It is likely to occur internal exposure for workers in Nuclear Power Plants (NPPs) due to the intake of radionuclide. To assess the internal exposure dose the measurement of activity for remain radionuclide is necessary. The Whole Body Counters (WBCs) are commonly used for measurement of remain radionuclide activity in human body. Korea Hydro & Nuclear Power Co., Ltd. (KHNP) conduct performance test of WBCs in all NPPs for every year to confirm the performance of equipment. The performance test is conducted using unknown sources and the participants of the comparison test submit the radionuclide and activity of the unknown sources measured by WBC as a result. The performance indicator and criteria for WBC recommended in the American National Standards Institute (ANSI) N13.30 report published in 2011 are applied. The performance indicator is Root Mean Squared Error (RMSE) and criteria is 0.25 or less. The results of performance test performed in 2022 for all WBC is meet the ANSI N13.30 criteria. And the RMSE values are confirmed from 0.01 to 0.23. This means that the residual radioactivity measurement results using WBC are reliable.

The decommissioning of a nuclear power plant is a project that consists of several stages, and various technologies are applied when performing various tasks at each stage. And it is essential to secure safety and economic feasibility. As the paradigm has changed due to digital transformation in various industries, digitalization is applied to the life cycle of nuclear power plant from construction, operation and decommissioning project. Element technologies are being developed for decommissioning plan establishment, process design, econtamination method, decommissioning work process, waste management, environmental monitoring and radiation dose simulation. The utilization of digital twin in the decommissioning stage is classified into three categories. ① Process Monitoring (decommissioning work procedure, work progress (plan/actual), real-time work status and etc.) ② Facility Monitoring (real-time sensing and video data monitoring, decommissioning SSCs information, work alarm and etc.) ③ Safety Monitoring (work safety, radiation exposure, fire monitoring, work risk and etc.) A system suitable for the decommissioning stage and work should be developed in consideration of the target of use, development function, and when to create data according to the purpose of the system. Simulation module according to user purpose should be provided. In addition, data-base management should be performed according to the decommissioning characteristics in consideration of the data associated with the existing operating system. The system to be developed should support the project management to comply with the domestic standards and regulations to be determined in the future. This will improve the competitiveness of domestic and foreign markets.

The dimensioning machine installed in the hot cell has been used for 20 years. It has been used for a long time so it was often malfunction due to aging and radiation. In addition, some parts of apparatus were discontinued and there were a lot of problems in maintenance and repair. In the old measuring system, the operator’s subjectivity was much involved. The process of control the focal length (distance between lens and specimen) by operator’s sense is a good example. The existing dimensioning machine was the Kh-7700 of Hirox Co., Ltd. As the equipment had been used for a longtime, additional utilities such as jigs, lighting module and servo motors have been customized and used. The same company’s apparatus was selected for the reasons that it did not need to manufacture a new utility so it could save the cost of radioactive waste disposal for existing utilities and its radiation resistance which has been used for 20 years in radiation environment. Lighting, standing, stage, controllers, cables and so on had been customized to provide remote control in hot cell. The installation was completed in March of this year in hot cell and has been successfully used until now. Through the improvement of dimensioning machine, an auto-focusing and multi-focusing were available. Therefore, they made it possible to produce high quality data and improve the accuracy of data. And this research could be a good example of how hot cell devices can be built and customized to achieve remote control.

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

In accordance with the notification of the Nuclear Safety and Security Commission (NSSC), environmental impact assessments around nuclear power plants are conducted annually and the results are disclosed to the public. The effects of direct radiation exposure from nuclear power plants as well as liquid effluents and gaseous effluents are taken into consideration in the evaluation of dose calculation for residents. In the United States, regulatory guidelines on direct radiation exposure are described in Reg. Guide 4.1, and the effects of direct radiation are evaluated through regulatory guidelines in Korea. We are going to review optimal evaluation method by reviewing the direct radiation exposure evaluation method currently being conducted in domestic nuclear power plants and the direct radiation exposure evaluation method in overseas nuclear power plants such as in the United States.

The correlation between accident management plan and radiation emergency plan of Shin-Kori Units 3 and 4 was compared and analyzed from the point of view of the adequacy of facilities, equipments, organization and manpower which are necessary for the related emergency response. It was found the equipment of accident management plan and emergency response facility of radiation emergency plan had different technical contents and scope of application, so there was no risk of mutual conflict and overlapping functions. However, since the accident impact assessment code in accident management plan and computer program of radiation emergency plan were different, it was necessary to ensure the agreement or linkage of the evaluation between them. When a radiation emergency is issued in accident management plan, the composition and mission of the accident response organization were mostly consistent with the contents of the radiation emergency plan, but some corrections and improvement items were identified. Accident management plan specified that the disaster response safety center belonged to the emergency operations facility (EOF), but the radiation emergency plan did not mention it at all. The main tasks of disaster response safety center were the movement, arrangement and connection of mobile emergency response facilities, on-site construction of other emergency response facilities, and on-site road restoration. According to the accident management plan, the movement, arrangement, and connection of mobile facilities (i.e., mobile generators, mobile pumps, multi-purpose communication relay facilities), which were considered very important for the prevention and mitigation of serious accidents, were under the supervision of the disaster response safety center. It was stipulated that the operation was carried out with the cooperation of a regular emergency organization, and that the start, operation and stop of mobile equipments were to be performed under the supervision of the emergency operation team supported by the regular emergency organization. Since this organization structure and assignment of duties could not be confirmed in radiation emergency plan, it was necessary to revise and improve the radiation emergency plan for the successful operation of mobile equipments and to link them with the accident management plan.

The Korea Atomic Energy Research Institute operates the Nuclear Cycle Experimental Research Facility which has radiation controlled area in the laboratory with the aim of realizing pyroprocessing technology. In this Facility, depleted Uranium feed material and a depleted Uranium mixed with some surrogate material are used for performing experiments. Therefore the facility is using uranium, users should be careful of radiation. This paper will explain the radiation protection of the Nuclear Cycle Experimental Research facility and will also explain how much alpha radiation comes out from the facility. The RMS (Radiation Monitoring System) detector is made by CANBERRA and the model name is ICAM. ICAM RMS is the detector which can detect Alpha Radiation by absorbing the air in the facility. The RMS detector is installed in the HVAC room on the third floor to check the air contamination through the chimney. The RMS is connected to the air ventilation line for detecting Alpha radiation in the whole facility. Experiment was performed for two weeks to check the radiation level and the air ventilation fan continued to operate 24 hours a day. the results are below the required value which is 0.1 Bq/m3, indicating that the facility is safe in terms of radiation safety management.

As part of the third ATM Challenge, we performed a series of atmospheric dispersion simulations for routine releases of Xe-133 from ordinarily operating nuclear facilities such as Medical Isotope Production Facilities (MIPFs), Nuclear Power Plants (NPPs), and Research Reactors (RRs) in the Northern Hemisphere using our ATM, Lagrangian Atmospheric Dose Assessment System (LADAS), with Numerical Weather Prediction (NWP) data produced by the Korea Meteorological Administration (KMA). The simulation time period is 6 months, from June to November in 2014, and we used the stack emission data except for CNL (Canada) and IRE (Belgium) in accordance with the scenario of the third ATM Challenge 2019. In addition, the simulations were done individually for all MIPFs, NPPs, and RRs. We utilized 3-hourly KMA’s Unified Model Global Data Assimilation and Prediction System (UM-GDAPS) data with 0.35°×0.23° horizontal resolution as input meteorological fields and extracted hourly time series results for Xe-133 activity concentrations with few different resolutions such as 0.5°×0.5°, 0.35°×0.23°, and 0.1°×0.1° at several IMS stations in the Northern Hemisphere which were in normal operation in 2014. Considering previously reported values of daily Xe-133 release amounts for CNL and IRE, measured signals at some IMS stations (such as CAX17, DEX33, SEX63, and USX75) were well reproduced from the simulation results.

n Korea, the decommissioning of nuclear power plants is being prepared, and a large amount of radioactive waste is expected to be generated. In particular, clearance level waste, which accounts for more than 90%, requires prevention of cross-contamination and prompt classification. In this study, the possible exposure route and the derivation of exposure dose for worker exposure management in a movable analysis system that can be analyzed onsite were studied. The movable radionuclide analysis system is divided into a preparatory room, a sample storage room, a radioanalysis room, a laboratory, and a waste storage room. It consists of one radioanalysis worker and one pre-treatment worker, and the main radiation exposure is expected to occur in the movement path in the sample storage room, radioanalysis room, and laboratory. The source term for the exposure evaluation, the annual usage dose presented in the radiation safety report in the movable radionuclide analysis system was used. The input data for the evaluation of the external exposure dose under normal circumstances (exposure situation, working hours, distance, etc.) is referenced at facility specifications. The internal exposure dose evaluation was assumed to be acute exposure (1 hour) assumed as internal pollution due to the drop in liquid sample during the pretreatment work. As an evaluation method, a method using a calculation formula and a method using an evaluation code was performed. For the evaluation of exposure dose using the calculation formula, a preliminary evaluation was performed using the point source method, the point kernel method, and intake and dose conversion factors. In addition, VISIPLAN and IMBA codes were used to evaluate exposure dose using the evaluation code, and the input data were supplemented for evaluation. As a result of the evaluation, the annual exposure dose limit of 20 mSv was satisfied for both normal and non-normal situations. In future research, it is planned to derive the evaluation results by particular scenarios for the detailed movement route and evaluation time according to the work process in the mobile radionuclide analysis.

Surface contaminants may attach to surfaces or objects in the radiation controlled area to cause radiation exposure, or spread out to the general environment by person and object exiting the radiation workplace. Accordingly, in radiological safety control, surface contamination monitoring is one of the important factors in workplace monitoring. When obtaining the measurement results for the monitoring, the results are accompanied by uncertainty since measurements contain numerous errors. Accordingly, the International Organization for Standardization (ISO) has published the ISO 7503 series which is comprehensive and detailed guidelines on the measurement and evaluation of surface contamination. ISO 7503-3 especially presents a mathematical model for the contamination measurement and provide calculation guidelines on measurement uncertainty evaluation, decision threshold and detection limit. This paper is focused on reevaluating and comparing the surface contamination monitoring method applied to radiation safety management practice and its results based on the measurement and evaluation method set by the International Organization for Standardization. The evaluation was performed in accordance with ISO 7503, and the current reporting method for measurement results was compared with the method recommended in ISO 11929 publication.

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

Currently, low and intermediate-level radioactive wastes and spent nuclear fuels are continuously generated in Korea. For the disposal of the radioactive wastes, the transport demand is expected to increase. Prior to transportation, it is necessary to evaluate the radiation risk of transportation to confirm that is not high. In Korea, there is no transportation risk assessment code that reflects domestic characteristics. Therefore, foreign assessment codes are used. In this study, before developing the overland transportation risk assessment code that reflects domestic characteristics, we analyzed the radiation risk assessment methodology in transportation accident codes developed in other countries. RADTRAN and RISKIND codes were selected as representative overland transportation risk assessment codes. For the two codes we analyzed accident scenarios, exposure pathways, and atmospheric diffusion. In RADTRAN, the user classifies accident severity for possible accident scenarios, and the user inputs the probability for each accident severity. On the other hand, in the case of RISKIND, the accident scenarios are classified and the probabilities are determined according to the NRC modal study (LLNL, 1987) in consideration of the cask impact velocity, cask impact angle, and fire temperature. In the case of RISKIND, the accident scenarios are applied only to transportation of spent nuclear fuel, and cannot be defined for low and intermediate-level radioactive waste. However, in the case of RADTRAN, since the severity and probability of accidents are defined by user, it can be applied to low and intermediate-level radioactive wastes. As the exposure pathways considered in transportation accident, both RADTRAN and RISKIND consider external exposure (cloudshine and groundshine), and internal exposure (inhalation, resuspension inhalation and ingestion). In the case of RADTRAN, additionally, external exposure due to loss of shielding (LOS) is considered. Atmospheric diffusion calculation is essential to determine the extent to which radioactive materials are diffused. In both RADTRAN and RISKIND, atmospheric diffusion calculations are based on Gaussian diffusion model. Users must input Pasquill stability class, release height, heat release, wind speed, temperature and mixing height, etc. Additionally, RADTRAN can input weather information relatively simply by inputting only the Pasquill stability class fraction and selecting the US average weather option. This study results will be used as a basis for developing radioactive waste overland transportation risk assessment code that reflects domestic characteristics.

Kr-85 has a half-life of 10.7 years and it stays in the atmosphere for a long time. However it does not accumulate as an noble gas but only emits beta particles. Therefore its contribution to environmental radiation dose is lower than any other radionuclides. Kr-85 is one of the main fission products produced by nuclear fission reaction and artificial radionuclide that does not exist in nature. For these reasons, monitoring Kr-85 from the atmosphere is meaningful so that the nuclear-related facilities are recommended to control and regulate environmental emissions. Post Irradiation Examination Facility (PIEF) which located in KAERI is a facility that conducts various material and chemical experiments using the irradiated nuclear fuels. Therefore, various radionuclides can present in gaseous effluent including Kr-85. To prevent the environmental hazards and guarantee the radiation safety of the public, nuclear facilities are recommended to be equipped with stack radiation/radioactivity monitoring system, so that the Kr-85 concentration in gaseous effluent is controlled within the regulatory criteria. Particularly, the Kr-85 concentration of gaseous effluent is commonly monitored by the stack monitoring system connected to the process ventilation system from the hot cell. The monitoring system supply the information such as beta count rate, dose rate and flow rate, etc. Due to the concentration of Kr-85 in gaseous effluent is subject to regulatory guide lines, a systemized procedure for calculating Kr-85 concentration of the stack exhaust is necessary. Furthermore, the emission should be monitored whether it satisfies the regulatory standard or does not. This paper performed discussion on the process of calculating the concentration of Kr-85 in the gaseous effluent of PIEF stack from the monitoring system (NGM209, MGP), and the amount of Kr- 85 over the last 2 years emissions was calculated. In addition to calculating effluent rate of radioactive Kr-85, the Minimum Detectable Concentration (MDC) and Decision Threshold (SD) were calculated. As a result, the calculated Kr-85 concentration was below the SD during the entire period. It is considered that there are no environmental emissions of Kr-85.

For the safety assessment of the high-level radioactive waste (HLRW) disposal, the thermodynamic data such as solubility products, formation constants of complexes, redox equilibrium constants of radionuclides, and their reaction enthalpy and entropy are required. In order to recommend and summarize the reliable data, thermodynamic databases (TDB) have been persistently developed through the OECD-NEA TDB projects and an updated TDB of actinides has been recently published in 2020. To date, reliable data for Pu reactions are scarce due to the possibility of coexistence of four different oxidation states, Pu(III-VI) by redox equilibria in solutions. To determine the thermodynamic data for the reaction of each Pu oxidation state, it is necessary to precisely control the oxidation state and quantitatively analyze all reactants, products and bi-products by using highly sensitive speciation techniques. Since 2004, the nuclear chemistry research team in KAERI has been focused on developing techniques for the sensitive chemical speciation by using laser-based spectroscopy and determining thermodynamic data of actinides such as U, Pu, Am. In this paper, chemical speciation and thermodynamic studies on Pu in KAERI are reviewed. A combination of a commercial spectrophotometer and a capillary cell was adopted for a sensitive chemical speciation of Pu(III-VI) in solutions. A sensitive detection of trace amount of Pu colloids was carried out with the laser-induced breakdown detection (LIBD) system. Pu(VI) complexation with hydroxide or carbonate ions were investigated under strong oxidation conditions controlled with hypochlorite (NaOCl). The solubility product of Pu(OH)3(am) and formation constant of Pu(III)-OH speices were determined by a combination of wet-chemistry experiments and several analysis methods of spectrophotometry, LIBD, radiometry under a strong reducing condition controlled by electrochemistry. More recently, we reported the reaction enthalpy and entropy data for the formation of Pu(OH)2+ and the dissolution of Pu(OH)3(am). A preliminary data for reaction between Pu(III) and organic matter will be presented.

Niobium (Nb) is present in Ni-based alloys and stainless steels used in nuclear reactors as structural materials. Nb-93 is a naturally occurring and stable isotope of niobium and Nb-94 (half-life = 20,000 years) is produced by neutron activation of Nb-93. Nb-94 can be present in waste streams from dismantling of nuclear power plants and treatment of the primary coolant circuit. Hence, the radioactive wastes containing active Nb-94 are disposed of in the repositories for low- and intermediate-level waste (LILW). Nb predominantly exhibits a pentavalent oxidation state (i.e., +V) within the stability field of water. Cementitious materials (concrete, mortar, and grout) are extensively utilized in LILW disposal systems as structural components and chemical agents for the stabilization of waste. Solubility defines the source term (i.e., upper concentration limit) in the repository system. However, the solubility behavior of Nb in cementitious systems at high pH remains ill-defined, and information available on the Nb solid phases controlling the solubility is scarce and often ambiguous. Sorption on cementbased materials is one of the main mechanisms controlling the retention of niobium(V) in a LILW repository, and distribution coefficients (Rd) are necessary to evaluate the retention capacity by sorption in the safety assessment of disposal systems. Available sorption data of Nb(V) on cement showed a large discrepancy in Rd, moreover, no sorption data is available for Nb(V) under conditions characterizing the first degradation stage of cement (young cement condition) at pH 13 – 13.5. In this context, the solubility of Nb was extensively investigated in porewater conditions representative of the cement degradation stage I, as well as in CaCl2-Ca(OH)2 systems. Special focus was given to the accurate characterization of the solubility-controlling solid niobium phases. We also studied the sorption of Nb(V) by hardened cement pastes (HCP) and calcium silicate hydrates (CSH, major hydrate of HCP). This work provides the results on Rd, sorption isotherm and sorption mechanisms of Nb(V). Besides, the impact of ISA (polyhydroxycarboxylic acid generated by the degradation of cellulose) on Nb(V) sorption and the dissolution of cement materials was investigated.

The sorption/adsorption behavior of radionuclides, usually occurring at the solid-water interface, is considered to be one of the primary reactions that can hinder the migration of radiotoxic elements contained in the spent nuclear fuel. In general, various physicochemical properties such as surface area, cation exchange capacity, type of radionuclides, solid-to-liquid ratio, aqueous concentration, etc. are known to provide a significant influence on the sorption/adsorption characteristics of target radionuclides onto the mineral surfaces. Therefore, the distribution coefficient, Kd, inherently shows a conditiondependent behavior according to those highly complicated chemical reactions at the solid-water interfaces. Even though a comprehensive understanding of the sorption behavior of radionuclides is significantly required for reliable safety assessment modeling, the number of the chemical thermodynamic model that can precisely predict the sorption/adsorption behavior of radionuclides is very limited. The machine-learning based approaches such as random forest, artificial neural networks, etc. provide an alternative way to understand and estimate complicated chemical reactions under arbitrarily given conditions. In this respect, the objective of this study is to predict the sorption characteristics of various radionuclides onto major bentonite minerals, as backfill materials for the HLW repository, in terms of the distribution coefficient by using a machine-learning based computational approach. As a background dataset, the sorption database previously established by the JAEA was employed for random forest machine learning calculation. Moreover, the hyperparameters such as the number of decision trees, the number of variables to divide each node, and random seed numbers were controlled to assess the coefficient of determination, R2, and the final calculation result. The result obtained in this study indicates that the distribution coefficients of various radionuclides onto bentonite minerals can be reliably predicted by using the machine learning model and sorption database.

Though many treatment technologies of contaminated water have been developed for a long time, it is still difficult to find a suitable method for large volumes of low radioactivity tritium-contaminated water. For this reason, most of the tritium-contaminated water been discharged to the biosphere or been stored in a special control area as radioactive waste. Activated carbon is a common material, but since there are few data on the treatment of tritium-contaminated water, its adsorption behavior to HTO is worth studied. In our study, for the tritium-contaminated water having a low radioactivity concentration (350-480 Bq/g), adsorption experiments were performed with activated carbon. The effects on the selective adsorption of HTO were investigated for temperature (5-55°C), hydrogen peroxide (1-10wt%) and activated carbon reuse (1-6 times) under non-equilibrium conditions. The treatment of activated carbon significantly reduced the radioactivity of tritium-contaminated water around 60 minutes of adsorption time. In order to clearly analyze the experimental results, positive factors and negative factors on the HTO selectivity were separately evaluated according to the adsorption time. Temperature and the reuse of activated carbon were evaluated as negative factors for HTO selectivity of activated carbon, whereas hydrogen peroxide (> 5wt%) was evaluated as a positive factor. By the evaluation method of separating the influencing factors into two types, the adsorption experimental results of HTO could be understood more clearly.

This study presents a rapid and quantitative radiochemical separation method for Nb isotopes in radioactive waste samples from the nuclear power plant with anion exchange resin after Fe coprecipitation. After radionuclides were leached from the radioactive waste samples with concentrated HCl and HNO3, the Nb isotopes were coprecipitated with Fe after filtering the leaching solution with 0.45 micron HA filter, while the Sr, Tc and Ni isotopes were in the solution. The Nb isotopes were separated in HCl medium with anion exchange resin. The purified Nb isotopes were measured using a low level liquid scintillation counter after installing quenching curve with standard Nb-94 isotopes. The separation method for Nb isotopes investigated in this study was applied to neutron dosimeter samples from the nuclear power plant after validating the Nb activity concentration with gamma spectrometry system.

According to the Nuclear Safety and Security Commission (NSSC) Notice No. 2021-26 “Delivery Regulations for the Low- and Intermediate Level Radioactive Waste (LILW)”, the activity of 3H, 14C, 55Fe, 58Co, 60Co, 59Ni, 63Ni, 90Sr, 94Nb, 99Tc, 129I, 137Cs, 144Ce, and gross alpha must be identified. Currently, the scaling factor of the dry active waste (DAW) for LILW is applied as an indirect evaluation method in Korea. The analyses are used the destructive methods and 55Fe, 60Co, 59Ni, 63Ni, 90Sr, 94Nb, 99Tc, and 137Cs, which are classified as nonvolatile nuclides, are separated through sequential separation and then measured by gamma detector, liquid scintillation counter (LSC), alpha/beta total counter (Gas Proportional Counter, GPC), and ICP-MS. We will introduce how to apply the existing nuclide separation method and improve the measurement method to supplement it.

A new method for chemical separation of light rare-earth elements (LREEs) using gas-pressurized extraction chromatography (GPEC) is described. GPEC is a microscale column chromatography system that features a constant flow of solvents (0.1 mL/min), which is created by pressurized nitrogen gas. The separation column with a Teflon tubing was packed with LN resin. We evaluated the separation of Ba, La, Ce, and Nd using various elution solvents. Here, we applied the natural isotopes of LREEs (La-139, Ce-140, and Nd-144) and barium (Ba-138) instead of radioactive isotopes for the preliminary test and reducing unnecessary radioactive waste. The column reproducibility of the proposed GPEC system ranged from 2.4% to 4.9% with RSDs of recoveries, and the column-to-column reproducibility ranged from 3.1% to 6.3% with RSDs of recoveries. This proposed GPEC method provides robust analysis and facilitates production of lesser chemical wastes and faster separation owing to the use of low solvent volume compared to traditional column chromatography.