Monte Carlo based absolute efficiency calibration of power reactor spent fuel NDA measurements

Experiments have been carried out at Paks Nuclear Power Plant for the purpose of spent fuel burnup characterization using gamma-ray spectrometry. Obtaining absolute quantitative information, e.g. fission product activities, from the spectra requires accurate knowledge of the detection efficiency. Due to nature of the measurement conditions, experimental calibration was unfeasible, and the Monte Carlo particle transport method was selected to calculate the efficiency. The model building process is presented, which involves X-ray radiography of the germanium detector, optimization of the dead layer on the outside of the crystal, and a solution to the challenging low-efficiency simulation problem. The simulations were validated using measurements carried out at the Paks reactor units together with verified experimental data from the SFCOMPO library. The constructed model proved to be accurate to the results determined by empirical methods within +/- 3 % error, promising to be a reliable basis for future applications that require spent fuel characterization.