Concrete is the primary building material for nuclear facilities, making it one of the most common forms of radioactive waste generated when decommissioning a nuclear facility. Of the total waste generated at the Connecticut Yankee and Maine Yankee nuclear power plants in the United States, concrete waste accounts for 83.5% of the total for Connecticut Yankee and 52% for Maine Yankee. In order to dispose of the low- to medium-level radioactive concrete waste generated during the decommissioning of nuclear power plants, it is necessary to analyze the radioactivity concentration of gamma nuclides such as Co-58, Co-60, Cs-137, and Ce-144. Gamma-ray spectroscopy is commonly used method to measure the radioactivity concentration of gamma nuclides in the radioactive waste; however, due to the nature of gamma detectors, gamma rays from sequentially decaying nuclides such as Co-60 or Y-88 are subject to True Coincidence Summing (TCS). TCS reduces the Full Energy Peak Efficiency (FEPE) of specific gamma ray and it can cause underestimation of radioactivity concentration. Therefor the TCS effect must be compensated for in order to accurately assess the radioactivity of the sample. In addition, samples with high density and large volume will experience a certain level of self-shielding effect of gamma rays, so this must also be compensated for. The Radioactive Waste Chemical Analysis Center at the Korea Atomic Energy Research Institute performs nuclide analysis for the final disposal of low- and intermediate-level concrete waste. Since a large number of samples must be analyzed within the facility, the analytical method must simultaneously satisfy accuracy and speed. In this study, we report on the results of evaluating the accuracy of the radioactivity concentration correction by applying an efficiency transfer method that appears to satisfy these requirements to concrete standard reference material.

Alpha activities can be used for categorization, transportation, and disposal of radioactive waste generated from the operation of nuclear facilities including nuclear power plants. In order to transport and dispose of such low- and intermediate-level radioactive waste (LILW) to the Wolsong LILW Disposal Center (WLDC) at Gyeongju, the gross alpha concentration of an individual drum should be determined according to the acceptance criteria. In addition, when the gross alpha concentration exceeds 10 Bq/g, the inventory of the comprising alpha emitters in the waste is to be identified. Gross alpha measurements using a proportional counter are usually straightforward, inexpensive, and high-throughput, so they are broadly used to assay the total alpha activity for environmental, health physics, and emergency-response assessments. However, several factors are thoughtfully considered to obtain a reliable approximate for the entire alpha emitters in a sample, which include the alpha particle energy of a particular radionuclide, the radionuclide that is used as a calibration standard, the uniformity of film in a planchet, time between sample collection and sample preparation, and time between sample preparation and counting. Korea Atomic Energy Research Institute (KAERI) have evaluated the inventory of radionuclides in low-level radioactive waste drums to send every year hundreds of them to the WLDC. In this presentation, we revisit the gross alpha measurement results of the drums transported to WLDC in the past few years and compare them with the concentrations of alpha emitters measured from alpha spectrometry and gamma spectrometry. This study offers an insight into the gross alpha measurement for radioactive waste regarding calibration source, self-absorption effect, composition of alpha emitters, etc.

To achieve permanent disposal of radioactive waste drums, the radionuclides analysis process is essential. A variety of waste types are generated through the operation of nuclear facilities, with dry active waste (DAW) being the most abundant. To perform radionuclides analysis, sample pretreatment technology is required to transform solid samples into solutions. In this study, we developed a dry ashing-microwave digestion method and secured the reliability of the analysis results through a validity evaluation. Additionally, we conducted a comparative analysis of the radioactivity of 94Nb nuclides with and without the chemical separation process, which reduced the minimum detectable activity (MDA) level by more than 65-fold for a certain sample.

The Korea Atomic Energy Research Institute (KAERI) employs a methodology for evaluating the concentration of radionuclides, dividing them into volatile and non-volatile nuclides based on their characteristics, to ensure the permanent disposal of internally generated radioactive waste. Gamma spectroscopy enables the detection and radiation concentration determination of individual nuclides in samples containing multiple gamma-emitting nuclides. Due to the stochastic nature of radioactive decay, the generated radiation signal can interact with the detector faster than the detected signal processing time, causing dead time in the gamma spectroscopy process. Radioactive waste samples typically exhibit higher radiation levels than environmental samples, leading to long dead times during the measurement process, consequently reducing the accuracy of the analysis. Therefore, dead time must be considered when analyzing radioactive waste samples. During the measurement process, dead time may vary between a few seconds to several tens of thousands of seconds. More long dead time may also result in a temporal loss in the analysis stage, requiring more time than the actual measurement time. Long dead time samples undergo re-measurement after dilution to facilitate the analysis. As the prepared solution is also utilized in the nuclide separation processes, minimizing sample loss during dilution is crucial. Hence, predicting the possibility of dead time exceeding the target sample in advance and determining the corresponding dilution factor can prevent delays in the analysis process and the loss of samples due to dilution. In this study, to improve the issues related to gamma analysis, by using data generated during the analysis process, investigated methods to predict long dead time samples in advance and determining criteria for dilution factors. As a result of comparing the dead time data of 5% or long with the dose of the solution sample, it was concluded that analysis should be performed after dilution when it is about 0.4 μSv/h or high. However, some samples required dilution even at doses below 0.4 μSv/h. Also, re-measurement after dilution, the sample with a dead time of less than 32% was measured with less than 5% when diluted 10 times, and more than 32% required more than 10 times dilution. We suppose that with additional data collection for analyzing these samples in the future, if we can establish clearer criteria, we can predict long dead time samples in advance and solve the problem of analysis delay and sample loss.

According to the ‘Regulations on the Delivery of Low and Medium Level Radioactive Waste’, Notification No. 2021-26 of the Nuclear Safety and Security Commission, a history of radioactive waste and a total amount of radioactivity in a drum are mandatory. At this time, the inventory of radionuclides that make up more than 95% of the total radioactivity contained in the waste drum should be identified, including the radioactivity of H-3, C-14, Fe-55, Co-58, Co-60, Ni-59, Ni-63, Sr- 90, Nb-94, Tc-99, I-129, Cs-137, Ce-144, and total alpha. Among nuclides to be identified, gamma-emitting nuclides are usually analyzed with a gamma ray spectrometer such as HPGe. When a specific gamma-ray is measured with a detector, several types of peaks generated by recombination or scattering of electrons are simultaneously detected in addition to the corresponding gamma-ray in gamma-ray spectroscopy. Among them, the full energy peak efficiency (FEPE) with the total gamma energy is used for equipment calibration. However, this total energy peak efficiency may not be accurately measured due to the coincidence summing effect. There are two types of coincidence summing: Random and True. The random coincidence summing occurs when two or more gamma particles emitted from multiple nuclides are simultaneously absorbed within the dead time of the detector, and this effect becomes stronger as the counting rate increases. The true coincidence summing is caused by simultaneous absorption of gamma particles emitted by two or more consecutive energy levels transitioning from single nuclide within the dead time of the detector. This effect is independent of the counting rate but affected by the geometry and absolute efficiency of the detector. The FEPE decreases and the peak count of region where the energy of gamma particles is combined increases when the coincidence summing occurs. At the Radioactive Waste Chemical Analysis Center, KAERI, samples with a dead time of 5% or more are diluted and re-measured in order to reduce the random coincidence summing when evaluating the gamma nuclide inventory of radioactive waste. In addition, a certain distance is placed between the sample and the detector during measurement to reduce the true coincidence summing. In this study, we evaluate the coincidence summing effect in our apparatus for the measurement of radioactive waste samples.