The fine group libraries with the energy structures listed in Table 8.1 must be produced and regularly updated starting from cross-section sets of the type of ENDF/B (see § 3.4).

Processing codes are used in order to generate from ENDF/B a library usable in codes for spectrum calculations.

The fine group cross-sections must be obtained from the point values of ENDF/B assuming a certain form of the flux within these fine groups (e. g. constant flux per unit lethargy in the fast range).

The thermal neutron scattering law data are represented in ENDF/B for each moderating molecule n, by a scattering law S„(a, /8, T) where /3 is proportional to the energy change, a is related to the momentum change, and T is the temperature in °K.

The differential scattering cross-section is given by:l23>

<т(Е -> E’, cos 0o, T) = 2 VI e"P/2S"(a’ & T)

where there are (NS + 1) types of atoms in the molecule (i. e. for H20, NS = 1), and

M„ = number of atoms of the nth type in the molecule,

P=(E’-E)lkT,

a = (E’ + E — 2 cos 0oVEE’)/Ao/cT,

An = mass of the nth type atom, A0 is the mass of the principal scattering atom in the molecule,

(Thn = bound atom scattering cross-section of the nth type atom,

Only the data for the principal scatterer of the molecule (e. g. H in H20), S0(a, /З, T) are tabulated in ENDF/B. The scattering properties for the other atom types (n = 1,2,…, NS) are represented by analytic functions.

Supplementary codes (e. g. FLANGE’24’) are used to produce the actual scattering kernels. This approach gives the possibility of obtaining scattering kernels for different energy group structures having stored the scattering law S„(a, /З, T). Various interpola­tion schemes are given to interpolate between the values of a, /3 and T. This scattering law is obtained starting from the phonon spectra of the scatterer under considera — tion.’2526’ (See also § 6.3.)

As the data stored in the transfer matrices have often only rather smooth variations it is possible in some processing codes to reduce the computer store requirement by means of polynomial fittings.

For nuclides like structural materials and fission products the resonance calculation is usually done by the processing codes assuming zero temperature and concentration.

For fuel and fertile materials this calculation is done later in the spectrum codes.

As an example of a processing code we will quote SUPERTOG<27) which averages the ENDF/B data over specified group widths. The flux per unit lethargy is assumed to be constant unless suitable weight functions are supplied by the user. When resonance data are available resonance contributions are calculated using Breit-Wigner expres­sions. The point cross-sections are integrated in order to obtain smooth cross-sections. Elastic scattering matrices are computed from Legendre coefficients of the scattering angular-distribution data. Inelastic scattering and (n, 2n) matrices are computed considering the excitation levels of the considered nuclides. This code can be used to produce libraries for codes for spectrum calculation or for other transport codes.

Other processing codes are described in refs. 28 to 32.