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

Low- and intermediate level waste (LILW) repository in Gyeongju, Korea is in operation and the radioactive waste should satisfy the waste acceptance criteria (WAC) of the repository. Among the WAC of the Gyeongju LILW repository, the leachability index criterion is considered to be the criterion that is directly related to the isolation of the radionuclides from biosphere. Cesium, strontium, and cobalt should satisfy the leachability index larger than six by following the ANS 16.1 leaching test method. Several research were performed for the leachability index of Cs, Sr, and Co by following the ANS 16.1 leaching test method. However, the test condition of the previous research is expected to be different to the condition of the actual waste. Due to the radioactivity of the radionuclide such as Cs, Sr and Co, most of the research applied the surrogate of those radionuclides. The concentration of those nuclides was generally measured by the inductively coupled plasma (ICP) equipment, however, high concentration compared to the disposal limit of those nuclides due to the detection limit of the ICP was applied. From the Freundlich and Langmuir adsorption isotherms, the adsorption of the nuclides differs according to the concentration of the nuclides. As the leachability index of the nuclides is affected by the adsorption of the nuclides on the binding material, the effect of nuclide concentration is expected to be not ignorable. Therefore, the leachability index difference according to the nuclide concentration should be compared to avoid over- or underestimation of the leachability index. In this study, the difference in the leachability index according to the concentration of nuclides is aimed to be checked. Cs, Sr, and Co, which should satisfy the leachability index criterion in the WAC of the Gyeongju repository, were selected as target nuclides. Three concentrations were selected to compare the leachability index: 0.1 mol, 0.001 mol and below the regulatory exemption concentration. The concentration of non-radioactive nuclides in the leachant was measured by ICPOES and ICP-MS while the concentration of radionuclides was measured by HPGe. The result of this study can be applied as background data enhancing the WAC or disposal concentration limit of the radionuclides in Gyeongju LILW repository.

It has been investigated on the management of Strontium-90 in KAERI. It is needed to separate the solute from the salt solution for the recovery of strontium after the chlorination of the strontium oxide in molten salt. A vacuum distillation technology was used for the separation of strontium from the molten salt in our previous study. Strontium chloride was successfully carbonated by reactive distillation of SrCl2 – K2CO3 – LiCl – KCl system. In this study, it was tried to develop another route to recover strontium from the salt solution by a solid-solid reaction for avoiding the entrainment of product and the salt-K2CO3 reaction. Reactive distillation experiments were carried out for SrCl2 - K2CO3 – LiCl – KCl system. The carbonation temperature and pressure were 520°C and 0.8 bar. After the carbonation reaction, the temperature was elevated to 820°C to remove KCl from the reaction product. SrCO3 and KCl peaks were found in the XRD analysis of the residual product. It could be concluded that SrCl2 can be successfully carbonated after salt removal by the solid-solid reaction.

The operation of nuclear power plants, nuclear waste depositories, and the decontamination and decommissioning of nuclear power plants all have the possibility of generating various kinds of radionuclides that can be formed as gaseous or liquid phases. Among the radionuclides, strontium is considered as most harmful substance due to its abundance in nuclear accident effluent, long half-life, high fission yield, high water solubility, and high mobility in aquatic environment. To remove strontium from aquatic environment, adsorption technique is mainly used with high economic feasibility, efficiency, and selectivity. Previously, we synthesized sodium titanates with mid-temperature hydrothermal method as selective strontium adsorbent in aqueous solution. Moreover, it was demonstrated that synthesized sodium titanates show high strontium adsorption rate with high selectivity with high surface area, pore diameter and volume. Herein, we investigated the surface structure of synthesized sodium titanates before and after strontium adsorption in aqueous solution using scanning electron microscopy (SEM), energy dispersive X-ray spectroscopy (EDS), and X-ray photoelectron spectroscopy (XPS) analysis. According to SEM and EDS experimental results, aquatic strontium can be adsorbed as surface precipitation with formation of cube-shaped structure, which is quite similar strontium titanate structure crystals onto the surface of sodium titanates. In addition, XPS experimental results revealed that the titanium ions on the surface of sodium titanates were oxidized during strontium surface precipitation process, and the sodium ion on the surface of sodium titanates were exchanged with aquatic strontium ions via ion exchange process during strontium adsorption process.

The mobility of radionuclides is largely determined by their radiological properties, geochemical conditions, and adsorption reactions, such as surface adsorption, chemical precipitation, and ion exchange. To evaluate the safety assessments of radionuclides in nuclear sites, it is essential to understand the behavior and mechanism of radionuclides onto soils. Since nuclear power plants are located in coastal areas, the chemical composition of groundwater can vary depending on the intrusion of seawater, altering the adsorption distribution coefficient (Kd) values of radionuclides. This study examines the impact of seawater on the Kd values of clay minerals for cesium (Cs) and strontium (Sr). The results of Cs+ adsorption experiments showed a broad range of Kd values from 36 to 1,820 mL/g at an initial concentration of 1 mg/L and a high sorption coefficient of 15-613 mL/g at an initial concentration of 10 mg/L. Montmorillonite, hydrobiotite, illite, and kaolinite were ranked in terms of their CEC values for Cs+ adsorption, with hydrobiotite having the highest adsorption at 1 mg/L. The results of Sr adsorption experiments showed a wide range of Kd values from 82 to 1,209 mL/g at an initial concentration of 1 mg/L and a lower adsorption coefficient of 6.68-344 mL/g at an initial concentration of 10 mg/L. Both Cs+ and Sr2+ demonstrated lower Kd values at higher initial concentrations. CEC of clays found to have a significant impact on Sr2+ Kd values. Ca2+ ions showed a significant impact on Sr2+ adsorption distribution coefficients, demonstrating the greater impact of seawater on Sr2+ compared to Cs+. These findings can inform future safety assessments of radionuclides in nuclear sites.

90Sr is considered a hazardous radionuclide due to its long half-life of 28.8 years, radiotoxicity, and potential to bioaccumulate in various organisms. In the environment, strontium typically exists as divalent cation Sr2+ or in different complexes, depending on the environmental physical and chemical factors. Despite its mobility, Sr2+ transport remains affected by adsorption from solid phases, such as soil and sediments. This research aimed to investigate the efficiency of a magnetic flocculant (MNP/IF) in separating suspended soil and Sr2+ from a soil suspension. MNP/IF was prepared via the electrostatic interaction between magnetite particles and an inorganic flocculant (IF) composed of CaCO3 and Na2SO4. Analysis of the physical properties of MNP/IF confirmed that MNP/IF was successfully imparted with magnetism and had excellent adsorption capacity for Sr2+. The optimal MNP/IF dosage for the sedimentation of suspended soil was determined to be 0.3 g/g (mass ratio of flocculant to soil). The lower the pH, the more favorable the flocculation-sedimentation process of the suspended soil by MNP/IF, since Ca2+ and Mg2+, which are the most common strong flocculators, were further eluted from IF under acidic conditions. Besides, MNP/IF exhibited outstanding removal performance for Sr2+, with maximum adsorption capacities of 163.6 mg/g observed during the flocculation-sedimentation reaction of suspended soil. The adsorption of Sr2+ exhibited consistency with the Langmuir model and followed pseudo-second-order kinetics. These findings suggest that MNP/IF can be used for the simultaneous removal of suspended soil particles and Sr2+ from a radioactive soil suspension.

It has been investigated on the management of the nuclides in KAERI. Strontium-90 is a high heatgenerating nuclide in spent nuclear fuel. It is needed to separate the salt from the salt solution for the recovery of strontium after the chlorination of the strontium oxide in molten salt. A vacuum distillation technology was used for the separation of strontium from the molten salt. It was investigated on operating conditions of reactive distillation process for the recovery of the strontium from the salt solution. At a reduced pressure, considerable amount of the carbonation agents such as K2CO3 and Li2CO3 were reduced during heating in the distiller due to the thermal decomposition. Therefore, the two step process was proposed, which is composed of a reaction step at an atmospheric pressure and a salt distillation step at a reduced pressure. In the reaction step, the condition of low temperature and high pressure is suitable to suppress the decomposition of the carbonation agent. In the salt distillation step, reduced pressure is preferable at a suitable temperature depending on the evaporation rate of the salt.

PURPOSES : The effects of strontium cations on the strength of hydrated cement composites as well as the morphology and chemical composition changes of cement hydrates due to strontium treatment are investigated in this study. Subsequently, the potential of a strontiumbased aqueous solution as a near-surface treatment method for hydrated cement composites is evaluated. METHODS : To supply strontium cations to a hydrated cement composite, a 30% strontium nitrate aqueous solution was used. Cement paste (w/c = 0.4) specimens were prepared and cured in the 30% strontium nitrate aqueous solution, which allows the strontium ions to penetrate into the specimen and treat the near-surface region. Compressive and flexural strength tests were performed on both specimens treated by the strontium ions and untreated specimens cured in deionized water, and the test results were compared. To investigate the changes in the morphology and chemical compositions of the cement hydrates due to the treatment, scanning electron microscopy (SEM) and energy dispersive spectrometry (EDS) analyses were performed. RESULTS : The results of the strength tests indicate that both the compressive and flexural strengths of the specimens cured in the 30% strontium nitrate aqueous solution improved significantly compared with the specimens cured in deionized water for the same duration. In particular, the strontium nitrate aqueous solution shows greater improvement in terms of the flexural strength compared with the compressive strength. The maximum penetration depth of strontium into the hydrated cement composite is 5 mm during the first 7 d of immersion and increases to 6.5 mm during the subsequent 14 d. The SEM results show that the microstructure of the hydrated cement composite is densified by the strontium nitrate aqueous solution treatment. The EDS results show that morphology and chemical compositions of the cement hydrates are altered. This implies that the strontium cations can be combined with C–S–H and C–S–A–H phases to form new types of hydrates. CONCLUSIONS : The experimentally results show that the strength of hydrated cement composites can be improved by modifying their near-surface regions via the strontium cation penetration technique. This implies that the strontium-based aqueous solution exhibits high potential for the maintenance and rehabilitation of concrete structures.

Cardiovascular disease remains as one of the most common causes of high morbidity and mortality worldwide, despite remarkable medical advances in recent decades. Non-invasive techniques play a preeminent role in prevention of cardiovascular disease by diagnosing it at an early stage and guiding optimal patient management. Nuclear imaging is one of the most powerful means available for noninvasive diagnosis and management of poorly perfused myocardial region resulting from the cardiovascular disease. Several radionuclides are available for monitoring blood flow to cardiac tissue. The most validated radionuclides for these measurements are 13N, 15O, 99mTc, 201Tl and 82Rb. Each of 13N, 15O and 201Tl require the presence of an on-site cyclotron, whereas, 82Rb and 99mTc require only a generator. Rubidium (Rb) is an alkali metal ion that acts biologically like potassium and accumulates in cardiac muscle tissue. Rb has a rapid blood clearance profile which allows the use of 82Rb. It also has an ultra-short physical half-life of 75 sec for non-invasive evaluation of regional cardiac blood flow. There are several advantages of 82Rb over other radionuclides. Having a short half-life significantly reduces the radiation dose to the patient. In addition, 82Rb is a positron emitter, which gives the full advantages of PET such as image quantification with superior sensitivity. Several reports have shown superior diagnostic performances of 82Rb-PET over conventional 99mTc-SPECT. 82Rb can be produced from a generator system by the decay of its 25.6-day half-life parent 82Sr. However, the 82Sr parent is difficult to prepare. In routine generator production, certain purity is required to meet the specification of the product. Since there has been no the use of 82Rb radionuclide for research or medical purpose in Korea, we have plans to produce 82Sr with certain purity and develop a 82Sr/82Rb generator system. These studies can also be applied to remove radioactive Sr from radioactive waste waters. Because ion exchange resin, used for purification of 82Sr from impurities, is also utilized to trap radioactive Sr2+ ions from radioactive waste water. After Fukushima Daiichi nuclear accident, interest in the treatment of radioactive waste water has surged. As one of main fission products of nuclear reactor, 90Sr has been regarded as a hazardous radionuclide with half-life of about 29 years. Therefore, the investigation on ion exchange resin is important for removal of 90Sr from radioactive waste water. Here, we optimized 82Sr purification method using ion exchange resin to establish the most suitable procedure.

It has been studied on the disposal area reduction for the used nuclear fuel by the management of high decay-heat nuclides, long-lived nuclides, and highly mobile nuclides. It was investigated on the management of the nuclides in KAERI. Strontium-90 is a high heat-generating nuclide in spent nuclear fuel. It is needed to separate the salt from the salt solution for the recovery of strontium after the chlorination of the strontium oxide in molten salt. Vacuum distillation was used for the separation of strontium from the molten salt. Potassium carbonate was chosen as a reactive distillation reagent for SrCl2 – LiCl – KCl system by the thermodynamic calculation. Reactive distillation experiments were carried out. The residual was mainly SrCO3 in the XRD analysis. It could be concluded that K2CO3 could be one of the suitable reagents for the reactive distillation. The salt in the long–lived nuclide powders should be removed to prepare the block for disposal. Experiments were carried out using W powders (surrogate) and U3O8 powders to develop a process for the removal of the residual salt from UOx powders. The salts were successfully removed from the W and U3O8 powders by distillation.

Concrete is used as the main engineering barrier in low and intermediate level radioactive waste disposal facilities. As the time passed, the radionuclides stored in repository may contact with groundwater and leak into the ecosystem through the rock media. In this process, the radionuclides can react with calcite via sorption or coprecipitation, because calcite is the major mineral of concrete. Under the various background conditions in repository, frequent dissolution-precipitation reactions can happen. Dissolution of Sr-coprecipitated calcite may be different from that of SrCO3(s) which can mislead the safety performance of radioactive Sr and the estimate of Sr mobility based on the solubility of SrCO3(s). Strontium is not only one of the fission products but also emits beta rays with a long half-life almost 29 years. The strontium may be released or retarded by the dissolution-precipitation reactions in repository. In this study, the dissolution of Sr-coprecipitated with respect to calcite was tested in various environment conditions. The Sr-coprecipitated calcite, (Sr,Ca)CO3(s) was synthesized by coprecipitation method in alkaline condition. The 250 mL of 0.1 M of CaCl2 solution was mixed with 250 mL of 1.14 mM SrCl2·6H2O solution. Then, independently prepared 500 mL of 0.1 M Na2CO3 solution was mixed with the mixed solution of CaCl2 and SrCl2. The precipitates could be made and they were aged for 3 days at room temperature. Then, the supernatant was separated by the centrifugation and the solid at the bottom was dried in an oven at temperature 80°C. After that, the Srcoprecipitated calcite powder was washed using the DI water several times and dried again before use. Characterization of solid powder was conducted by XRD and SEM, and the ICP-MS and ICP-AES were used to analyze the concentrations of Ca and Sr. The batch dissolution experiment was conducted with a solid-to-solution ratio of 10 g/L groundwater in polyethylene tubes. The oxidative groundwater was synthesized by simulating the chemical composition of KAERI Underground Research Tunnel (KURT) DB-3 groundwater. Different temperatures and pHs were prepared and tested for the release of Sr and Ca from the coprecipitated (Sr,Ca)CO3(s) to compare the results with the release of Sr and Ca from SrCO3(s) and CaCO3(s), respectively. Such as, these results will be used to provide better understanding of Sr release and mobility in various repository environments.

Radioactivity of radiostrontiums, Sr-89 and Sr-90, which are both pure beta-emitters, are generally measured via Cherenkov counting. However, the determination of Cherenkov counting efficiencies of radiostrontiums requires a complicated procedure due to the presence of Y-90 (also a pure betaemitter) which is the daughter nuclide of Sr-90. In this study, we have developed a machine learning approach using a linear regression model which allows an easier and simultaneous determination of the Cherenkov counting efficiencies of the radiostrontiums. The linear regression model was employed because total net Cherenkov count (Ct) from the three beta-emitters at time t after the separation of Y- 90, can be expressed as a linear combination of their respective time-varying radioactivities with their respective coefficients (parameters) being their counting efficiencies: Ct = εSr-90[ASr-90·exp(–λSr-90·t)] + εSr-89[ASr-89·exp(–λSr-89·t)] + εY-90[ASr-90·exp(1–λSr-90·t)], where ε is a counting efficiency, A is an initial activity, λ is a decay constant and t is time after the separation of Y-90, Thus, if we train the model with multiple Cherenkov counts measured from the three beta emitters, then we can obtain their estimates for counting efficiencies (so-called parameters) straightforward. For this, the model has been trained by two methods: Ordinary Least Squares (OLS) and Bayesian linear regression (BLR), for which two software packages, PyMC3 and Stan were employed to compare their performances. The results showed that the accuracy of the OLS was worse than that of the BLR. Particularly, the counting efficiency of Sr-90 was estimated to be smaller than 0, which is an unrealistic value. On the other hand, the estimates of the BLR gave realistic values which are close to the true values. Additionally, the BLR was able to provide a distribution for each counting efficiency (so-called “posterior”) from which various types of inference can be made including median and credible interval in the Bayesian statistics which is analogous to, but different from confidence interval in the Frequentist statistics. In the results of the BLR, the Stan package gave more accurate estimates than the PyMC3 package. Therefore, it is expected that counting efficiencies of the radiostrontiums including radioyttrium can be determined at the same time, more easily and accurately, by using the BLR with the Stan package and that the activities of radiostrontium also can be determined more easily by using the BLR if we know their counting efficiencies in advance. It is worth noting that the usage of the linear regression model in this study was different from the usual one where the trained model is used to predict a response value (count) from a set of unseen regressor values (activities).

Magnesium potassium phosphate cements (MKPCs) are prepared by the acid-base reaction of dead burned magnesia (MgO) and monopotassium phosphate (KH2PO4). Low-pH cementitious materials such as MKPCs are currently of interest for the geological disposal of nuclear waste. MKPCs have advantages such as high early strength, high bonding strength, small drying shrinkage, low permeability, and high sulfate resistance. According to the results of previous studies, it is known that cesium, strontium, and cobalt are immobilized in the form of MgCsxK1−xPO4·6H2O, MgxSr1−xKPO4·6H2O, and Co3(PO4)2, respectively, in MKPCs. However, these results were predicted based on thermodynamic data, not directly observed precipitates to clearly show the evidence. Therefore, in this study, we directly analyzed the immobilized forms of Cs, Sr, and Co, respectively. CsNO3, Sr(NO3)2, and Co(NO3)2·6H2O powders (0.3 mol each) were mixed individually in each of the MKPC suspensions. The suspensions in which KH2PO4 was dissolved were pH 4.3 and the dissolution of MgO decreased the H+ concentration, raising the pH close to 11. The hydration products according to pH evolution in the MKPC suspensions were analyzed, and the change in the concentration of ions in the aqueous solution was also measured. An aqueous solution was obtained using a syringe filter (0.45 μm) to analyze the ion concentrations in the solution of the suspension. The collected solutions were diluted with nitric acid and analyzed using inductively coupled plasma mass spectrometry. To characterize the solid phases, the suspensions were obtained with a pipette at specific times and filtered under a vacuum in a Buchner funnel. Because the amounts of hydration products including Cs, Sr, and Co were small, it was not observed by XRD and TGA analysis, but their components could be analyzed by SEM-EDS. The final precipitate forms of Cs, Sr, and Co in the MKPC matrix are MgCsPO4·6H2O, SrHPO4, and Co3(PO4)2·8H2O, respectively.

Strontium-90 is a high heat-generating nuclide in spent nuclear fuel. The removal of the nuclide separation is indispensable to reduce the burden of storage and disposal of high-level radioactive waste. Korea Atomic Energy Research Institute has developed the molten salt immersion technique to separate the strontium by the chlorination of the strontium oxide in molten salt. It is needed to separate the salt for the recovery of strontium from the salt solution after the chlorination reaction. In this study, it was investigated on the recovery of the strontium from the salt. Vacuum distillation was used for the separation of strontium from the molten salt. The vapor pressures of the candidate salts were calculated by HSC chemistry and the apparent evaporation rates (AER) were measured at 830°C to evaluate the salts for strontium recovery. The candidate salts were LiCl, KCl, MgCl2, NaCl and CaCl2. The AERs of MgCl2 and NaCl were 1.9 and 1.3 g/cm2-h, respectively. Those two salts can be separated from the strontium compound even though the AER values are much lower than those of LiCl-KCl (~ 8 g/cm2-h). CaCl2 salt was rarely evaporated (AER < 0.03 g/cm2-h) and it is not suitable to use as a strontium recovery salt. Therefore, MgCl2, NaCl, LiCl and KCl can be regarded as candidates for a strontium recovery salt.

Heavy metal pollution has a harmful impact on human health and is regarded as a vital problem. Preparation of a novel, low cost bio-sorbent for heavy metal sorption is the main target of this research. Non-living Chlorella Vulgaris Alga/Date pit activated carbon composite (1:1), (CV/AC), is a novel bio-sorbent prepared by the wet-chemical method for sorption of Pb (II) and Sr (II) from aqueous media. The optimum pH for sorption reaction is 5 and the equilibrium time is achieved within 1 h. The sorption efficiencies are 90.5% for Pb(II) and 95.7% for Sr(II) with initial concentration Co 10 mg L– 1 at 298 K. The monolayer sorption capacities of CV/AC composite at 298 K and pH = 5 were 6.34 ± 0.059, 5.97 ± 0.22 mg g– 1. The saturation capacities were 98.5 and 125 mg g– 1 for Pb (II) and Sr (II), respectively after 10 days. The sorption process is a spontaneous and endothermic reaction. It follows a pseudo-2nd-order mechanism. The results are suggestive of the need to adopt CV/AC composite as a potential bio-sorbent of Pb (II) and Sr (II) for waste water treatment.