**1. NAME AND TITLE**

MINX: Multigroup Interpretation of Nuclear X-Sections from ENDF/B Standard CCCC-III
Interface Formats.

**AUXILIARY ROUTINES**

IDAY, ICLOCK, TIME: Non-standard user routines.

**2. CONTRIBUTOR**

Los Alamos National Laboratory, Los Alamos, New Mexico.

**3. CODING LANGUAGE AND COMPUTER**

Fortran IV and Assembler Language; IBM 360/370 (A), CDC-7600 (B).

**4. NATURE OF PROBLEM SOLVED**

MINX calculates fine-group averaged infinitely dilute cross sections, self-shielding factors, and
group-to-group transfer matrices from ENDF/B-IV data. Its primary purpose is to generate pseudo-composition-independent multigroup libraries in the standard CCCC-III interface formats for use in
the design and analysis of nuclear systems. MINX incorporates and improves upon the resonance
capabilities of existing codes such as ETOX and ENDFUN and the high-Legendre-order transfer
matrices of ETOG and PSR-13/SUPERTOG. Group structure, Legendre order, weight function,
temperature, dilutions, and processing tolerances are all under user control. Paging and variable
dimensioning allow very large problems to be run.

**5. METHOD OF SOLUTION**

Infinitely dilute, unbroadened point cross sections are obtained from resolved resonance parameters
using a modified version of the RESEND program. The SIGMA1 kernel-broadening method is used
to Doppler broaden and thin the tabulated linearized pointwise cross sections at 0^{o}K (outside of the
unresolved energy region). Effective temperature-dependent self-shielded pointwise cross sections are
derived from the formulation in the ETOX code. The primary modification to the ETOX algorithm
is associated with the numerical quadrature scheme used to establish the mean values of the fluctuation
intervals. The selection of energy mesh points, at which the effective cross sections are calculated,
has been modified to include the energy points given in the ENDF/B file or, if the energy-independent
formalism was employed, points at half-lethargy intervals. Infinitely dilute group cross sections and
self-shielding factors are generated using the Bondarenko flux weighting model with the gross spectral
shape under user control. The integral over energy for each group is divided into a set of panels
defined by the union of the grid points describing the total cross section, the reaction cross section of
interest and the gross spectral shape. The integration scheme actually employed in MINX is adaptive
Simpson's procedure for which the initial estimate is based on the unionized grid described above.
The computation of elastic and discrete group-to-group matrices is based upon a semi-analytic scheme
which treats the rapidly fluctuating cross-section behavior analytically. Where this laboratory-system-based scheme becomes difficult to implement (e.g., light nuclei, inelastic thresholds), an alternative
numerical integration in the center-of-mass system is employed. Multigroup transfer matrices for
processes in which the outgoing neutron energy and angular distribution is uncoupled are computed
by direct numerical integration.

**6. RESTRICTIONS OR LIMITATIONS**

The principal restriction is the computing time available for a given desired accuracy. The paging
techniques and variable dimensioning make efficient use of available core storage; however, the
extensive use of BCD files can make excessive demands on the number of I/O transfers required.

**7. TYPICAL RUNNING TIME**

Typical running times are difficult to define because they are sensitive functions of the a) accuracy
required, b) number of resonances, c) number of groups, d) Legendre expansion order, e) number of
temperature and dilutions, etc.

**8. COMPUTER HARDWARE REQUIREMENTS**

MINX is operable on the IBM 360/370 computers (A) or the CDC 7600 computers (B).

**9. COMPUTER SOFTWARE REQUIREMENTS**

Fortran IV and Assembler Language compilers are required.

**10. REFERENCE**

C. R. Weisbin, P. D. Soran, R. E. MacFarlane, D. R. Harris, R. J. LaBauve, J. S. Hendricks,
J. E. White and R. B. Kidman, "MINX: A Multigroup Interpretation of Nuclear X-Sections from
ENDF/B," LA-6486-MS (September 1976).

**11. CONTENTS OF CODE PACKAGE**

Included are the referenced document and one (1.2MB) DOS diskette which contains the source
code and sample problem input and output, the auxiliary routines.

**12. DATE OF ABSTRACT**

February 1977; revised and updated March 1984.

**KEYWORDS: ** ENDF FORMAT; MULTIGROUP CROSS SECTION PROCESSING;
NEUTRON CROSS SECTION PROCESSING