**4. Summary and conclusions**

**3.3. The detection efficiency evaluation**

130 Nuclear Material Performance

The detector intrinsic efficiency commonly is conditioned mainly by the material of the detector, the radiation energy and the physical thickness of the detector in the direction of the incident radiation [34]. A small dependence on source-detector distance is present due to the average path length of the radiation through the detector will amend somewhat with this area. The counting efficiencies can be classified by the nature of the events recorded. If all phenom‐ ena from the detector will be recorded, then the total efficiency will be of interest. Therefore, all interactions, no matter if the energies are small, are considered to be recorded. The peak efficiency presumes that only those interactions that deposit the full energy of the incident radiation are recorded. If the total area under the peak is integrated, then the number of full energy events can be achieved. In **Figure 11**, it represents the FEP and total peak efficiencies obtained from the simulated spectra, for Geom1 and Geom2 geometries. The fact that the efficiency in Geom2 is smaller than in Geom1 even if the second detector has a higher intrinsic

efficiency is due to the smaller collimator acceptance in the case of Geom2.

**Figure 11.** The peak efficiencies and total efficiencies simulated with GEANT 3.21 for Geom1 and Geom2.

This chapter explores the specific gamma-ray spectrometry phenomena in their deepness in different work conditions. Thus, the studies, simulations and experimental results were carried out and were presented in an integrated view in the sections of the chapter.

The utility of numerical method as it is ETNA software to compute the efficiency transfer factors for divers measurement geometries used in routine laboratory measurements was tested for NaI(Tl) and HPGe detector. Starting from a reference efficiency measured for a point source, peak efficiencies were evaluated for point sources placed at several detector-source distances, moreover for disc sources or for volume sources with different compositions and densities. This was done specifically for each detector involved in the study. The methodology shows that the efficiency transfer factors are accurately computed using ETNA software. The obtained results are valuable and can be used without restriction if all the details of the detectors and measurements are accurately known.

An important contribution in the development of gamma-ray spectrometry methods was made with the examination realized for the response function characterization of the two gamma-ray spectrometry systems: ISOCART from Ortec and WS1100 Segmented Gamma Scanner from Canberra. These systems are used especially for the measurement and charac‐ terization of radioactive waste. Based on GEANT 3.21 toolkit, a simulation program was developed to simulate the spectra expected to be obtained by the two systems, for volume sources, and for the 50–2000 keV energy range. Many spectra (hundreds) were simulated and then combined to obtain the spectrum expected in real measurements.

Considering the national regulations, the radioactivity and the nuclide composition of the waste must be identified prior to their transfer outside the site of their burial or placement in storage areas. The most important step on the characterization process is the establishment of the radionuclide content, most often achieved through non-destructive measurements (NDA). The radiological characterization of radioactive waste should ensure their correct classification and a reasonable use of interim storage and final disposal. The radioactivity overestimation leads to a reduction in waste storage capacity, and the underestimation creates problems in terms of safety. The release from regulatory control will reduce the volume of storage waste and enable their beneficial use.

Future studies are required to develop calibration techniques and evaluation of measurement for parallelepiped shape containers with radioactive waste using gamma spectrometric measurement systems. The counting geometry of parallelepiped container is completely different from the counting geometry of a small cylindrical radioactive source, and conse‐ quently, the efficiency calibration is more difficult to be estimated. Knowing the efficiency calibration, which varies greatly with the source-detector distance, the geometry and the absorption factors, is essential for the assay of radioactive waste. Because it is almost impossible to estimate the efficiency calibration curve based on the experimental measurement, simula‐ tion programs based on the Monte Carlo codes need to be developed. The evaluation of response function of HPGe detectors for parallelepiped counting geometry needs to be done.
