Radiological characterization is important in decommissioning and dismantling of nuclear facilities, in order to assess the radioactivity concentration, classify the wastes, and secure workers’ safety. The Some components such as Reactor Pressure Vessel (RPV) in nuclear facilities has dose rate higher than Sv/hr, thus in-situ gamma spectroscopy systems suffer from a very high count rate which causes energy resolution degradation, photo-peak shift, and count loss by pile-up and dead-time. The system must be operated in a very high count rate, in order to measure spectra precisely and to quantify radionuclide contents. In order to apply in-situ measurement in high radiation dose rate environment, the sensor, front-end electronics, and data acquisition (DAQ) should be carefully selected and designed as well as precise design of collimators and radiation shield. In this paper, the components of the detector system were selected and performance was evaluated in a high count rate before design the collimator and shield. A LaBr3 coupled with a PMT having short decay time constant (16 nsec) was selected for high count rate application, and two different amplifiers (a conventional charge sensitive preamplifier with 50 usec decay time constant, and wide-band voltage amplifier) were tested. As DAQs, DT5781 (14 bit, 100 MS/s, CAEN) of Pulse Height Analysis (PHA) which is conventionally used signal processing method in the gamma spectroscopy, and DT5730 (14 bit, 500MS/s, CAEN) of Pulse Shape Discrimination (PSD) which is similar to Charge to Digital Convertor (QDC) were used. The number of photons incident to the detector was varied by changing the detector-source distance with Certificate Radiation Material (CRM), and compared to the output count rate. The count rate capability, and energy resolution with different amplifier and DAQ was evaluated. Additionally, the performance of DAQs in extremely high count rate was evaluated with signal data generated by the emulator which can simulate the detector signal waveforms fed into the DAQ based on the measured spectrum.

The amount of radioactive waste generated during decommissioning directly affects the disposal cost of waste. Most of the radioactive waste generated is a concrete waste. Therefore reducing the amount of concrete waste can ensure the economic feasibility of the decommissioning project. The activated concrete in a concrete waste can reduce waste only by physical cutting. Therefore it is most important to accurately identify and categorize radionuclides, radioactivity levels, and radioactivity distribution. In the case of radioactive concrete, radiological characteristics are generally evaluated by laboratory analysis after sampling. However it is difficult to apply to all facilities (accelerator & NPP, etc.) because it is a destructive method. Therefore it is necessary to secure verified in-situ measurement technology that can be applied to operational monitoring or decommissioning plans. In this study, the applicability of cyclotron facilities was evaluated based on the evaluation algorithm derived from the Peak to Compton (PTC) method of in-situ measurement technology. And the reliability of the PTC method was verified through qualitative analysis and quantitative analysis. In the case of qualitative analysis, the analysis results of KAERI which has core technology are compared. To this end SAEAN and KAERI conducted field application tests on the front concrete shielding wall of the cyclotron facility at the same time. After removing the background spectrum from the measured spectrum the PTC method was applied to calculate the Q-value for the counting rate in the peak area per counting rate in the Compton continuum area was calculated. As a result the Q-values of SAE-AN and KAERI were 0.52 and 0.24 respectively, and the result of deriving activation distribution(β) by substituting this for the β-Q correlation equation was found that 14.78 and 12.94. As a result of evaluating the activation by the thickness of the shielding wall it was found that 89.1% (SAE-AN) and 91.9% (KAERI) of the total radioactivity were exist at a depth of 5 cm. And it was found that 97.7% and 99.05% of the total radioactivity exists at a depth of 10 cm. The relative error between SAE-AN and KAERI is 1.35%, indicating that the analysis results of the two institutions are highly consistent. A core drill was performed on the concrete shielding wall in the cyclotron facility for the technical verification of the quantitative analysis method. A core sample (6 cm in diameter, 10 cm in depth) was cut to a depth of 2 cm and analyzed in the laboratory. The activation distribution(β) was calculated based on the radioactivity level of each depth sample, and it was found to be 16.99. The relative error between the quantitative analysis and the on-site measurement results was 14.95% confirming that the accuracy is relatively high.