JAEA-Data/Code 2007-004 - Welcome to Research Group for ...

Block-8-1 Required if IBC6=1 /1/
NMICR
Number of nuclides of which effective cross-sections are written on MICREF file
Block-8-2 Required if IBC6=1 /NMICR/, (18A4)
NAMMIC Names of nuclides of which effective cross-sections are written on MICREF file.
Enter by 3 characters and one arbitrary character. The microscopic cross-sections
of the specified nuclide are written for every material in which the nuclide is
contained.
Sample input: U08_B00_PU9_ (the fourth character is ineffective)
Block-9 Required if IBC10=1 /NMAT/
(IBTYPE(i), i=1,NMAT)
= 0 Non-depleting material
= 1 Depleting material
Enter IBTYPE(i) to material i to specify if it is depleting or not. NMAT is number
of materials, and material is numbered in the order appearing in the material
specifications. (cf. Sect.2.9). Specify IBTYPE(i)=0 to a material composed of the
nuclides which do not appear in the selected burn-up chain model.
Block-10 Required if IBC3=±3 and no fissionable nuclide is contained in any material /1/
FLXLVL When the system does not contain any fissionable nuclide but depleting nuclide
like burnable poison, depletion calculation is available under fixed flux condition
by specifying IBC3=±3. Enter average flux level (n/cm 2 /s) to the burnable region.
Unit of burn-up (MWd/t, MWd) is replaced by the integrated absorption reaction
rate. The use of this option requires to specify IBC2=±3 or =±4.
Block-11-1 Required if IBC9>0 /A4/
NAMFP Name of nuclide of which number densities are specified by input values.
Sample: XE5_ (the first 3 characters are effective)
Block-11-2-1 Required if IBC9>0 /1/
MPOS = 0 End of Block-11-2 entries
> 0 Material number for which the atomic number density of the nuclide is
replaced. Material is numbered in the order appearing in the material
specification (Sect.2.9).
Block-11-2-2 Required if IBC9>0 and MPOS>0 /IBC1+1/
143