**RSIC CODE PACKAGE CCC-084**

**1. NAME AND TITLE**

SHADRAC: Kernel Integration Code Shield
Heating and Dose Rate Calculation in Complex Geometry.

The code is a complete rewrite of machine
language computer codes CCC-5/C-17 and CCC-6/L-63, now obsolete.

**2. CONTRIBUTORS**

USAF Nuclear Aerospace Research Facility,
General Dynamics,

Air Force Weapons Laboratory, Kirtland Air
Force Base,

**3. CODING LANGUAGE AND COMPUTER**

FORTRAN IV; IBM 7090 and 7094.

**4. NATURE OF PROBLEM SOLVED**

SHADRAC calculates the neutron and/or
gamma-ray spectra, heat generation rate, and/or dose rate at each of a group of
point detectors, due to each of a group of point sources. The sources may be
divided into sets, with each set having a unique source spectra. The spectrum,
heating rate, and/or dose rate for each detector, summed over each source-point
set and over the entire source group, may also be computed. Complex geometry
may be treated.

**5. METHOD OF SOLUTION**

Point-to-point kernels, based upon the
differential energy spectra for a point isotropic source in an infinite medium,
are integrated over various sources. The data used is based on the
moments-method solution of the fast-neutron or gamma-ray transport equation.
The stepping-point method is used to solve for the path lengths from source to
detector in each region.

The gamma-ray absorption coefficients are
based on interpolations of the photoelectric and pair production cross sections
so that the coefficients may be computed for all media of the system. The
effective atomic number is interpolated from a table of atomic numbers versus the
absorption coefficient per electron.

The mode of distributing the source points
is chosen (either equal interval or according to Gaussian quadrature abscissa)
which locates the coordinate planes that are perpendicular to the coordinate
axes. The intersections of these planes are source point locations.

Modifications in SHADRAC resulting in
improvement over earlier models are: direct computation of unscattered flux,
removal of all neutron energy modes except the first, greater capabilities in
the use of the source spectra, library tape storage of material data, coding of
gamma-ray data into the program, and improvement of the output format.

**6. RESTRICTIONS OR LIMITATIONS**

Enough physical and source description
capability is provided by the program so that there should be little
uncertainty except that associated with the point-to-point kernels as applied
to specific geometries. Inhomogeneities generally increase the error since it
is necessary to use arbitrary prescriptions for combining homogeneous media
data.

**7. TYPICAL RUNNING TIME**

No study has been made at this time to
determine typical running time. Estimated running time of sample problem on the
IBM 7090: 0.05 hour.

**8. COMPUTER HARDWARE REQUIREMENTS**

IBM 7090 and 7094 32 K computer with 7 tape
units.

**9. COMPUTER SOFTWARE REQUIREMENTS**

The code was designed for and is operable in
the IBM FORTRAN IV Operating System, using the ALTIO package, and is compatible
with the CDC 6600 FORTRAN IV system. Four pool tapes, in addition to standard
input, output, and systems tapes, are assigned. Allowance has been made for a
library to be read in from tape. It is currently being read as standard input
data.

**10. REFERENCE**

J. A. Moore, J. B. Eggen, C. W. Austin, D.
H. Huckaby, and R. A. Miller, "Shield Heating and Dose Rate Attenuation
Calculation (SHADRAC)," NARF-DC-MEMO 1.097 (March 1966).

**11. CONTENTS OF CODE PACKAGE**

Included are the referenced document and one
(1.2MB) DOS diskette which contains the source code, a library of cross
sections and input and output for a sample problem.

**12. DATE OF ABSTRACT**

January 1968; updated July 1981, February
1985.

**KEYWORDS: **KERNEL; HEATING; ENVIRONMENTAL DOSE; COMPLEX GEOMETRY