MCNP calculations of correction factors for radionuclide neutron source emission rate measurements using the manganese bath.

The manganese bath method for measuring the neutron emission rate of radionuclide sources requires corrections to be made for emitted neutrons which are not captured by manganese nuclei. The Monte Carlo particle transport code MCNP has been used to calculated the fractions of neutrons that escape from the boundaries of the tank, that are captured by the steel cavity in which the source is housed, and that are captured by fast neutron reactions in oxygen and sulphur in the solution. A comparison has been made with the calculations from in-house transport codes, and the measurements with the long counter, which highlighted a number of disagreements. These have been investigated and satisfactorily explained. Validation work has also been performed to demonstrate the reliability of the model. In future, the MCNP model will be used, in place of the in-house codes and the long counter, to calculate the correction factors.