1. NAME AND TITLE
FEDGROUP-C86 Rev. 3: Code System for Processing Evaluated Nuclear Data in
ENDF/B, KEDAK or UKNDL Formats into Constants
for Reactor Physics Calculations
FEDAUX-C86: Auxiliary programs for printing, deleting, merging, rewriting, converting, mixing, and updating libraries.
FLANGE-AE: Thermal neutron processing from ENDF data.
CVRT: Converts source file for VAX implementation.
WIMS-IJS1: Extended version of the WIMS group constants library.
Available as D00147D881001.
J. Stefan Institute, Ljubljana, Slovenia through the NEA Data Bank, France.
3. CODING LANGUAGE AND COMPUTER
Fortran IV Extended; CDC CYBER-72 and VAX-8810.
4. NATURE OF PROBLEM SOLVED
Data contained in the evaluated nuclear data files are processed to produce spectrum averaged multigroup constants for applications in transport calculations. Cross sections (shielded and unshielded), scattering matrices and some other constants are prepared. Special emphasis has been placed on thermal reactor applications, particularly on the WIMS cross section library.
FEDGROUP-C, was initially derived from PSR-123/FEDGROUP-3, operable in an IBM environment, to run on a CDC Cyber 174. FEDGROUP-C86 was released because of an increased interest in the use of FEDGROUP-C as an evaluated data processing code and a link to the WIMS library.
FEDGROUP C86 developed from its ancestor FEDGROUP-2 by Dr. P. Vertes, Budapest, Hungary. Different versions of FEDGROUP originating from other laboratories (for example: FEDGROUP-3 by Dr. P. Vertes, Budapest, Hungary, FEDGROUP-R by Dr. A. Holubar, Rez, Czech and Slovak Federal Republic, etc.) are generally incompatible. They only have their name and their ancestor FEDGROUP-2 in common.
Types of evaluated files which can be processed are: ENDF/B, KEDAK, UKNDL and Livermore
ENDF in ENDF/B format.
5. METHOD OF SOLUTION
KEDAK, UKNDL, ENDF-4 and ENDF-5 data formats are accepted. An internal binary file is constructed which has the same format regardless of the source data library.
The data averaging spectrum may be specified by the user. A standard spectrum consisting of the fission spectrum, 1/E and Maxwellian spectrum can be requested.
Doppler broadening is performed from the resonance parameters. In the resolved resonance range the so-called psi-chi method is used.
The calculation of inelastic scattering matrices for the thermal region is performed with FLANGE-AE. Auxiliary programs perform a variety of cross-section library manipulations, including the
updating of a WIMS library with data calculated by FEDGROUP-C.
6. RESTRICTIONS OR LIMITATIONS
By default the size of the work array declared in the main program is 30000 words. Due to dynamic programming this is sufficient for most purposes, even when preprocessed ENDF files are used when resonance reconstruction (and Doppler broadening) have been performed by other codes. Preparing data sets with a large number of groups can be done by performing the calculations a few groups at a time and merging the data subsequently.
For inelastic scattering into continuum an evaporation spectrum is assumed in constructing the
scattering matrix. The same assumption is applied to generate the scattering matrices for the multiple
neutron scattering reactions such as (n,2n) and (n,3n). This restriction must be considered when
preparing data for fast reactor or shielding applications while it is considered less restrictive in the case
of thermal reactors.
7. TYPICAL RUNNING TIME
Running time is strongly case-dependent. The sample case which includes executable code
creation, complete processing of 125U data from ENDF/B-V and a WIMS library update takes less than
20 min. on micro-VAX 3900. The installation of this version plus the execution of the three test
problems requires about 5000 CPU seconds and 1000 I/O seconds of a CDC830.
8. COMPUTER HARDWARE REQUIREMENTS
The package has been successfully installed on several CDC machines and VAX machines from
micro VAX 3900 to VAX 8810.
9. COMPUTER SOFTWARE REQUIREMENTS
The code has run using the Fortran IV Extended compiler, under NOS/BE 1.4 operating system.
The VAX version runs under VMS.
a. Included in documentation:
A. Trkov, "Notes on FEDGROUP-C86 (Rev.2)," IJS-DP-5446 (June 9, 1989).
A. Trkov, "Notes on FEDGROUP-C86 (Rev.3)," IJS-DP-6228 (November 29, 1991).
A. Trkov and A. Perdan, "FEDGROUP Package Upgrade to Version C86," IJS-DP-4368 (August 1986).
A. Trkov, A. Perdan and M. Budnar, "FEDGROUP-C84 - An Improved and Modified CDC Version of the Program Package for Processing Evaluated Nuclear Data in KEDAK, UKNDL and ENDF/B Format," INDC(YUG)-9/GV (April 1984).
A. Trkov and A. Perdan," Modifications to the FLANGE-AE Package," IJS-DP-3470 (March
b. Background information:
D. E. Cullen, "Report on the IAEA Cross Section Processing Code Verification Project," INDC(NDS)-170/NI (May 1985).
P. Vertes, "FEDGROUP-3, A Program System for Processing Evaluated Nuclear Data in
ENDF/B, KEDAK or UKNDL Format to Constants to be Used in Reactor Physics Calculation,"
11. CONTENTS OF CODE PACKAGE
Included are the referenced documents and one (1.44MB) DOS diskette.
12. DATE OF ABSTRACT
October 1984, updated January 1988, March 1993.
KEYWORDS: ENDF FORMAT; MULTIGROUP CROSS SECTION PROCESSING; NEUTRON CROSS SECTION PROCESSING; REACTOR PHYSICS