Adjoint Monte Carlo neutron transport using cross-section probability table representation

The probability table representation of cross-sections is generally used to deal with neutron interactions in the unresolved energy range. In the frame of neutron transport methods, the capability of the probability table representation of cross-sections on the whole neutron energy range has been mentioned by Cullen (1974) and it has been already demonstrated for the Monte Carlo transport calculations by Zheng et al. (1998). Such an advantage is also illustrated here with a simple neutron propagation configuration dealt with the TRIPOLI-4 Monte Carlo transport code.This article gives a new expression of the integral operator kernels for adjoint Monte Carlo neutron multigroup transport including the probability table representation of cross-sections. This formalism is applied to the adjoint two energy group neutron transport in an infinite homogeneous medium.Therefore, the same physical advantage as in forward neutron transport should be expected for adjoint neutron transport.