Development and Verification of Nuclear Calculation Methods for ...

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3. 2. The CROSS Data Processing Programme
Two nearly identical versions of the CROSS programme exist. The
CROSS MACRO version provides arbitrary energy-condensed multi-group,
macroscopic and homogenized cross sections for later application in overall
reactor investigations. The CROSS MICRO version determines optional
energy-condensed multi-group microscopic cross sections for later use in
unit cell calculations.
The CROSS programme is written in FORTRAN 4, and the overlay
structure is shown in fig. 3.
3.2.1. Macroscopic. Homogenized Cross Sections
The detailed space-dependent flux spectrum within the fuel cell determined
by means of LA3ER is at each time step applied for space and
arbitrary energy condensation to provide macroscopic, homogenized cross
sections in less than 85 groups for the unit cell as a function of irradiation.
The slowing-down calculation in LASER is based on the slowing-down
density concept, and the elastic transfer cross sections between the groups
therefore do not occur explicitely in the programme. For 2 and 3 group
problems with the energy boundaries 0 - 1.855 eV - 5. 53 keV - 10 MeV the
transfer cross section from one group to the group just below it is obtained
from a neutron conservation principle. For multi-group problems the
CROSS MACRO programme calculates the detailed multi-group elastic
transfer cross sections by a conventional method (ref. 8).
In the thermal range the LASER programme provides free gas scattering
(Wigner-Wilkins) kernels (ref. 9) for all elements. For light water the
bound proton scattering kernel of Nelkin (ref. 10) and for heavy water the
Honeck extension (ref. 11) are furthermore provided.
In the fast and the thermal ranges the anisotropic scattering is accounted
for by transport-corrected self-scattering cuss sections
4' 6 = l8 ' g - silastic- < 3 -"
i
where E* ' * is the transfer cross section from group g to group g , and
^elastic is *° e total elastic scattering cross section in group g. j» 0 • -JJ- ,
where A is the atomic mass, is the average cosine of the scattering angle
per collision in the laboratory system.
For the light- and the heavy-water Nelkin kernels in the thermal range
the above transport correction is replaced by