan M.
2009-01-01
BACKGROUND: ClinicalTrials.gov is a publicly accessible, Internet-based registry of clinical trials managed by the US National Library of Medicine that has the potential to address selective trial publication. Our objectives were to examine completeness of registration within ClinicalTrials.gov and to determine the extent and correlates of selective publication. METHODS AND FINDINGS: We examined reporting of registration information among a cross-section of trials that had been registered at ...
Neutron cross-sections database for amino acids and proteins analysis
Biological materials may be studied using neutrons as an unconventional tool of analysis. Dynamics and structures data can be obtained for amino acids, protein and others cellular components by neutron cross sections determinations especially for applications in nuclear purity and conformation analysis. The instrument used for this is the crystal spectrometer of the Instituto de Engenharia Nuclear (IEN-CNEN-RJ), the only one in Latin America that uses neutrons for this type of analyzes and it is installed in one of the reactor Argonauta irradiation channels. The experimentally values obtained are compared with calculated values using literature data with a rigorous analysis of the chemical composition, conformation and molecular structure analysis of the materials. A neutron cross-section database was constructed to assist in determining molecular dynamic, structure and formulae of biological materials. The database contains neutron cross-sections values of all amino acids, chemical elements, molecular groups, auxiliary radicals, as well as values of constants and parameters necessary for the analysis. An unprecedented analytical procedure was developed using the neutron cross section parceling and grouping method for data manipulation. This database is a result of measurements obtained from twenty amino acids that were provided by different manufactories and are used in oral administration in hospital individuals for nutritional applications. It was also constructed a small data file of compounds with different molecular groups including carbon, nitrogen, sulfur and oxygen, all linked to hydrogen atoms. A review of global and national scene in the acquisition of neutron cross sections data, the formation of libraries and the application of neutrons for analyzing biological materials is presented. This database has further application in protein analysis and the neutron cross-section from the insulin was estimated. (author)
Neutron cross-sections database for amino acids and proteins analysis
Voi, Dante L.; Ferreira, Francisco de O.; Nunes, Rogerio Chaffin, E-mail: dante@ien.gov.br, E-mail: fferreira@ien.gov.br, E-mail: Chaffin@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Rocha, Helio F. da, E-mail: hrocha@gbl.com.br [Universidade Federal do Rio de Janeiro (IPPMG/UFRJ), Rio de Janeiro, RJ (Brazil). Instituto de Pediatria
2015-07-01
Biological materials may be studied using neutrons as an unconventional tool of analysis. Dynamics and structures data can be obtained for amino acids, protein and others cellular components by neutron cross sections determinations especially for applications in nuclear purity and conformation analysis. The instrument used for this is the crystal spectrometer of the Instituto de Engenharia Nuclear (IEN-CNEN-RJ), the only one in Latin America that uses neutrons for this type of analyzes and it is installed in one of the reactor Argonauta irradiation channels. The experimentally values obtained are compared with calculated values using literature data with a rigorous analysis of the chemical composition, conformation and molecular structure analysis of the materials. A neutron cross-section database was constructed to assist in determining molecular dynamic, structure and formulae of biological materials. The database contains neutron cross-sections values of all amino acids, chemical elements, molecular groups, auxiliary radicals, as well as values of constants and parameters necessary for the analysis. An unprecedented analytical procedure was developed using the neutron cross section parceling and grouping method for data manipulation. This database is a result of measurements obtained from twenty amino acids that were provided by different manufactories and are used in oral administration in hospital individuals for nutritional applications. It was also constructed a small data file of compounds with different molecular groups including carbon, nitrogen, sulfur and oxygen, all linked to hydrogen atoms. A review of global and national scene in the acquisition of neutron cross sections data, the formation of libraries and the application of neutrons for analyzing biological materials is presented. This database has further application in protein analysis and the neutron cross-section from the insulin was estimated. (author)
Nelson, R; R. Vogt; Frawley, A. D.
2012-01-01
We explore the available parameter space that gives reasonable fits to the total charm cross section to make a better estimate of its true uncertainty. We study the effect of the parameter choices on the energy dependence of the J/\\psi\\ cross section.
The photonuclear absorption cross section of Pb, σ(TOT:Esub(γ), is studied in the 145-440 MeV Δ resonance range using a quasi-monochromatic photon beam obtained by monoenergetic positon in-flight annihilation. This study is deduced of the cross section measurement for at least j neutron emission σsup(j))Esub(γ). The cross sections of reactions with 1 or 0 neutron are evaluated as the same values as the experimental errors. The variation of the photonuclear absorption cross section for a nuclear σ(TOT:Esub(γ)/A is mass independent for A<=4-6. It seems that the damping between σ(TOT:Esub(γ)/A and the cross section of the free nucleon is caused by the Fermi movement of the nucleons. In conclusion: it seems that the excitation of the nucleus in the Δ resonance region is produced on free nucleons and there are no collective states
The total collision cross section in the glory region
Chapter 1 presents a calculation of approximate total cross sections in the glory region from noble gas potentials. The relations between the main features of the total cross section and the properties of the potential to which these are sensitive are extensively investigated in chapter II. A beam apparatus has been developed, which allows for accurate measurements on the total cross section. All effects due to the finite angular and velocity resolution of the apparatus can be eliminated from the data to yield actual total cross sections as a function of the relative velocity. This facilitates a comparison to total cross sections predicted by potentials available in the literature. A brief description of the apparatus and of the data reduction is given in chapter III. The total cross section data obtained for various noble gas combinations are presented and analysed in chapter IV, where also a large number of potentials proposed in the literature is tested. In chapter V the quenching of the glories in the case of a non-spherical interaction is analysed. Subsequently, total cross section data for some atom-molecule systems are discussed. (Auth.)
Cross-Sectional Drawing Techniques And The Artist
Berry, William A.
1980-07-01
Although Democritus, a Greek pholosopher of the fifth century B.C. described the use of cross-sections in analyzing a solid form, this method was not extensively developed in art until the Renaissance. The earliest treatise documenting the integration of the cross-section and linear perspective is Piero della Francesca's De prospective pingendi (c. 1480), in which a drawing of the human head is mathematically conceived and plotted by means of cross-section contours. Piero's method anticipates contemporary biostereometric techniques and current theories of visual perception. Outside of theoretical treatises the complete cross-section rarely occurs in art, though certain pictorial elements such as the religious halo can be interpreted as cross-sections. The chan-ging representation of the halo in art of the Medieval, Renaissance and Baroque periods parallels the development of the artist's concepts and techniques for representing form and space. During the Renaissance and Baroque periods the widespread use of contour hatching, a drawing technique based on the cross-section, indicates that the cross-section concept has played a greater role in pictorial representation than has generally been recognized.
AMPX-77, Modular System for Coupled Neutron-Gamma Multigroup Cross-Sections from ENDF/B-5
1 - Description of program or function: The AMPX system is a system of computer programs (modules) capable of producing coupled multigroup neutron-gamma-ray cross section sets. The system is one of the standards for producing multigroup neutron, gamma-ray production, gamma-ray interaction, and coupled neutron-gamma cross-section sets from ENDF data. AMPX-produced cross sections can be used directly with a variety of diffusion theory, discrete ordinates, and Monte Carlo radiation transport computer codes. A one-dimensional Sn calculation capability is provided for general use and for cross section collapsing. Treatments are included for resonance self-shielding effects. 2 - Method of solution: The system includes a full range of features needed to: (1) produce multigroup neutron, gamma-ray production, and/or gamma-ray interaction cross-section data, (2) resonance self-shield, (3) spectrally collapse, (4) convert cross-section libraries from one format to another format, (5) execute a one- dimensional (1-D) discrete-ordinates calculation, and (6) perform miscellaneous cross section-operations. 3 - Restrictions on the complexity of the problem: The principal restriction is the availability of adequate core storage. All large modules are variably dimensioned. Certain modules will automatically use external storage (disk,tape), if in-core storage is inadequate. While these procedures are of little consequence on today's large computers with 'virtual memory' capabilities, they can be important when small-core PC's or workstations are used
Electron impact ionization cross sections of beryllium-tungsten clusters*
Sukuba, Ivan; Kaiser, Alexander; Huber, Stefan E.; Urban, Jan; Probst, Michael
2016-01-01
We report calculated electron impact ionization cross sections (EICSs) of beryllium-tungsten clusters, BenW with n = 1,...,12, from the ionization threshold to 10 keV using the Deutsch-Märk (DM) and the binary-encounter-Bethe (BEB) formalisms. The positions of the maxima of DM and BEB cross sections are mostly close to each other. The DM cross sections are more sensitive with respect to the cluster size. For the clusters smaller than Be4W they yield smaller cross sections than BEB and vice versa larger cross sections than BEB for clusters larger than Be6W. The maximum cross section values for the singlet-spin groundstate clusters range from 7.0 × 10-16 cm2 at 28 eV (BeW) to 54.2 × 10-16 cm2 at 43 eV (Be12W) for the DM cross sections and from 13.5 × 10-16 cm2 at 43 eV (BeW) to 38.9 × 10-16 cm2 at 43 eV (Be12W) for the BEB cross sections. Differences of the EICSs in different isomers and between singlet and triplet states are also explored. Both the DM and BEB cross sections could be fitted perfectly to a simple expression used in modeling and simulation codes in the framework of nuclear fusion research. Contribution to the Topical Issue "Atomic Cluster Collisions (7th International Symposium)", edited by Gerardo Delgado Barrio, Andrey Solov'Yov, Pablo Villarreal, Rita Prosmiti.Supplementary material in the form of one pdf file available from the Journal web page at http://dx.doi.org/10.1140/epjd/e2015-60583-7
Neutron inelastic cross section measurements for 24Mg
OLACEL A.; Borcea, C.; DESSAGNE Philippe; Kerveno, M.; NEGRET A.; PLOMPEN Arjan
2014-01-01
The gamma production cross sections from the neutron inelastic scattering on 24Mg were measured for neutron energies up to 18 MeV at GELINA (Geel Linear Accelerator), the neutron source operated by EC-JRC-IRMM, Belgium. The level cross section and the total inelastic cross section were determined. We used the GAINS (Gamma Array for Inelastic Neutron Scattering) spectrometer with 7 large volume HPGe detectors placed at 110◦ and 150◦ with respect to the beam direction. The neutron flux was dete...
Thermal neutron capture cross-sections and neutron separation energies
Thermal radiative neutron capture cross-sections have been re-evaluated as part of an ongoing project at the National Nuclear Data Center at Brookhaven National Laboratory at Upton, New York, to update the Neutron Cross-sections compendia, Vol. 1, Parts A and B, Neutron Resonance Parameters and Thermal Capture Cross-sections, published by Academic Press in 1981 and 1984, respectively. Neutron separation energies are evaluated as part of an ongoing project at the Atomic Mass Data Center in Orsay, France. The adopted data are compared with new results derived from this evaluation
Neutron activation cross section measurements and evaluations in CIAE
The cross sections of 28 reactions have been measured by the activation method since 1995 in CIAE. At the same time the cross sections of 40 reactions which we have measured since 1989 have been compiled and evaluated. A brief description of experimental measurement of activation cross sections is given. The data measured after 1995 by ourselves are listed in Table 4 and our evaluations for 40 reactions are listed in Table 5, respectively. A graphical intercomparison with available experimental data isi given in appendix. (author)
Neutron total scattering cross sections of elemental antimony
Neutron total cross sections are measured from 0.8 to 4.5 MeV with broad resolutions. Differential-neutron-elastic-scattering cross sections are measured from 1.5 to 4.0 MeV at intervals of 50 to 200 keV and at scattering angles distributed between 20 and 160 degrees. Lumped-level neutron-inelastic-scattering cross sections are measured over the same angular and energy range. The exPerimental results are discussed in terms of an optical-statistical model and are compared with respective values given in ENDF/B-V
A method for measuring light ion reaction cross-sections
An experimental procedure for measuring reaction cross-sections of light ions in the energy range 20-50 MeV/nucleon, using a modified attenuation technique, is described. The detection method incorporates a forward detector that simultaneously measures the reaction cross-sections for five different sizes of the solid angle in steps from 99.1% to 99.8% of the total solid angle. The final reaction cross-section values are obtained by extrapolation to the full solid angle
Evaluation of neutron induced reaction cross sections on Rh isotopes
Evaluations of neutron nuclear data on 101,102,103,105Rh in the incident energies up to 20 MeV were performed, using theoretical nuclear reaction model code CCONE. The calculated cross sections of stable 103Rh are in good agreement with measured inelastic scattering, capture, (n, 2n), (n, p), (n, α) and (n, nα) reaction cross sections. The production cross section for the meta-state of 99Tc with half-life of 6.0 h was evaluated for the estimation of nuclear medicine use and resulted in 2.4 mb at a maximum. (author)
Resonance interaction effects in photonucleon reaction cross sections
The fine structure of a giant dipole resonance in the photonuclear reaction cross section is investigated. Developed is a diagram of parametrization of cross sections, angular distribution and polarization for two resonances, one of which is directly excited by gamma-quantum, the second - due to internal and external mixing with the first state. It is shown, that for several reaction channels the interaction effects significantly the energy dependence of the cross sections and results in qualitative effects in the photonuclear angular distributions and polarization of photonucleons
Comparison of fission and capture cross sections of minor actinides
Nakagawa, T
2003-01-01
The fission and capture cross sections of minor actinides given in JENDL-3.3 are compared with other evaluated data and experimental data. The comparison was made for 32 nuclides of Th-227, 228, 229, 230, 233, 234, Pa-231, 232, 233, U-232, 234, 236, 237, Np-236, 237, 238, Pu-236, 237, 238, 242, 244, Am-241, 242, 242m, 243, Cm-242, 243, 244, 245, 246, 247 and 248. Given in the present report are figures of these cross sections and tables of cross sections at 0.0253 eV and resonance integrals.
Neutron-induced fission cross-section of 231Pa
A first series of fission cross-section measurements for incident neutron energies between 0.6 and 3.4 MeV has confirmed a first chance threshold value around 1b. In contrast to our findings for the fission cross-section in 233Pa, both the direct and the surrogate cross-section data lead to the same result. This seems to support the assumption, that only in cases, where ingoing and outgoing particle are similar, particle-transfer reactions give results that are in agreement with those obtained from direct compound nuclear reactions
Total cross sections of beauty and charmed mesons on protons
Using a simple scaling law we predict the values of the total cross sections σ(B±p), σBd,s0, σ(bar Bd,s0P), σ(Dd,s±P), σ(D0p), σ(bar D0p) from known total Kp cross sections. We assume that mesons with the same light valence quark, q, and differing only by their heavy valence quark content, Q, have total cross sections on protons which scale as the inverse of the nth power of the reduced mass of the meson. We predict that σ(Q bar q)p > σ(bar Qq)p
Comparison of fission and capture cross sections of minor actinides
The fission and capture cross sections of minor actinides given in JENDL-3.3 are compared with other evaluated data and experimental data. The comparison was made for 32 nuclides of Th-227, 228, 229, 230, 233, 234, Pa-231, 232, 233, U-232, 234, 236, 237, Np-236, 237, 238, Pu-236, 237, 238, 242, 244, Am-241, 242, 242m, 243, Cm-242, 243, 244, 245, 246, 247 and 248. Given in the present report are figures of these cross sections and tables of cross sections at 0.0253 eV and resonance integrals. (author)
Measurements of fission cross-sections. Chapter 4
The steps involved in the measurement of fission cross sections are summarized and the range of techniques available are considered. Methods of fission detection are described with particular emphasis on the neutron energy dependent properties of the fission process and the details of fragment energy loss which can lead to energy-dependent changes in detector efficiency. Selected examples of fission cross-section measurements are presented and methods of data reduction, storage, analysis and evaluation, are examined. Finally requested accuracies for fission cross section data are compared to estimated available accuracies. (U.K.)
Photodetachment cross-section of the negatively charged hydrogen ion
Frolov, Alexei M.
2015-01-01
Photodetachment cross-section $\\sigma_{ph}(p_e)$ of the negatively charged hydrogen ion H$^{-}$ is determined with the use of highly accurate variational wave functions constructed for this ion. Photodetachment cross-sections of the H$^{-}$ ion are also studied for very small and very large values of the photo-electron momentum $p_e$. Maximum of this cross-section and its location have been evaluated to high accuracy. In particular, we have found that $[\\sigma_{ph}(p_e)]_{\\max} \\approx$ 3.862...
Cross Section to Multiplicity Ratios at Very High Energy
Block, M M
2014-01-01
Recent data from the LHC makes it possible to examine an old speculation that at very high energy the total multiplicity and the cross section in elementary particle interactions vary in parallel with energy. Using fits incorporating the new data, it appears that the ratios of the total, elastic, and inelastic cross sections to the average multiplicity N can in fact approach constants at very high energy. The approach to the limit is however quite slow for the total and inelastic cross sections and is not yet reached at LHC energies. The elastic ratio sigma^{el}/N at 7 TeV, however, is not far from its asymptotic value.
Calculation of the neutron induced fission cross-section of 233Pa up to 20 MeV
Since very recently, direct measurements of the 233Pa(n,f) cross-section are available in the energy range from 1.0 to 8.5 MeV. This has stimulated a new, self-consistent, neutron cross-section evaluation for the n+233Pa system, in the incident neutron energy range 0.01-20 MeV. Since higher fission chances are involved also the lighter Pa-isotopes had to be re-evaluated in a consistent manner. The results are quite different compared to earlier evaluation attempts. Since 233Pa is a key isotope in the thorium based fuel cycle the quality of its reaction cross-sections is important for the modeling of future advanced fuel and reactor concepts. The present status of the evaluated libraries is that they differ by a factor of two in the absolute fission cross-section and also in the threshold energy value
Evaluation of cross sections of 63Cu and 65Cu for JENDL high energy file
Cross sections of 63Cu and65Cu for neutron and proton induced reactions have been evaluated up to 3 GeV for the High Energy File of Japanese Evaluated Nuclear Data Library (JENDL-HE). Different theoretical model codes were employed in this evaluation. For intermediate energy region between 20 and 150 MeV, GNASH based on statistical Hauser-Feshbach and preequilibrium models was used. Transmission coefficients calculated with DWUCK were used in the GNASH calculations of particle and photon emission cross sections and isotope production cross sections up to 150 MeV. The input parameters for the model codes were determined through analysis of experimental data in this energy region. For high-energy region between 150 MeV and 3 GeV, the JQMD-GEM code based on Quantum Molecular Dynamics (QMD) with statistical decay model (GEM) was employed. For energy region below 20 MeV, the existing JENDL-3.3 data were adopted. Several isotope production cross sections above 20 MeV were analyzed using the GMA code based on the generalized least-squares method or empirical fits in which experimental data were available. The results were compared with available experimental data as well as integral experiments such as thick target neutron yields. (author)
Contribution to the study of the unresolved resonance range of the neutrons cross sections
This document presents the statistical description of neutron cross sections in the unresolved resonance range. The modeling of the total cross section and of the 'shape - elastic' cross section is based on the 'average R-Matrix' formalism. The partial cross sections describing the radiative capture, elastic scattering, inelastic scattering and fission process are calculated using the Hauser-Feshbach formalism with width fluctuation corrections. In the unresolved resonance range, these models depend on the average resonance parameters (neutron strength function Sc, mean level spacing Dc, average partial reaction widths Γc, channel radius ac, effective radius R' and distant level parameter R-barc∞). The codes (NJOY, CALENDF...) dedicated to the processing of nuclear data libraries (JEFF, ENDF/B, JENDL, CENDL, BROND... ) use the average parameters to take into account the self-shielding phenomenon for the simulation of the neutron transport in Monte-Carlo (MCNP, TRIPOLI... ) and deterministic (APOLLO, ERANOS...) codes. The evaluation work consists in establishing a consistent set of average parameters as a function of the total angular momentum J of the system and of the orbital moment of the incident neutron l. The work presented in this paper aims to describe the links between the S-Matrix and the 'average R-Matrix' formalism for the calculation of Sc, R-barc∞, ac and R'. (author)
Takács, S.; Takács, M. P.; Ditrói, F.; Aikawa, M.; Haba, H.; Komori, Y.
2016-09-01
The cross sections of alpha particles induced nuclear reactions on natural germanium were investigated by using the standard stacked foil target technique, the activation method and high resolution gamma spectrometry. Targets with thickness of about 1 μm were prepared from natural Ge by vacuum evaporation onto 25 μm thick polyimide (Kapton) backing foils. Stacks were composed of Kapton-Ge-Ge-Kapton sandwich target foils and additional titanium monitor foils with nominal thickness of 11 μm to monitor the beam parameters using the natTi(α,x)51Cr reaction. The irradiations were done with Eα = 20.7 and Eα = 51.25 MeV, Iα = 50 nA alpha particle beams for about 1 h. Direct or cumulative activation cross sections were determined for production of the 72,73,75Se, 71,72,74,76,78As, and 69Ge radionuclides. The obtained experimental cross sections were compared to the results of theoretical calculations taken from the TENDL data library based on the TALYS computer code. A comparison was made with available experimental data measured earlier. Thick target yields were deduced from the experimental cross sections and compared with the data published before.
The effect of self-shielding of resonance cross sections on the tritium breeding ratio was investigated for three promising fusion blanket designs with liquid lithium, lithium oxide and lithium-lead breeders. Calculations were performed using ANISN and MCNP transport codes with the ENDF/B-V based nuclear data libraries. It is found that the self-shielding effect cannot be neglected in the blanket design if the blanket is neutron leaky in the case when the blanket is thin or with lower Li-6 enrichment in Li. This may result in an underestimate of the tritium breeding ratio if the cross sections are infinitely diluted. This is due to the resonances in the structure materials in which the absorption cross sections are enhanced in the infinitely diluted case. Thus the effect of self-shielding of resonance cross sections should be considered in neutronics calculations of fusion reactors. It is shown that the MCNP results are better reproduced by those from the transport code with the infinitely diluted library. This is probably due to the weight function used to generate the library and to the number of groups considered. Thus for fusion applications it is recommanded to collapse broad group cross sections with the spectrum obtained from an accurate calculation based on many fine groups. (author)
Local Deplanation Of Double Reinforced Beam Cross Section Under Bending
Baltov, Anguel; Yanakieva, Ana
2015-12-01
Bending of beams, double reinforced by means of thin composite layers, is considered in the study. Approximate numerical solution is proposed, considering transitional boundary areas, where smooth quadratic transition of the elasticity modulus and deformations take place. Deplanation of the cross section is also accounted for in the areas. Their thickness is found equalizing the total stiffness of the cross section and the layer stiffness. Deplanation of the cross section of the transitional area is determined via the longitudinal deformation in the reinforcing layer, accounting for the equilibrium between the internal and the external moment, generated by the longitudinal stresses in the cross section. A numerical example is given as an illustration demonstrating model's plausibility. The model allows the design and the calculation of recycled concrete beams double reinforced by means of thin layers. The approach is in agreement with modern design of nearly zero energy buildings (NZEB).
Electronic stopping cross sections for use in ion range calculation
Theoretical and empirical methods of determining the electronic stopping cross sections are discussed. The values used by various authors in ion range calculations are outlined. Recommendations are made for future range calculations. (author)
Nonelastic-scattering cross sections of elemental nickel
Neutron total cross sections of elemental nickel were measured from 1.3 to 4.5 MeV, at intervals of approx. 50 keV, with resolutions of 30 to 50 keV and to accuracies of 1 to 2.5%. Neutron differential-elastic-scattering cross sections were measured from 1.45 to 3.8 MeV, at intervals and with resolutions comparable to those of the total cross sections, and to accuracies of 3 to 5%. The nonelastic-scattering cross section is derived from the measured values to accuracies of greater than or equal to 6%. The experimental results are compared with previously reported values as represented by ENDF/B-V, and areas of consistency and discrepancy, noted. The measured results are shown to be in good agreement with the predictions of a model previously reported by the authors. 4 figures, 1 table
Longitudinal Vibrations of Rheological Rod With Variable Cross Section
Katica（Stevanovic）HEDRIH; AleksandarFILIPOVSKI
1999-01-01
Longitudinal vibrations of rheological rod with variable cross section are examined.Particular solutions and eigenfunction are accomplished for natural vibrations of the rod with hereditary material of standard hereditary body.Some examples are given.
Fission cross section for 242Am.met
The neutron-induced fission cross section for 242Am.met (152y) was measured at the Livermore 100-MeV electron linac in the neutron energy range of 0.01 eV to 20 MeV. Fission fragments were detected using a hemispherical fission chamber. The neutron flux was measured below 10 keV using lithium glass scintillators. Above 10 keV, the 242Am.met fission cross section was measured relative to the 235U fission cross section. Below 20 eV, the data were fit with a sum of single-level Breit-Wigner resonances. Results for the distribution of fission widths, the average fission width, and the average level spacing are presented. The fission cross section in the 100 keV to 20 MeV range is compared with previous measurements
Differential cross sections of positron–hydrogen collisions
Rong-Mei, Yu; Chun-Ying, Pu; Xiao-Yu, Huang; Fu-Rong, Yin; Xu-Yan, Liu; Li-Guang, Jiao; Ya-Jun, Zhou
2016-07-01
We make a detailed study on the angular differential cross sections of positron–hydrogen collisions by using the momentum-space coupled-channels optical (CCO) method for incident energies below the H ionization threshold. The target continuum and the positronium (Ps) formation channels are included in the coupled-channels calculations via a complex equivalent-local optical potential. The critical points, which show minima in the differential cross sections, as a function of the scattering angle and the incident energy are investigated. The resonances in the angular differential cross sections are reported for the first time in this energy range. The effects of the target continuum and the Ps formation channels on the different cross sections are discussed. Project supported by the Nanyang Normal University Science Foundation of China (Grant No. ZX2013017) and the National Natural Science Foundation of China (Grant Nos. 11174066, 61306007, and U1304114).
Local Deplanation Of Double Reinforced Beam Cross Section Under Bending*
Baltov Anguel
2015-12-01
Full Text Available Bending of beams, double reinforced by means of thin composite layers, is considered in the study. Approximate numerical solution is proposed, considering transitional boundary areas, where smooth quadratic transition of the elasticity modulus and deformations take place. Deplanation of the cross section is also accounted for in the areas. Their thickness is found equalizing the total stiffness of the cross section and the layer stiffness. Deplanation of the cross section of the transitional area is determined via the longitudinal deformation in the reinforcing layer, accounting for the equilibrium between the internal and the external moment, generated by the longitudinal stresses in the cross section. A numerical example is given as an illustration demonstrating model’s plausibility. The model allows the design and the calculation of recycled concrete beams double reinforced by means of thin layers. The approach is in agreement with modern design of nearly zero energy buildings (NZEB.
Absolute cross sections for dissociative electron attachment to HCCCN
New absolute cross sections for dissociative electron attachment to HCCCN (cyanoacetylene or propiolonitrile) in the range of 0–10 eV electron energy are presented here, which have been determined from a new analysis of previously reported data (Graupner et al 2006 New J. Phys. 8 117). The highest cross sections are observed for the formation of CN− at 5.3 eV and CCCN− at 5.1 eV; approximately 0.06 Å2 and 0.05 Å2 respectively. As part of the re-analysis, it was necessary to determine absolute cross sections for electron-impact ionization of HCCCN with the binary-encounter Bethe method. These electron-impact ionization absolute cross sections for HCCCN are also presented here; the maximum value was found to be ∼6.6 Å2 at ∼80 eV. (paper)
Scaling of Cross Sections for Ion-atom Impact Ionization
The values of ion-atom ionization cross sections are frequently needed for many applications that utilize the propagation of fast ions through matter. When experimental data and theoretical calculations are not available, approximate formulas are frequently used. This paper briefly summarizes the most important theoretical results and approaches to cross section calculations in order to place the discussion in historical perspective and offer a concise introduction to the topic. Based on experimental data and theoretical predictions, a new fit for ionization cross sections is proposed. The range of validity and accuracy of several frequently used approximations (classical trajectory, the Born approximation, and so forth) are discussed using, as examples, the ionization cross sections of hydrogen and helium atoms by various fully stripped ions
The neutron cross-sections of Xe135
Measurements of the total and absorption cross-sections of Xe135 reviewed briefly. The low-energy cross-section is very large and dominated by a single resonance at 0.084 eV; the spin state for this level is not known, this being one of the major uncertainties in the data. The resonance parameters given in the literature were found to give a good fit to the total cross-section but failed to reproduce the preferred 2200 m/sec. value of σγ. A new set of parameters was therefore deduced, by a least-squares analysis, which gave this preferred value of σγ and fitted the shape of the total cross section curve. To obtain this fit it was necessary to re-normalise the curve of σT by 4%. The new parameters are listed, and a discussion of the probable accuracy of the data is included. (author)
Radiative neutron capture cross sections on 176Lu at DANCE
Roig, O.; Jandel, M.; Méot, V.; Bond, E. M.; Bredeweg, T. A.; Couture, A. J.; Haight, R. C.; Keksis, A. L.; Rundberg, R. S.; Ullmann, J. L.; Vieira, D. J.
2016-03-01
The cross section of the neutron capture reaction 176Lu(n ,γ ) has been measured for a wide incident neutron energy range with the Detector for Advanced Neutron Capture Experiments at the Los Alamos Neutron Science Center. The thermal neutron capture cross section was determined to be (1912 ±132 ) b for one of the Lu natural isotopes, 176Lu. The resonance part was measured and compared to the Mughabghab's atlas using the R -matrix code, sammy. At higher neutron energies the measured cross sections are compared to ENDF/B-VII.1, JEFF-3.2, and BRC evaluated nuclear data. The Maxwellian averaged cross sections in a stellar plasma for thermal energies between 5 keV and 100 keV were extracted using these data.
Macroscopic cross section measurements in materials by neutron radiography technique
Macroscopic cross-section of materials play an important role in the study of material properties. Number of materials are used for shielding against penetrating radiation like X-rays, gamma rays and neutrons and exhibit different attenuation cross-sections. Neutron radiography technique is a multi discipline non-destructive technique with a large number of applications. The technique was applied to study and analyze the behavior of different shielding materials against thermal neutrons. Samples as step wedges of graphite, copper, brass and acrylic etc. were fabricated. The test samples were exposed to a beam of thermal neutrons at neutron radiography facility and the transmittance of neutrons through different materials was measured. Gamma-ray contribution and scattered radiation were subtracted from the observed neutron intensities to calculate the neutron macroscopic cross-section. Calculated values of the macroscopic cross-section were compared with the values given in the literature. (author)
Scaling of Cross Sections for Ion-atom Impact Ionization
Kaganovich, I D; Startsev, E
2003-01-01
The values of ion-atom ionization cross sections are frequently needed for many applications that utilize the propagation of fast ions through matter. When experimental data and theoretical calculations are not available, approximate formulas are frequently used. This paper briefly summarizes the most important theoretical results and approaches to cross section calculations in order to place the discussion in historical perspective and offer a concise introduction to the topic. Based on experimental data and theoretical predictions, a new fit for ionization cross sections is proposed. The range of validity and accuracy of several frequently used approximations (classical trajectory, the Born approximation, and so forth) are discussed using, as examples, the ionization cross sections of hydrogen and helium atoms by various fully stripped ions.
Models for Photon-photon Total Cross-sections
Godbole, RM; Grau, A.; Pancheri, G.
1999-01-01
We present here a brief overview of recent models describing the photon-photon cross-section into hadrons. We shall show in detail results from the eikonal minijet model, with and without soft gluon summation.
Electron Swarm Parameters and Electron Collision Cross Sections
Electron collision cross section data for atoms and molecules and electron swarm data in respective gases are important for quantitative modeling of related plasmas. This fact and wide application of plasmas in various fields boos data collection and evaluation activities worldwide. We have been measuring electron swarm parameters (drift velocity, longitudinal diffusion coefficient, ionization/attachment coefficients, and so on) over a wide E/N range (where E is the electric field and N the gas number density) in a number of gases. We also derived a set of electron collision cross sections for each gas so that the set was consistent with our experimental swarm data. Our speciality in studying molecular target is to measure swarm parameters not only in the pure molecular gas but also in dilute molecular gas-argon gas mixtures, the mix rations of the molecule are 0.5-5.0%. The swarm parameters in pure molecular gas depend primarily on the elastic momentum transfer cross section of the molecule and its vibrational excitation cross sections. Those in the mixtures, on the other hand, depend mainly on the elastic momentum transfer cross section of major argon atom and the vibrational cross sections of minor admixed molecule. Alternative use of swarm parameters in pure molecular gas and those in the mixtures enable us to derive the momentum transfer cross section and vibrational cross sections for the molecule separately. Combination of the Ramsauer-Townsend minimum of argon atom and sharp structures in vibrational cross sections of the molecule frequently gives rise prominent E/N dependences in swarm parameters, which can be used to determine the position and magnitude of resonances in the vibrational excitation cross sections. Detailed accounts of the procedure, including estimated uncertainty in our electron swarm data and also in the resultant set of electron collision cross sections, will be given in the presentation by referring to our recent results. Stress will be
Absolute Total np and pp Cross Section Determinations
Arndt, R A; Laptev, A B; Strakovsky, I I; Workman, R L
2008-01-01
Absolute total cross sections for np and pp scattering below 1000 MeV are determined based on partial-wave analyses of NN scattering data. These cross sections are compared with most recent ENDF/B and JENDL data files, and the Nijmegen partial-wave analysis. Systematic deviations from the ENDF/B and JENDL evaluations are found to exist in the low-energy region.
Singly differential cross sections with exchange for Ps-fragmentation
Ray, Hasi
2008-01-01
Ps ionization in Ps-atom scattering is of fundamental importance. The singly differential cross sections (SDCS) provides more accurate information to test a theory than integrated or total ionization cross section since the averaging over one parameter is not required. We evaluate the SDCS for Ps-ionization with respect to the longitudinal energy distribution of the break-up positron and electron in Ps-H and Ps-He scattering and compare them with the recently available experimental and theore...
Photoproduction models for total cross section and shower development
Cornet Fernando
2015-01-01
Full Text Available A model for the total photoproduction cross section, based on the ansatz that resummation of infrared gluons limits the rise induced by QCD minijets in all the total cross-sections, is used to simulate extended air showers initiated by cosmic rays with the AIRES simulation program. The impact on common shower observables, especially those related with muon production, is analysed and compared with the corresponding results obtained with previous photoproduction models.
Measurement of fusion cross section with neutron halo nuclei
Fusion cross sections of 11Be, 10Be and 9Be have been measured on 209Bi target at 30-70MeV. Due to the neutron halo effect of 11Be, a large enhancement or suppression of the fusion cross section around the Coulomb barrier was theoretically predicted. Comparing the excitation function of 11Be with 10Be at near the Coulomb barrier region, no significant difference has been observed. ((orig.))
Top Quark Pair Production Cross Section at the Tevatron
Peters, Reinhild Yvonne [Manchester U.
2015-09-25
The top quark, discovered in 1995 by the CDF and D0 collaborations at the Tevatron proton antiproton collider at Fermilab, has undergone intense studies in the last 20 years. Currently, CDF and D0 converge on their measurements of top-antitop quark production cross sections using the full Tevatron data sample. In these proceedings, the latest results on inclusive and differential measurements of top-antitop quark production cross sections at the Tevatron are reported.
Elastic cross sections for electron-carbon scattering
Liu Jun-Bo; Wang Yang; Zhou Ya-Jun
2007-01-01
We used the close-coupling optical (CCO) approach to investigate the open-shell carbon atom. The elastic cross sections have been presented at the energies below 90eV, and the present CCO results have been compared with other theoretical results. We found that polarization and the continuum states have significant contributions to the elastic cross sections. The present calculations show that the CCO method is capable of calculating electron scattering from open-shell atoms.
Thermal neutron capture cross sections of tellurium isotopes
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
Thermal neutron capture cross sections of tellurium isotopes
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 γ-ray cross sections obtained from experiments on targets contained natural Te and γ intensities per capture of individual Te isotopes. Isomeric ratios for the thermal neutron capture on the even tellurium isotopes are also given
Thermal neutron capture cross sections of tellurium isotopes
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.