Fixed bugs in tallying total source in openmc
Fixed bugs in tallying total source in openmc

Two bugs are: 1) openmc_total_src needs to be reset to zero every time
compute_xs() is called. The tallies are cummulative themselves, and if
we don't reset openmc_total_src at every batch, the openmc_total_src
term is essentially accumulating accumulative tallies. 2)
openmc_total_src should be the tally for the scattering reaction rate
+ the tally for the fission reaction rate dividied by keff.

lilulu/openmc,lilulu/openmc,lilulu/openmc