Concerning the apprehensions about naturally occurring radioactive materials (NORM) residues, the International Atomic Energy Agency (IAEA) and its member nations have acknowledged the imperative to ensure the radiation safety of NORM industries. Residues with elevated radioactivity concentrations are predominantly produced during NORM processing, in the form of scale and sludge, referred to as technically enhanced NORM (TENORM). Substantial quantities of TENORM residues have been released externally due to the dismantling of NORM processing factories. These residues become concentrated and fixed in scale inside scrap pipes. To assess the radioactivity of scales in pipes of various shapes, a Monte Carlo simulation was employed to determine dose rates corresponding to the action level in TENORM regulations for different pipe diameters and thicknesses. Onsite gamma spectrometry was conducted on a scrap iron pipe from the titanium dioxide manufacturing factory. The measured dose rate on the pipe enabled the estimation of NORM concentration in the pipe scale onsite. The derived action level in dose rate can be applied in the NORM regulation procedure for on-site judgments.

In this study, we introduce the validation of the analysis guidelines through preliminary experiments of the draft analysis guidelines before analyzing waste materials (non-combustible). This validation data was applied the accuracy and efficiency of the separation and analysis for the waste such as steel generated from NPP. Steel (non-flammable) was leached the mixed acid and the leaching solution was separated by using the separation guidelines. Steel was corroded with radioactive RM (Co-60, Cs-137) and mixed acid. After drying, the corroded steel was measured the initial radioactivity by a HPGe detector (10,000 seconds). The sample was inserted in a beaker and leached with mixed acid (10 M HNO3 + 4 M HCl) for 2 hours. In this solution, it added 2 ml of H2O2 to increase the leaching effect. The ultrasonic device was adjusted so that the temperature does not exceed 60°C. After elution, the surface of the sample was washed with pure water. The weight of the sample was measured accurately, and recorded the weight loss rate after infiltration. The leaching sample was measured radioactivity by a HPGe detector (10,000 seconds). It was calculated the recovery rate based on the difference in total radioactivity before and after leaching. Before the test, radioactive RM (Co-60, Cs-137) was radioactive deposited by corrosion, but Cs- 137 was not detected in the initial gamma measurement and only Co-60 nuclides were deposited. The recovery rate test results were confirmed to be about 100%.

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.

Radioactive contamination distribution in nuclear facilities is typically measured and analyzed using radiation sensors. Since generally used detection sensors have relatively high efficiency, it is difficult to apply them to a high radiation field. Therefore, shielding/collimators and small size detectors are typically used. Nevertheless, problems of pulse accumulation and dead time still remain. This can cause measurement errors and distort the energy spectrum. In this study, this problem was confirmed through experiments, and signal pile-up and dead time correction studies were performed. A detection system combining a GAGG sensor and SiPM with a size of 10 mm × 10 mm × 10 mm was used, and GAGG radiation characteristics were evaluated for each radiation dose (0.001~57 mSv/h). As a result, efficiency increased as the dose increased, but the energy spectrum tended to shift to the left. At a radiation dose intensity of 400 Ci (14.8 TBq), a collimator was additionally installed, but efficiency decreased and the spectrum was distorted. It was analyzed that signal loss occurred when more than 1 million particles were incident on the detector. In this high-radioactivity area, quantitative analysis is likely to be difficult due to spectral distortion, and this needs to be supplemented through a correction algorithm. In recent research cases, the development of correction algorithms using MCNP and AI is being actively carried out around the world, and more than 98% of the signals have been corrected and the spectrum has been restored. Nevertheless, the artificial intelligence (AI) results were based on only 2-3 overlapping pulse data and did not consider the effect of noise, so they did not solve realistic problems. Additional research is needed. In the future, we plan to conduct signal correction research using ≈10×10 mm small size detectors (GAGG, CZT etc.). Also, the performance evaluation of the measurement/analysis system is intended to be performed in an environment similar to the high radiation field of an actual nuclear facility.

To evaluate the inventory of radionuclides for the disposal of waste generated from nuclear power plants, indirect assessment methods such as the scaling factor method or average radioactivity concentration method can be applied. A scaling factor represents the average concentration ratio between key radionuclides and difficult-to-measure (DTM) radionuclides, while the average radioactivity concentration refers to the average concentration of DTM radionuclides, regardless of the concentration of key radionuclides or within specific ranges of key radionuclide concentrations. These indirect assessment methods can be statistically derived through the analysis of representative drums. This study will address how to apply these scaling factors and average radioactivity concentrations. Firstly, the concentration of gamma-emitting radionuclides will be analyzed using a drum radionuclide analyzer, and the concentration of DTM radionuclides will be determined by applying scaling factors specific to each DTM radionuclide. In the case of using the average radioactivity concentration method, the average concentration of DTM radionuclides will be applied independently of the concentration of gamma-emitting radionuclides. It is crucial to perform radioactive decay correction based on the date of generation or disposal when applying scaling factors or average radioactivity concentration. Additionally, for repackaged 320 L drums, determining which drum among the two 200 L drums inside should serve as the reference is of utmost importance

For the disposal of radioactive waste from nuclear facilities, assessing their radioactivity inventories is essential. As a result, countries with nuclear facilities are implementing assessment schemes tailored to their respective policies and available resources for radioactive waste management. This paper specifically describes the assessment scheme for radioactivity inventory applied to metal waste generated during the dismantling of the Japan Power Demonstration Reactor (JPDR), a 1.25 MW BWR. The distinctive aspect of the Japanese approach lies in the fact that, for a pair of a key nuclide and a difficult-to-measure (DTM) nuclide that lack a significant correlation in their concentrations, the mean activity concentration method was used. In this method, an arithmetic average of all measurements of the DTM nuclide from representative drums, including MDAs (Minimum Detectable Activities), was assigned to the concentration of the DTM nuclide for all drums, regardless of the concentration of its paired key nuclide. Conversely, for a specific pair of a key nuclide and a DTM nuclide with a significant correlation, the scaling factor method was applied, as is common in many other countries. This Japanese case can serve as a valuable reference for Korea, which does not have the option of using the mean activity concentration method in its assessment scheme.

To safely dispose of highly radioactive spent resin and concentrate waste generated through nuclear power plant operations, it is essential to meet the physicochemical properties requirements of the packages and ensure the accuracy and reliability of radiological characteristics determination. Both spent resin and concentrate are packaged in high-integrity containers (HICs) after drying and are homogeneous waste products generated in the primary system and liquid radioactive waste treatment system. Meeting the physicochemical properties requirements does not appear to be difficult. However, to achieve reliable radiological characterization of high-integrity container packages, it is necessary to take a representative sample and perform accurate radiological analysis. Therefore, this paper discusses the methodology for evaluating the radionuclide inventory of high radioactive resin and concentrate packages, as well as the essential element technology and considerations. For relatively high radioactive resin and concentrate packages, the radionuclide inventory for each package should be evaluated with high reliability through direct radiological analysis of the representative samples collected for each package. This can contribute to the efficient operation of radioactive waste disposal facilities. Radionuclide-specific concentrations directly analyzed for each package will be managed in a database. As analytical data accumulates and direct measurements of high-integrity container package such as the radwaste drum assay system (RAS) become feasible, statistical techniques such as correlation analysis between easy-tomeasure (ETM) nuclides and difficult-to-measure (DTM) nuclides can lead to the development of efficient and reasonable indirect evaluation methods, such as scaling factor and the mean activity concentration method. As for the element technology, a remote representative sampling technique should be developed to safely and effectively take representative samples of highly radioactive materials, including granulated or hardened concentrate waste. Considerations should also be given to determining the sample quantity representing each package, as well as establishing radiation calibration and measurement methods appropriate to the radiation levels of the representative samples.

To address the pressing societal concern in Korea, characterized by the imminent saturation of spent nuclear fuel storage, this study was undertaken to validate the fundamental reprocessing process capable of substantially mitigating the accumulation of spent nuclear fuel. Reprocessing is divided into dry processing (pyro-processing) and wet reprocessing (PUREX). Within this context, the primary focus of this research is to elucidate the foundational principles of PUREX (Plutonium Uranium Redox Extraction). Specifically, the central objective is to elucidate the interaction between uranium (U) and plutonium (Pu) utilizing an organic phase consisting of tributyl phosphate (TBP) and dodecane. The objective was to comprehensively understand the role of HNO3 in the PUREX (Plutonium Uranium Redox Extraction) process by subjecting organic phases mixed with TBPdodecane to various HNO3 concentrations (0.1 M, 1.0 M, 5.0 M). Subsequently, the introduction of Strontium (Sr-85) and Europium (Eu-152) stock solutions was carried out to simulate the presence of fission products typically contented in the spent nuclear fuel. When the operation proceeds, the complex structure takes the following form. 􀜷􀜱􀬶 􀬶􀬾(􀜽􀝍) + 2􀜰􀜱􀬷 􀬿(􀜽􀝍) + 2􀜶􀜤􀜲(􀝋􀝎􀝃) ↔ 􀜷􀜱􀬶(􀜰􀜱􀬷)􀬶 ∙ 2􀜶􀜤􀜲(􀝋􀝎􀝃) Subsequently, separate samples were gathered from both the organic and aqueous phases for the quantification of gamma-rays and alpha particles. Alpha particle measurements were conducted utilizing the Liquid Scintillation Counter (LSC) system, while gamma-ray measurements were carried out using the High-Purity Germanium Detector (HPGe). The distribution ratio for U, Eu (Eu-152), and Sr (Sr-84) was ascertained by quantifying their activity through LSC and HPGe. Through the experiments conducted within this program, we have gained a comprehensive understanding of the selective solvent extraction of actinides. Specifically, uranium has been effectively separated from the aqueous phase into the organic phase using a combination of tributyl phosphate (TBP) and dodecane. Subsequently, samples containing U(VI), Eu(III), and Sr(II) underwent thorough analysis utilizing LSC and HPGe detectors. Our radiation measurements have firmly established that the concentration of nitric acid enhances the selective separation of uranium within the process.

In this research, the dose rate was measured using a backpack-type scan survey device at 4 sites in sites around Nuclear Power Plants (Kori, Wolsong, Hanbit, Hanul), and the radioactivity ratio for each nuclide was evaluated using an high-purity germanium (HPGe) detector. Kori, Wolsong and Hanul power plants were measured within 2 km of the power plant, and Hanbit power plants were measured about 6.7 km from the power plant. As a result of measuring the dose rate with a backpacktype scan survey device, the average dose rate was the lowest in the measurement site 1 at 0.090 μSv/h, and the highest in the measurement site 4 at 0.145 μSv/h. All measurement points showed the domestic environmental dose rate level. The data obtained by the scan survey was visualized using the classed post and gridding functions of the surfer program. As a result of measurement with the HPGe detector, 137Cs was not detected, and only natural nuclides were detected. Among the detected natural nuclides, the radioactivity ratio was the highest for 40K with an average of 94.56%, and the lowest for 214Pb with an average of 0.26%. The results of this research can be used as basic data for radiation environment surveys around nuclear power plants. Further studies are needed to evaluate the radiation impacts by region and environment through periodic measurements.

Radiation workers who handle radioisotopes, radioactive waste, nuclear material etc. may be contaminated with radioactive material due to inhalation, resulting in internal radiation exposure. For preventing radiation damage and monitoring the exposure of workers, KAERI operates a Body Radiation Measurement Laboratory. According to Article 5 of the Nuclear Safety and Security Commission (NSSC) Notice No. 2017-77, “Regulation on Measurement and Calculation of Internal Radiation Dose,” The nuclear energy-related business operator with workers etc. shall establish and operate procedures and methods including the following Subparagraphs to secure the reliability of measurement of the internal radiation dose : operation and calibration of measuring instrument, inspection procedures, uncertainty of measurement, lower limit of detection and geometric configuration used for measurement. In accordance with the provision, Whole Body Counter utilized in the Body radiation Measurement Laboratory has periodic calibration / QA procedures to ensure reliability. This paper performed reliability validation of the measurement system of the Body Radiation Measurement Laboratory in the KAERI based on the performance criteria for radio-bioassay criteria presented in ISO 28218 and ANSI HPS N13.30-2011(R2017). The first criteria is MTL (Minimum Testing Level). ISO 28218 provides MTLs for each measurement category, type and nuclide. For reliable results, it is recommended to use calibration sources with higher radioactivity than the values given. The MTL for fission products in total body counting is 3 kBq and for the last 3 years the laboratory has been using sources of 6-7 kBq (Co-60, Cs-137 etc.). The second criteria is RMSE (Root Mean Square Error). It is a measure of total error defined as the square root of the sum of the square of the relative precision (SB) and the square of the relative bias (Br). The RMSE shall be lower than or equal to 0.25. The largest RMSE in the last 3 years is 0.12, and average value is 0.065, which meets the criteria. In this study, we verified the reliability of the radioactivity measurement system (WBC) based on the radio-bioassay standards presented in ISO 28218 and ANSI HPS N13.30-2011(R2017). The values were obtained using 3 years of calibration count data, and it was found that both MTL, RMSE for each nuclide met the standards with a large margin of error and were in good operating condition. This study can be applied to the maintenance, performance check, and reliability verification of similar in vivo radio-bioassay methods.

In this study, radioactivity of Cs-134, Cs-137, and Eu-154, which are gamma-emitting nuclides among fission products of spent fuel, was analyzed as a tool to measure the burnup of spent fuel nondestructively. This nuclide has a unique gamma-ray energy, allowing the amount of the isotope to be estimated based on the intensity of the gamma-ray at a specific energy. The SCALE 6.2 ORIGAMI (ORIGen AsseMbly Isotopics) module and the latest ORIGEN-arp library were used for computational analysis. The spent fuel samples were selected as WH14×14 with an enrichment of 1.5~5.0wt%, a burnup of 10~60 GWD/MTU, and a cooling time of 0~40 years. The analysis results were benchmarked using SFCOMPO experimental data provided by OECD/ NEA, including isotope inventory and uncertainty measured by destructive radiochemical analysis, fuel assembly design data required for benchmark model development, reactor design information, and operating history information. 16 similar spent fuels were selected from SFCOMPO data and the calculation results of Cs-134, Cs-137, and Eu-154 were compared. The average error of the Cs-134 radioactivity calculation result was 2.81%, and the maximum error was 6.70%. The average errors of Cs-137 and Eu-154 were 2.42% and 4.95%, respectively, and the maximum errors were 5.40% and 14.91%, respectively.

n this research, the dose rate was measured using backpack-type scan survey device at 4 sites on Jeju Island, and the radioactivity ratio for each nuclide was evaluated using an high-purity germanium (HPGe) detector. As a result of measuring the dose rate with a backpack-type scan survey device, the average dose rate was the lowest in the measurement site 3 at 0.049 Sv/h, and the highest in the measurement site 1 at 0.066 Sv/h. The average dose rate of the 4 sites on Jeju Island was 0.056 Sv/h, and the dose rate on Jeju Island was lower than that of other regions. The data acquired by scan survey were interpreted using classed post and gridding function of surfer program. The radioactivity ratio of each nuclide in the gamma spectrum measured by the HPGe detector was the highest for K-40 with an average of 87.62%, and the lowest for Pb-214 with an average of 0.63%. In the case of the Jeju Island site, Cs-137 was detected, and the average radioactivity ratio of Cs-137 was 3.27%, which was the background level. The results of this research can be used as basic data on the radioactivity ratio for each nuclide and dose rate at the Jeju Island site. Further studies on the assessment of dose rates and radioactivity ratios in other regions are needed.

Discharge limits for nuclear power plant gaseous effluents are presented as dose constraints or on the basis of radioactivity or radioactivity concentration. Accordingly, the operator evaluates the amount of radioactive material discharged from a specific nuclear power plant to the environment and periodically reports them to regulators. Multi-step sampling and analysis and calculation are performed during the radioactivity evaluation process of radioactive effluent, and the uncertainty generated in each step causes the uncertainty of the final radioactivity. Considering that the purpose of evaluating radioactivity discharged from nuclear power plants to the environment is to verify the satisfaction of discharge limits and safety margins, it is necessary to accurately evaluate the discharged radioactivity as much as possible, understanding of the uncertainty contained in the reported value of radioactivity and efforts to reduce it. In this study, modelling of the radioactivity evaluation procedure in gaseous effluent discharged as batch mode from nuclear power plant has performed, a generalized framework was established to evaluate the uncertainty based on ISO/IEC Guide 98-3 (GUM: 1995) involved in the whole process, and the uncertainty contained in the calculated radioactivity of each radionuclide (group) was evaluated and its characteristics. In addition, through probabilistic evaluation, the actual probabilistic distribution and statistical characteristics of radioactive effluent releases reported as a single value were confirmed. As a result, the range of values expected to be included in the confidence level of approximately 95% of the distribution of values for radioactivity in a gaseous effluent discharged as a batch mode from nuclear power plant was calculated. And, the priority of each input parameter turned out to be (1) gaseous waste volume, (2) sample bottle volume, and (3) measured radioactivity of the sample. In addition, the probability distribution of the radioactivity was simulated by Monte Carlo method. As such, the mean, minimum, and maximum values in confidence level of 95% were obtained, and they were reasonably matched the calculated value within 5% deviation. It was shown that radioactivity to the environment, which has been reported as a single value, has a specific probabilistic distribution form.

Reliable evaluation of radioactivity inventory for the nuclear power plant components and residual materials is very important for decontamination and decommissioning. This can make it possible to define optimum dismantling approaches, to determine radioactive waste management strategies, and to estimate the project costs reasonably. To calculate radioactivity of the nuclear power plant structure, various information such as interest nuclide, cross-section, decay constant, irradiation time, neutron flux, and so on is required. Especially irradiation time and neutron flux level are very changeable due to cycle specific fuel loading pattern, the plant overhaul, cycle length. However most of the radioactivity calculations have generally been performed assuming one representative or average neutron flux during the lifetime of the nuclear power plant. This assumption may include excessive conservatism because the radioactivity level has the characteristics of saturation and decay. Therefore, considering these variables as realistically as possible could prevent overestimation. In order to perform realistic radioactivity calculation, we developed monthly relative power contribution factor applying plant-specific operation history and cycle-specific neutron flux. The factors were applied to the radioactivity calculation. The calculation results ware compared with measured values of the neutron monitors that were actually installed and withdrawn from the nuclear power plant. As a result of the comparisons, there are good agreements between the calculated values and measured values. These accurate calculation results of radioactivity could contribute to the establishment of radioactive waste dismantling strategies, the classification of radioactive waste, and the deposit of disposal costs for safe and reasonable decommissioning of the nuclear power plant.

For the transport of spent nuclear fuel, it is necessary to evaluate the amount of radioactivity for each assembly and the total amount of radioactivity for each cask. Currently, KHNP is evaluating the radioactivity using the Express mode of the OrigenArp program in the SCALE6.1 code. Express mode is a method to evaluate the radioactivity assuming that it has been burned with the same power per cycle, and Detail mode is a method to evaluate the actual combustion history such as power and cooling time for each cycle. For a total of 3,795 assemblies, including 1,391 assembliess for Kori Unit 1, 1,427 assemblies for Hanbit Unit 2, and 977 assemblies for Hanul Unit 3, the radioactivity was evaluated in Express mode and Detail mode, respectively, and the results were compared. As a result of the evaluation, it was confirmed that the results of the Express mode were evaluated more conservatively by 2.5~12.9% than that of the Detail mode. Accordingly, KHNP established a plan to change the evaluation method from Express mode to Detail mode in order to improve the accuracy of the radioactivity assessment results and eliminate conservatism.

As Kori-1 permanently shut down in Korea, it is expected that a large amount of radioactive waste will be generated during decommissioning of nuclear power plants. Radioactive concrete waste is contaminated up to depth of 100 mm with radionuclides such as 137Cs and 60Co. The radioactive waste should be accurately classified to reduce the cost of disposing of radioactive waste. Therefore, the specific radioactivity of waste must be precisely evaluated by gamma-ray measurements emitted from the radionuclides. In general, the effectiveness of the radioactivity measurement and process is confirmed using certified reference material (CRM) composed of water or agar. However, the decommissioning waste differs from this CRM in apparent density and chemical elements, so the specific radioactivity is underestimated or overestimated. Therefore, reference material composed of the same apparent density and chemical elements as the sample is required to improve the quality of radioactivity measurement. The purpose of this study is to develop a concrete reference material for the nuclear decommissioning waste. The concrete reference material composed of SiO2, CaO, and Al2O3 were manufactured in compliance with ISO Guide 35. 10 bottles were randomly selected for homogeneity test, and 2 samples for analysis were taken from each bottle. The specific radioactivity was measured using an HPGe detector with an efficiency of 30%. The results of the homogeneity test of 137Cs and 60Co satisfied the requirements of ISO Guide 35. Coincidence summing and selfabsorption effects were corrected using the Monte Carlo efficiency transfer code and Monte Carlo NParticle transport code. The reference values of 137Cs and 60Co in the concrete reference material were evaluated in the range of 1,000–1,100 Bq·kg−1 and extended uncertainty was around 7%.

The purpose of full system decontamination before decommissioning a nuclear power plant is to reduce radiation exposure of decommissioning workers and to reduce decommissioning waste. In general, full system decontamination removes the CRUD nuclides deposited on the inner surface of the reactor coolant system, chemical and volume control system, residual heat removal system, pressurizer, steam generator tube, etc. by chemical decontamination method. The full system decontamination process applied to Maine Yankee and Connecticut Yankee in the USA, Stade, Obrigheim, Unterweser, Nekawestheim Unit 1 in Germany, Mihama Unit 1 and 2 in Japan, Jose Cabrera Unit 1 in Spain, and Barseback Unit 1 and 2 in Sweden are HP/CORD UV, NP/CORD UV, and DfD. In this study, the quantity of 60Co radioactivity removal, metal removal, ion exchange resin and filter generation according to reactor power, surface area and volume of the full system decontamination flow path, and the decontamination process were compared and analyzed. In addition, the quantity of 60Co radioactivity removal by each nuclear power plant was compared and analyzed with the evaluation results of the 60CO radioactivity inventory of the Kori Unit 1 full system decontamination loops conducted by SAE-AN Enertech Corporation.

A method of quantitatively analyzing radioactivity of uranium waste in the In-situ measurement using Bayesian inference was proposed. When applying the traditional efficiency calibration method, which uses standard sources or Monte Carlo simulation, the radioactivity error is large depending on the degree of spread of the radioactive contamination especially in large sample such as a 200 L drum. In addition, the existing method has a limitation in that it is difficult to reflect the uncertainty according to the location of the source. In this preliminary study, to overcome the limitations of the existing method, a Bayesian statistical-based radioactivity quantitative analysis model was proposed that can increase the accuracy of analysis even in situations where radioactive contamination of uranium waste is non-uniformly distributed. As a result of evaluating the simulated waste with the proposed Bayesian method, the accuracy was improved more than about 6 times compared to the classical efficiency calibration method.