The primary objective of radiological environmental monitoring after a radiological emergency at a nuclear facility is acquisition of background data for the determination of protective actions for the population and the comprehensive assessment of the impact on the population residing in proximity to the nuclear facility. The responsible entities engaged in the conduct of the radiological environmental monitoring encompass government organization and nuclear licensees, operating in strict adherence to the national radiological disaster prevention framework. In accordance with the national radiological disaster prevention framework, radiation environmental monitoring is executed through the deployment of emergency response organization, and recurrent exercise drills aimed at augmenting responsible capabilities. In the context of radiation environmental monitoring, it is necessary to specify measurement parameters, monitoring location, and methodological protocols for each stage, considering potential exposure pathways. In terms of equipment, it is important to utilize mobile assets such as aerial or vehicle surveys for rapid and accurate radiation environment monitoring. Radiation disaster drills are regularly conducted, and the radiation environment monitoring field is also regularly trained to enhance response capabilities. The scale of these drills may vary, ranging from exclusive participation by nuclear licensees to joint exercises conducted by governmental agencies. This iterative process of periodic drills and equipment enhancements has led to a progressive augmentation of environmental monitoring capabilities, ensuring a well-coordinated orchestration of radiation monitoring within the framework of radiation protection. Notwithstanding these achievements, challenges in public communication regarding the decision to take protective actions and the dissemination of information to the public. Considering that the purpose of radiation environmental monitoring extends beyond safeguarding public health; it also serves to alleviate public anxiety. In the future, public communication between these stakeholders should also be included in disaster drill programs to ensure proper consultation between each stakeholder during drills and to build understanding and trust in radiation environmental monitoring. This is expected to improve the quality of radiation environmental monitoring response capabilities.

The Korea Atomic Energy Research Institute (KAERI) has facilities that are operated for the purpose of treating radioactive wastes and storing drums before sending them to a disposal site. Domestic regulations related to nuclear facility require radiological dose assessment resulting from release of gaseous radioactive effluent of nuclear facilities. In this study, ICRP-60-based dose conversion factors were applied to evaluate the radiation dose to residents in the event of operation and accident for the radioactive waste management facilities in KAERI. The radioactive gaseous effluent generated from each facility diffuse outside the exclusion area boundary (EAB), causing radiation exposure to residents. To evaluate the external exposure dose, the exposure pathways of cloudshine and radioactive contaminated soil were analyzed. The internal exposure dose was estimated by considering the exposure from respiration and ingestion of agricultural and livestock products. The maximum individual exposure dose was evaluated to be 1.71% compared to the dose limit. The assumed situation used for accidental scenarios are as follows; A fire inside the facility and falling of radioactive waste drum. It was a fire accident that caused the maximum exposure dose to individual and population living within an 80 km radius of the site. At the outer boundary of the low population zone (LPZ), the maximum effective dose and thyroid equivalent dose were estimated as 8.92 E-06% and 5.29 E-06%, respectively, compared to the dose limit. As a result of evaluating the radiological exposure dose from gaseous emissions, the radioactive waste treatment facilities and its supplementary facilities meet the regulations related to nuclear facility, and are operated safely in terms of radiological environmental impact assessment.

The dismantling nuclear power plant is expected to continue to change the radiation working environment compared to the operating nuclear power plant. Contamination monitors and survey meters currently in use have limitations in accurate analysis source term and dose rates for continuous changes in radiation fields at dismantling sites. Due to these limitations, the use of semiconductor detectors such as HPGe and CZT detectors with excellent energy resolution and portability is increasing. The CZT detector performs as well as the HPGe detector, but there is no proven calibration procedure yet. Therefore, in this study, the HPGe calibration method was reviewed to derive implications for the CZT detector calibration method. The operating principle of a semiconductor detector that measures gamma emission energy converts them into electrical signals is the same. Two calibrations of HPGe detectors are performed according to the standard calibration procedure for semiconductor detectors for gamma-ray measurement issued by the Korea Association of Standards & Testing Organizations. The first is an energy calibration that calculates gamma-ray peak position measurements and relational expressions using standard source term that emit gamma-rays. The channel values for energy are measured using certified reference source term to determine radionuclides by identifying channels corresponding to the measured peak energy values. The second is the measurement efficiency of measuring the coefficient calibration device, which measures gamma rays emitted from the standard source term. The detector efficiency by sample or distance is measured in consideration of the shape, size, volume, and density of the calibration device. The HPGe detector performs calibration once every six months through a verified calibration method and is being used as a source term analyzer at the power plant. The CZT detector may also establish a procedure for identifying peak positions through energy calibration and calculating radioactivity through efficiency calibration. This will be a way to expand the usability of semiconductor detectors and further monitor radiation in a more effective way.

The first commercial operation of Kori-1, which commenced in April 1978, was permanently shut down in June 2017, with plans for immediate dismantling. The decommissioning process of nuclear power plants generates a substantial amount of radioactive waste and poses significant radiation exposure risks to workers. Radioactivity is widely distributed throughout the primary coolant system of the reactor, including the reactor pressure vessel (RPV), steam generator (SG), and pressurizer. In particular, the SG has a considerable size and complex geometry, weighing approximately 326 tons and having a volume of 400 m3. The SG tubes are known to contain high levels of radioactivity, leading to significant radiation exposure to workers during the dismantling process. Therefore, this study aims to evaluate the workers’ radiation exposure during the cutting of SG tubes, which account for approximately 95% of the total radiation dose in the SG. Firstly, the CRUDTRAN code, developed to predict the behavior of soluble and particulate corrosion products in a PWR primary coolant system, is used to estimate the radioactive inventory in the SG tubes. Based on decontamination factors (DF) obtained in the SG tubes through overseas experience, the expected reduction in radioactivity during the Kori-1 reactor’s full-system decontamination (FSD) process is considered in the CRUDTRAN results. These results are then processed to estimate the radioactivity in both the straight and bent sections of the tubes. Subsequently, these radioactivity values are used as inputs for the MicroShield code to calculate the worker radiation exposure during the cutting of both straight and bent sections of the tubes. The cutting process assumes that each SG tube section is cut in a separate, shielded area, and the radiation exposure is assessed, taking into account factors such as cutting equipment, cutting length, working hours, and working distance. This study evaluates the worker radiation exposure during the cutting of SG tubes, which are expected to have a significantly high radioactivity due to chalk river unidentified deposit (CRUD). This assessment also considers the reduction in radioactivity within the steam generator tubes resulting from the FSD process. Consequently, it enables an examination of how worker radiation exposure varies based on the extent of FSD. This study may provide valuable insights for determining the scope and extent of the FSD process and the development of shielding methods during the dismantling of SG tubes in the future.

One of the important components of a nuclear fuel cycle facility is a hot cell. Hot cells are engineered robust structures and barriers, which are used to handle radioactive materials and to keep workers, public, and the environment safe from radioactive materials. To provide a confinement function for these hot cells, it is necessary to maintain the soundness of the physical structure, but also to maintain the negative pressure inside the hot cell using the operation of the heating, ventilation, and air conditioning (HVAC) systems. The negative pressure inside the hot cells allows air to enter from outside hot cells and limits the leakage of any contaminant or radioactive material within the hot cell to the outside. Thus, the HVAC system is one of the major components for maintaining this negative pressure in the hot cell. However, as the facility ages, all the components of the hot cell HVAC system are also subject to age-related deterioration, which can cause an unexpected failure of some parts. The abnormal operating condition from the failure results in the increase of facility downtime and the decrease in operating efficiency. Although some major parts are considered and constructed in redundancy and diversity aspects, an unexpected failure and abnormal operating condition could result in reduction of public acceptance and reliability to the facility. With the advent of the 4th Industrial Revolution, prognostics and health management (PHM) technology is advancing at a rapid pace. Korea Hydro & Nuclear Power, Siemens, and other companies have already developed technologies to constantly monitor the integrity of power plants and are applying the technology in the form of digital twins for efficiency and safety of their facility operation. The main point of PHM, based on this study, is to monitor changes and variations of soundness and safety of the operation and equipment to analyze current conditions and to ultimately predict the precursors of unexpected failures in advance. Through PHM, it would be possible to establish a maintenance plan before the failure occurs and to perform predictive maintenance rather than corrective maintenance after failures of any component. Therefore, it is of importance to select appropriate diagnostic techniques to monitor and to diagnose the condition of major components using the constant examination and investigation of the PHM technology. In this study, diagnostic techniques are investigated for monitoring of HVAC and discussed for application of PHM into nuclear fuel cycle facilities with hot cells.

The nuclear fuel that melted during the Fukushima nuclear accident in 2011 is still being cooled by water. In this process, contaminated water containing radioactive substances such as cesium and strontium is generated. The total amount of radioactive pollutants released by the natural environment due to the nuclear accident in Fukushima in 2011 is estimated to be 900 PBq, of which 10 to 37 PBq for cesium. Radioactive cesium (137Cs) is a potassium analog that exists in the water in the form of cations with similar daytime behavior and a small hydration radius and is recognized as a radioactive nuclide that has the greatest impact on the environment due to its long half-life (about 30 years), high solubility and diffusion coefficient, and gamma-ray emission. In this study, alginate beads were designed using Prussian blue, known as a material that selectively adsorbs cesium for removal and detection of cesium. To confirm the adsorption performance of the produced Prussian blue, immersion experiments were conducted using Cs standard solution, and MCNP simulations were performed by modeling 1L reservoir to conduct experiments using radioactive Cs in the future. An adsorption experiment was conducted with water containing standard cesium solution using alginate beads impregnated with Prussian blue. The adsorption experiment tested how much cesium of the same concentration was adsorbed over time. As a result, it was found that Prussian blue beads removed about 80% of cesium within 10-15 minutes. In addition, MCNP simulation was performed using a 1 L reservoir and a 3inch NaI detector to optimize the amount of Prussian blue. The results of comparing the efficiency according to the Prussian volume was shown. It showed that our designed system holds great promise for the cleanup and detection of radioactive cesium contaminated seawater around nuclear plants and/or after nuclear accidents. Thus, this work is expected to provide insights into the fundamental MCNP simulation based optimization of Prussian blue for cesium removal and this work based MCNP simulation will pave the way for various practical applications.

The cyclotron is an apparatus used for the production of radioactive isotopes through nuclear reactions, resulting in the inevitable emission of neutrons. Consequently, surrounding components become activated. The purpose of this study was to investigate the radiological characteristics of Havar foil, a periodic replacement part of the Targetry system. In this study, radioactivity and radiation dose were calculated based on the time of Havar foil replacement and equipment dismantling. The time to dismantle the equipment was set at one year after the equipment was shut down, based on the recently used 1g of Havar foil. All simulations were performed using the FLUKA program. First, in the simulation results, 11 elements (Re, W, Tc, Nb, Cu, Ni, Co, Fe, Mn, Cr, V) were converted into 36 radioisotopes by activation based on the replacement period. Based on radioactivity levels, major isotopes included 52Mn (77.63%), 56Co (13.36%), 96Tc (2.4%), and 95Tc (1.80%). Based on radiation dose rates, 52Mn (82.66%) and 56Co (13.45%) exhibited the highest levels. Furthermore, the dose rates at distances of 10 cm, 50 cm, and 100 cm were found to be 1.36E+1 mSv/hr, 2.24E+00 mSv/hr, and 8.80E-01 mSv/hr, respectively. Second, as of the time the equipment was dismantled, 20 radioactive isotopes of 10 elements, excluding short-lived nuclides, were generated. In terms of radioactivity, 56Co (45.83%), 55Fe (19.73%), 57Co (14.48%), and 54Mn (13.50%) were prominent. Regarding radiation dose rates, 56Co (92%) and 54Mn (7.32%) exhibited higher levels. Dose rates at distances of 10 cm, 50 cm, and 100 cm were calculated at 5.31E-01 mSv/hr, 8.80E-02 mSv/hr, and 3.47 E-02 mSv/hr, respectively. Third, according to the radioactive waste classification standards in the replacement and decommissioning stages, Havar foil was predicted to be low-level radioactive waste in terms of radioactivity. In addition, it was derived that a cooling period of approximately 12 years is necessary to satisfy the allowable dose for clearance level waste. In conclusion, Havar foil, which is periodically generated as radioactive waste, can cause radiation exposure to replacement workers. Therefore, special and careful management is required for the Havar foil of the cyclotron.

The objective of this study is to investigate the safety awareness and effectiveness of the education and training for employees engaged in radiological emergency organization of the Korea Atomic Energy Research Institute (KAERI). In 2022, the questionnaire for the education satisfaction survey was revised to regulary evaluate the effect of edcation on perceptions of importance on emergency preparedness for nuclear research facilities. In line with, a standard questionnaire was created which covers 3 factors and 9 attributes, and the evaluation indicatior is based on a 5-point Likert scale. In 2023, the education on radiological emergency preparedness was conducted for 235 emergency staff. From May 24 to July 13, 2023, data was collected from a total of 235 emergency response personnels, including 28 new staffs and 207 maintenance staffs. Aa a result of response analysis, it was identified that education for radiological emergency response had a significant correlation with the promoting safety culture. It was found that senior emergency personnel with more years of experience are highly interested in radioactive disaster prevention and actively participate in and training. On the other hand, it was presented that new and less experienced groups tend to have a relatively high scored of the risk perception of nuclear research facilitites. Therefore, it is necessary to improve the practical curriculum in order to increase the participation of junior disaster prevention personnel in education and training, ensuring that they correctly recognize the risk of research facilities. This results are expected to be used to improve the quality of education and drills for radiological emergency response at KAERI.

To ensure the maintenance of the nuclear emergency response system, it is important to periodicaly conduct hazard assessments using up-to-date input variables. The results of this review are apllied to drills and exercises, enabling the inspection of emergency plan and response procedures. Therefore, this study aims to analyze off-site consequences according to the occurrence time of the Design Basis Accident (DBA) for the Hanaro Fuel Fabrication Facility (HFFF) by using the recent site-specific meteorological data and to review the appropriateness of urgent protective measures. MELCOR and SafeHanaro computer codes were used for radiation source-term estimation and environmental impact assessment, respectively. It was assumed that radioactive materials are released into environment for 2 hours due to the fire during the nuclear fuel sieving process. The following 12 scenarios for each occurrence time period was selected (0 am, 2 am, 4 am, 6 am, 8 am, 10 am, 12 pm, 2 pm, 4 pm, 6 pm, 8 pm, 10 pm) and the effective dose and thyroid dose in earlyand intermediate-phase were assessed. As a result, the most severe exposure-induced accident scenario is found to be as occurring at 0 am on July 15th, with the Most Exposed Individual (MEI) positioned 200 meters downwind from the facility. The committed effective dose for MEI is identified as to be 2.97E-02 mSv which has a significant margin against the IAEA's (Generic Intervention Level) GIL and (Generic Criteria) GC. During the passage of the radio-active plume, the estimated effective dose and thyroid dose due to inhalation were 2.97E-02 mSV (99.99%) and 5.06E-05 mSv (99.77%), respectively. External exposure appeared to be negligible. Meanwhile, the thyroid dose is noticeably below the criteria for decision-making for distribution of Potassium Iodide (KI). Accordingly, in order for local residents to participate in the exercise and drills, it is essential to develop scenarios considering simultaneous emergencies at multi-facilities and latenight accidents. In conclusion, this results will be used to improve the exercise plans for enhancing the nuclear or radiological emergency competencies of the KAERI.

Radioactive iodine released from nuclear power plants has been recognized to pose significant risks and environmental hazards. In response to these challenges, extensive investigations into iodine sorbents have been conducted with a particular focus on the utilization of layered double hydroxides (LDH) as a promising candidate. Herein, we have focused on the investigation of LDH materials featuring diverse transition metals for their synthesis, with specific emphasis on CoAl LDH for its proficiency in removing iodine species, particularly IO3 –. Nevertheless, a comprehensive understanding of the removal mechanisms employed by these LDH materials remained elusive. Hence, the primary aim of this study is to elucidate the intricacies of the removal mechanisms through sorption tests, spectroscopic techniques, and theoretical chemistry analyses, subsequently contrasting the experimental outcomes with computational results. For the experimental facet, the synthesis of CoAl LDH was conducted utilizing 0.15 mol L−1 of Co(NO3)2⋅6H2O and 0.06 mol L−1 of Al(NO3)3⋅9H2O to attain a molar ratio (M2+:M3+) of 2.5:1. Subsequently, pH-dependent IO3 – sorption tests were carried out, coupled with X-ray absorption near-edge structure (XANES) and extended X-ray absorption fine structure (EXAFS) spectroscopy, facilitating the elucidation and discourse of the removal mechanism. The theoretical chemistry in this research harnessed ab initio molecular dynamics (AIMD) simulations for structural modeling, atomic density profiles, radial distribution function, analysis of oxide species, and MD-EXAFS spectrum analysis. In summary, this study aims to elucidate iodine removal mechanisms using diverse experimental results, culminating in the revelation that ion-exchange with NO3 – present in the interlayer predominates as the principal mechanism for IO3 – removal. Notably, a distinct spectral feature at approximately 33,190 eV emerged, defying identification through XANES and EXAFS analyses conducted under experimental conditions. In the AIMD simulations, meticulous scrutiny of individual iodine atoms uncovered the prevalence of I−O and I−O−H molecular species, marked by interactions between O and H atoms, with a coordination number of I−O = ~3. This transformation was primarily instigated by proton hopping. As a result, the comparative investigation reveals the dominance of IO3 – intercalation in the CoAl LDH material with the potential to undergo a transformation to the I−O−H molecule upon interaction with protons.

The radioactive contamination in the ocean has raised significant concern on the environmental impact among Asian and Pacific countries since the Fukushima Daiichi Nuclear Power Plant accident (Mar 11, 2011). The first step in determining the contamination by the radioactive material is monitoring anomalies of environmental radioactivity of interest. As a result, each country has its own environmental radioactivity surveillance program. Strontium-90 (half-life 28.8 y) is one of the radionuclides of high interest in the environment, owing to its high fission production rate and biological accumulation resulting from similar chemical behavior with calcium. The level of Strontium-90 in the seawater is very low, with a global average of about 1 mBq kg-1. Consequently, it requires large volume of seawater sample, typically ranging from 40 L to 60 L. The purification of 90Sr from seawater sample is challenging due to the high salinity and presence of stable Sr (about 7 ppm). Therefore, the conventional method for determining 90Sr is time-consuming and labor-intensive work. The author reported an advanced method, which is a more analyst-friendly and simpler method compared to the current method, for the determination of 90Sr in seawater. This method focuses on the separation of 90Y, which is equilibrium with 90Sr, utilizing a commercialized extraction resin. As a result, it takes less than 3 hours to determine 90Sr in 50 L of seawater sample and requires less labor. Additionally, this approach could be applied to the analysis of 90Sr in radioactive waste

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.

High-temperature molten salts not only demonstrate exceptional thermal and chemical stability but also offer significant advantages in catalyzing chemical reactions. Consequently, they have garnered attention as a promising medium for next-generation nuclear reactors and a wide range of electrochemical processes. Nevertheless, the challenging experimental conditions in molten salts make applying conventional analytical methods to understand reaction mechanisms a formidable task. This underscores the imperative need for more intuitive approaches to investigate molten salt chemistry. One of the simplest yet potent methods involves real-time visual monitoring of the reaction system as chemical reactions progress. In light of this, we have developed an experimental system enabling real-time visual monitoring of the internal dynamics of molten salt media. This system can capture high-resolution videos and images within molten salts, surpassing existing methodologies. We have applied this system in various electrochemical experiments using the molten LiCl-KCl eutectic salt medium. Among them, this study primarily focuses on two challenging experimental scenarios that became comprehensible through our proposed system’s application: (1) the transpassivation of Zr metal and the agglomeration of potassium hexachlorozirconate (K2ZrCl6) solid salt, and (2) the solvation of electrons during the oxidation of Li metal within the molten LiCl-KCl eutectic salt.

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.

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.

This program aims to build a specialized and converged educational platform for the training of students in the back-end nuclear fuel cycle and cultivate integrated human resources encompassing majors, generations, and fields. To achieve this, we have established an infrastructure for integrated education and training in the radiochemistry and back-end nuclear fuel cycle and operated specialized educational courses linked with special lectures, experimental practices, and field trips. Firstly, to construct an integrated educational and training infrastructure for the back-end nuclear fuel cycle, we formed a committee of experts from both inside and outside the institution and built an advanced radiochemistry laboratory equipped with physical and chemical analysis instruments. Through a comprehensive educational program involving theory, experiments, and discussions, we have established an integrated curriculum across adjacent majors and interdisciplinary studies. We also operate short-term education and experimental training programs (e.g., summer and winter schools for the back-end nuclear fuel cycle). Secondly, the program has connected leading researchers domestically and internationally, as well as the next generation of scholars. The program offers long-term educational opportunities and internships targeting both undergraduate and graduate students. To support this, we continuously offer expert colloquiums and individual research internships. Through regular committee meetings and workshops, we focus on nurturing the integrated talents necessary for the back-end nuclear fuel cycle. Through this program, students from various fields are being trained as competent integrated human resources capable of addressing various issues in the back-end nuclear fuel cycle. It is expected that this will enable us to supply specialized technical personnel in the back-end nuclear field in line with mid-to-long-term demands.

Radiation from spent nuclear fuel (SNF) is one of key factors affecting the dissolution process of SNF and the source term from repository. The dissolution rate of uranium dioxide (UO2) matrix of SNF is expected to control the release of radionuclides from SNF in contact with water under geological disposal conditions. Based on the oxidative dissolution mechanism, the solubility of UO2 can increase significantly if the reducing environment near the fuel surface is altered by water radiolysis caused by radiation from SNF. Therefore, the analysis of water radiolysis products such as radicals (·OH, ·OH2, eaq, ·H) and molecules (H3O+, H2, H2O2) is perquisite for studies on the rate of such dissolution process to determine oxidation/dissolution mechanism and related rate constants. In this study we examined the two-known spectroscopic methods developed for H2O2 determination; one is the luminol-based chemiluminescence (luminol-CL) method and the other is the spectrophotometry using ferrous oxidation-xylenol orange complexation (FOX). Their applicability for quantitative analysis of H2O2 in potential aqueous samples from SNF dissolution studies was evaluated in terms of the analytical dynamic range (ADR), the limit of detection (LOD) and the interfering effects of various metal ions possibly present in real samples. The luminol-CL method exploits the chemiluminescence reaction caused by luminol; when in the presence of a metallic catalyst (e.g., Cu2+, Co2+), luminol emits a blue light (425 nm) at pH 10- 11 in response to oxidizing agents such as hydrogen peroxide. Although a flow-through reaction system is routinely employed to enhance the analytical sensitivity we achieved the ADR up to ~200 μM and LOD < 1 μM by a batch-wise CL detection using conventional cuvette cells and an intensified charge-coupled device (ICCD). Interestingly, it turned out that the interfering effects of other metal ions (e.g., UO2 2+, U4+, Fe2+ and Fe3+) is minimal, which should be advantageous for the luminol-CL method to be employed for samples potentially containing other metal ions. On the other hand, the FOX method spectrophotometrically analyzes H2O2 based on the difference in color (or absorption spectra) of Fe-xylenol orange (XO) complexes. Initially, the Fe2+-XO complex was provided in working solutions at pH 3, which was subsequently mixed with samples having H2O2 and allowed for quantitative oxidation of Fe2+ to Fe3+. Typically, by monitoring the absorbance of Fe3+-XO complex at 560-580 nm (λmax) the ADR up to ~100 μM and LOD ~1.6 μM were achieved. However, it is found that interfering effects from M3+ and M4+ ions are significant; these interfering metal ions can form XO complexes so as to directly contribute the measured absorbance. In contrast, the influence from M2+ ions was found to be negligible. To summarize we conclude that both methods can be applied for H2O2 determination for aqueous samples taken from SNF dissolution tests. However, prior to applying the FOX method the metal ion composition in those samples should be thoroughly identified not to overestimate the H2O2 concentration of samples. More details of underlying chemical reactions in both methods will be discussed in the presentation.

This study demonstrated a rapid and simple method for the determination of seven anions including halides and oxyhalides from the KURT underground water sample by capillary electrophoresis with UV detection. In nuclear waste disposal, some anions such as iodine, selenium, and technetium have been of great concern due to its high mobility and toxicity with a long half-life. It has been needed for a reliable analysis of anionic speciation because the high mobility of anions is easily affected by environmental conditions especially pH and salinity of underground water. Here this project is to develop a fast separation of seven anions including iodide, iodate, and selenite using capillary electrophoresis. The electroosmotic flow (EOF) was suppressed using a poly (ethyleneglycol) -coated capillary (DB-WAX capillary). As a result, anions migrated depending on their mobility under a reverse polarity condition (-15 kV) and the analysis time was within 15 min. UV detection was used at 200 nm. The RSDs for migration time were between 0.7% and 1.3% except for selenite of 5.1%. The RSDs for peak area were obtained between 2.9% and 7.4%. The calibration curves were linear from 10 to 200 mg/L with correlation coefficients greater than 0.9952. The LODs were 7.3, 10.9, 11.3, 12.9, 13.0, 13.9, and 17.4 mg/L for iodide, nitrate, bromide, selenite, bromate, tungstate and iodate. The KURT underground water sample spiked with seven anions at 50 mg/L were analyzed. The recoveries of spiked KURT sample ranged from 93.4% to 99.3%. The proposed method was successfully applied to determine seven anions in underground water sample.

Chelating agents, such as ethylenediaminetetraacetic acid (EDTA), diethylenetriaminepentaacetic acid (DTPA), and nitrilotriacetic acid (NTA) are widely used in industry and agriculture as water softeners, detergents, and metal chelating agents. In wastewater treatment plants, a significant amount of chelating agents can be discharged into natural waters because they are difficult to degrade. Since those compounds affect the mobility of radionuclides or heavy metals in decontamination operations at nuclear facilities and radioactive waste disposal, quantification of the amount of ligand is very important for safe nuclear waste management. To predict the behavior of the main complexation in sample matrices of radioactive wastes, it is essential to evaluate the distribution of the metal-chelating species and their stabilities in order to develop analytical techniques for quantifying chelating agents. We have investigated to collect information on the pH speciation of metal chelation and the stability constants of metal complexes depending on three chelating agents (EDTA, DTPA, and NTA). For example, Zhang’s group recently reported that the initial coordination pH of Cu(II) and EDTA4− is delayed with the addition of Fe(III), and the pH range for the stable existence of [Cu(EDTA)]2− is narrowed compared to when it is alone in the sample matrix. The addition of Fe(III) clearly impacts the chemical states of the Cu(II)-EDTA solution. Additionally, Eivazihollagh’s group demonstrated differences in the speciation and stability of Cu(II) species between Cu(II) and three chelating ligands (EDTA, DTPA, and NTA). This study will be greatly helpful in identifying the sample matrix for binding major chelating agents and metals as well as developing chemically sample pretreatment and separation methods based on the sample matrix. Finally, these advancements will enable reliable quantitative analysis of chelating agents in decommissioning radioactive wastes.

During the initial cooling period of spent nuclear fuel, Cs-137 and Sr-90 constitute a large portion of the total decay heat. Therefore, separating cesium and strontium from spent nuclear fuel can significantly decrease decay heat and facilitate disposition. This study presents analytical technique based on the gas pressurized extraction chromatography (GPEC) system with cation exchange resin for the separation of Sr, Cs, and Ba. GPEC is a micro-scaled column chromatography system that allows for faster separation and reduction volume of elution solvent compared to conventional column chromatography by utilizing pressurized nitrogen gas. Here, we demonstrate the comparative study of the conventional column chromatography and the GPEC method. Cation exchange resin AG 50W-X12 (200~400 mesh size) was used. The sample was prepared at a 0.8 M hydrochloric acid solution and gradient elution was applied. In this case, we used the natural isotopes 88Sr, 133Cs, and 138Ba instead of radioactive isotopes for the preliminary test. Usually, cesium is difficult to measure with ICP-OES, because its wavelengths (455.531 nm and 459.320 nm) are less sensitive. So, we used ICP-MS to determine the identification and the recovery of eluate. In this study, optimized experimental conditions and analytical result including reproducibility of the recovery, total analysis time and volume of eluents will be discussed by comparing GPEC and conventional column chromatography.