**1. NAME AND TITLE**

TDOWN-IV: A Code System to Generate Composition- and Spatially-Dependent Neutron Cross
Sections for Multigroup Neutronics Analysis.

**2. CONTRIBUTOR**

The General Electric Company, Sunnyvale, California.

**3. CODING LANGUAGE AND COMPUTER**

Fortran-Y, GMAP; HONEYWELL 6600.

**4. NATURE OF PROBLEM SOLVED**

TDOWN-IV generates spatial- and composition-dependent neutron cross section sets for both core
and shielding analysis from a Bondarenko-type, generalized cross section library called GMUG, or,
alternatively, from the CCCC standard interface files ISOTXS and BRKOXS. Additional cross section
libraries may be output from TDOWN-IV by condensing the primary output library to fewer neutron
groups. Input to TDOWN-IV can be as simple as a description of material compositions with an input
flux spectrum or as complex as a description of a multiregional two-dimensional reactor or shield and
associated cells with several zero- or one-dimensional diffusion or transport theory flux solutions (k_{eff}
or fixed source) to provide the flux for spectral adjustments of the cross section sets.

**5. METHOD OF SOLUTION**

TDOWN-IV uses a rapid cross section adjustment procedure and a generalized library. For each
material, the library contains the group values of: infinitely-dilute cross sections; resonance
selfshielding factors which are tabulated for temperatures and a range of sigma_{o} - the material
composition-dependent parameters; P_{1}-scattering matrices; and, if the GMUG library is used, neutron
cross sections for photon production.

Selfshielding of a cross section is accomplished by computing sigma_{o} for either homogeneous
compositions or multizone cells and using interpolation routines to determine the selfshielding factor
for the sigma_{o} and input temperature. Selfshielding is an iterative procedure with the convergence
criteria specified in the input. Spectral adjustments to the elastic removal are carried out directly if
the flux sets are input. Zero- and one-dimensional flux solutions may be specified to generate the flux
sets. The flux solutions are iterated until spectral adjustments converge to the input criteria.

**6. RESTRICTIONS OR LIMITATIONS**

There is a maximum of 100 energy groups on the GMUG library. There is no group limit when
using the CCCC libraries. Output cross section sets from the compositions are not automatic; each
set is selected from particular compositions and the user then applies any spectral set to correct the
elastic removal. If the problem requires flux solutions and, consequently, regional descriptions,
compositions may be defined outside of the regions.

**7. TYPICAL RUNNING TIME**

Problems with all the flux spectra input use a maximum of a few minutes on the Honeywell 6000
computer. A 50-energy group problem generating 27 materials in 11 compositions with the flux
spectra computed by one transport theory cell and two one-dimensional diffusion theory flux solutions,
taking a total of 14 flux iterations, required .21 hours of computer processor time.

**8. COMPUTER HARDWARE REQUIREMENTS**

TDOWN-IV is operable on the Honeywell 6600 computers. It requires 27 K words of computer
memory for program storage; for data storage, the program acquires whatever it needs. In addition
to the card reader and printer devices, one tape handler, six random access scratch files on peripheral
devices such as discs, and a variable number of sequential files for file input and output are required.

**9. COMPUTER SOFTWARE REQUIREMENTS**

The Operating System is GECOS Software Release J with the optimized Fortran-Y compiler.

**10. REFERENCE**

R. Protsik, E. Kujawski, and C. L. Cowan, "TDOWN-IV - A Code to Generate Composition and
Spatially Dependent Neutron Cross Sections for Multigroup Neutronics Analysis," GEFR-00485
(September 1979).

**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.

**12. DATE OF ABSTRACT**

April 1984.

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