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
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)
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)
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.
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
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
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
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
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
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. (authors)
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.
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.)
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
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
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
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
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.
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
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.
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)
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)
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.
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.
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)
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)
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)
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
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 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
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.
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)
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
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
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)
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...
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)
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.
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
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.
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)
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.)
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
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.
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