Tally modifying of MCNP and post processing of pile-up simulation with time convolution method in PGNAA

Time convolution method has been employed for pile-up simulation in prompt gamma neutron activation analysis with an Am-Be neutron source and a 137Cs gamma source. A TALLYX subroutine has been written to design a new tally in the MCNP code. This tally records gamma particle information for the detector cell into an output file to be processed later. The times at which the particles are emitted by the source have been randomly generated following an exponential decay time distribution. A time convolution program was written to process the data produced and simulate more realistic pile-up. This method can be applied in optimization studies.