The recent release of contaminated water from the Fukushima Daiichi Nuclear Power Plant highlights the need for accurate tritium measurement, particularly near the minimum detectable activity (MDA) of 5 Bq·L−1 set by South Korea’s Nuclear Safety and Security Commission. This study aims to improve low-level tritium measurement accuracy by optimizing the region of interest (ROI) for quench curve determination. These adjustments are crucial for separating tritium signals from background noise. Quench standards were prepared and measured using a liquid scintillation counter (LSC). Three ROIs were analyzed to assess the impact of channel selection on measurement precision: A 20-148 channel range optimized via figure of merit (FOM) analysis, a 20-250 channel range covering tritium’s full beta spectrum, and a broad 1-1024 channel range. Quench curves were obtained by fitting the counting efficiency of each ROI to the quench standards. Tritium samples with six different activity levels were prepared, and their radioactivity was calculated using the quench curves. Selecting appropriate ROIs for quench curve determination is critical for measuring low-concentration tritium accurately. This approach reduces uncertainty and emphasizes reliable methods to improve the precision and consistency of tritium measurements.

There are analytical methods used for measuring activity when light photons are emitted for scintillating-based analytical application. When this electron returns to the original stable state, it releases its energy in the form of light emission (visible light or ultraviolet light), and this phenomenon is called scintillation. Scintillator is a general term for substances that emit fluorescence when exposed to radiation such as gamma-rays. Radioactivity is all around us and is unavoidable because of the ubiquitous existence of background radiations emitted by different sources. The scintillator contributes to these sensing, and it is expected that the inspection accuracy and limit of detection will be improved and new inspection methods will be developed in the future. Moreover, scintillators are chemical or nanomaterial sensors that can be used to detect the presence of chemical species and elements or monitor physical parameters on the nanoscale. In this study, it includes finding use in scintillating-based analytical sensing applications. A chemical and nanomaterial based sensors are self-contained analytical tools that could provide information about the chemical compositions or elements of their environment, that is, a liquid or even gas condition. Herein, we present an insightful review of previously reported research in the development of high-performance gamma scintillators. The major performance-limiting factors of scintillation are summed up here. Moreover, the 2D material has been discussed in the context of these parameters. It will help in designing a prototype nanomaterial based scintillators for radiation detection of gamma-ray.

The presence of organic components in spent scintillation liquid should be considered during all steps of radioactive waste processing for final disposal. Scintillation liquids often referred to as cocktails are generated form radiochemical analyses of radionuclides, which mainly consists of mixtures of liquid organic materials such as toluene and xylene. Typical features of these liquid organic materials are volatility, combustibility and toxicity. These are the reason why special attention must be paid to the management of liquid organic radioactive wastes. To select an appropriate waste management strategy and to design the treatment process of spent scintillation cocktails, it is required to investigate the nature of the waste such as specific radioactivity and moisture content. The analysis results of spent scintillation liquid generated at Wolsong nuclear power plants will be discussed. An overview of the technical approaches available for the treatment of organic radioactive waste will be additionally provided.

When self-disposing of radioactive waste, it is important to follow the acceptable concentration standards for each nuclide set by the Nuclear Safety and Security Commission (NSSC). Gamma-emitting nuclides can be easily analyzed with a simple pretreatment process, but beta-emitting nuclides require a chemical separation procedure to be analyzed for radiochemistry analysis. When analyzing betaemitting nuclides for the purpose of self-disposal, there may be difficulties in radiation detection after the chemical separation process. This is because the concentration of beta nuclides in the sample may be low and some of them may be lost during the chemical separation. Therefore, measurement method of gross-beta activity can be used instead of that of each nuclide to access the compliance of selfdisposal criteria. While a proportional counter is commonly used to measure gross-beta activity, liquid scintillation counting can also be used to measure gross-beta, and we plan to compare the results of both methods.

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.

Radioactive wastes that are generated as a result of operating NPPs, contain 63Ni and 59Ni that should be analyzed in accordance with the notice of Nuclear Safety and Security Commission (NSSC) for the acceptance of Korea Radioactive Waste Agency (KORAD). Analyzing 63Ni and 59Ni has few challenges to determine activities of each nuclide in radioactive waste sample that contains both nuclides. As is well known, 63Ni can be analyzed by liquid scintillation counter (LSC) detecting its emitted beta rays, however, beta rays emitted from 59Ni are overlapped on the spectrum. Therefore, to discriminate those two nuclides, spectrum channel should be divided according to its dedicating part of the spectrum. For instance, 59Ni contribute to spectrum channel 30–250, on the other hand, 63Ni contributes to spectrum channel 30–450. In other word, 63Ni solely can be analyzed on the channel from 260 to 450. To analyze both 63Ni and 59Ni using this channel division method, detection efficiency must be measured in advance; efficiency of 63Ni and 59Ni at ch. 30–250, and efficiency of 63Ni at ch. 260–450, then the activity can be calculated using the corresponding efficiency. In this study, for verifying the feasibility of channel division method, 5 simulated samples were prepared with different ratio of 63Ni/59Ni. The ratio varies as 1, 2, 10, 20 and 100 spiking standard source of 63Ni and 59Ni. Each sample was mixed with scintillation cocktail and detected for 90 minutes by LSC (300SL, Hidex) after the stabilization of solutions. As a result, calculated 63Ni activities for all sample were averaged as 97% of spiked activity. However, calculated 59Ni activity were 101%, 103%, 128%, 140%, 260%, respectively. The result indicates that 59Ni cannot be discriminated by channel division method when it exists in the sample with high 63Ni over 10 times then 59Ni such as radioactive waste sample. However, the results also show that the channel division method for analyzing 63Ni activity was successful verifying it can determine the activity of 63Ni regardless of the affect of 59Ni on the spectrum.

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.

Liquid scintillation cocktail is liquid waste, which consists of an organic solvent, scintillator, surfactant, and radionuclide. Large volumes of liquid scintillation waste are generated each year, and both the organic compound and radionuclide content can negatively affect on the health and the environment. Therefore, the liquid scintillation waste should be treated in an appropriate way. In this study, to facilitate the treatment of liquid scintillation waste, the sulfate-radical advanced oxidation process (SR-AOP) was performed for the mineralization of liquid scintillator waste. In SR-AOP, highly reactive sulfate radicals, which react more selectively and efficiently with organic compounds, are produced in situ by cleaving the peroxide bond in the persulfate molecule. For the experiment, 100 times diluted ULTIMA GOLD-LLT (initial TOC=699,800 ppm) was used as a liquid scintillation waste. The TOC removal efficiency of liquid scintillation waste by the OXONE (potassium peroxymonosulfate, PMS, 2KHSO5+KHSO4+K2SO4) and sodium persulfate (PS) with varying dosages (4–12 mM) was tested, and the effects of Co2+ and Cu2+ catalysts were compared at a range of pHs (3, 7, and 9). The experimental results demonstrated that 91% TOC removal of ULTIMA GOLD-LLT could be achieved for SR-AOP at initial pH=9, Co2+=1.2 mM (catalyst), PMS=4.8 mM (oxidant) for 60 min reaction. Compared to traditional Fenton AOP which is effective only at low pH, PMS based SR-AOP with Co2+ catalyst is effective at wide range of pHs and less dependent on the treatment efficiency of the operational pH. Therefore, it can be useful for the mineralization of liquid scintillation waste which is difficult to treat with a general treatment method due to the mixture of various organic compounds.