1. NAME AND TITLE OF DATA LIBRARY
FENDL-2.0: Compendium of Reference and Processed Sub-libraries Derived from International Evaluated Nuclear Data Files for Fusion Applications.
2. NAME AND TITLE OF DATA RETRIEVAL PROGRAMS
No retrieval program is included in the package. Data are in ENDF-6 and well known processed formats: REAC, GENDF (output of NJOY/GROUPR and NJOY/GAMINR), MATXS, and continuous energy format for use in MCNP calculations.
IAEA Nuclear Data Section, Vienna, Austria.
4. HISTORICAL BACKGROUND AND INFORMATION
The IAEA Nuclear Data Section (NDS), in cooperation with national nuclear reaction data centers and research groups, has established the international Fusion Evaluated Nuclear Data Library (FENDL). This library is a comprehensive collection of high-quality nuclear data, selected from the various existing national data libraries, covering the necessary nuclear input for physics and engineering studies of the nuclear performance of the International Thermonuclear Experimental Reactor (ITER) project and other fusion-related development projects. Within the framework of the FENDL project, the IAEA/NDS played the critical organizational role in planning and coordinating the assembly, cross section processing and benchmark data testing of the FENDL library following the recommendations of a series of international meetings.
The FENDL project started in 1987 and the first version of the library, FENDL-1, was officially released early in 1995. Extensive integral data testing of FENDL-1 led to an improved version designated as FENDL-2 which was finalized for general distribution in January 1999. FENDL-2 represents contributions from ENDF/B-VI (United States), JENDL-3 and JENDL-FF (Japan), BROND-2 (Russian Federation), and EFF-3 (European Union).
FENDL-2.0 obsoletes previous versions, and it is strongly recommended that all users request the new package.
5. APPLICATION OF THE DATA
The FENDL library is a comprehensive source of processed and benchmark tested nuclear data tailored to the requirements of the ITER EDA Project and other fusion-related development projects.
6. SOURCE AND SCOPE OF DATA
The present version of FENDL consists of the following sublibraries:
FENDL/A-2.0 [Ref. 2] (October 1998) neutron-activation cross sections below 20 MeV for stable and unstable nuclides with half-lives longer than ½ day, selected from different available sources, for 13006 activation reactions for 739 target nuclides and processed into PENDF (pointwise ENDF-6 format) using the IAEA pre-processing codes LINEAR/RECENT/FIXUP (PSR-351/PREPRO-96) [Ref. 9], to be used in calculations of activation resulting from the transport of plasma-source neutrons and secondary gamma rays. The pointwise data have been processed further into four representations:
MCNP: Data were processed and submitted by F. M. Mann [Ref. 3] into continuous energy format for input to the Monte-Carlo neutron/photon transport code CCC-200/MCNP
VITJ_E: VITAMIN-J 175 multigroup data were processed and submitted by F. M. Mann. [Ref. 3] using the VITAMIN-E spectrum for input to the REAC*2/3 (CCC-443) transmutation code
VITAMIN-J 175 multigroup data processed the same as above for input to FOUR ACES code. Data in LIBOUT format.
VITJ_FLAT: 175 group data in ENDF-6 format, processed with a flat weighting spectrum [Ref. 3] using the GROUPIE pre-processing code [Ref. 9]
FENDL/C-2.0: (March 1997), data for the fusion reactions D(d,n), D(d,p), T(d,n), T(t,2n), He-3(d,p) extracted from ENDF/B-6 and processed [Ref. 6].
FENDL/D-2.0: (March 1997), nuclear decay data for 1867 nuclides and isomers in ENDF-6 format as well as processed formats for REAC and MCNP. This data is identical to the EAF-4.1 decay library and is complementary to the activation cross section sublibrary FENDL/A-2.0. [Ref. 4].
FENDL/DS-2.0: (March 1996), neutron activation data for dosimetry by foil activation. This is identical with file 1 (neutron activation cross-sections) of the International Reactor Dosimetry File IRDF-90 version 2 of 1993 [Ref. 5], given as multigroup data in 640 group extended SAND-2 format with covariance data.
FENDL/E-2.0: (October 1998), basic evaluated data, in ENDF-6 format, for coupled neutron-photon transport calculations. It includes the basic evaluations for neutron interaction and photon production for 57 elements and isotopes, selected from ENDF/B-VI, JENDL-3.2 and JENDL-FF, BROND-2, EFF-3 [Ref. 7],and a photon-atom interaction data library for 34 elements taken from ENDF/B-VI [Ref. 10]. The carryover FENDL/E-1.0 data have been processed by R. E. MacFarlane using the data processing code NJOY (PSR-355) [Ref. 11]. In contrast, the FENDL-2 nuclides were processed by the regional nuclear data projects. The results are presented in the following two sublibraries:
FENDL/MG-2.0: (October 1998), multigroup data in GENDF and MATXSR format (output of the NJOY modules GROUPR, GAMINR and MATXSR) for use in discrete-ordinates calculations (TRANSX, ANISN, DANTSYS-3.0, DOORS-3.2 etc.) [Ref. 8]. MATXS files for 57 nuclides are available.
FENDL/MC-2.0: (October 1998), data processed into the ACE (continuous energy) format needed for input to the Monte Carlo code MCNP4B (output of the NJOY module ACER) [Ref. 8]. Data available for 57 nuclides.
7. DISCUSSION OF THE DATA RETRIEVAL PROGRAM
No data retrieval program is included in the package. See note under item 2 above.
8. DATA FORMAT AND COMPUTER
The data are in ascii format which can be used on any computer platform. However, because of the large size of this package, they are distributed only in Unix compressed tar files.
9. TYPICAL RUNNING TIME
Run times vary.
a. Documentation available with library:
A.B. Pashchenko, H. Wienke, and D.W. Muir, "FENDL-2: An Improved Nuclear Data Library for Fusion Applications," pp 1150-1154, International Conference on Nuclear Data for Science and Technology, edited by G. Reffo, A. Ventura, and C. Grandi, Trieste, Italy, 19-24 May 1997.
A.B. Pashchenko, H. Wienke, J. Kopecky, J.-Ch. Sublet, and R.A. Forrest, "FENDL/A-2.0, Neutron Activation Cross Section Data Library for Fusion Applications," IAEA-NDS-173, Rev. 1 (October 1998).
F.M. Mann, R.A. Forrest and H. Wienke, "FENDL2/A-MCNP, FENDL2/A-VITJ_E and FENDL2/A-VITJ_FLAT: The Processed FENDL-2 Neutron Activation Cross Section Data Files," (summary documentation by A.B. Pashchenko and H. Wienke) IAEA-NDS-174, (March 1997).
R.A. Forrest and F.M. Mann, "FENDL/D-2.0, Decay Data Library for Fusion Applications," (summary documentation by A.B. Pashchenko and H. Wienke) IAEA-NDS-178, Rev. 0 (March 1997).
N. P. Kocherov and P.K. McLaughlin, "The International Reactor Dosimetry File (IRDF-90 Version 2)," IAEA-NDS-141, Rev. 3 (March 1996).
R.M. White, D.A. Resler and G.M. Hale, "FENDL/C-2.0, Charged-Particle Reaction Data Library for Fusion Applications," (summary documentation by A.B. Pashchenko and H. Wienke) IAEA-NDS-177 (March 1997).
"FENDL/E-2.0, Evaluated Nuclear Data Library of Neutron Nuclear Interaction Cross Sections and Photon Production Cross Sections and Photon-Atom Interaction Cross Sections for Fusion Applications," (summary documentation by A.B. Pashchenko and H. Wienke) IAEA-NDS-175, Rev. 3 (October 1998).
"FENDL/MG-2.0 and FENDL/MC-2.0, The Processed Cross Section Libraries for Neutron-Photon Transport Calculations," (summary documentation H. Wienke and M. Herman), IAEA-NDS-176, Rev.1 (October 1998).
b. Other useful documentation:
D. E. Cullen, "The 1996 ENDF Pre-processing Codes," IAEA-NDS-39, Rev. 9 (November 1996).
"EN6-PHOTO and JEF-2/PHOTO Photo-Atomic Interaction Data Library," (summary documentation by H. D. Lemmel) IAEA-NDS-58, Rev. 4 (September 1994).
R. E. MacFarlane, "The NJOY Nuclear Data Processing System, Version 91," LA-12740-M (1994).
Sz. Czifrus, "Processing of ENDF-6 Format Resonance Region Covariance Data Using a New Algorithm," KERNTECHNIK 60, 152-156 (1995).
11. CONTENTS OF LIBRARY
Included are the referenced documents in 10.a and the FENDL-2.0 libraries written in Unix compressed tar files. The entire package is 261MB when compressed and requires an additional 1 gigabyte of disk space for the uncompressed files.
12. DATE OF ABSTRACT
KEYWORDS: MULTIGROUP CROSS SECTIONS; CHARGED PARTICLE CROSS SECTIONS; DOSIMETRY CROSS SECTIONS; REACTION CROSS SECTIONS; MATXS FORMAT; MCNP FORMAT; RADIOACTIVE DECAY SPECTRA