c homogenous core model
c
c single cell for cross section generation
c
c Core specifications
c
c U15N fueled core
c Nb cladding
c Sodium coolant
c
c conditions
c assumed temperature of 1000 K
c
c N(UN) = 6.2e-02 1/b-cm
c N(Nb) = 8.4 g/cc = 5.4e-02 1/b-cm
c N(Na) = 0.5 g/cc = 1.3e-02 1/b-cm
c
c volume fractions in core
c
c fuel       50%
c structure  10%
c coolant    40%
c
c smear density
c
c Ntot = 2(6.2e-2)(0.5) + (5.4e-2)(0.10) + (1.3e-2)(0.40)
c Ntot = 7.26e-02 1/b-cm
c
c
c fuel composition
c 25/(25+28) enrichment = 0.90
c 
c Ns(25)  = 0.0279 1/b-cm
c Ns(28)  = 0.0031 1/b-cm
c Ns(15N) = 0.0310 1/b-cm
c Ns(Nb)  = 0.0054 1/b-cm
c Ns(Na)  = 0.0052 1/b-cm 
c 
c
c
c
c cell cards
c
c
1   1   7.26e-02 -10 -22 +23   tmp=1.010e-7 imp:n=1 $ core region
999 0   (+10:+22:-23)                       imp:n=0 $ void
c
c
c end cell cards

c start surface cards
c
c
*10     cz    +15
*22     pz    +15
*23     pz    -15
c
c end of surface cards

c start of material cards
c 
m1      07015.10c   0.0310
        92235.10c   0.0279
        92238.10c   0.0031
        11023.10c   0.0052
        41093.10c   0.0054
m238    92238.10c   1.0
m235    92235.10c   1.0
c
c end of material cards
c
c
c beginning of data cards
c
ksrc 0 0 0
mode n
kcode 1000 1.0 200 800
prdmp 800 800 800
print
c
c
c TALLY CARDS
c
c
c calculate the flux in the core for normalization
c
F14:N      1
c
c calculate the integrated cross section for the homogenous material
c  2 : scattering
c -6 : fission
c -2 : absorption - fission
c -7 : prompt fission neutron yield
c -8 : energy released from fission
c
c
F24:N      1
FM24       (1 1 (2) (-6) (-2) (-7) (-8))
c
c calculate spectrum averaged cross sections for important actinides
c purpose to calculate the breeding ratio and fertile fission fraction
c 235U and 238U are of importance 
c 102 : radiative capture
c
F34:N      1
FM34       (1 238 (102) (-6))
           (1 235 (-6))