The Evaluation of Neutron Exposure Using Ex-vessel Neutron Dosimetry in a Commercial Nuclear Reactor Vessel

An accurate evaluation of the neutron exposure to the materials comprising the beltline region of the reactor vessel is performed through a neutron dosimetry analysis where a fluence calculation is done by both measurement of the dosimeter materials removed from surveillance capsules and a neutron transport calculation. Once all the capsules have been completely removed from the reactor vessel in a typical Korean PWR, ex-vessel sensor sets are installed as a substitute capsule in an axial direction in the reactor cavity and then removed for an analysis. All transport calculations were carried out by using the DORT 3.1 discrete ordinate code and the BUGLE-96 cross-section library which provides 67 group coupled neutron-gamma ray cross-section data set. In order to determine the calculated reaction rates for each dosimeter sensor, the dosimeter cross-section library, SNLRML, was used. The results that have been performed on one of our reactors during the consecutive fuel cycles showed that the differences between the measurements and calculations by the ex-vessel neutron dosimetry sensors are less than 20% for each capsule, which satisfy the acceptable criterion.