Within the air purification system of a nuclear power plant, specific radioactive isotopes are extracted from gases through adsorption onto activated carbon. To properly dispose of used activated carbon, it is essential to determine the concentration of radioactive nuclides within it. This study discusses the application of the pyrolysis method for analyzing the concentrations of 3H and 14C in spent activated carbon. The pyrolysis was conducted using Raddec’s Pyrolyser, with adjustments made to parameters such as temperature profiles, airflow rates, sample quantities, and trapping solution volumes. The evaluation method for the pyrolysis of activated carbon to analyze 3H and 14C involved adding 3H and 14C sources to the activated carbon before use and subsequently assessing the recovery rates of the added sources in comparison to the analysis results.

In order to establish disposal plans for sludge, which is one of the untreated waste materials from domestic nuclear power plants, it is necessary to determine the radioactivity concentration of radioactive isotopes. In this study, we aim to evaluate the gross alpha radioactivity of sludge containing radioactive contaminants after pre-treatment, in order to assess the level of sludge waste and obtain analytical data for discussing disposal methods. Samples of sludge generated from nuclear power plants were pre-treated, solutionized, and prepared as analysis samples for evaluating the gross alpha radioactivity.

Typically, the bottom of the effluent treatment facility at a nuclear power plant contains sediment, which is low-contamination waste consisting of sludge, gravel, sand, and other materials from which radioactive contaminants have been removed. Among these sediments, sludge is an irregular solid form consisting of small particles that are coagulated together, with radioactive isotopes containing cobalt attached. Currently, there is a record of disposing of dry active waste from domestic nuclear power plants, and efforts are underway to gather basic data for the disposal of untreated waste such as sludge, spent filter, and spent resin. In particular, the classification and disposal methods of waste will be determined based on the radioactivity concentration. Therefore, plans are being made to determine the radioactivity concentration of radioactive isotopes and establish disposal plans for sludge samples. In this study, pre-treatment and solutionization were carried out for the analysis of radioactive isotopes in sludge sampels from nuclear power plants. The deviation of the gamma radioisotope analysis results was derived to obtain an optimal sample quantity that represents the sludge.

For the disposition of radioactive wastes generated from nuclear power plant, radioisotope inventory must be analyzed to determine an activity concentration of radionuclides. Radionuclides in low- and intermediate-low-level of radioactive wastes, however, can be easily classified to easyto- measure (ETM) and difficult-to-measure (DTM) nuclides. ETM nuclides are gamma emitting nuclides that is relatively easy to measure because they do not need to be destroyed for the preprocessing. On the other hands, DTM nuclides are alpha and beta emitting nuclides that need to be destroyed for the preprocessing and also need chemical separation. Currently, measurement methods for DTM nuclides are developed and in this paper measurement methods of Fe-55, Ni-59, Ni-63, Sr-90 and Tc-99 will be introduced.

To analyze the radioactivity of 3H and 14C in miscellaneous radioactive wastes generated from nuclear power plants, a wet digestion method using sulfuric acid is currently used. However, sulfuric acid is classified as a special management material, and there is no disposal method for contaminated radioactive waste. Therefore, research on a thermal decomposition method that can analyze the DAW radioactive waste samples without using sulfuric acid is necessary. In this study, we will cover the final sample amount, sample injection method, and prevention of organic ignition to meet the minimum detection limit requirements of the analysis equipment. Through this research, optimal conditions for the thermal decomposition method for analyzing the radioactivity of 3H and 14C in DAW radioactive wastes generated from nuclear power plants can be derived.

For the final disposal of radioactive waste, concentration of gamma nuclides such as Co-58, Co-60, Cs-137, Nb-94 have to be determined to meet nuclear regulatory requirements. In general, gamma nuclide analysis can be performed with simple sample pretreatment without complicated chemical separation processes due to the characteristics of the nuclide and high resolution of the measuring equipment. However, when the concentration of Co-60 is high in a specific radioactive waste generated at the NPP, the background is increased by the compton continuum of Co-60. That makes it difficult to evaluate accurately Nb-94, which is in the lower energy band than the gamma ray energy region of Co-60 and especially Cs-137, which is used as a key nuclide of scaling factor. In this study, We consider the problem of MDA dissatisfaction or overestimation due to the increased background by Co-60.

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.

Many radionuclides emit two or more gamma rays in a cascade once they decay. At this time, gamma rays are detected at the same time, and the signals are overlapped and measured as one added signal. This is called the summing coincidence effect, and it causes an error of more than 10% depending on the detection efficiency, measurement conditions, and target nuclide. It is known to be greater as the efficiency of the detector increases and as the distance between the source and the detector decreases. It is necessary to consider the summing coincidence effect since the efficiency of the HPGe detector owned by the KHNP CRI is as high as 65%. In this study, We would like to propose an appropriate gamma nuclide analysis method for radioactive waste generated from NPP by evaluating the influence on the summing coincidence effect.

To analyze the activity concentration of radionuclides in radioactive sludge samples generated from low- and intermediate-low-level radioactive waste from domestic nuclear power plant, a pretreatment process that dissolves and homogenizes the sample is essential. However, this pretreatment process requires the use of hydrofluoric acid, which makes analysis difficult and challenges users to handle harmful chemicals. Therefore, we aim to minimize the use of hydrofluoric acid by measuring gamma nuclides in the sludge sample without pretreatment process and compare the differences of measurement results according to the sample matrix with and without pretreatment process. We will collect about 0.1 g of the sludge sample, and dissolve it using an acid treatment process after using microwave decomposition. We will then use gamma spectroscopy to check the concentration of nuclides present in the sludge before and after dissolution and consider the effect of the sample matrix.

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.

Radioactivity of radiostrontiums, Sr-89 and Sr-90, which are both pure beta-emitters, are generally measured via Cherenkov counting. However, the determination of Cherenkov counting efficiencies of radiostrontiums requires a complicated procedure due to the presence of Y-90 (also a pure betaemitter) which is the daughter nuclide of Sr-90. In this study, we have developed a machine learning approach using a linear regression model which allows an easier and simultaneous determination of the Cherenkov counting efficiencies of the radiostrontiums. The linear regression model was employed because total net Cherenkov count (Ct) from the three beta-emitters at time t after the separation of Y- 90, can be expressed as a linear combination of their respective time-varying radioactivities with their respective coefficients (parameters) being their counting efficiencies: Ct = εSr-90[ASr-90·exp(–λSr-90·t)] + εSr-89[ASr-89·exp(–λSr-89·t)] + εY-90[ASr-90·exp(1–λSr-90·t)], where ε is a counting efficiency, A is an initial activity, λ is a decay constant and t is time after the separation of Y-90, Thus, if we train the model with multiple Cherenkov counts measured from the three beta emitters, then we can obtain their estimates for counting efficiencies (so-called parameters) straightforward. For this, the model has been trained by two methods: Ordinary Least Squares (OLS) and Bayesian linear regression (BLR), for which two software packages, PyMC3 and Stan were employed to compare their performances. The results showed that the accuracy of the OLS was worse than that of the BLR. Particularly, the counting efficiency of Sr-90 was estimated to be smaller than 0, which is an unrealistic value. On the other hand, the estimates of the BLR gave realistic values which are close to the true values. Additionally, the BLR was able to provide a distribution for each counting efficiency (so-called “posterior”) from which various types of inference can be made including median and credible interval in the Bayesian statistics which is analogous to, but different from confidence interval in the Frequentist statistics. In the results of the BLR, the Stan package gave more accurate estimates than the PyMC3 package. Therefore, it is expected that counting efficiencies of the radiostrontiums including radioyttrium can be determined at the same time, more easily and accurately, by using the BLR with the Stan package and that the activities of radiostrontium also can be determined more easily by using the BLR if we know their counting efficiencies in advance. It is worth noting that the usage of the linear regression model in this study was different from the usual one where the trained model is used to predict a response value (count) from a set of unseen regressor values (activities).

For the final disposal of radioactive waste generated during the operation of nuclear power plants, concentrations of 14 radionuclides including gross alpha have to be determined to meet nuclear regulatory requirements. In order to determine the gross alpha radioactivity in radioactive waste, the sample must be preprocessed into a solution which is usually a strong acid. When this solution is used to prepare the gross alpha measurement sample, it produces a lot of salt, which makes an accurate measurement difficult. Also it causes corrosion of a planchet, which causes problems in the disposal of waste in the future. For these reasons, an acid treatment of the solution was added to the existing preprocess procedure, which is also expected to improve the measurement error. Although the gross alpha measurement is known to be easy to perform and able to give rapid results, it cannot be used for quantitative analysis. This is because the energies emitted by the individual alpha nuclides are assumed to be produced from a single alpha emitted by the individual alpha nuclides are assumed to be produced from a single alpha emitter used as the standard calibration source. Also, due to self-absorption of alpha particles a counting rate depends on the thickness (or weight) of the residues on the planchet. In this study, we compared gross alpha radioactivity with and without an acid treatment to prepare gross alpha measurement samples. The weights of the treated samples increased by at most 5% after about 12 hours of evaporation to dryness, and then saturated or slightly decreased, while the weights of the untreated samples increased up to 20% over time. In addition, the radioactivities of the untreated samples were about two times those of the treated samples. This is considered to be due to differences both in the geometric shapes of the samples and the weights of their residues which resulted from whether acid treatment was applied or not. The results of this study showed that an acid treatment was beneficial in reducing both production corrosion and salts which could result in more reliable and constant measurements of gross alpha activity. The results showed that acid treatment was beneficial in reducing corrosion and measurement errors.