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).

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.

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.

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.

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.

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

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.

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 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.

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 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.

In the event of a radiological emergency at a nuclear facility, the exchange of information on the accident situation is very important in the response process. For this reason, international organizations such as the IAEA and the EU operate systems to exchange information in the event of a radiological emergency. In south korea, the emergency response information exchange system (ERIX) developed by KINS is operated for use by the national radiological emergency response organization. The ERIX enables the exchange of emergency response information between organizations such as the government, nuclear operators, local authorities, KINS and KIRAMS. The KAERI has developed the KAERI emergency response information exchange system (KAERIX) for the exchange of accident information and emergency response information between the emergency response organizations of the KAERI in the event of a radiological emergency. This system is web-based using HTML, runs on internal network and is only available to KAERI staff. Recently, as the need for optimizing and upgrading KAERIX has arisen, improvements have been derived. The main improvement is optimizing KAERIX for Microsoft Edge to minimize errors. At present, it is optimized for Internet Explorer, but optimizing it for Microsoft Edge mode has become essential due to Microsoft discontinuing support for Internet Explorer. Another major improvement involves adding functions in ERIX to KAERIX, such as displaying the deletion/ correction status of input information and providing notifications for important information registration. Ultimately, KAERIX will be upgraded and optimized in 2024, reflecting improvements.

The nuclear licensee must ensure that the nuclear or radiological emergency preparedness and response organization is explicitly defined and staffed with adequate numbers of competent and assessed personnel for their roles. This paper describes the responsibilities of medical and support personnel for the medical action of casualties in the event of a radiological emergency at the KAERI. Currently, there is one medical personnel (nurse) in KAERI, and a total of eight medical support personnel are designated for medical response in the event of a radiological emergency. These medical support personnel are designated as one or two of the on-site response personnel at each nuclear facility, operating as a dedicated team of A, B (4 people each). In the event of a radiological emergency, not all medical support personnel are mobilized, but members of the dedicated medical team, which includes the medical support personnel of the nuclear facility where the accident has occurred, are summoned. Medical and support personnel will first gather in the onsite operational support center (OSC)/technical support center (TSC) to prepare and stand by for the medical response to injured when a radiological emergency is declared. They should take radiation protective measures, such as wearing radiation protective clothing and dosimeters, before entering the onsite of a radiological emergency, because injuries sustained during a radiological emergency may be associated with radioactive contamination. In the event of an injury, direct medical treatment such as checking the patient’s vitals, first aid, and decontamination will be carried out by medical personnel, while support personnel are mainly responsible for contacting the transfer hospital, reporting the patient’s condition, accompanying the ambulance, filling out the emergency medical treatment record, and supporting medical personnel. In order to respond appropriately to the occurrence of injuries, we regularly conduct emergency medical supplies education and medical training for medical support personnel to strengthen their capabilities.

South Korea’s first commercial nuclear reactor, Kori Unit 1, was permanently shut down in 2017, and preparations are currently underway for its decommissioning. After the permanent shutdown, the spent nuclear fuel from the reactor core is removed and stored in a spent fuel storage facility. Subsequently, steps are taken for its permanent disposal, and if a permanent disposal site is not determined, it is stored in an interim storage facility (or temporary storage facility). Therefore, the activation criteria for radiation emergency plans vary depending on the movement of spent nuclear fuel and the storage location. In this study, it reviewed emergency plans in the U.S. NRC Regulatory Guide (Draft) titled ‘Emergency Planning for Decommissioning Nuclear Power Reactors’ to determine the requirements for radiation emergency plans needed for decommissioned nuclear power plants. Additionally, by examining emergency plans applied to decommissioning nuclear power plants in the United States, this study identified emergency plan requirement that could be applicable to future decommissioned nuclear power plants in South Korea. This study will contribute to the establishment of appropriate radiation emergency plans for decommissioning nuclear power plants in Korea for providing accurate information on overseas cases and relevant guidelines.

When the parent radionuclide decays, the progeny radionuclide is produced. Accordingly, the dose contribution of the progeny radionuclide should be considered when assessing dose. For this purpose, European Commission (EC) and International Atomic Energy Agency (IAEA) provide weighting factors for dose coefficient. However, these weighting factors have a limitation that does not reflect the latest nuclide data. Therefore, in this study, we analyzed the EC and IAEA methodology for derivation of weighting factor and used the latest nuclide data from ICRP 107 to derive weighting factors for dose coefficient. Weighting factor calculation is carried out through 1) selection of nuclide, 2) setting of evaluation period, and 3) derivation based on ICRP 107 radionuclide data. Firstly, in order to derive the weighting factor, we need to select the radionuclides whose dose contribution should be considered. If the half-life of progeny radionuclides sufficiently short compared to the parent radionuclide to achieve radioactive equilibrium, or if the dose coefficient is greater of similar to that of the parent radionuclide and cannot be ignored, the dose contribution of the progeny radionuclide should be considered. In order not to underestimate the dose contribution of progeny radionuclides, the weighting factors for the progeny nuclides are taken as the maximum activity ratio that the respective progeny radionuclides will reach during a time span of 100 years. Finally, the weighting factor can be derived by considering the radioactivity ratio and branch fraction. In order to calculate the weighting factor, decay data such as the half-life of the radionuclide, decay chain, and branch fraction are required. In this study, radionuclide data from ICRP 107 was used. As a result of the evaluation, for most radionuclides, the weighting factors were derived similarly to the existing EC and IAEA weighting factors. However, for some nuclides, the weighting factors were significantly different from EC and IAEA. This is judged to be a difference in the half-life and branch fraction of the radionuclide. For example, in the case of 95Zr, the weighting factor for 95mNb showed a 35.8% difference between this study and previous study. For ICRP 38, when 95Zr decays, the branch fraction for 95mNb is 6.98×10-3. In contrast, for ICRP 107, the branch fraction is 1.08×10-2, a difference of 54.7%. Therefore, the weighting factor for the dose coefficient based on ICRP 107 data may differ from existing studies depending on the half-life and decay information of the nuclide. This suggests the need for a weighting factor based on the latest nuclide data. The results of this study can be used as a basis for the consideration of dose contributions for progeny radionuclides in various dose assessments.

The demand for transportation is increasing due to the continuous generation of radioactive wastes. Especially, considering the geographical characteristics of Korea and the location characteristics of nuclear facilities, the demand for maritime transportation is expected to increase. If a sinking accident happens during maritime transportation, radioactive materials can be released into the ocean from radioactive waste transportation containers. Radioactive materials can spread through the ocean currents and have radiological effects on humans. The effect on humans is proportional to the concentration of radioactive materials in the ocean compartment. In order to calculate the concentration of radioactive materials that constantly flow along the ocean current, it is necessary to divide the wide ocean into appropriate compartments and express the transfer processes of radioactive materials between the compartments. Accordingly, this study analyzed various ocean transfer evaluation methodologies of overseas maritime transportation risk codes. MARINRAD, POSEIDON, and LAMER codes were selected to analyze the maritime transfer evaluation methodology. MARINRAD divided the ocean into two types of compartments that water and sediment compartments. And it was assumed that radionuclides are transfered from water to water or from water to sediment. Advection, diffusion, and sedimentation were established as transfer process for radionuclides between compartments. MARINRAD use transfer parameters to evaluate transer processes by advection, diffusion, and sedimentation. Transfer parameters were affected by flow rate, sedimentation rate, sediment porosity, and etc. POSEIDON also divided the ocean into two types that water and sediment compartment, each compartments was detaily divided into three vertical sub-compartment. Advection, diffusion, resuspension, sedimentation, and bioturbation were established as transport processes for radionuclides between compartments. POSEIDON also used transfer parameters for evaluating advection, diffusion, resuspension, sedimentation, and bioturbation. Transfer parameters were affected by suspended sediment rates, sedimentation rates, vertical diffusion coefficients, bioturbation factors, porosity, and etc. LAMER only considered the water compartment. It divided the water compartment into vertical detailed compartments. Diffusion, advection and sedimentation were established as the nuclide transfer processes between the compartments. To evaluated the transfer processes of nuclides for diffusion and advection, LAMER calculated the probability with generating random position vectors for radionuclides’ locations rather than deterministic methods such as MARINRAD’s transfer parameters or POSEIDON’s transfer rates to evaluate transfer processes. The results of this study can be used as a basis for developing radioactive materials’ ocean transfer evaluation model.

Radiation workers, especially those dealing with Uranium isotopes, can potentially intake Uranium -containing materials through their respiratory and digestive systems. According to the “Regulations on the Measurement and Calculation of Internal Exposure” from Nuclear Safety and Security Commission (NSSC), those who intend to work in or enter the nuclear facilities with a risk of exceeding 2 mSv exposure per year should be examined the internal exposure. However, when it comes to in-vitro bioassay, Uranium intake through drinking water can affect the quantitative analysis. The International Commission on Radiological Protection (ICRP) reported in ICRP Publication 23 (Report on the Task Group on Reference Man) that the reference man excretes Uranium in the urine (0.05-0.5 μg/day) and feces (1.4-1.8 μg/day). Korea Atomic Energy Research Institute (KAERI) set the 90.5 ng/day as the 238U background of workers handing Uranium based on the daily Uranium intake of Koreans. In this research, we examined the possible effects of Uranium in drinking water on internal exposure by analyzing the concentration of Uranium in bottled waters from various water sources sold in the domestic market and a water from the water purifier. The 238U concentration results of analyzing 11 bottled waters and 1 purified water, were ranged from 0 to 10.2 μg/L. All the results were satisfied the standard of 30 μg/L according to “Regulations for Drinking Water Quality Standards and Inspection” enacted by the Ministry of Environment. However, various concentrations were shown depending on the water sources. Assuming that these concentrations of water are consumed by drinking 1 L per day, the internal dose assessment result is 0 to 0.94 mSv. On the other hand, if it is assumed to be inhaled, it can be an overestimated because the dose coefficient of inhalation, Type M is higher than that of ingestion, f1=0.02 which are the values recommended by ICRP Publication 78 (Individual Monitoring for Internal Exposure of Workers) when the Uranium compound is unspecified. In case of two workers at KAERI, the daily excretion of urine was 151 and 120 ng/day respectively in the first quarter monitoring. However after changing the kind of drinking water in the second quarter monitoring, it dropped to 17.4 and 15.4 ng/day respectively. Through this study, it is confirmed that the Uranium background in urine can be analyzed differently depending on the kind of drinking water consumed by each worker. Depending on the Uranium concentration of drinking water, the internal exposure dose assessment can be overestimated or underestimated. Therefore, the Uranium concentration and intake amount according to the kind of drinking water should be considered for in-vitro bioassays of Uranium handlers. Furthermore, if necessary, the Uranium isotope ratio analysis in urine and the handling information should be comprehensively considered. In addition, in order to exclude the effect of intake through the digestive system, replacing the kind of drinking water can be considered. The additional analysis such as in-vivo bioassay and 24 hours urine analysis rather than spot samples can be also recommended.

Detectors utilized for nuclear material safeguards have been using scintillation detectors which are inexpensive and highly portable, and electrically cooled germanium detectors which are expensive but have excellent energy resolution. However, recently IAEA, the only international inspectorate of nuclear material safeguards for the globe, have replaced the existing scintillation detector and electrically cooled germanium detector with a CdZnTe detector owing to the improved performance of room-temperature semiconductors significantly. In this paper, we will examine the spectrum features of the CdZnTe detector such as spectrum shape, energy resolution, and efficiency in the energy region of interest, which are the important characteristics for measuring Uranium enrichment. For this purpose, it would be conducted to compare its spectrum features using CdZnTe, NaI, HPGe detectors. The main energies of interest include 185.7 keV and 1,001 keV, which are the decay energies of uranium 235 and uranium 238. The results of this study will provide a better understanding of the spectral features of various detectors used in uranium enrichment analysis, and are expected to be used as basic data for future related software development.

As nuclear decommissioning ventures become increasingly complex, the role of digitalization in facilitating and enhancing these operations is becoming indispensable. This transition to a more digitized approach presents a myriad of advantages, including: augmented avenues for data acquisition, analysis, and visualization to bolster dismantling strategies; simulations in virtual environments for operator training; precise forecasting of future waste emergence, culminating in refined cost estimations; and more immersive decommissioning visualizations for both operators and external stakeholders. Salient benefits conferred by the integration of digital technologies in decommissioning encompass improved collaboration, enriched knowledge transfer, clarity regarding present technological constraints, insights into key influencing factors, clearer criteria for technology selection, and a profound understanding of the potential challenges and merits of a broader incorporation of digital tools in decommissioning endeavors. Of paramount importance is the opportunity presented for superior workforce training and safety measures, exemplified by ALARAbased planning. Amidst the myriad facets of digital adoption, 3D modeling of nuclear facilities derived from laser-scanned point clouds stands out as a pivotal domain in the digitalization. The transformation of intricate point cloud data into a comprehensible 3D mesh remains the crux of this paper. The process of mesh generation, despite being simpler than its counterpart of converting to a 3D solid model, is crucial for multiple reasons. The resultant 3D mesh offers an enhanced visual representation compared to a sparse point cloud, paving the way for improved spatial perception. Furthermore, it serves as a rudimentary tool for approximating component volumes and the ensuing waste, thereby playing an instrumental role in waste manipulation strategies, notably in collision detection. This paper delves deep into the nuances of mesh generation, conducting an parametric study of mesh conversion algorithms, including down-sampling rates. Through this rigorous examination, we endeavor to shed light on optimal methodologies, hoping to catalyze advancements in the digitalization of nuclear decommissioning processes.

The operation of nuclear facilities involves the potential for on-site contamination of soil, primarily resulting from pipe leaks and other operational incidents. Globally, decommissioning process for commercial nuclear power plants have revealed huge-amounts of soil waste contaminated with Cs-137, Sr-90, Co-60, and H-3. For example, Connecticut Yankee in the United States produced approximately 52,800 ton of contaminated soil waste, constituting 10% of the total waste generated during its decommissioning. Environmental remediation costs associated with nuclear decommissioning in the US averaged $60 million per unit, representing a significant 10% of the whole decommissioning expenses. Consequently, this study undertook a preliminary investigation to identify important factors for establishing a site remediation strategy based on radionuclide- and site-specific media- characteristics, focusing the efficiency enhancement for the environmental remediation. The factors considered for this investigation were categorized into physical/environmental, socioeconomic, technical, and management aspects. Physical/environmental factors contained the site characteristics, contamination levels, and environmental sensitivity, while socio-economic factors included the social concerns and economic costs. Technical and management factors included subcategories such as technical considerations, policy aspects, and management factors. Especially, technical factors were further subdivided to consider the site reuse potential, secondary waste generation by site remediation, remediation efficiency, and remediation time. Additionally, our study focused the key factors that facilitate the systematic planning for the site remediation, considering the distribution coefficient (Kd) and hydrogeological characteristics associated with each radionuclide in specific site conditions. Therefore, key factors in this study focus the geochemical characteristics of site media including the particle size distribution, chemical composition, organic and inorganic constituents, and soil moisture content. Moreover, the adsorption properties of site media were examined concerning the distribution coefficient (Kd) of radionuclides and their migration characteristics. Furthermore, this study supported the development of a conceptual framework, containing the remediation strategies that incorporate the mobility of radionuclides, according to the site-specific media. This conceptual framework would necessitate the spatial analysis techniques involving the whole contamination surveys and radionuclide mobility modeling data. By integrating these key factors, the study provides the selection and simulation of optimal remediation methods, ultimately offering the estimated amounts of radioactive waste and its disposal costs. Therefore, these key factors offer foundational insights for designing the site remediation strategies according the sitespecific information such as the distribution coefficient (Kd) and hydrogeological characteristics.