Efficiency calibration of a coaxial HPGe detector-Marinelli beaker geometry using an 152Eu source prepared in epoxy matrix and its validation by efficiency transfer method

In this study, an in-house 152Eu calibration source was produced from a custom epoxy matrix with a density of ρ = 1.14 g cm−3, which is chemically stable and durable form after its solidification. The homogeneity of 152Eu in matrix was obtained better than 98%. For a Marinelli beaker geometry, an efficiency calibration procedure was applied to a n-type, coaxial, 78.5% relative efficient HPGe detector in the energy range of 121.7–1408.0 keV by using in-house 152Eu calibration source. Then the measured efficiencies for Marinelli geometry were compared with the results calculated by MEFFTRAN and ANGLE softwares for the validation. Although MEFFTRAN and ANGLE have two different efficiency transfer algorithms to calculate the efficiencies, they usually need to use a reliable and accurate reference efficiency values as input data.Hence, reference efficiency values were obtained experimentally from a multinuclide standard source for the same detector- Marinelli geometry. In the present source characterization, the corrections required for self-absorption and true coincidence summing effects for 152Eu gamma-rays were also obtained for a such close counting geometry condition. The experimental results confirmed the validity of efficiency calculations obtained by MEFFTRAN and ANGLE softwares that are calculation tools. Keywords: Gamma-ray spectrometry, Marinelli geometry, Epoxy, 152Eu, In-house source preparation, Efficiency transfer, MEFFTRAN, ANGL

Investigation of thermal neutron detection capability of a CdZnTe detector in a mixed gamma-neutron radiation field

The aim of this study was to investigate the thermal neutron measurement capability of a CdZnTe detector irradiated in a mixed gamma-neutron radiation field. A CdZnTe detector was irradiated in one of the irradiation tubes of a 241Am-Be source unit to determine the sensitivity factors of the detector in terms of peak count rate (counts per second [cps]) per neutron flux (in square centimeters per second) [cps/neutron·cm−2·s−1]. The CdZnTe detector was covered in a 1-mm-thick cadmium (Cd) cylindrical box to completely absorb incoming thermal neutrons via 113Cd(n,γ) capture reactions. To achieve, this Cd-covered CdZnTe detector was placed in a well-thermalized neutron field (f-ratio = 50.9 ± 1.3) in the irradiation tube of the 241Am-Be neutron source. The gamma-ray spectra were acquired, and the most intense gamma-ray peak at 558 keV (0.74 γ/n) was evaluated to estimate the thermal neutron flux. The epithermal component was also estimated from the bare CdZnTe detector irradiation because the epithermal neutron cutoff energy is about 0.55 eV at the 1-mm-thick Cd filter. A high-density polyethylene moderating cylinder box can also be fitted into the Cd filter box to enhance thermal sensitivity because of moderation of the epithermal neutron component. Neutron detection sensitivity was determined from the measured count rates from the 558 keV photopeak, using the measured neutron fluxes at different irradiation positions. The results indicate that the CdZnTe detector can serve as a neutron detector in mixed gamma-neutron radiation fields, such as reactors, neutron generators, linear accelerators, and isotopic neutron sources. New thermal neutron filters, such as Gd and Tb foils, can be tested instead of the Cd filter due to its serious gamma-shielding effect