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 FENDL/MG description

The multigroup neutron, photon-atom and coupled neutron-photon cross section data were generated with the nuclear data processing code NJOY94.

The multigroup data are available in two formats:

  • GENDF - output of the GROUPR module (neutron and coupled neutron-photon cross section data only).
  • MATXS - for use in coupled neutron-photon transport calculations with discrete-ordinates codes like anisn, one- and two-dant. These MATXS files were generated by post-processing the above mentioned GENDF files with the MATXSR module of NJOY, along with the GENDF files of photon-atom interaction data. The latter were generated with the NJOY modules RECONR/GAMINR from the evaluated photon-atom cross section data file

A file named 'NJOY-INPUT.DAT' contains the NJOY input files for processing the data into multigroup structure.

Specifications for the multigroup processing to produce GENDF and MATXS files:

- Neutron groups: 175, in Vitamin-J structure
- Gamma groups: 42, in Vitamin_J structure
- Vitamin-E neutron weight function (IWT=11 in NJOY)
- Gamma weight function: 1/E with roll-offs (IWT=3 in NJOY)
- Legendre order for neutron and photon scattering: P6 for transport correction to P5
- Temperature: 300 Kelvin ( plus 600, 900 and 1200 K for Ga-nat and -JFF data)
- Dilution factors: (see Table 1)
- Up to 7 digits of accuracy for resonance reconstruction.
- Reactions included:
All reactions present in evaluations
Energy balance heating (MT = 301)
Kinematic heating (Mt = 443)
Damage (MT = 444)
Thermal data only for H-2, Be-9, C-12, N-14, O-16, Al-27, V-51, Fe-56, Zr-nat, Ga-nat, Nb-93, Mo-nat, Sn-nat, W-nat, Au-197
Gas production data (see Table below)

Status of the multigroup files for different dilution factors ( in barns).

Nuclide 10**10 10**5 10000 1000 300 100 30 10 3 1 0.3 0.1 .001
H-1 X                        
H-2 X   X X   X   X   X      
H-3 X                        
He-3 X                        
He-4 X                        
Li-6 X                        
Li-7 X                        
Be-9 X   X X X X X X   X   X X
B-10 X                        
B-11 X                        
C-12 X   X X X X X X   X   X X
N-14 X   X X X X X X   X   X X
N-15 X                        
O-16 X   X X X X X X   X   X X
F-19 X                        
Na-23 X                        
Mg-Nat X                        
Al-27 X                        
Si-28 X         X   X   X      
Si-29 X         X   X   X      
Si-30 X         X   X   X      
P-31 X                        
S-Nat X     X X X X X          
Cl-Nat X     X X X X X          
K-Nat X     X X X X X          
Ca-Nat X     X X X X X X X      
Ti-Nat X     X X X X X X X      
V-51 X   X X X X X X   X   X X
Cr-50 X     X X X X X X X      
Cr-52 X     X X X X X X X      
Cr-53 X     X X X X X X X      
Cr-54 X     X X X X X X X      
Mn-55 X X   X   X   X   X      
Fe-54 X X X X   X   X          
Fe-56 X X X X   X   X X X X X  
Fe-57 X X X X   X   X          
Fe-58 X X X X   X   X          
Co-59 X X X X   X   X          
Ni-58 X     X X X X X X X      
Ni-60 X     X X X X X X X      
Ni-61 X     X X X X X X X      
Ni-62 X     X X X X X X X      
Ni-64 X     X X X X X X X      
Cu-63 X   X   X X X X          
Cu-65 X   X   X X X X          
Ga-nat X   X X X X X X   X   X X
Zr-nat X   X X X X X X   X   X X
Nb-93 X   X X X X X X   X   X X
Mo-Nat X   X X X X X X   X   X X
Sn-Nat X   X X   X   X   X      
Ta-181 X   X X   X   X          
W-nat X   X X X X X X   X   X X
Au-197 X   X X X X X X   X      
Pb-206 X     X   X   X   X      
Pb-207 X     X   X   X   X      
Pb-208 X     X   X   X   X      
Bi-209 X   X X   X   X          

Status of gas production cross sections present in the multigroup libraries.

Nuclide MT203 MT204 MT205 MT206 MT207
H-1 H-2 H-3 He-3 He-4
H-2 X   X    
Be-9         X
C-12         X
N-14 X X X   X
O-16 X X     X
Al-27 X       X
Si-28 X X     X
Si-29 X       X
Si-30 X       X
V-51 X X X   X
Fe-56 X X X X X
Ga-nat X X X X X
Zr-nat X X X X X
Nb-93 X X     X
Mo-nat X X X X X
Sn-nat X X X X X
W-nat X X     X
Au-197 X       X

 
 
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