This is correct that F5 tally is score of born and scattered neutrons. But scattering is happened in moderator only. Try to set a zero importance to a moderator. 

This is about the expectations from output file. By using F5 tally, you can not obtain uncollided photons. For this aim, try to use F4 tally and carefully readt the collided and uncollided particle amounts from output file. Otherwise, you can observe it.

This template shows how to use an FM multiplier to get a spectrum averaged total cross section.  The RXN number for neutron total cross section is -1 (-5 for photons). [Formatting is off, but copy and paste content below "*'s" to text editor.]

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Determine Spectrum Average Cross Section for Carbon

C

C  Neutron Template

C

C        1         2         3         4         5         6         7         8

C ******************************************************************************

C *                               CELL CARDS                                   *

C ******************************************************************************

C

1     0   -1     IMP:N=1

2     0    1     IMP:N=0

C

C        1         2         3         4         5         6         7         8

C ******************************************************************************

C *                              SURFACE CARDS                                 *

C ******************************************************************************

C

1     so   0.282094791     $Sphere with surface area of 1 cm^2

C        1         2         3         4         5         6         7         8

C ******************************************************************************

C *                               DATA CARDS                                   *

C ******************************************************************************

C

MODE N

RAND  GEN=2 SEED=19073486328125 STRIDE=152917

NPS   2e7

C

C ********************

C *  SOURCE CARDS    *

C ********************

C

SDEF  ERG=D2 PAR=1

C

SI2   E low  E high   $place energy group boundaries here

C

SP2   0.0000000 1.0000000

C

C

C ********************

C *  MATERIAL CARDS  *

C ********************

C

C   Carbon from Nuclear Wallet Cards - Electronic Version                 

C    Isotope   atomic abundance    delta (MeV)                            

C     C-12        98.93%              0.0000                              

C     C-13         1.07%              3.1250                              

C                                                                         

C   Atomic Weight:  12.011 g/mol                                          

C         Density:  1.8 to 2.1 g/cc (1.9 to 2.3 g/cc for graphite)        

C                  - Will use 2.1 (midpoint for graphite)                 

C

C   Summary of MatMCNP (Version 3.0) Calculations:

C

C  Isotope  Number Fraction    Weight Fraction       Atoms/b-cm

C   C-12        0.989300           0.988416          0.1041666

C   C-13        0.010700           0.011584          0.0011266

C

C  The total compound atom density (atom/b-cm):  0.1052933

C

C

C  This material contains an isotope that is often modified by

C  an S(alpha,beta). Check MCNP Manual Appendix G to see if an

C  S(alpha,beta) card (i.e., an MTn card) is required.

C

M1      06000.80c   1.000000

C

C  Caution: The natural zaid is used for Carbon.

C

C  To convert a particle flux to rad[Material]

C  use FM  8.0323725E-10 6    -4  1 for neutrons

C   or FM  8.0323725E-10 6    -5 -6 for photons.

C

C

C ********************

C * NEUTRON TALLIES  *

C ********************

C

FQ0   E  F

C

C

FC2    Neutron Fluence in Each Group  $Used to normalize bin values for xsec

F2:N   1

C

FC12   Spectrum Averaged Total Cross Section for the Group

F12:N  1

FM12   1.0  1  -1 

C 

C  Total fluence should be 1.0 n/cm**2.

C  A "C" value of 1.0 means that the FM card is a microscropic cross section.

C  The "-1" is the total neutron cross section MT value.

C

C

Now that I have read the document you attached, I will suggest that you include the proper S(alpha, beta) treatment for materials like water and graphite. Those treatments modify the thermal neutron scattering cross sections of materials that moderate neutrons. As mentioned in the comments to the carbon material card, those were found in Appendix G of the MCNP5 manual. I am not certain where they are found in the latest (MCNP6 1.1) version of the code.

F4