Nuclear data forms
FISPACT-II requires connection to several nuclear data libraries and forms before it can be used to calculate inventories. While any libraries in the correct ENDF-6 format could be used (with suitable processing), the development of FISPACT-II over the last several years has run in parallel with the development of the TALYS-based Evaluated Nuclear Data Library TENDL project and those libraries are the recommended source of cross section data forms. Together FISPACT-II and TENDL's nuclear data forms make up the simulation platform that is a complete package tailored for all application needs: nuclear fission and fusion, nuclear fuel cycle, accelerator physics, isotope production, material characterisation, storage and life cycle, earth exploration, astrophysics, homeland security and more.
The following data libraries are required:
- Cross section data for neutron, proton, deuteron, alpha and gamma-induced reactions
- Fission yields data for neutron, proton, deuteron, alpha and gamma-induced reactions
- Variance-covaraiance data for neutron-induced reactions
- Probability tables data for neutron-induced reactions in the resonance energy ranges
- Decay data
- Radiological data:
- Biological hazard data
- Legal transport data
- Clearance data
To streamline, simplify and control any feature of all the nuclear data assimilation processes, the code development philosophy has been to follow in all aspects as much as possible the format described in the ENDF-6 format manual. Three processing codes are used in sequence and in parallel to produce, process, check, and compare the nuclear data forms: NJOY12-050, PREPRO-2015 and CALENDF-2010. All the processing steps cannot be handled by only one or even two of those unique processing codes, a combination of the three is needed to extract the data forms that are the most useful in all applications. A schematic of the processing sequences is shown in the figure below.
Cross Section Data
The principal sources of cross-section data are the di erent generations of the TALYS-based Evaluated Nuclear Data Libraries. The latest TENDL-2015 is the recommended evaluated data source for use in any type of nuclear technology applications. The principal advances of this new library are in the unique target coverage, over 2800 nuclides; the upper energy range, 200 MeV; variance-covariance information for all nuclides; and the extension to cover all important projectiles: neutron, proton, deuteron, alpha and gamma, and last but not least the proven capacity of this type of library to transfer regularly to technology the feedbacks of extensive validation, veri cation and benchmark activities from one release to the next. TENDL is the sixth generation of such a library and as such has bene ted from the previous releases since TENDL-2008, as well as a wide range of modern V&V.
The cross sections are provided in a set of group structures including:
- UKAEA-1102 general-purpose neutron group structure
- CCFE-162 incident charged particle group structure
- CCFE-709 fusion neutron group structure
- CASMO-586 fission reactor group structure
- Several legacy groups are available with deprecated EAF libraries
While the neutron data is contained in the finer 709 and 1102 multi-groups, the 162 scheme for the non-resonant p, d, α and γ-induced cross-sections is used. The data format used is fully compliant with the ENDF-6 manual specification handled on an isotopic basis and so allows many existing utility codes further to manipulate, visualise or check any aspects of the pre-processed files. The data files are produced using a complex but robust, complementary sequence of modules of the processing codes NJOY12-050 and PREPRO-2015. During the processing outputs from veri cation and validation steps are regularly taken in order to establish the validity of all computed derived data. To be able to account for Doppler broadening effects the processed files are given at three reactor temperatures: 293.6, 600 and 900 degree Kelvin and two astrophysical temperatures: 5 and 30 keV.
Fission Yield Data
The fission yield data need to be provided for each actinide and incident particle. The files are supplied in an ENDF-6 format and are read by FISPACT-II with no further processing. A library is provided based on the JEFF-3.1.1 library for neutron-induced fission. Only 19 of the many nuclides that have ssion have any ssion yield data in JEFF-3.1.1 and these cover only a reduced energy range. For the remainder the UKFY4.2 library then further extends the range before a neighbouring fission yield is used. This UKFY4.2 library using Wahl's systematics is also used for all other particle induced fission yields.
GEF-based fission-fragment yield libraries in ENDF-6 format are also provided: GEFY-5.2 as independent and cumulative ssion-fragment yields with multi-chance fission. The uncertainties are given and reflect the uncertainties of the model. They are determined from calculations with perturbed model parameters. 109 spontaneous and 119 neutron induced ssion, including target in isomeric state are provided on a fine 49 incident energy grid structure up to 20 MeV.
Variance and Covariance Data
Above the upper energy of the resolved resonance range, for each of the 2800+ isotopes a Monte Carlo method in which the covariance data come from uncertainties of the nuclear model calculations is used. A complete description of the procedure is given in [dx.doi.org/10.1016/j.nima.2008.02.003 this reference]. For all isotopes, the initial "best" set of results is produced by a TALYS calculation with an adjusted input parameter set. This set of results is stored in a set of sampeld ENDF files MF-3 to MF-10. For each isotope, many TALYS runs with random nuclear model parameters are performed, which are used to generate and correlations. As well as correlation within the same reaction channels, correlation between reaction channels is included. All information on cross section covariance is stored in the MF-33 format, starting at the end of the resonance range up to 200 MeV. Short-range, self-scaling variance components are also specified for each MT type.
The data format used to store the variance-covariance information has been made fully compliant with the ENDF-6 format description and the files are read directly by FISPACT-II without any further processing.
The CALENDF nuclear data processing system is used to convert the evaluation defining the cross-sections in ENDF-6 format (i.e., the resonance parameters, both resolved and unresolved) into forms useful for applications. Those forms used to describe neutron cross-section uctuations correspond to cross section probability tables", based on Gauss quadratures and effective cross-sections. The CALENDF-2010 code provides those probability tables in the energy range from 0.1 eV up to the end of the resolved or the unresolved resonance range. Probability table data in 709 group formats are provided for the majority of isotopes of the TENDL library. These data are used to model dilution effects from channel, isotopic or elemental interferences. To account for Doppler broadening effects the tables are given at three temperatures: 293.6, 600 and 900 degree Kelvin.
In addition to cross-sections the other basic quantities required by an inventory code are information on the decay properties (such as half-life) of all the nuclides considered. These data are available in a handful of evaluated decay data libraries. FISPACT-II is able to read the data directly in ENDF-6 format; it requires no pre-processing to be done. The eaf_dec_2010 library, based primarily on the JEFF-3.1.1 and JEF-2.2 radioactive decay data libraries with additional data from the latest UK evaluations UKPADD6.10, contain 2233 nuclides. However, to handle the extension in incident particle type, energy range and number of targets, many more are needed. A new 3875-nuclide decay library UKDD-12 has been assembled from eaf_dec_2010 complemented with all of JEFF-3.1.1, a handful of ENDF/B-VII.1 and other decay files to cover the range of daughters of TENDL and short lived ssion products.
There remain compatibility issues between the isomer de nitions arising from the cross section library, through the RIPL-3 database and the newly assembled decay library. Historical incompatibilities in isomeric state number (g, m, n, o, . . . ) and energy levels between radionuclide daughter products of reactions and the associated decay data files will need to be addressed in a future release.
Other Nuclear Data Libraries
FISPACT-II is compatible with all fully ENDF-6 compliant nuclear data forms which have been suitably processed and provided in the required multigroup stuctures. Several libraries are distributed with the code, including the most recent neutron-incident, fission yield and decay data from ENDF/B, JENDL and JEFF.
ENDF/B Nuclear Data Libraries
The Cross Section Evaluation Working Group (CSEWG) released the ENDF/B-VII.1 library on 22 December 2011. The ENDF/B-VII.1 library is the US latest recommended evaluated nuclear data le for use in nuclear science and technology applications, and incorporates advances made in the five years since the release of ENDF/B-VII.0, including: many new evaluation in the neutron sublibrary (423 in all) and over 190 of these contain covariances, new ssion product yields for 31 isotopes and a greatly expanded decay data sublibrary for 3817 radionuclides.
For more details, visit the NNDC offical website.
JENDL Nuclear Data Libraries
The purpose of JENDL-4.0 is to provide a Japanese standard library for fast breeder reactors, thermal reactors, fusion neutronics and shielding calculations, and other applications. The data libraries used have been updated to the JENDL-4.0u level of August 2013 for both the neutron reaction and ssion yields sublibrary. JENDL FP Decay Data File 2011 contains decay data of 1284 FP nuclides (of which 142 nuclides are stable) that includes recent TAGS (Total Absorption Gamma-ray Spectroscopy) information.
For more details, visit the JAEA official website.
JEFF Nuclear Data Libraries
The Joint Evaluated Fission and Fusion File is an evaluated library produced via an international collaboration of Data Bank member countries co-ordinated by the JEFF Scientific Co-ordination Group, under the auspices of the NEA Data Bank. The new JEFF-3.2 general purpose library has been released on March 5, 2014 in ENDF-6 format and contains incident neutron data for 472 nuclides or elements from 1-H-1 to 100-Fm-255.
For more details, visit the OECD-NEA official website.
CENDL Nuclear Data Libraries
The CENDL-3.1 neutron-induced cross section data was not released in the most recent FISPACT-II 3.00 version, but is available in the 709 group structure ENDF-6 format files directly from the FISPACT-II online nuclear data repository. The compressed file is approximately 26 Mb in size. It is recommended that users install the uncompressed files within their FISPACT-II installation directory under ENDFdata.
Please note: Ag, Ca, Cd, Cl, Cu, Ge, Hg, K, S, Sn, Tl, V, W and Zn are given as elemental evaluations which are not suitable for many activation analyses.
For more details we refer the interested reader to the ND2007 paper on CENDL-3.1.