1. NAME AND TITLE OF DATA LIBRARY
MGCLIB: 137 and 26 Neutron Multigroup Cross Section Library with the Bondarenko Type Shielding Table.
2. NAME AND TITLE OF DATA RETRIEVAL PROGRAMS
JCTSCON: Reaction Cross Section Card-Image-to-Binary Conversion Program.
SMFCON: Scattering Transfer Cross Section Card-Image-to-Binary Conversion Program.
MAIL: Effective Macroscopic Cross Section Generator.
REMAIL: ANISN/KENO Library Converter.
Japan Atomic Energy Research Institute (JAERI), Tokai-mura, Naka-gun, Ibaraki-ken, Japan.
4. HISTORICAL BACKGROUND AND INFORMATION
The MGCLIB and the retrieval programs listed above are part of the Japan Atomic Energy Research Institutes (JAERI) MGCL-PROCESSOR, that is a subsystem of the criticality safety evaluation code system JACS, which includes the processing system MGCL-ACE (see 6.).
5. APPLICATION OF THE DATA
The basic function of MGCLIB is to generate effective neutron cross section sets in either 137 or 26 group structures for use in the discrete ordinates codes ANISN-JR or DOT 3.5 or in the Monte Carlo codes KENO-IV or MULTI-KENO.
6. SOURCE AND SCOPE OF DATA
The data in MGCLIB were generated from evaluated cross sections from ENDF/B-IV and JENDL first using RESEND-D to generate point cross sections and then collapsing to 137 groups. Two different models were used to produce data needed to calculate effective cross sections. The resulting data include infinite dilution cross sections, resonance self-shielding factors, resonance material-shielding factors, and moderator mass-effect tables. The scattering matrix part was generated using SUPERTOG and FLANGE or PIXSE. Up scatter is treated for energy groups below 1.855cV.
A 26 group library, which was collapsed from the 137 group data, is also provided with MGCLIB. The 26 group library contains data for 90 materials; however the 137 group library provided herein contains data for only 67 materials.
The data are divided into two parts. The JCTS part (one-dimensional data) and the SMF part (two-dimensional data), the latter representing the scattering matrix.
7. DISCUSSION OF THE DATA RETRIEVAL PROGRAM
The MGCLIB data are packaged in card image form and must first be converted to binary. JCTSCON and SMFCON are provided for this purpose (each part is converted in a separate run). In order to make an effective cross section library for use in a discrete ordinates code, the converted JCTS and SMF libraries are processed in a single run of MAIL. A KENO library can be converted from the ANISN library using REMAIL.
8. DATA FORMAT AND COMPUTER
Card images; FACOM-380 (D00118FM38000).
9. TYPICAL RUNNING TIME
About 5 sec CPU to generate effective macroscopic cross sections in a material region with 10 nuclides by MAIL. Required running time for REMAIL is negligible.
a. Included in the documentation:
Yoshitaka Naito, Shin-ichiro Tsuruta, Tetsuo Matsumura, and Tamotsu Ohuchi, "MGCL-PROCESSOR: A Computer Code System for Processing Multigroup Constants Library MGCL," JAERI-M 9396 (March 1981).
b. Background information:
L. M. Pertie and N. F. Cross, "KENO-IV An Improved Monte Carlo Criticality Program," ORNL-4938 (1975).
K. Koyama et al., "ANISN-JR, A One-Dimensional Discrete Ordinates Code for Neutron and Gamma-Ray Transport Calculations," JAERI-M 6954 (1977).
W. A. Rhoades, "DOT 3.5 Two Dimensional Discrete Ordinates Radiation Transport Code," CCC-276 (1977).
ENDF/B Summary Document BNL-NCS-17541 2nd ed. (1975).
I. I. Bondarenko, ed., "Group Constants for Nuclear Reactor Calculations," Consultants Bureau, New York (1964).
O. Ozer, "Program RESEND," BNL-17134 (1972).
R. Q. Wright et al., "SUPERTOG: A Program to Generate Group Constants and Pl Scattering Matrices from ENDF/B," ORNL-TM-2679 (1969).
11. CONTENTS OF THE LIBRARY
Included are the referenced document in (10.a) and one DC6150 cartridge tape in TAR format, which contains 26 and 137 group JCTS and SMF data files, the retrieval programs listed in 2, and sample input and output from running a 26 group library case through the retrieval programs.
12. DATE OF ABSTRACT
KEYWORDS: ANISN FORMAT; MULTIGROUP CROSS SECTIONS BASED ON ENDF/B; NEUTRON CROSS SECTIONS