Radiological safety assessment of gamma spectroscopy based spent fuel burnup analysis at the ITU TRIGA Mark II...

This study investigates radiological safety and photon attenuation behavior during spent fuel inspection and burnup analysis at the ITU TRIGA Mark II Research Reactor. A lead-shielded fuel inspection system equipped with an HPGe detector was used to perform remote gamma spectroscopy on spent fuel elements, ensuring operator safety. Dose rate measurements were taken using a handheld radiation monitor at both the surface and at a distance of 1 m from the inspection system for each fuel element. The highest contact dose rate was observed for the B4 element (6.68 ± 0.60 μSv/h), while the lowest was for F30 (2.4 ± 0.31 μSv/h); all values remained within regulatory safety limits. Gamma spectra were analyzed to identify fission products and assess their contributions to the dose rates. Using buildup factors and attenuation coefficients, the unshielded dose rate at the fuel center was estimated. These findings confirm that the inspection system provides sufficient shielding and that radiological safety regulations are maintained during fuel handling operations.