Detectors used for nuclear material safeguards activities are using scintillator detectors to quickly calculate the uranium enrichment at various nuclear material handling facilities. In order to measure the uranium enrichment, a region of interest is set around 185.7 keV which is the main gamma emission energy of uranium-235 in which the proportional relationship between the amount of uranium-235 and the net count is used. It is necessary to perform channel/energy calibration that a specific channel of the multi-channel analyzer is set to 185.7 keV. Most detector manufacturers have a built-in calibration source so that it is automatically performed when the detector starts to operate. In addition, the scintillator detector requires attention because the channel/energy gain may change depending on the ambient temperature so that a calibration source is used to compensate for this. In this paper, the spectral features are examined from among the scintillator detectors seeded with calibration sources used for safeguards activities. For this purpose, FLIR’s Identifinder-2 R400 T2 model and Canberra’s NAID model were used. HM-5 contains about 15nCi of Cs-137 and a photoelectric peak occurs at 662.1 keV. NAID contains about Am-241 of 55 nCi which alpha decays and subsequently emits gamma rays of 59.5 keV and 26.3 keV. The major difference among the detectors occurs in the background spectrum due to the difference in the source. From that kind of spectral features, it can be confirmed that the equipment is operating properly only when the spectrum by the corresponding calibration source is accurately known. The results of this study will enable a better understanding of the characteristics of scintillator detectors used for uranium enrichment analysis. Therefore, it is expected to be used as basic research for related software utilization as well as development in the future.

Dose-rate monitoring instruments are indispensable to protect workers from the potential risk of radiation exposure, and are commonly calibrated in terms of the ambient dose equivalent (H*(10)), an operational quantity that is widely used for area monitoring. Plastic scintillation detectors are ideal equipment for dosimetry because of their advantages of low cost and tissue equivalence. However, these detectors are rarely used owing to the characteristics caused by low-atomic-number elements, such as low interaction coefficients and poor gamma-ray spectroscopy. In this study, we calculated the G(E) function to utilize a plastic scintillation detector in spectroscopic dosimetry applications. Numerous spectra with arbitrary energies of gamma rays and their H*(10) were calculated using Monte Carlo simulations and were used to obtain the G(E) function. We acquired three different types of G(E) functions using the least-square and first-order methods. The performances of the G(E) functions were compared with one another, including the conventional total counting method. The performance was evaluated using 133Ba, 137Cs, 152Eu, and 60Co radioisotopes in terms of the mean absolute percentage error between the predicted and true H*(10) values. In addition, we confirmed that the dose-rate prediction errors were within acceptable uncertainty ranges and that the energy responses to 137Cs of the G(E) function satisfied the criteria recommended by the International Commission.

For spent nuclear fuel transferred to dry storage facilities, it is difficult to apply safeguards approaches and long-term integrity verification due to the structural characteristics of the facility. There is a need to check the integrity of the nuclear fuel assembly before transferring it to a dry storage facility and are need to provide information on whether there are any defects. At the Korea Institute of Nuclear Nonproliferation and Control, as a non-destructive testing technology for ensuring Continuity of Knowledge (CoK) of the dry storage facilities, a methodology for reconstructing images by neutron tomographic technique from spent nuclear fuel using a He-4 gas scintillation detector was presented. It is thought that the He-4 gas scintillation detector-based technology can be used to verify the defect of the nuclear fuel assembly. This methodology must be accompanied by accurate neutron measurements. The place where the technique was conducted is surrounded by a concrete wall. Concrete contains water molecules, which can affect neutron measurements. In this study, reconstruction images based on neutron measurements and MCNP simulations are compared to verify the effects of the concrete. Neutron measurements were performed by measuring Cf-252 neutron sources in a 1/10 lab-scale TN- 32 cask with six He-4 gas scintillation detectors as an array. Neutron sources are fixed at each point in the cask, and the He-4 detector array is rotated from 0° to 360° at 10° intervals to reconstruct the image using the filtered back-projection (FBP) method. Also, in MCNP reconstructed images, there are two versions depending on whether concrete wall. The source image and ring shape were found in the measurement-based thermal neutron reconstruction image, which was similar to the simulation image that considering the concrete effects. On the other hand, in the simulation reconstruction image without the concrete, only the shape of the source was found. Thus, the effect of concrete should be considered when performing the neutron tomographic techniques using He-4 gas scintillation detectors.

일반적으로 방사선 선원의 강도는 거리의 역자승 법칙을 따른다. 그러나 방사선 선원과 검출기와의 거리가 가까울수록 거리의 역자승 법칙 실험은 이론과 실험의 일치하지 못하는 오류를 가져오게 된다. 본 연구에서는 방사선 선원과검출기와의 거리에 따른 거리의 역자승 법칙이 실제 실험에서는 정확하게 성립하지 않는 이유를 실험적으로 확인하였다. 그리고 이 문제를 해결하기 위하여 측정된 방사능을 보정하기 위하여 보정계수를 실험적으로 얻었다. 측정에 사용한 검출기는 2˝×2˝ø NaI(Tl) 신틸레이션 검출기를 사용하였고, 방사선에너지의 변화에 따른 효과를 확인하기 위하여감마선 선원 60Co(1.174 MeV, 1.333 MeV)와 137Cs(0.662 MeV)에 대한 실험도 병행하였다. 측정에서 얻어진 거리의역자승 법칙의 결과들을 보정계수를 이용하여 측정값들을 보정한 결과 거리의 역자승 법칙과 매우 일치하는 경향을 보였고, 오류에 대한 원인을 실험적으로 확인하였다. 이러한 결과는 유한한 체적을 가진 검출기를 사용하여 방사선의 강도가 거리의 역자승에 반비례하는 실험을 할 경우 모두 해당되는 문제이므로 본 연구의 결과는 방사선계측 분야에 매우 유용하게 사용되어질 것으로 사료된다.