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= 0 fission (Σ x, g =Σ f, g )
= 1 capture (Σ x, g = Σ c, g )
= 2 absorption (fission + capture)
3 NMESH Number of spatial points (mesh intervals) to calculate the reaction rates
Block-3-2 Required if IOPT(2)>0 /3*NMESH/
((MESH(i,n),i=1,3), n=1,NMESH)
Spatial position by mesh number, where i=1 corresponds to X-direction, i=2 to
Y-direction and i=3 to Z-direction. In one- or two-dimensional case, enter zero
values into irrelevant data. If the output of the PIJ is processed, specify R-Region
number as if one-dimensional calculation.
Block-3-3 Required if IOPT(2)>0 /IGMAX/
(FG(g),g=1,IGMAX)
Transmission (used in Eq.(2.10-2)) has to range between 0.0 and 1.0.
IGMAX is number of energy groups (IGMAX=NEF+NET for fine group
calculation and IGMAX=NERF+NERT for few group one, cf. Sect.2.2)
Note:
If the exact values are required around the cut-off energy, print out the collision
probabilities calculated in an isolated cell composed of a detector covered by a
filter. The probability P is is the probability that a neutron born in region i escapes
from the outer surface of the filter. Then, calculate G i by G i =4V i /S*Σ i P is where G i
= G d (i is detector region) the probability that a neutron impinging into filtered
detector from the outer surface with isotropic angular distribution.
Repeat Block-3.1 through Block 3-3 IOPT(2) times.
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Block-4-1 Input for spectrum index, required if IOPT(3)>0 /6, IGMAX/
1 MPOS Material number in which the nuclides for spectrum index calculation are involved.
Material number is the order of appearance in the material specification in Sect.2.9.
2 U235 The position of the first nuclide in the material. Count the position in the order of
material specification.
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