Enhancing the capacitive deionization performance requires the inner structure expansion of porous activated carbon to facilitate the charge storage and electrolyte penetration. This work aimed to modify the porosity of coconut-shell activated carbon (AC) through CO2 activation at high temperature. The electrochemical performance of CO2- activated AC electrodes was evaluated by cyclic voltammetry, charge/discharge test and electrochemical impedance spectroscopy, which exhibited that AC-800 had the superior performance with the highest capacitance of 112 F/g at the rate of 0.1 A/g and could operate for up to 4000 cycles. Furthermore, in the capacitive deionization, AC-800 showed salt removal of 9.15 mg/g with a high absorption rate of 2.8 mg/g min and Ni(II) removal of 5.32 mg/g with a rate close to 1 mg/g.min. The results promote the potential application of CO2- activated AC for desalination as well as Ni-removal through capacitance deionization (CDI) technology.

췌석은 만성췌장염에서 흔히 동반되는 소견으로 반복되는 복통과 췌장염의 원인이 될 수 있다. 췌석은 췌관 협착과 자주 동반되고 췌관내 박혀 있는 경우가 많아 바스켓을 이용하여 제거하기 어렵다. 체외충격파쇄석술(ESWL)로 치료하기도 하나 반복 시술이 필요하고 성공률도 높지 않았다. 최근에 개발된 SpyGlass™ DS II (Boston Scientific, Marlborough, MA, USA)는 직경이 3.5 mm로 가늘어 췌관이 확장되어 있을 때 안으로 삽입이 가능하게 되었다. 그리고 직접 췌석을 보면서 전기수압쇄석술(EHL)이나 레이저 유도 쇄석술(laser lithotripsy)를 시행하며 췌석을 제거해 볼 수 있게 되었다. 본고에서는 SpyGlass™ DS II와 EHL을 이용하여 10 mm 이상의 다발성 췌석을 제거하는 방법을 소개하고자 한다.

The biocarbon (SKPH) was obtained from Sargassum spp., and it was evaluated electrochemically as support for the CO2 reduction. The biocarbon was synthesized and activated with KOH, obtaining a high surface area (1600 m2 g− 1) due to the activation process. Graphitic carbon formation after pyrolysis was confirmed by Raman spectroscopy. The XRD results show that SKPH has an amorphous structure with peaks corresponding to typical amorphous carbonaceous materials. FTIR was used to determine the chemical structure of SKPH. The bands at 3426, 2981, 2851, and 1604 cm− 1 correspond to O–H, C-H, and C-O stretching vibrations, respectively. Then, it compares SKPH films with different carbon films using two electrolytic systems with and without charge transfer. The SKPH film showed a capacitive behavior in the KOH, H2SO4, and, KCl systems; in the acid medium, the presence of a redox couple associated with carbon functional groups was shown. Likewise, in the [Fe(CN)6]−3 and Cu(II) systems, the charge transfer process coupled with a capacitive behavior was described, and this effect is more noticeable in the [Fe(CN)6]−3 system. Electrodeposition of copper on SKPH film showed two stages Cu(NH 3)2+ 4 /Cu(NH 3)+ 2 and Cu(NH 3)+ 2 ∕Cu in ammonia media. Hydrogen formation and the activity of CO2 are observed on SKPH film and are favored by the carbon’s surface chemistry. Cu/SKPH electrocatalyst has a catalytic effect on electrochemical reduction of CO2 and inhibition of hydrogen formation. This study showed that the SKPH film electrode responds as a capacitive material that can be used as an electrode for energy storage or as metal support.

본 연구에서는 전장이 400m인 24,000TEU급 컨테이너선박을 대상으로 선박 내 빌지펌핑의 성능에 대한 케이스 스터디를 수행하 였다. 본 연구의 대상인 24,000TEU급 컨테이너선박의 빌지시스템의 경우 선급의 규칙에 맞게 설계되었지만, 선박 내 설치되어 있는 빌지펌 프의 정격유량 및 최대유량 조건에서도 SOLAS Reg.II-1/35-1의 2 m/s 요건을 만족시키지 못하였다. 특히 1번 ~ 4번 화물창에 대해, 해수로 가득차 있다고 가정한 상태에서 해수를 모두 배출하는 동안에 빌지 주관에서의 평균유속을 계산할 결과, 2번 화물창, 3번 화물창 및 4번 화물창은 평균유속이 2 m/s 미만으로 기준에 적합하지 않은 것으로 나타났다. 본 연구에서는 이를 해결하기 위해 2번, 3번 및 4번 화물창의 150A 빌지 지관을 200A 배관으로 교체하여 계산을 수행하였으며, 그 결과 화물창 내의 해수를 모두 배출하는 동안의 빌지 주관에서의 해 수평균유속이 각각 2.479m/s , 2.476,m/s 및 2.459m/s 로 기준을 만족시키는 것을 알 수 있었다.

Medical cyclotrons have been used for dedicated medical of commercial applications such as positron emission tomography (PET) for the past tens of years. These cyclotron facilities have produced positron-emitting radionuclides (i.e. 11C, 13N, 15O, 18F, etc.). Among them, 18F, produced by 18O(p,n)18F reaction is the most widely used which has longer half-life (around 110 m) and lower energy of emitted positrons (around 0.63 MeV). Secondary neutrons produced during 18O(p,n)18F reaction could cause neutron activation of structures, systems, and components of cyclotron facilities. Therefore, International Atomic Energy Agency (IAEA) had addressed that during the operation of cyclotrons, concrete walls become radioactive over time and this radioactivity needs to be characterized for planning of the facility decommissioning. Moreover, several prior studies had estimated the neutron activation and levels of radioactivity of concrete wall of cyclotron facilities. Although those studies assessed the neutron activation of actual cyclotron facilities, however, the purpose of assessment was only for decommissioning each individual facility. Also, the assumptions, conditions or insights of conclusion may be limited to each individual case. For these reasons, this study focused on analysis of effects of major factors (e.g. concrete type, impurity contents of structural materials, etc.) about neutron activation of cyclotron facilities. In this study, the well-known methodology of neutron activation estimation was established and neutron activation products of concrete wall of cyclotron vault was calculated. Also, sensitivity analyses were conducted to figure out the effects of major factors of neutron activation and production of radioactive wastes during decommissioning of the facility. The methodology and results were validated by two steps: comparing with prior studies and comparing with another computer code. Concrete type did not affect that the decision of level of radioactivity waste criteria. Because of relatively longer half-lives, impurity contents of structural materials especially Co and Eu were turned out one of the most important factors for planning the facility decommissioning. It is hard to simply figure out the radioactivity levels of cyclotron facilities, however, rough predictions of minimum period for decay-in-storage as radioactive waste management can be possible with using information of thermal neutron spectra and major impurity nuclides (e.g. 59Co, 151Eu and 153Eu) for minimization of radioactive waste production and relief of charge of radioactive waste management.

Bentonite has been considered as a buffer material in a deep geological repository for high-level radioactive waste (HLW). Bentonite may come into contacted with various chemical solutions during the long-term storage. In particular, solutions containing K+ can affect stability of bentonite (e.g., illitization). The bentonite can be gradually saturated with the inflow of groundwater, and the temperature can rise simultaneously due to the decay of HLW. This study aimed to evaluate the bentonite stability in contacted with very highly concentrated K+ solutions with different pHs at 150°C. Batch reaction tests using KJ-II bentonite were performed for 30–150 days in teflon-stainless steel reactors. De-ionized (DI) water (pH = 6.0) and 1 M KCl (pH = 6.0), and 1 M KOH (pH = 12.5) solutions were used as reaction solutions. After completing batch reaction tests, the reacted samples were analyzed using various analytical techniques. For DI water, chemical, mineralogical, and physicochemical properties of reacted samples were similar to those of unreacted samples. For 1 M KCl solutions, cation exchage for Ca by K and slight changes in mineralogical properties of reacted samples were observed, but there are no significant changes in the physicochemical properties. In contrast, for 1 M KOH solutions, changes in chemical, mineralogical, and physicochemical properties of reacted samples were observed. Results of X-ray diffraction (XRD) analysis indicated dissolution of montmorillonite and formation of zeolite minerals, which were confirmed by thermogravimetricdifferential thermal analysis (TGA-DTA) and fourier transform infrared (FTIR) analysis. These results suggest that highly concentrated K+ (1 M) solution combined with high pH (12.5) and high temperate (150°C) may affect bentonite alteration. These prelimiary experiments were intended to qualitatively evaluate the mechanism and influncing factors of the buffer material alteration under extreme experimental conditions, and it is revealed that the conditions do not reflect the actual repository environment.

To minimize the short-term thermal load on the repository facility, heat generating nuclides such as Cs-137 and Sr-90 should be separated from the spent nuclear fuel for efficiency of repository facility. In particular, Sr-90 must be separated because it generates high heat during the decay process. Recently, Korea Atomic Energy Research Institute (KEARI) is developing a waste burden minimization technology to reduce the environmental burden caused by the disposal of spent nuclear fuel and maximize the utilization of the disposal facility. The technology includes a nuclide management process that can maximize disposal efficiency by selectively separating and collecting major nuclides such as Cs, Sr, I, TRU/RE, and Tc/Se from spent nuclear fuel. Among the major nuclides, Sr nuclides dissolve in chloride phase during the chlorination process of spent nuclear fuel and recovered in the form of carbonate or oxide via reactive distillation. In this process, Ba nuclides are also recovered along with Sr nuclides due to their chemical similarity. In this study, we prepared group II nuclide ceramic waste form, Ba(x)Sr(1-x)TiO3 (x=0, 0.25, 0.5, 0.75, 1), using the solid-state reaction method by considering the various ratio of Sr/Ba nuclides generated from nuclide management process. The established waste form fabrication process was able to produce a stable waste form regardless of the ratio of Sr/Ba nuclides. To evaluate the stability of group II waste form, physicochemical properties such as leaching and thermal properties were evaluated. Also, the radiological properties of the Ba(x)Sr(1-x)TiO3 waste forms with various Sr/Ba ratios were evaluated, and the estimation of centerline temperature was carried out using the experimental thermal property data. These results provided fundamental data for long-term storage and management of group II nuclides waste form.

Important medical radionuclides for Positron Emission Tomography (PET) are producing using cyclotrons. There are about 1,200 PET cyclotrons operated in 95 countries based upon IAEA database (2020). Besides, including PET cyclotrons, demands for particle accelerators are continuously increasing. In Korea, about 40 PET cyclotrons are in operating phases (2020). Considering design lifetime (about 30-40 years) and actual operating duration (about 20-30 years) of cyclotrons, there will be demands for decommissioning cyclotron facilities in the near future. PET cyclotron produces radionuclides by irradiating accelerated charged particles to the targets. During this phase, nuclear reactions (18O(p,n)18F etc.) produce secondary neutrons which induce neutron activation of accelerator itself as well as surrounding infrastructures (the ancillary subsystems, peripheral equipment, concrete walls etc.). Generally, experienced cyclotron personnel prefer an unshielded cyclotron because of the repair and maintenance time. In unshielded cyclotron, water cooling systems, air compressor, and other equipment and structures could be existed for operating purposes. Almost all the equipment and structures are consisted of steel, and these affect neutron distribution in vault especially thermal neutron on the concrete wall. In addition, most of them can be classified as very low level radioactive wastes by Nuclear Safety and Security notice (NSSC Notice No. 2020-6). However, few studies were estimating radioactivity concentrations (Bq/g) of surrounding structures using mathematical calculation/simulation codes, and they were not evaluating the effect of surrounding structures on neutron distribution. In this study, by using computational neutron transport code (MCNP 6.2), and source term calculation code (FISPACT- II), we evaluated effect of the interaction between surrounding structures (including surrounding equipment) and secondary neutrons. Discrepancies of activation distribution on/in concrete wall will be occur depending on thickness of structure, distance between structures and walls, and consideration of interaction between structures and neutrons. Throughout this study, we could find that the influence of those structures can affect neutron distribution in concrete walls even if, thickness of the structure was small. For estimating activation distribution in unshielded cyclotron vault more precisely, not only considering cyclotron components and geometry of target, but also, considering surrounding structures will be much more helpful.

Low- and intermediate-level radioactive wastes have been disposed of in the first-phase deep underground silo disposal at Gyeongju in South Korea. These radioactive wastes contain harmful radionuclides such as Uranium-238 (238U), which can pose long-term and deleterious effects on humans and the natural environment. Ethylenediaminetetraacetic acid and isosaccharinic acid, which can be formed via cellulosic waste degradation under high alkaline conditions might considerably enhance the transport behavior of 238U with the intrusion of rainwater and groundwater. In this study, the engineered barriers (concrete and grout) and natural barriers (sedimentary rock and granite) were used to investigate the 238U transport behavior in artificial cementitious porewater of State I (pH 13.3) and State II (pH 12.5) based on groundwater or rainwater. The surface properties and geochemical compositions of barrier samples were characterized using XRD, XRF, SEM-EDX, and BET. The transport behaviors of 238U in various solution conditions were observed by sorption distribution coefficient (Kd) at a range of initial chelating agents concentration (10-5-10-2 M). The sorption behavior of 238U was retarded more in the engineered rock barriers than in the natural rock barriers. The mobility enhancement of 238U was more significant in State I than in State II. In comparison with the absence of chelating agents, negligible changes in the Kd values of 238U were observed at less than initial chelating agent concentrations of 10-4 M. However, the Kd values of 238U were significantly reduced at initial chelating agent concentrations higher than 10-3 M. Therefore, these experimental findings show that the transport behavior of 238U into the geo- and bio-sphere could be accelerated by the presence of chelating agents and the type of cement degradation states.

Cellulose-based wastes can be degraded into short-chain organic acids at the cementitious radioactive waste repository. Isosaccharinic acid (ISA), one of the main degradation products, can form the chelate complex with metals and radionuclides, and these complexes have a potential that can accelerate to move the radionuclides to far-field from the repository. This study characterized the amount of generated ISA from typical cellulosic materials in the repository. Two different degradation experiments were conducted under alkaline conditions (saturated with Ca(OH)2 at pH 12.4): i) cellulosic material mixture under an opened condition (partially aerobic), and ii) cellulosic material under an anaerobic condition in a nitrogen-purged glove box. In the first case, three different types of cellulosic materials–paper, cotton, and wood– were mixed at the same ratio, and the experiments were carried out at three different temperatures (20°C, 40°C, and 60°C). It revealed that both the cellulose degradation rate and generated ISA concentration were high at high reaction temperatures, and various soluble degradation products such as formic acid and lactic acid were generated. The cellulose degradation in this work seems to still stay at a peeling-off process. In the second study, each type of cellulosic material was applied in its own batch experiments, and the amount of generated ISA was in the order of paper > wood > cotton. The above two experiments are supposed to be a long-term study until the generated ISA reaches an equilibrium state.

Bentonite containing >50wt% montmorillonite is being considered as a buffer material in a deep geological repository to dispose of high-level radioactive wastes (HLRW). Bentonite is considered a buffer material because of its exceptional properties such as high swelling capacity, low hydraulic conductivity, and high radionuclide sorption capacity. The bentonite buffer can be exposed to heat from the radioactive decay of HLRW and to groundwater. Water in bentonite buffer can be converted to steam under elevated temperature and pressure conditions. Previous studies reported contrasting results showing that steam treatment could decrease the swelling capacity due to changes in the surface properties from hydrophilic to hydrophobic or could not change. The contrasting results were probably because different studies used different experimental conditions and methods. Therefore, the effect of steam treatment on the bentonite properties is still unclear. The purpose of this study is to determine how the bentonite properties change after steam treatment, in particular swelling and hydrophilic properties. Two types of bentonite were used for steam treatment and analysis; Gyeongju Ca-bentonite (KJ- II) and Wyoming Na-bentonite (GCL-B). Steam treatment was performed at 150°C in an oven for various periods (7, 30, 60, and 90 days). Free swell test, X-ray fluorescence (XRF) analysis, surface-area measurement (BET), thermal gravimetric analysis (TGA), cation exchange capacity (CEC), and water uptake test were performed on steam-treated bentonite for various periods and raw bentonite. After steam treatment, some properties of steam-treated bentonite changed when compared to raw bentonite. Free swell index, which means the swelling capacity, decreased significantly as the results of previous studies. CEC and BET surface area values depended on the bentonite type. For Wyoming Na-bentonite, in which the dominant interlayer cation is a monovalent cation, CEC and BET surface area values were increased. On the other hand, Gyeongju Ca-bentonite, in which the dominant interlayer cation is a divalent cation, has no change in the above two properties. Results of XRF analysis, TGA, and water uptake test showed that these properties of both bentonites did not change after steam treatment. The results of this study confirmed that steam treatment affected the swelling and physicochemical properties of bentonite, in particular Na-bentonite. Further studies will focus on the surface properties of bentonite to investigate whether the surface properties have changed from hydrophilicity to hydrophobicity, or whether the montmorillonite structure has changed.

In order to enter a nuclear power plant, access approval is required in advance, and biometric information such as fingerprints of visitors must be registered when issuing a key card, and only those certified through biometric equipment can enter the nuclear facilities (Protected area II). Fingerprint recognizers and facial recognizers are installed and operated in domestic nuclear facilities for access control. Domestic nuclear facilities establish and implement a protection system in accordance with physical protection requirements under the “Act on Physical Protection and Radiological Emergency” and “Physical Protection Regulations” of each nuclear facility. Detailed implementation standards are specified in Regulation Standard (RS) documents established and distributed by KINAC. Biometrics are mentioned in a KINAC RS-104 (Access Control) document. In this study, it was analyzed what points should be considered in order to prepare for performance tests and establish plans for biometric devices. In order for the results of performance evaluation of biometric devices to obtain high reliability and to be applied to nuclear facilities in the future, standardized performance evaluation targets, procedures, standards, and environments must be created. In order to collect samples such as fingerprints for performance evaluation, the size roll of the sample shall be determined, and the appropriateness of the sample size shall be evaluated in consideration of reliability and error range. In addition, the analysis results for the characteristics (gender, age, etc.) of the sample should be presented. When collecting samples, conflicts with other laws such as personal information protection should be considered, and the reliability of the performance test result data should be analyzed and presented. Quality evaluation should also be performed on forged biometric information data such as silicon fingerprints. In addition, when establishing a performance evaluation plan, a systematic evaluation procedure should be established by referring to domestic and foreign certification and evaluation systems such as the Korea Internet & security Agency (KISA). In order to improve the completeness of the access control system using the biometrics of nuclear facilities, it is necessary to test the performance of biometric devices and to install and operate only devices that have the ability to defend against counterfeit technology. In this study, it was analyzed what points should be considered in order to prepare for performance tests and establish plans for biometric devices.

There are highly toxic radio-isotopes and high heat emitting isotopes in spent nuclear fuels which could be a burden in a deep geological repository. Some preliminary study in order to see if there are some advantages in terms of waste burden, in case that the spent fuel is appropriately processed and then disposed of in a final repository, has been carried out at KAERI. This study is focused on the proliferation resistance for various processing alternatives for them. The evaluation criteria and their indicators for proliferation resistance analysis are selected and then evaluated quantitatively or quantitatively for the alternatives. The processing alternatives are grouped into three categories according to the level of decrease of burden for final disposal and named them as Level I, Level II and Level III technolgy alternatives. Level I alternative is to maximize the long-term safety in the final repository from the removal of I- 129, semi-volatile radioisotope, which is the greatest impact on the long-term safety of the repository. Level II alternative is to remove the strontium-90, high heat emitter, in addition to the removal in Level I. The Level III is to additionally remove uranium from main stream of the level II to reduce the volume of the high level wastes to be disposed. The intrinsic radiation and chemical barriers against the nuclear proliferation are selected and analyised for the alternatives. It is resulted from the proliferation resistance analysis that all three options showed excellent resistance to nuclear proliferation for the two barriers. However, Level III technology including electrochemical refining process is relatively a little weaker than others. Overall, it could be an effective means to reduce the burden of disposal if the spent fuels are appropriately conditioned for final disposal. Further detailed studies are, however, needed to finalize its feasibility.

In August 2021, in response to the rapidly changing trade environment, including the advancement of Information Communication Technology (ICT) and its services, the European Union (EU) implemented the Dual-Use Items Control Regulation 821/2021 to introduce an Internal Compliance Program (ICP) to the EU countries. Accordingly, the exporters should comply with the regulation to strengthen their transactions review systems. Sweden, Germany, France, and the United Kingdom have implemented ICPs and outreach activities for dual use items. In particular, France explicitly stipulates the introduction of ICP in the law to manage and supervise it. While Sweden, Germany, and the United Kingdom strengthen the supervisory authority of regulatory agencies then companies are encouraged to autonomously introduce ICPs. Before introducing the ICP for the trigger list items (the items) to the Republic of Korea (ROK), a comprehensive export license system for them should be firstly considered based on EU Regulations. Also the comprehensive export license might be implemented by expanding the subject for the existing license on technology export of nuclear plant into the items. The ROK does not introduce an ICP as it does not recognize a self-classification on the items in accordance with the nuclear export control law. However, in preparation for the export to the EU countries that have intentions to introduce nuclear plants, it is necessary to analyze the export control programs of Sweden, Germany, France, and the United Kingdom. Like the programs of Sweden, Germany and the United Kingdom, the EU regulations might be adopted to reduce the regulation burden in the ROK. With the reference of Sweden, the authority could support the Export Control Manager Certification (ECMC) system accredited by civil association then its outreach activities could be diverse and extended. Basically, the ECMC system could consist of Part I, II, III and IV and an applicant could be accredited by a civil association as the ECM after completing the courses of Part I and II. The ECMC courses might be as follow; 1) Part I: the Basic common course for beginner 2) Part II: the National export control system for the items 3) Part III: the International export control regulations 4) Part IV: Re-Certification within the certain period In this paper, we analyzed the export control programs in Sweden, Germany, France, and the United Kingdom and suggested the ECMC system that might be applied to the ROK as above.

Drought stress is a condition that occurs frequently in the field, it reduces of the agricultural yield of field crops. The aim of the study was to screen drought-adapted genotype of Italian rye grass. The experiments were conducted between the two Italian ryegrass (Lolium multiflorum L.) cultivars viz. Hwasan (H) and Kowinearly (KE). The plants were exposed to drought for 14 days. The results suggest that the morphological traits and biomass yield of KE significantly affected by drought stress-induced oxidative stress as the hydrogen peroxide (H2O2) level was induced, while these parameters were unchanged or less affected in H. Furthermore, the cultivar H showed better adaptation by maintaining several physiological parameter including photosystem-II (Fv/Fm), water use efficiency (WUE) and relative water content (RWC%) level in response to drought stress. These results indicate that the cultivar H shows improved drought tolerance by generic variation, improving photosynthetic efficiency and reducing oxidative stress damages under drought stress. These findings can be useful to the breeder and farmer for improving drought tolerance in Italian rye grass through breeding programs.

광양 마로산성은 6세기대 백제에 의해 축성된 산성으로 지금까지 총 5차례의 발굴조사 를 통해 산성 내·외부에 대한 전면적인 조사가 이루어졌다. 그 결과 다종·다양한 유구와 유물이 확인되었고, 특히 광양의 백제때 지명인 ‘馬老’와 치소·관청을 상징하는 ‘官’이 결 합된 ‘馬老官’명 기와가 출토되면서 이 지역의 고대 治所城으로 비정되어 왔다. 이 연구는 마로산성의 발굴성과를 검토하여 고대 사회 치소성의 개념을 고고학적으로 접 근한 것이다. 이를 위해 마로산성 내부시설의 입지, 규모, 출토유물 등의 집중 분석하여 상 관관계를 검토하였고, 그 결과 다음과 같은 결론을 도출하였다. 첫째, 관아·군사·저장·제의 관련 시설의 존재이다. 이러한 시설들은 치소성으로서 반드시 갖추어야 할 필수 시설인데 행정의 핵심시설인 東軒은 Ⅱ-4호건물지, 외부 경계와 조망을 위한 將臺는 Ⅱ-1호건물지, 지배계급의 생활 공간인 內衙는 Ⅰ-2호건물지, 조세와 관련된 창고 시설은 Ⅱ-2호,Ⅱ-3호 건물지로 판단된다. 둘째, 祭場으로서의 기능이다. 마로산성에서는 집수시설과 수혈유구 등을 통해 진단구, 공헌물로 볼 수 있는 특정 종류의 유물들이 반복적으로 출토되고 있어 마로현의 제장으로서 기능했음을 알 수 있다. 셋째, 상위 계층의 신분표상품의 존재이다. 마로산성에서는 금동과대, 금동투조장식, 중 국제 자기, 동경, 벼루 등 다양한 신분표상품들이 다수 출토되었다. 이러한 신분표상품들이 집중되고 있는 점은 마로산성이 마로현의 치소성으로서 활용되었음을 방증하는 것이다. 넷째, ‘馬老官’名 명문와의 존재이다. ‘마로관’명 기와는 광양의 백제때 지명과 행정치소 인 관이 복합된 기와로서 마로산성이 마로현의 치소성이었음을 보여주는 가장 확실한 자료 이다.