In Korea, extensive industry-academia-research research has already established many facilities and technologies for materials and chemical experiments on non-radioactive substances. However, few facilities have been built to analyze the physical and chemical properties of substances irradiated through neutron irradiation. Korea is planning to decommission Kori-1 and Wolsong-1 in 2027. Extensive analysis of low-level and intermediate-level materials is required to begin decommissioning these nuclear power plants. The material’s composition and level can be identified by analyzing the structure’s characteristics, and a cutting and decontamination plan can be established based on this. In addition, by conducting a nuclide analysis on the waste generated after cutting, suitability for disposal can be secured, and stable treatment can be performed. Accordingly, the Korea Decommissioning Research Institute (KRID) plans to secure infrastructure (hot cells) to analyze the characteristics of intermediate-level decommissioning waste. The goal is to secure high-dose/high-radiation decommissioning waste processing technology through Korea’s first intermediate-level hot cell, support domestic nuclear power plant decommissioning projects, and secure and verify procedures related to nuclide analysis of intermediate-level using hot cells. In addition, by possessing these material properties and nuclide analysis technology, KRID can have a foundation to conduct continuous research on low- and intermediate-level radioactive materials and decommissioning. The purpose of KRID’s establishment is to use this foundation in the future to improve the technological level of the nuclear industry as a whole through linkage between industry, academia, and research institutes.

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.

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.

The ultimate objective of deep geological repositories is to achieve complete segregation of hazardous radioactive waste from the biosphere. Thus, given the possibility of leaks in the distant future, it is crucial to evaluate the capability of clay minerals to fulfill their promising role as both engineered and natural barriers. Selenium-79, a long-lived fission product originating from uranium- 235, holds significant importance due to its high mobility resulting from the predominant anionic form of selenium. To investigate the retardation behaviors of Se(IV) in clay media by sorption, a series of batch sorption experiments were conducted. The batch samples consisted of Se(IV) ions dissolved in 0.1 M NaCl solutions, along with clay minerals including kaolinite, montmorillonite, and illite-smectite mixed layers. The pH of the samples was also varied, reflecting the shift in the predominant selenium species from selenious acid to selenite ion as the environment can shift from slightly acidic to alkaline conditions. This alteration in pH concurrently promotes the competition of hydroxide ions for Se(IV) sorption on the mineral surface as the pH increases and impedes the selective attachment of selenium. The acquired experimental data were fitted through Langmuir and Freundlich sorption isotherms. From the Freundlich fit data, the distribution coefficient values of Se(IV) for kaolinite, montmorillonite, and illite-smectite mixed layer were derived, which exhibited a clear decrease from 91, 110, 62 L/kg at a pH of 3.2 to 16, 6.3, 12 L/kg at a pH of 7.5, respectively. These values derived over the pH range provide quantitative guidance essential for the safety assessment of clay mineral barriers, contributing to a more informed site selection process for deep geological repositories.

Molten chloride salts have received considerable research attention as potential nuclear fuel and coolant candidates for molten salt reactors. However, there are several challenges, especially for structural materials due to the selective dissolution of chromium (Cr) in the molten chloride salts environment. Understanding the compatibility of uranium (U), which is used as nuclear fuel in molten salt reactors, with Cr in molten chloride salts is critical for designing the molten salt reactor structure. Therefore, in this study, the cyclic voltammetry (CV) was used to investigate the electrochemical behaviors of U and Cr. The diffusion coefficients and formal potentials were obtained. The electrochemical properties of uranium and chromium were investigated by CV in molten NaCl-MgCl2 salt at 600°C. Tungsten rods for working and counter electrode, and Ag/AgCl for reference electrode were utilized in this experiment. UCl3 made from the chemical dissolution of U rods and CrCl2 (Sigma-Aldrich, 99.99%) were used. Diffusion coefficients (D) of U and Cr were calculated by measuring reduction peak current of U3+/U and Cr2+/Cr from CV curves and using the Berzins-Delahay equation; D (U3+/U) = 3.0×10-5 cm2s-1 and D (Cr2+/Cr) = 3.3×10-5 cm2s-1. The formal potentials were also calculated by using the reduction peak potential obtained from CV results; E0’ (U3+/U) = -1.173 V and E0’ (Cr2+/Cr) = -0.321 V. The ionization tendency was investigated by comparing each reduction peak potential. The reduction peak potential Ep,c was increasing order of Ep,c (U3+/U) < Ep,c (Cr2+/Cr) < Ep,c (U4+/U3+). It can be seen that in the presence of U4+ and Cr metals, the Cr in the alloy can dissolve into Cr2+, but in the presence of U3+ and Cr metals, the Cr in the alloy does not dissolve into Cr2+. By analyzing the CV curve, diffusion coefficients and formal standard potentials were obtained. The result of comparing reduction peak potentials suggests that the nuclear fuel using U4+ should be inhibited to prevent the selective dissolution of Cr.

A comprehensive understanding of actinide coordination chemistry and its structure is essential in many aspects of the nuclear fuel cycle, such as fuel reprocessing, waste management, reactor safety, and non-proliferation efforts. Managing radioactive waste generated during the nuclear fuel cycle has recently become more important, accordingly increasing the importance of designing appropriate waste forms and storage solutions for long-term waste disposal. Compared to the increase in the need for understanding the chemistry of major radioactive elements, the information on the local structure of the radioactive elements, especially actinides, remains unknown. To probe this issue, X-ray absorption fine structure (XAFS) can be applied. By analyzing the EXAFS (extended X-ray absorption fine structure) and XANES (X-ray absorption near edge structure), the local structure around atoms can be determined. However, the radioactive properties of the nuclides hindered the measurement of EXAFS and XANES, due to the difficulties of preparation, containment, and transfer of the sample. To measure the EXAFS of various compounds regarding the back-end nuclear fuel cycle, laboratory-based EXAFS (hiXAS, HP spectroscopy) has been introduced which can measure the EXAFS and XANES at the energy range of 5-18 keV. Compounds of Copper (Cu foil, CuO samples), Zirconium (Zr foil), and Europium (Eu2O3) were used for the verification of the laboratory -based EXAFS at a given energy range. The measured EXAFS spectrum of various compounds exhibit good agreement with the theoretical data, showing an R-factor of less than 0.02. It was found that each graph has a first peak corresponding to 2.55Å for Cu foil (Cu-Cu), 1.93Å for CuO samples (Cu-O), 3.23Å for Zr foil (Zr-Zr), and from 2.32Å to 2.34Å for Eu2O3 (Eu-O), which agree well with other values from the literature. From the result, it can be implied that this equipment can be used especially in the back-end nuclear fuel cycle field to enhance the understanding of local structure in radiochemistry.

Bis (2-ethylhexyl)phosphoric acid (HDEHP) is a renowned extractant, favored for its affinity to selectively remove uranium via its P=O groups. We previously synthesized HDEHP-functionalized mesoporous silica microspheres for solid-phase uranium adsorption. Herein, we investigated the kinetic and isothermal behavior of uranyl ion adsorption in mesoporous silica microspheres functionalized with phosphate groups. Adsorption experiments were conducted by equilibrating 20 mg of silica samples with 50 mL of uranium solutions, with concentrations ranging from 10 to 100 mgU L−1 for isotherms and 100 mgU L−1 for kinetics. Three distinct samples were prepared with varying HDEHP to TEOS molar ratios (x = 0.16 and 0.24) and underwent hydrothermal treatment at different temperatures, resulting in distinct textural properties. Contact times spanned from 1 to 120 hours. For x = 0.16 samples, it took around 50 and 11 hours to reach equilibrium for the hydrothermally treated samples at 343 K and 373 K, respectively. Adsorbed quantities were similar (99 and 101 mg g-1, respectively), indicating consistent functional group content. This suggests that the key factor influencing uranium adsorption kinetics is pore size of the silica. The sample treated at 373 K, with a larger pore size (22.7 nm) compared to 343 K (11.5 nm), experienced less steric hindrance, allowing uranium species to diffuse more easily through the mesopores. The data confirmed the excellent fit of pseudo-second-order kinetic model (R2 > 0.999) and closely matched the experimental value, suggesting that chemisorption governs the rate-controlling step. To gain further insights into uranium adsorption behavior, we conducted an adsorption isotherm analysis at various initial concentrations under a constant pH of 4. Both the Langmuir and Freundlich isotherm models were applied, with the Langmuir model providing a superior fit. The relatively high R2 value indicated its effectiveness in describing the adsorption process, suggesting homogenous sorbate adsorption on an energetically uniform adsorbent surface via a monolayer adsorption and constant adsorption site density, without any interaction between adsorbates on adjacent sites. Remarkably, differences in surface area did not significantly impact uranium removal efficiency. This observation strongly suggests that the adsorption capacity is primarily governed by the loading amount of HDEHP and the inner-sphere complexation with the phosphoryl group (O=P). Our silica composite exhibited an impressive adsorption capacity of 133 mg g-1, surpassing the results reported in the majority of other silica literature.

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.

As existing nuclear power plants reach the end of their lifespan, 22 nuclear power plants in korea are scheduled to be permanently shut down and decommissioned by 2050. Chelates are used as decontamination agents during nuclear power plant operation and decommissioning, and as a result, decommissioning waste contains chelates. Chelates contained in radioactive waste are complexed with radionuclides and increases their mobility. So, qualitative and quantitative analysis of chelates contained in radioactive waste is necessary. However, the spectroscopic method (UVVis), previously used for chelate analysis in Korea takes too much time for analysis and cannot analyze two or more chemically similar chelates at the same time. Due to these problems, new methods for analyzing chelate must be developed. Overseas, many cases of chelate analysis using advanced analysis equipment have been reported. CEA in France has developed a chelate analysis method for application to radioactive waste using HPLC-MS (J. Chromatogram. A, 1276, 20-25, 2013). In this method, the existing method of measuring EDTA using a complex of Fe and EDTA was improved to measuring a complex of Ni and EDTA. Based on such overseas cases, we would like to develop an analysis method for chelates in radioactive waste. For this purpose, we will verify similar overseas papers and develop pretreatment methods for mixtures of chelates (EDTA, DTPA, NTA) and metals (Fe, Ni, Cu, etc.) in various media. Finally, we will develop a separation analysis technology for multi-component chelates in nuclear decommissioning waste based on LCMS.

Chelating agents in low and intermediate radioactive wastes can form complexes with radionuclides and increase the mobility of the radionuclides. According to the Korea Radioactive Waste Agency (Acceptance criteria for low and intermediate radioactive waste, WAC-SIL-2022-1), if the amount of residual chelating agents in the waste are greater than 0.1%, the chemical names and residual amounts should be specified; if greater than 1%, the waste must be solidified and contain no more than 8%. The existing method for analyzing chelates in radioactive waste was based on UV–Visible spectrophotometry (UV-Vis), but the new method is based on liquid chromatography/mass spectrometry (LC-MS). The analysis was performed in aqueous solution before applying to real samples. Since the real sample may contain several heavy metals, it is expected that the chelates will exist as complexes. Therefore, 1.0×10-4 mol L-1 of EDTA (Ethylenediaminetetraacetic acid), DTPA (Diethylenetriaminepentaacetic acid), NTA (Nitrilotriacetic acid), and excess metals in aqueous solution were analyzed using HPLC using RP (Reverse Phase) column and HILIC (Hydrophilic interaction) column. When the RP column was used, each substance eluted without separation at the beginning of the analysis. However, when analyzed using a HILIC column, the peaks of each substance were separated. LC-MS measurements using HILIC conditions resulted in separations with better sensitivity.

In the decommissioning process of nuclear power plants, Ni-59, Ni-63 and Fe-55 present in radioactive waste are crucial radionuclides used as fundamental indicators in determining waste treatment methods. However, due to their low-energy emissions, the chemical separation of these two radionuclides is essential compared to others. Therefore, this study aims to evaluate the suitability of various pre-treatment methods for decommissioning waste materials by conducting characteristic assessments at each chemical separation stage. The goal is to find the most optimized pre-treatment method for the analysis of Ni-59, Ni-63 and Fe-55 in decommissioning waste. The comparative evaluation results confirm that the chemical separation procedures for Fe and Ni are very stable in terms of stepwise recovery rates and the removal of interfering radionuclides. However, decommissioning waste materials, which mainly consist of concrete, metals, etc., possess unique properties, and a significant portion may be low-radioactivity waste suitable for on-site disposal. Considering that the chemical behavior and reaction characteristics may vary at each chemical separation stage depending on the matrix properties of the materials, it is considered necessary to apply cascading chemical separation or develop and apply individual chemical separation methods. This should be done by verifying and validating their effectiveness on actual decommissioning waste materials.

I-129 is one of the imporant nuclides that must be determined in the disposal process of radioactive waste in many countries. This radionuclide emits gamma-ray and x-ray photons within the energy range of 29 to 39 keV, consequently, an x-ray detector with high resolution performance is required for the analysis of I-129 activity. An n-type coaxial HPGe detector exhibits higher efficiency characteristics compared to a planar-type HPGe detector, however, its resolution is lower than a planar type. So it is difficult to completely deconvolute and fit the gamma-ray and xray peaks in the spectrum using a general gamma-ray spectrum analysis program such as GammaVision. To address this problem, in a previous study introduced the developed algorithm for the fitting and analysis of I-129 gamma-ray and x-ray spectum by fixing their emission ratios. In this study, we improved the algorithm by considering the variation of the efficiency in the HPGe spectrum, which reflects the actual HPGe crystal condition. And algorithm tests were performed using measured I-129 sample spectra with interfering nuclides acting as background curve are introduced.

For safe disposal of radioactive wastes, accurate analysis of nuclear isotopes is important. It is known that there are 14 nuclides that have to identify nuclide-specific concentration levels. 63Ni, one of non-volatile nuclear isotopes which is included in those 14 nuclides, has to follow chemical separation for exact analysis. As various analysis methods were developed, various methods for analyzing 63Ni also emerged. Past method has used measurement specimens of 59Ni, after 59Ni measurement has been done. It used HClO4, known as strong oxidizing agent, to dissolve DMG, an organic substance used to form 59Ni precipitates. Nowadays, we analyze 59Ni and 63Ni simultaneously, which enables short analysis time, without use of HClO4. But high accuracy is just as important as short measurement time and efficiency. So, this paper compare 63Ni specific activity value used new method with the value, past method used, using real sample’s data. As a result, all sample data from new method’s relative 63Ni specific activity is within the uncertainty range of past ones based on past specific activity value. Consistency of new method’s result and past method’s data increased the reliability of the data and accuracy of those methods.

Currently, non-volatile nuclides such as 94Nb, 99Tc, 90Sr, 55Fe, and 59/63Ni are used a sequential separation. In this study, we developed a separation for 99Tc and 90Sr by a carbonate precipitation. Sodium Carbonate (Na2CO3) was inserted in the aqueous sample from a Dry Active Waste (DAW) and a carbonate precipitation was produced. The precipitate is composed of di- or tri-valent element such as Co, Sr, Fe, Ni and the supernatant is composed of mono-valent element (Cs) and anion materials (ReO4 -, TcO4 -). In DAW, it was confirmed that the recovery of 90Sr (precipitate) and 99Tc (supernatant) were > 90%, respectively. The precipitate and supernatant separated by using a Sr-resin and an anion-exchange resin, respectively. The separated samples were measured by a Liquide Scintillation Counter (LSC, 90Sr) and Induced-Coupled Plasma-Mass Spectroscopy (ICPMS, 99Tc).

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%.

Molten Salt Reactor (MSR) is one of the 4th generation nuclear power systems which is its verified technology in physically and chemically. Among the various salts used for MSR system, the eutectic composition of NaCl-MgCl2 system maintains the liquid state at around 450°C, in the same time, it has high solubility for nuclear fuel chlorides. This characteristic has high advantage for lowering the operating temperature for the MSR, which could reduce the problem of hightemperature corrosion by salt for structural materials significantly. In particular, since MgCl2 has the similar standard reduction potential with nuclear fuel, is used as a surrogate for, many basic researches have been conducted for verifying characteristic of MgCl2. It is well-known that main short-advantage of MgCl2 is hygroscopic properties. MgCl2 changes to MgCl2-xH2O state easily by absorbing moisture in air condition. The hydrated MgCl2 is producing MgOHCl by thermally decomposing at high temperature, the formed MgOHCl corrodes structural materials, even small amount of MgOHCl gives significant damage. Therefore, the purification of MgCl2 has been required for long-term operation of MSR using MgCl2 as a base salt. In this study, the purification of eutectic composition salt for NaCl-MgCl2 has been mainly performed by considering its thermodynamic properties and electrochemical characteristic, and the experimental results have been discussed.

Most of the C-14 produced is in the organic form, generated as methane (14CH4), methanol (14CH3OH), formaldehyde (14CH2O), and formic acid (14CO2H2). When analyzing C-14, it is transformed into the form of 14CO2, and its concentration is determined using LSC. Typical examples include the wet oxidation method, the combustion or Pyrolysis. The wet oxidation method uses strong acids and involves repeated operations, which generates large amounts of acid waste and secondary radioactive waste. The combustion method uses high temperatures, which requires an oxygen device. Pyrolysis also requires high temperature in a vacuum and catalysts. Catalysts are expensive because they are platinum-based. To compensate for these shortcomings, a C-14 analysis method using UV irradiation was developed. In this study, 100 mL of distilled water mixed with formic acid (CO2H2), potassium persulfate (K2S2O8), and silver nitrate (AgNO3) was irradiated with a 320-390 nm UV lamp to conduct a CO2 production reaction experiment. The UV range was measured using a photometer (UV Power puck II). The beaker was made of quartz in 150 mL size with three inlets : a temperature measurement, a sample inlet, and a collection tube connector. We changed the UV lamp used from a 450 W halogen lamp to a 100 W LED, which has a lower temperature and is safer. As a result of the experiment, CO2 bubbles were generated in the collection tube, due to the UV irradiation react, which uses oxidizer and catalysts. The maximum temperature of the solution irradiated with the LED UV lamp was less than 56°C. It confirmed the rate of bubble generation changed depending on the lamp distance, the amount of sample, oxidizer, and catalyst. In an experiment to confirm the reaction caused by heat, it was found that although a reaction occurred due to heat, the reaction was significantly lower than when using a UV lamp. The reproducibility experiment was conducted three times in total under the same conditions. It showed the same pattern. In the future, we plan to select mock samples, collect 14CO2 in Carbo- Sorb, and analyze them using LSC. The results of this research will be used as a technology to recover C-14 more safely and efficiently and will also be used to expand its application to the treatment of other wastes such as waste liquid and waste resin through simulated samples.

The need for the development of sustainable, efficient, and green radioactive waste disposal methods is emerging with the saturation of spent nuclear waste storage facilities in the Republic of Korea. Conventional radioactive waste management methods like using cement or glass have drawbacks such as high porosity, less chemical stability, high energy consumption, carbon dioxide production, and the generation of secondary wastes, etc. To address this gigantic issue of the planet, we have designed a study to explore the potential of alternative materials having easy processability, low carbon emissions and more chemical stability such as ceramic (hydroxyapatite, HAP) and alkali-activated materials (geopolymers, GP) to capture the simulated radioactive cobalt ions from the contaminated water and directly solidify them at low temperatures. Physical and mechanical properties of HAP alone and 15wt% GP incorporated HAP (HAP-GP- 15) composite were studied and compared. The surface of both materials was fully sorbed with an excess amount of Co(II) ions in the aqueous system. Co(II) sorbed powders were separated from aqueous media using a centrifuge machine operating at 5,000 RPM for 10 minutes and dried at 100°C for 8 hours. The dried powders were then placed in stainless steel molds, and shaped into cylindrical pellets using a uniaxial press at a pressure of 1 metric ton for 1 minute. The pellets were sintered at 1,100°C for 2 hours at a heating rate of 10°C/min. Following this, the water absorption, density, porosity, and compressive strength of the polished pellets were measured using standard methods. Results showed that HAP has a greater potential for decontamination and solidification of Co(II) due to its higher density (2.65 g/cm3 > 1.90 g/cm3), less open porosity (16.2±2.9% < 42.4 ±0.9%) and high compressive strength (82.1±10.2 MPa > 6.9±0.8 MPa) values at 1,100°C compared to that of HAP-GP-15. Nevertheless, further study with different constituent ratio of HAP and GP at various temperatures is required to fully optimize the HAP-GP matrix for waste solidifications.

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.

Noble metal phase, present in used fuel, are fission products that can be found as metallic precipitates in used nuclear fuel. They exist as small particles (nm~um) in grain boundaries of the used fuels. Since they are particles deposited between the grain structures, they can be considered as defects in the pellet structure. Thermal expansion of fuels with noble metal is slightly higher than that of bare fuels. The fuels at high temperature, such as immediately after being discharged from nuclear reactors, may be subject to fuel failure if sufficient cooling is not provided. Recent research has shown that the noble metals can migrate into the rim space between the pellet and the cladding, and be deposited in the inner layer of the claddings. therefore, the mechanical integrity of the cladding can be degraded by noble metals, as well as the pellets. The concentration of the noble metal phase should be considered to evaluate the effect of the noble metals on the fuel integrity, after discharge from the reactors. SCALE/ORIGEN code was used to evaluate the noble metals in fuel assembly-scale, and the radial distribution in the fuel assembly. The radial distribution of the reactor power was derived from the SCALE/TRITON, considering Westinghouse 17×17. Square cell model was chosen for the geometry and 1/4 model was applied to reduce the computation time.