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group flux and for the fast group flux should be different when both have to be solved simultaneously
in the whole energy range.
Our scenario to solve the eigenvalue problem in the whole energy range in the cell calculation is
a combination of the procedures described in the previous two sections, as follows;
Step 1. Set an initial guess of fission distribution and normalize it.
Step 2. Starting at the highest energy group g=1, calculate the emission rate of a group H ig by
Eq.(7.4.3-1).
Calculate fluxes of a group by a matrix inversion.
Repeat Step 2 for all fast groups.
Step 3. Calculate the slowing-down source to the thermal groups.
Step 4. Calculate the thermal fluxes by repeating the thermal iteration as described in Sect.7.4.1 until
the convergence is attained or the fixed iteration count is reached.
Step 5. Calculate and re-normalize the fission distribution using the new flux distribution and
modify the distribution by an SOR.
Repeat Step 2 through Step 5 until the fission distribution converges. The eigenvalue is obtained as the
re-normalization factor of fission distribution calculated in Step 5 at the final power iteration.
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