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Member KzzmTbft or Kzzmc00t
(( σ s
(g,g’),g’=1,NGT),g=1,NGT)
( σ up (g),g=1,NGT)
( σ c
(g),g=1,NGT)
( σ t
(g),g=1,NGT)
( σ f (g),g=1,NGT)
(ν (g),g=1,NGT)
P 0 scattering matrix
σ if INT(2)>0
s, g→g
'
Note: Some nuclides may not have thermal scattering data.
Up-scattering cross-sections
Capture cross-sections
Total cross-sections
Fission cross-sections
Neutron yields per fission in thermal range
/NGT*(NGT+5) or /5*NGT/
Member mmmmBMIC /MAXNG*6*MMK/
((( σ eff (g,x,i),g=1,MAXNG),x=1,6),i=1,MMK)
where σ eff (g,x,i) is the effective microscopic cross-sections of energy group g,
reaction x, nuclide i.
MAXNG is the maximum number of groups (=107) fixed by the parameter in a
include file of the SRAC sources. If g-value exceeds the number of fine groups,
null values are filled.
The reaction types are σ c by x=1, σ f by x=2, σ e (elastic) by x=3, σ er (elastic
removal) by x=4, σ n,2n by x=5 and νσ f by x=6.
MMK is the number of the whole nuclides contained in the material if IC20=0 (cf.
Sect.2.2) is specified. Otherwise (burn-up calculation), MMK is the number of
whole depleting nuclides implicitly determined by the burn-up chain model even at
the zero burn-up step. As this member has no burn-up tag, the cross-section values
are updated by those at the current burn-up step.
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