고치벌과의 송충살이고치벌아과에 속하는 Aleiodes thirakupti (Butcher et al., 2012)을 국내 최초로 보고한다. 본 종에 대한 진단, 분포 및 삽 화를 제공한다.

과실파리, 굴파리의 유충에 기생하는 Areotetes van Achterberg & Li, 2013(벌목: 고치벌과: 꽃파리고치벌아과)는 중국에서 처음으로 보고 된 바 있다. 현재까지 Areotetes속에는 4종이 보고되어 있다. 금번 연구에서 Areotetes속의 1종, Areotetes carinuliferus van Achterberg & Li, 2013를 한국에서 처음으로 보고하며, 본 종의 형태 진단, 분포, 도해도를 작성하였고, 추가적으로 미토콘드리아 COI 데이터를 제공한다.

This study demonstrates a different approach method to fabricate antimony selenide (Sb2Se3) thin-films for the solar cell applications. As-deposited Sb2Se3 thin-films are fabricated via electrodeposition route and, subsequently, annealed in the temperature range of 230 ~ 310oC. Cyclic voltammetry is performed to investigate the electrochemical behavior of the Sb and Se ions. The deposition potential of the Sb2Se3 thin films is determined to be -0.6 V vs. Ag/AgCl (in 1 M KCl), where the stoichiometric composition of Sb2Se3 appeared. It is found that the crystal orientations of Sb2Se3 thin-films are largely dependent on the annealing temperature. At an annealing temperature of 250 oC, the Sb2Se3 thin-film grew most along the c-axis [(211) and/or (221)] direction, which resulted in the smooth movement of carriers, thereby increasing the carrier collection probability. Therefore, the solar cell using Sb2Se3 thin-film annealed at 250 oC exhibited significant enhancement in JSC of 10.03 mA/cm2 and a highest conversion efficiency of 0.821 % because of the preferred orientation of the Sb2Se3 thin film.

Designing and producing a low-cost, high-current-density electrode with good electrocatalytic activity for the oxygen evolution reaction (OER) is still a major challenge for the industrial hydrogen energy economy. In this study, nanostructured Fe-doped CuCo(OH)2 was discovered to be a precedent electrocatalyst for OER with low overpotential, low Tafel slope, good durability, and high electrochemically active surface sites at reduced mass loadings. Fe-doped CuCo(OH)2 nanosheets are made using a hydrothermal synthesis process. These nanosheets are clumped together to form a highly open hierarchical structure. When used as an electrocatalyst, the Fe-doped CuCo(OH)2 nanosheets required an overpotential of 260 mV to reach a current density of 50 mA cm−2. Also, it showed a small Tafel slope of 72.9 mV dec−1, and superior stability while catalyzing the generation of O2 continuously for 20 hours. The Fe-doped CuCo(OH)2 was found to have a large number of active sites which provide hierarchical and stable transfer routes for both electrolyte ions and electrons, resulting in exceptional OER performance.

Measurement of the physical properties of high-temperature molten salts is important for the efficient design and operation of molten salt reactors (MSR) in which the reactor coolant and nuclear fuel are in a homogeneous liquid state. Although some crucial physical properties such as viscosity, thermal conductivity, density, etc., have been drawing much attention, relative data, especially for molten chloride salts, are scarce. Thus, it is urgent to prepare the viscosity data as one of the key transport properties in thermal hydraulics analysis. However, it is not an easy task to measure the molten salt viscosity with high accuracy due to end effect, a small gap between the chamber and spindle, thermal expansion of the chamber and spindle at high temperatures in a rotational viscometer. Additionally, molten salt temperatures inside furnace are not uniform due to the large temperature gradient inside the chamber, and therefore the assumption of laminar condition can be violated. In this study, geometric factors, which can be a major interference in the torque measurement, were considered for the accurate determination of the viscosity. We established a high-temperature molten salt viscosity measurement system with Brookfield rotational viscometer. KNO3 molten salt was used as a model substance at a temperature range of 650–773 K. In-house designed spindles and chambers were made of corrosion-resistant alumina. Thermal expansion has a significant influence on the size and shape of the chamber and spindle. The effect of thermal expansion on the conventional correction method was examined with temperature variation and distribution. Gap size variation was also investigated in order to improve the accuracy.

This study presents a rapid and quantitative sequential separation method for H-3 and C-14 isotopes with distillation apparatus in environmental samples released from nuclear facilities. After adding 200 mg of granulated potassium permanganate and 500 mg of sodium hydroxide in 100 mL of sample solution, the sample solution was heated until approximately 10 mL of distillate, and the distillate fraction was removed. The sample solution was heated again until a minimum 10 mL of additional distillate was collected. 10 mL of distillate was transferred to the LSC vail and the measurement sample for H-3 was made by adding 10 mL of Ultima Gold LLT to the LSC vial. After adding 2.5 g of potassium persulfate, 2 mL of 1M silver nitrate and 15 mL of concentrated nitric acid to the remained sample solution, the sample solution was heated for 90 minutes and C-14 isotopes were adsorbed into 10 mL of Carbo-Sorb solution in glass vial. The measurement sample for C-14 was made by adding 10 mL of Permafluor to the C-14 fraction in glass vial. The purified H-3 and C-14 samples were measured by the liquid scintillation counter after quenching correction. The average recoveries of H-3 and C-14 with CRM were measured to be 96% and 85%, respectively. The sequential separation method for H-3 and C-14 investigated in this study was applied to activated charcoal filter produced from nuclear power plants after validating the reliability by result of proficiency test (KOLAS-KRISS, PT-2021-51).

According to the article 18 of NSSC notice “Regulations on the delivery of low-and intermediatelevel radioactive wastes”, the consignor shall establish and implement the quality assurance program about waste management to ensure conformity with the criteria set forth in the regulations and detailed criteria proposed by the disposal facility operator, including matters related to characterization of the waste concerned. To meet the above requirement, commercially available laboratory information management system, STARLIMS from Abbott Informatics was introduced in the late of 2019 and was customized to our standardized test method in 2020. In that time, Electronic Lab Notebook (ELN), which is an electronic system to create, store, retrieve, and share fully electronic records, was tailored to replace paper lab notebook. Scientific Data Management System (SDMS), which is computer system used to capture, centrally store, catalog, and manage data, was installed. Due to the parsing ability of SDMS, human error like mistake while data entry was reduced by extracting data from measurement sheet and exporting measurement data to designated area of ELN and this feature made work efficiency improved. Afterward, validation of STARLIMS was conducted following the procedure of user acceptance testing including Operational Qualification and Performance Qualification. As a result of these activities, STARLIMS has been officially operated and applied to means to manage test results since 2021. In 2021, for user-friendly environment, updating STARLIMS was also conducted by applying SDMS to import data from other radiometric measurements including gas proportional counter (GPC), liquid scintillation counter (LSC), and low-energy Ge detector (LEGe) besides HPGe detector for gamma measurement. From implementation to operation, it is confirmed that STARLIMS has been providing reliable and stable platforms to manage laboratory information regarding measurement records and playing a significant role in tool to meet the quality assurance required.

Density of chloride molten salts is an essential physical property in the reactor core design and thermal-hydraulic design simulation, especially in molten salt reactor (MSR) design currently under development in Korea. NaCl-MgCl2-UCl3 pseudo-ternary system is one of the various candidate chloride-based salt mixtures because it has relatively-low melting point, very low vapor pressure, high thermal conductivity, etc. However, to the best of our knowledge, the density data of NaCl-MgCl2- UCl3 have not yet been measured or published worldwide, and therefore the ballpark figures of the density should be given for the preliminary reactor design. In our present study, the density estimation of NaCl-MgCl2-UCl3 based on the pseudo-binary data, i.e., NaCl-MgCl2, MgCl2-UCl3, and NaCl- UCl3, reported in the literature previously were performed using the Redlich-Kister model. Binary interaction parameter for MgCl2-UCl3 was higher than that for NaCl-MgCl2 and lower than that for NaCl-UCl3. As an example, calculated density of 0.62 NaCl: 0.18 MgCl2: 0.20 UCl3 at 873 K was 2.578 g·cm−3. In our further study, the methodology using Redlich-Kister model will be applied to more complex multicomponent systems and to other physical properties such as viscosity, thermal conductivity, surface tension, etc.

In 2005, groundwater contamination due to unplanned releases of radioactive materials from the US. Nuclear Power Plants (NPPs) such as Braidwood and Indian Point was confirmed. The following year, in 2006, The Nuclear Regulatory Commission (NRC) established a task force team to investigate the history of unplanned release of all NPP in the US. As a results 217 events of unplanned release including leaks and spills were identified in the US NPPs. The NRC regulates the radioactivity concentration of off-site groundwater by setting a reporting levels (RLs), and if exceeds the RLs, the licensee must report within 30 days. When the off-site groundwater is used as drinking water or non-drinking water, the RLs for tritium in groundwater are 740 Bq·L−1 or 1,110 Bq·L−1, respectively. Whereas the NRC does not set the RLs for on-site groundwater. The Nuclear Energy Institute (NEI) issued the guidance document “Industry groundwater protection initiative” NEI 07-07 in 2007. And the members of the NEI promised with regulatory body and local governments to implement groundwater monitoring/protection program according to the NEI 07-07. The document states that when the on-site groundwater is used as drinking water, the RL (740 Bq·L−1) for off-site groundwater will be applied and the licensee voluntarily reported to the NRC. And also, NPPs are setting the Investigation Level (IL) below the RP and the IL is various among the NPPs. The IL is the standard by which detailed investigations are implemented when the level (radioactivity concentration) is exceeded.

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. KHNP evaluates the dose of residents around nuclear power plants using the K-DOSE60 program that reflects ICRP-60. K-DOSE60 calculates the expected exposure dose for residents by modifying the atmospheric dispersion and deposition factors evaluation module (XOQDOQ), gaseous effluent evaluation module (GASDOS) and liquid effluent evaluation module (LIQDOS) developed by the US NRC. The current evaluation program is the Bounding Assessments method, which evaluates under the assumption that residents reside at the exclusion area boundary (EAB), and has a disadvantage in that the estimated exposure dose is evaluated too conservatively. In the EPRI, instead of the conservative method that is conventionally performed for the residents’ dose evaluation method, a plan to improve the accuracy of the dose evaluation reflecting the site characteristics was reviewed. In addition, improvements were derived through the review of NPPs operation status, experience cases and the latest technology.

The dose was evaluated for the workers transporting the spent resin drums from a spent resin mixture treatment facility. The treatment technology of spent resin mixture waste based on microwave was developed to compensate for the shortcoming of the existing one. The mechanism of the facility for the treatment is divided into separation, desorption, condensation and adsorption process. The treated spent resin that has passed through the microwave reactor flows into the spent resin storage tank. As the treatment time elapses, if spent resin accumulates in the spent resin storage tank, it is moved to the drum of the volume of 200 L. The drum must be moved by the worker, in which case radiation exposure to the drum transport worker occurs. It requires the dose evaluation for drum transport workers in terms of radiation safety. Dose evaluation was performed in consideration of the change in the composition ratio and weight of the spent resin mixture, where the working time for transportation was considered from 10 to 120 minutes in 10-minute increment. In the case of 100 kg of the spent resin mixture, the dose range was derived as 4.62×10−3 – 5.90×10−2 mSv for the 100 kg of spent resin, 4.72×10−3– 5.58×10−2 mSv for the 80 kg of spent resin and 20 kg of zeolite and activated carbon, and 5.38×10−3 – 6.32×10−2 mSv for the 60 kg of spent resin and 40 kg of zeolite and activated carbon. In the case of 150 kg of the spent resin mixture, the dose range was derived as 6.83×10−3 – 8.20×10−2 mSv for the 150 kg of spent resin, 7.13×10−3 – 8.22×10−2 mSv for the 120 kg of spent resin and 30 kg of zeolite and activated carbon, and 8.28×10−3 – 8.86×10−2 mSv for the 90 kg of spent resin and 60 kg of zeolite and activated carbon. The estimated maximum doses for each weight (100 kg and 150 kg of mixture) were confirmed to be 3.16×10−1% and 4.43×10−1% of the annual average dose limit of 20 mSv for radiation workers.

Korea Institute of Radiological and Medical Sciences provides proton irradiation service of up to 40 MeV using cyclotron. The use of such a cyclotron was approved in advance to satisfy the Nuclear Safety Act, and radiation safety was evaluated in this process. The Monte Carlo method is generally used to evaluate the shielding safety of high-energy accelerators, and MCNP 6.2 was used in the previous evaluation. In this study, in order to verify the results of previous evaluation, the calculation results of MCNP 6.2 and Particle and Heavy Ion Transport code System (PHITS) 3.24 are compared. PHITS is a general-purpose Monte Carlo particle transport simulation code that is used in many studies in the fields of accelerator technology, radiotherapy, space radiation, etc. In the previous evaluation, the effective dose by neutrons and photons generated by the collision of 40 MeV 20 μA of protons with a 10.5 mm thick beryllium target was evaluated, and in this study, this was reproduced with PHITS. As the radiation exposure evaluation for the user or pubic is evaluated based on the radiation dose and energy distribution generated around the target, the effective dose and energy distribution received by the water phantom with a radius of 1 cm on the front, side, and back of the target were calculated. T-Track, a tally of PHITS, was used to calculate effective dose, which is similar to F4 tally of MCNP 6.2 using a dose conversion factor. For the dose conversion factor, the value suggested as AP irradiation in Publication 103 was used. As a result of the calculation, the effective dose by neutrons at the front, side and back of the target was 1.42×105, 2.09×104, and 1.39×104 mSv·h−1, respectively, which was similar to 2.00×105, 1.84×104, and 2.59×104 calculated using F4 tally in MCNP. Moreover, the results of calculating the effective dose by photons using PHITS were 4.81×10, 3.10×10, and 2.66×10, respectively, and the results of calculating MCNP were 4.49×102, 6.45×10, and 9.64×10. The average energies of neutrons were 11.2, 0.69, and 0.31 MeV when calculated by PHITS, respectively, and 13.8, 7.8, and 4.6 when calculated by MCNP. Moreover, the average energies of photons were 1.98, 0.98, and 0.86 when calculated by PHITS, respectively, and 3.9, 3.2, and 2.6 when calculated by MCNP.

In Korea, research on the introduction of dry storage facility is being conducted as an alternative to saturation of temporary storage facilities for spent nuclear fuel. The introduction of dry storage facilities requires a radiological impact assessment on the workers of the facility, and for this, an appropriate exposure scenario must be derived through work procedure analysis. In this study, the procedure for storing spent nuclear fuel in dry storage facilities was analyzed based on the case of evaluating the radiological impact of workers in dry storage facilities abroad. We investigated cases of radiological impact assessment on workers at on-site dry storage facilities by PNNL, Dominion, and P. F. Weck. PNNL and Dominion analyzed the storage work procedure of the VSC (Vertical Storage Cask) method using CASTOR V/21, TN-32, respectively, and conducted a radiological impact assessment. P. F. Weck analyzed the storage work procedure of various spent nuclear fuel casks for VSC and HSM (Horizontal Storage Module), conducted a radiological impact assessment. As a result of comparing the procedure for storing spent nuclear fuel by case, it was found that the storage procedure was determined by the storage method and the cask type. In the case of VSC method, canister-type casks and basket-type casks are used, and the storage procedure are partially different according to each. Canister-type cask requires repackaging from transfer overpack to storage overpack, but basket-type cask doesn’t require that procedure. In the case of the HSM method, only the canister type cask was found to be used. However, the storage procedure was different depending on the type of HSM system. Depending on the type of HSM system, the necessity of cask for on-site transport was different. In this study, we investigated and analyzed the work procedure according to the storage method of dry storage facilities abroad. It was found that the dry storage procedure of spent nuclear fuel different according to the storage method and type of cask. The results of this study can be used as basic when deriving the exposure scenario for spent nuclear fuel dry storage workers suitable for the domestic situation.

The Co-60 is a radioactive material widely used in domestic and foreign medical, industrial, health and research fields. Currently, world market for the Co-60 is about 80 MCi/yr and is expected to grow to 150 MCi/yr by 2025. For the Co-60, Nordion of Canada occupies about 80% of the world market. In the case of Korea, a small amount of sources with relatively low radioactivity intensity are produced using research reactors, but most of the Co-60 is entirely dependent on imports. Accordingly, although the technical feasibility of the Co-60 production technology using the PHWR was evaluated, it was evaluated as a negative result on the additional construction of a hot cell, core management, safety analysis and economic feasibility. Canada, the main producer of the Co-60, is also conducting research on the Co-60 production technology using PWR with GE-Hitachi and Westinghouse as the number of PHWR is expected to decrease. In Korea, it is necessary to preoccupy the Co-60 production technology and auxiliary technology using the PWR by utilizing excellent technology, and active research is being conducted to secure unique nuclear power technology that does not depend on foreign countries. Therefore, in this study, the thickness and weight of the radioactive shielding required for handling (transport) of Co-60 produced using the PWR were calculated.

RADTRAN is a code that assesses the radiation risk of radioactive material transportation. RADTRAN assumes that the package is a point source or a line source regardless of package type and corrects the external dose rate using a shape factor which depends on the critical dimension of the package. However, the external dose rate calculated using a shape factor may be different from the actual external dose rate. Therefore, it is necessary to analyze the effect of the shape factor on the external dose rate. In this study, the effect of the shape factor on the external dose rate in RADTRAN was analyzed by comparison with MCNP. This study analyzed change in external dose rate depending on the distance from the package and the critical dimension. The distance from the package was in the range of 1–800 m. The shape of the package was assumed to be cylindrical with a radius of 1 m, and the critical dimensions of the package were assumed to be 2, 4, and 8 m. Attenuation and build-up in the air were not considered to consider only the effect on the shape factor. When simulating the exposure situation using MCNP, the package was assumed to be a volume source, and flux by distance from the package was calculated using F5 tally. The dose rate at 1 m from the package was normalized to 2 mSv·hr−1. As a result of the analysis, the external dose rates of the package were higher in RADTRAN than in MCNP. For the critical dimension of 2, 4, and 8 m, when the distance from package is 1–10 m, the RADTRAN was 1.83, 4.08, and 5.27 times higher on average than MCNP, respectively. And when the distance from the package was 10–100 m and 100–800 m, RADTRAN was 1.10, 2.02, 3.01 times and 1.04, 1.92, 2.43 times higher than MCNP, respectively. It was found that the larger the distance from the package is and the smaller the critical dimension of the package is, the less conservatively RADTRAN assessed. It is because the shape of the package gets closer to the point source as the distance from the package increases, and the shape factor decreases as the critical dimension of the package decreases. The result of this study can be used as the basis for radiation risk assessment when transporting radioactive materials.

There are many Systems, Structures, and Components (SSCs) in Nuclear Power Plants (NPPs). The systems include radiological waste treatment system, spent fuel pool cooling, emergency core cooling systems, etc. The structures include reactor building, piping vaults, radioactive waste storage facilities, etc. The components include valves, pumps, piping segments, etc. Radionuclides exist in some of these SSCs and unplanned release may occur when leaks or spills from them. And also Work Practice (WP) is another reason of unplanned release in NPPs. The WP is defined as an action taken by individuals during maintenance, operational or support activities, which could result in or prevent a spill or leak of a radioactive solid, liquid or gas that has a credible mechanism for contamination of groundwater. According to the results of the Electric Power Research Institute (EPRI) survey, a total of 323 unplanned release event occurred at US NPPs from 1970 to 2014. Among them, 219 events were counted to have occurred at pressurized water reactors (PWRs). In addition, it was confirmed that 41 of the 44 PWR sites (about 93%) in the US, operated at the time of the survey period, had experienced at least one unplanned release events of licensed material which impacted groundwater. This means that the US PWR sites have experienced an average of approximately 5 unplanned release event per site. The source with the most unplanned releases, including SSCs and WP, was miscellaneous systems with a percentage of about 33% (72 events). Miscellaneous systems include pipes, and it was confirmed that unplanned releases mainly occurred in pipes such as the main steam system, condensate and feedwater system, and emergency core cooling system. And the percentage was high in the order of WPs (21%, 45 events), radioactive effluents (20%, 43 events), refueling water storage (8%, 17 events), radioactive waste/material operations (7%, 16 events), spent fuel storage (5%, 12 events), unknown (4%, 9 events), and structures (2%, 5 events). The history of the unplanned release of the US NPPs will be considered when revising major SSCs in the domestic NPP groundwater monitoring program.

In order to monitor the contamination of groundwater due to unplanned release of radioactive materials and the spread to off-site environments, the nuclear power plants (NPPs) conduct groundwater monitoring program (GWMP) in Korea. The GWMP should be established based on the groundwater flow model reflecting the conceptual site model (CSM) of the NPP’s site. In this study, in order to optimize the GWMP, the existing CSM and the groundwater flow model of the domestic NPPs site was updated by reflecting the latest groundwater level. As part of the CSM improvement, the hydrogeological units were subdivided more detailed from three to six through the review of hydrogeological characteristics of the NPPs site. In addition, major variables that affect groundwater flow, such as water conductivity, have been updated. The groundwater flow model was revised overall as the CSM was improved. In particular, the excavation depth of the structure and backfill area generated during the construction stage of the NPP structures was accurately reflected, and the drainage boundary conditions were realistically reflected. To verify the revised groundwater flow model, steady-state correction was performed using the groundwater level measured in April, 2021. As a results of the steady-state correction, the standard error of estimate, root mean square (RMS), normalized RMS, and the correlation coefficient were 0.32 m, 1.692 m, 5.608%, and 0.964, respectively. This means that the groundwater flow model is reasonably constructed. The CSM and groundwater flow model improved in this study will be used to optimize the monitoring location of groundwater in NPPs.

Considering the characteristics of nuclear power plants in order to decommission nuclear power plants safely and economically, this thesis provides a methodology for optimizing the technology for developing decommissioning characteristic evaluation system using simulation technology for core facilities of the plants based on 3D that reflects various factors. The results of pollution assessment and radiation assessment for the Kori Unit 1 reactor building, auxiliary building, and each major device are displayed in 3D drawings and viewer, and the radiation dose rate and radiation assessment results are displayed separately for each major location. Furthermore, this D/B development method which includes inserting result values of characteristic evaluation and the quantity of waste is one of the main technology to optimize the system which enables users to select decommissioning processes and predict the quantity of waste. (Refer to the presented 3D models of the containment building, D/B, tag search module, the scale calculation result of models after visualizing the result value of 3D based decommissioning characteristic evaluation) The methodology for optimizing decommissioning characteristic evaluation result value DB development system using 3D models of the first major nuclear power plant allows the display of decommissioning characteristic values in virtual reality, the selection of decommissioning process, the establishment of the decommissioning procedure. Hence, this study is expected to provide reliable guidelines for managing a decommission business efficiently in the near future and can be used in the related field if needed.

Detritiation of low-level tritiated water has become global issue after Fukushima accident. Several attempts have been made to reduce the radioactivity of Fukushima tritiated water below legal limit of nuclear plant effluents (~104 Bq·L−1). Various technologies such as water distillation, electrolysis, and catalytic exchange were tested to treat the tritiated water, however, those demand enormous expense to achieve the goal due to low process efficiency. It highlights that the performance enhancement of current technologies is necessary to improve economic feasibility. We have quantitatively evaluated the separation performance of various polymers toward low-level tritium (~105 Bq·L−1) through batch experiments. The polystyrene with grafted by 20 types of functional groups (Tris (2-aminoethyl) amine, dimethylaminomethyl, isocyanate, mercaptomethyl, aminomethyl, hydroxymethyl, triphenylphosphine, morpholine, 2-chlorotrityl amine, 4-(dimethylamino) pyridine, poly (vinyl chloride) carboxylated, poly (4-vinyl pyridine), p-toluenesulfonic acid, p-toluenesulfonyl hydrazide, piperidine, acetyl polystyrene, 2-chlorotrityl chloride, piperazine, diethylene triamine, poly (vinyl chloride)) were suspended in HTO solution (initial activity = 4.7 × 105 Bq·L−1). After the equilibration, the suspension was filtered using 3 kDa membrane filter and the activity in filtrates were quantified by LSC (HIDEX-300 SL). The results demonstrate the detritiation efficiency and separation factors of functional groups toward tritium. Carboxylic group (COOH) showed the most reactive performance as detritiation efficiency of ~4%. Compared to other functional groups, styrene sulfonyl groups including sulfonyl amide (SNH2) and sulfonyl hydrazide (SNHNH2) revealed promising performance for tritium separation as separation factor of 10.97 and 3.85, respectively. However, sulfonyl hydroxide (SOH) which is known as reactive functional group to tritium exchange showed the poor performance (detritiation efficiency: 0.68%; separation factor: 3.02). This study could suggest the promising functional groups for detritiation of low-level tritiated water which can be utilized to enhance the performance of current technologies. For example, reactive functional groups can be grafted on the surface of packing material within distillation tower resulting in the increasing detritiation efficiency.