1. NAME AND TITLE
UNIFY-ECN: A Program to Calculate Fast Neutron Data for Structural Materials.
Chinese Nuclear Data Center, Beijing, China.
3. CODING LANGUAGE AND COMPUTER
Fortran 77; CDC Cyber 825/170.
4. NATURE OF PROBLEM SOLVED
Based on the unified model the UNIFY code is used for the calculation of the fast neutron data for
structural materials, which involves: (1) cross section- total cross section, all kinds of reactions
channels, the cross section of the discrete levels and continuum emission, (2) angular distribution-
elastic scattering angular distribution and its Legendre coefficients and transition matrix elements,the
Legendre coefficients of the discrete levels in the inelastic scattering channels, (3) energy spectra, (4)
double differential cross section of the inelastic channel and of the neutron outgoing channels.
5. METHOD OF SOLUTION
Gaussian integration is used in the program for all kinds of numerical integrations.
6. RESTRICTIONS OR LIMITATIONS
The UNIFY code can be used for the incident neutron energies from 1 KeV to 20 MeV. The
numbers of the incident energies could not exceed 50.
7. TYPICAL RUNNING TIME
For En=14 MeV, 200 seconds of CPU computer time is needed, while En=20 MeV requires 500
seconds. But for lower energies it is very fast.
8. COMPUTER HARDWARE REQUIREMENTS
UNIFY-ECN runs on the CDC Cyber 825/170.
9. COMPUTER SOFTWARE REQUIREMENTS
The code was written in Fortran 77 and requires the ICSSCU subroutine from the Internal
Mathematics and Statistics Library (IMSL). ICSSCU is a subroutine whose function is to fit cubic
Shi Xiangjun and Zhang Jingshang, "Description of UNIFY-ECN Code and Guide For Users,"
informal document, Chinese Nuclear Data Center (1990).
11. CONTENTS OF CODE PACKAGE
Included are the referenced document and one DS/DD (360 K) 5.25 inch diskette.
12. DATE OF ABSTRACT
KEYWORDS: CROSS SECTION PROCESSING; NEUTRON; PARAMETRIC MODELS