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.

본 연구는 자연계에서 가장 흔하게 관찰되는 두 그린 러스트(green rust) 광물인 carbonate green rust (CGR)과 sulfate green rust (SGR)을 공침법(co-precipitation)을 통해 각각 합성하고, 이들의 형성 메커니즘 및 이화학적 특성들을 체계적으로 규명하였다. X-선 회절(XRD) 분석 및 리트벨트 정련 수행 결과, 본 합성 조 건에서 이차광물상 없이 이중층수산화물로서 CGR과 SGR이 합성됨을 확인하였다. 또한, 각각의 구조 파라미 터는 CGR의 경우 a(=b)축 = 3.17Å, c축 = 22.52 Å이고, SGR의 경우 a(=b)축 = 5.50Å, c축 = 10.97 Å이며, 이 들의 미결정 크기는 각각 (003)면 기준 57.8 nm와 (001)면 기준 40.1 nm로 밝혀졌다. 주사전자현미경/에너지 분산형 분광분석(SEM/EDS) 결과, CGR과 SGR은 모두 육각 판상의 전형적인 이중층수산화물 결정 형상을 보이지만 탄소(C)와 황(S)의 함량은 서로 다르게 나타났다. 퓨리에 변환 적외선(FT-IR) 분광 분석결과, 탄산 염(CO3 2-)와 황산염(SO4 2-) 이온들이 각각 CGR과 SGR의 층간 음이온으로 밝혀졌고, 이는 XRD를 활용한 광 물상 동정 결과와 잘 일치한다. 철 용액으로의 수산화이온(OH-) 주입 시간에 따른 혼합 용액의 pH와 Eh, 그 리고 잔류 철 농도의 비율(Fe(II):Fe(III)) 측정 결과, 시간에 따른 차이는 있지만 두 green rusts 모두 1단계 전구체 형성, 2단계 중간 생성물로의 상변환, 그리고 3단계 green rust로의 상변환과 에이징에 의한 결정성장 으로 이어지는 결정 형성 메커니즘을 보이는 것으로 판단된다. 본 연구는 공침법을 통해 CGR과 SGR을 안 정적으로 합성하고 이들의 형성 메커니즘과 이화학적 특성을 규명함으로써, green rust를 활용한 응용 연구 및 산업 활용에 원천 기초자료를 제공할 것으로 기대된다.

Zhenzigou 연-아연 광상은 중국 동북지역에선 가장 규모가 큰 연-아연 광상 중의 하나로 지체구조상 Jiao Liao Ji belt내 Qingchengzi mineral field에 위치한다. 이 광상의 주변지질은 시생대의 그래뉼라이트 (granulite)와 이를 관입한 고원생대의 미그마타이트질 화강암과 고-중원생대의 소딕(sodic) 화강암을 부정합으로 피복한 고원생대의 Liaohe 층군 및 이들을 관입한 중생대의 섬록암과 몬조나이틱 화강암으로 구성된다. 이 광상은 고원생대의 Liaohe 층군내 Langzishan 층 및 Dashiqiao 층내에서 층상 광체 및 맥상 광체로 산출되며 층준규제 퇴적분기형 또는 퇴적분기형 광상에 해당된다. 이 광상에서 산출되는 백색운모는 층상 광체에서만 산출되며 모암의 종류, 변질 정도, 광석광물의 유무 및 광체 형태에 따라 4 가지 형(I 형 백색운모 : 약변질 (쇄설성 돌로마이트질 대리암), II 형 백색운모 : 강변질(돌로마이트질 쇄설성 암석), III 형 백색운모 : 층상 광체(돌로마이트질 쇄설성 암석), IV 형 백색운모 : 층상 광체(쇄설성 돌로마이트질 대리암))으로 분류된다. I 형 백색운모는 약 변질정도를 갖는 쇄설성 돌로마이트질 대리암내 돌로마이트를 열수교대작용에 의한 돌로마 이트화작용에 의해 형성된 돌로마이트와 함께 산출된다. II 형 백색운모는 강 변질정도를 갖는 돌로마이트질 쇄설성 암석내 열수교대작용에 의한 칼리장석의 변질물이나 돌로마이트화작용에 의해 형성된 돌로마이트, 철 백운석, 석영과 함께 산출된다. III 형 백색운모는 층상 광체를 갖는 돌로마이트질 쇄설성 암석내 열수교대작 용에 의한 칼리장석의 변질물이나 돌로마이트화작용에 의해 형성된 철백운석, 방해석, 석영과 함께 산출된다. IV 형 백색운모는 층상 광체를 갖는 쇄설성 돌로마이트질 대리암내 열수교대작용에 의한 칼리장석의 변질물이나 돌로마이트, 석영과 함께 산출된다. 이들 백색운모의 화학조성은 각각 (K0.92-0.80Na0.01-0.00Ca0.02-0.01Ba0.00 Sr0.01-0.00)0.95-0.83 (Al1.72-1.57Mg0.33-0.20Fe0.01-0.00Mn0.00Ti0.02-0.00Cr0.01-0.00V0.00Sb0.02-0.00Ni0.00Co0.02-0.00)1.99-1.90(Si3.40-3.29Al0.71-0.60)4.00O10(OH2.00-1.83 F0.17-0.00)2.00, (K1.03-0.84Na0.03-0.00Ca0.08-0.00Ba0.00Sr0.01-0.00)1.08-0.85(Al1.85-1.65Mg0.20-0.06Fe0.10-0.03Mn0.00Ti0.05-0.00Cr0.03-0.00V0.01-0.00 Sb0.02-0.00Ni0.00Co0.03-0.00)1.99-1.93(Si3.28-2.99Al1.01-0.72)4.00O10(OH1.96-1.90F0.10-0.04)2.00, (K1.06-0.90Na0.01-0.00Ca0.01-0.00Ba0.00Sr0.02-0.01)1.10-0.93 (Al1.93-1.64Mg0.19-0.00Fe0.12-0.01Mn0.00Ti0.01-0.00Cr0.01-0.00V0.00Sb0.00Ni0.00Co0.05-0.01)2.01-1.94(Si3.32-2.96Al1.04-0.68)4.00O10(OH2.00-1.91 F0.09-0.00)2.00 및 (K0.91-0.83Na0.02-0.01Ca0.02-0.00Ba0.01-0.00Sr0.00)0.93-0.83(Al1.84-1.67Mg0.15-0.08Fe0.07-0.02Mn0.00Ti0.04-0.00Cr0.06-0.00V0.02-0.00 Sb0.02-0.01Ni0.00Co0.00)2.00-1.92(Si3.27-3.16Al0.84-0.73)4.00O10(OH1.97-1.88F0.12-0.03)2.00 로써 이론적인 이중팔면체형 운모류 값보다 Si가 높고 K, Na, Ca는 낮으며 모두 백운모에 해당된다. 특히, Zhenzigou 연-아연 광상에서 산출되는 백색운 모의 화학조성 변화는 팬자이틱 또는 Tschermark 치환[(Al3+)VI+(Al3+)IV <-> (Fe2+ 또는 Mg2+)VI+(Si4+)IV] 메카 니즘에 의해 일어났으며 백색운모의 Fe는 Fe2+와 Fe3+ 로써 존재하지만 주로 Fe2+ 우세함을 의미한다. 따라서 Zhenzigou 연-아연 광상의 층상 광체에서 산출되는 백색운모들은 고원생대의 화성활동 및 녹색편암상의 변성 작용에 의한 열수교대작용으로 기존에 산출되었던 광물들의 재용융 및 재침전 과정에서 형성되었으며 이들 백색운모의 화학조성 변화는 열수교대작용 동안 모암인 돌로마이트 및 쇄설성 암석의 함량 차이, 변질 정도 및 광석광물의 유무에 따른 팬자이틱 또는 Tschermark 치환[(Al3+)VI+(Al3+)IV <-> (Fe2+ 또는 Mg2+)VI+(Si4+)IV] 메 카니즘에 의해 일어났음을 알 수 있다.

The unified scheme of Seyfert galaxies hypothesizes that the observed differences between the two categories of Seyfert galaxies, type 1 (Sy1) and type 2 (Sy2) are merely due to the difference in the orientation of the toroidal shape of the obscuring material in the active galactic nuclei. We used in this paper, a sample consisting of 120 Seyfert galaxies at 1.40 × 109 Hz in radio, 2.52 × 1017 Hz in X-ray and 2.52 × 1023 Hz in γ-ray luminosities observed by the Fermi Large Area Telescope (Fermi- LAT) in order to test the unified scheme of radio-quiet Seyfert galaxies. Our main results are as follows: (i) We found that the distributions of multiwave luminosities (Lradio, LX-ray, and Lγ-ray) of Sy1 and Sy2 are completely overlapped with up to a factor of 4. The principal component analysis result reveals that Sy1 and Sy2 also occupy the same parameter spaces, which agrees with the notion that Sy1 and Sy2 are the same class objects. A Kolmogorov-Smirnov test performed on the sub-samples indicates that the null hypothesis (both are from the same population) cannot be rejected with chance probability p ~ 0 and separation distance K = 0.013. This result supports the fact that there is no statistical difference between the properties of Sy1 and Sy2 (ii) We found that the coefficient of the best-fit linear regression equation between the common properties of Sy1 and Sy2 is significant (r > 0.50) which plausibly implies that Sy1 and Sy2 are the same type of objects observed at different viewing angle.

In worldwide, tens of thousands of units of particle accelerators have been used and more than 97% of those accelerators are used for dedicated medical of commercial applications. Radionuclide production cyclotron produce several positron-emitting radionuclides such as 18F by 18O(p,n)18F reaction which generates secondary neutrons. It is of note that these neutrons cause neutron activation in structures and components of cyclotron facilities. Therefore, International Atomic Energy Agency had addressed that a well-developed estimate of the neutron activation induced radioactive inventory of accelerator facilities is needed for the proper planning and safe implementation of decommissioning using proven methods or codes that can be used to perform activation calculations. Moreover, IAEA suggested that during the operation of cyclotrons, concrete walls become radioactive over time and this radioactivity needs to be fully characterized as part of early decommissioning planning. In this study, Neutron activation in the medical cyclotron facilities was evaluated with the MCNP and FISPACT-II code to analyze the generation of decommissioning radioactive wastes during facilities dismantling. For the reference case, residual radioactivity concentration of each activation product (e.g. 60Co, 152Eu, etc.) was calculated and the sum of fractions of the activity concentration of each radionuclide divided by its clearance level was exceeded 1.0 at each calculation point which means radioactive waste generations during decommissioning of the facility. Several points show the calculated sum of fractions (SoF) at inside wall were bigger than the surface wall. The reason of these phenomena is that the slowdown of the incident neutron energy at the inside wall due to neutron attenuation and larger thermal neutron flux than surface wall. It is of note that each activation reaction cross-section was dominant at thermal neutron energy band. Sensitivity analysis was conducted to analyze the effects of design characteristics (e.g. beam energy and current, operation period, and workload). The SoF was exceeded 1.0 at the least activation condition (i.e. 9 MeV, 10 μA) if the operation period was 10 years. For the realistic condition such as 13 MeV, only 10 μA of beam current case shows the SoF was under union. On the other hand, 19 MeV, 60 μA, and 10 years operation case shows the SoF as 20.4 which means the clearance rule can be applied only after 21 years of decay-in-storage. The result of this study can be used for proper planning of decommissioning and/or new installation of cyclotron facilities include considerations of radioactive waste management.

Numerous low-and intermediate level radioactive wastes were generated from the decommissioning processes of nuclear power plants. Radionuclides such as Co and Cs contained in decommissioning wastes should be immobilized to prevent the release of radionuclides from the wastes due to its harmful impacts on ecosystem by high radioactivity and long half-life. Ethylenediaminetetraacetic acid (EDTA) used as decontamination agent can be contained in cement waste during decommissioning process of nuclear power plants. In addition, EDTA can be stably and strongly bound with radionuclides, resulting in the acceleration of the nuclide release from solidified cement matrix. Here, we investigated the effects of EDTA on leaching behaviors of Co and Cs immobilized in the cement specimen. The leaching tests were performed according to the ANS 16.1 “Measurement of the leachability of solidified low-level radioactive wastes by a short-term test procedure”. From the results, an increase in the EDTA content in the cement specimen led to an increase in Co leaching, whereas a decrease in Cs leaching. Leaching of Cs was dominantly controlled by diffusion from the pore space of the cement specimen to the solution. The effective diffusion coefficient and leachability index of nuclide were determined using the diffusion-release models of ANS 16.1. The results of present study can be used in the safety assessment for disposal of the radioactive waste generated by decommissioning of nuclear power plants.

FTIR (Fourier Transform Infrared) and Raman spectra of KJ-II bentonite provided by Clariant Korea were compared with those of MX-80 bentonite. The FTIR spectra were obtained using a Nicolet 5 FTIR spectrometer (Fisher Scientific) equipped with a diamond ATR (Attenuated Total Reflection) module. The spectra were collected for 64 scans with a resolution of 4 cm−1. Raman spectra were obtained using an optical microscope (Olympus, BX43) and a spectrometer (Andor, SR- 500). The laser beam was focused using an objective lens with a magnifying power of 50. The backscattered light from the sample was collected into an optical fiber with a core diameter of 0.4 mm. The Raman signals were recorded with CCDs (Andor, DV401A-BV for 532 nm laser wavelength and DV420A-OE for 638 and 785 nm laser wavelengths). Each pixel of CCD received the signal for 1 s and its 1000 times accumulated data were collected. The FTIR spectra of the two bentonite samples are very similar. The FTIR spectra of both bentonites showed absorption bands at 3623, 3399, 3231 cm−1 in the higher wavenumber region and at 1637, 1443, 1117, 997, 914, 887, 847, 797, 611, 515, 414 cm−1 in the lower wavenumber region. A sharp band at 3623 cm−1 and the strong band at 997 cm−1 correspond to the OH stretching of structural hydroxyl groups and the Si-O stretching vibration, respectively. In addition to these clear bands, several absorption bands observed in this experiment are well matched with the results reported in various literatures. Unlike the FTIR spectrum, it is not easy to observe the Raman bands of bentonite. The reason is that strong fluorescence interferes with the Raman spectrum. The two bentonite samples showed different fluorescence intensities. In the case of MX-80 bentonite, no clear Raman band was observed due to the influence of very strong fluorescence. KJ-II bentonite showed a relatively weak fluorescence intensity and Raman bands were partially visible at around 147, 260, 397, 709, and 1279 cm−1. In particular, the band at 1279 cm−1 is wide and sturdy. It was observed that the non-powder samples showed a better quality spectra. The Raman characteristics of KJ-II bentonite, which depend on the incident laser wavelength and the sample pretreatment, are discussed in detail.

Accurate understanding of structural integrity and chemical reactivity of UO2 disposed in deep underground sites is of importance. Owing to the specific condition of the site location, UO2 may have substantially different properties from the conventional prediction. In this study, we demonstrate that the oxidation resistivity of UO2 is considerably modified by gadolinium (Gd), which is the element of neutron absorber and a byproduct of nuclear decay of radioactive U-235. Using density functional theory calculations, we investigate how the oxidation mechanism of UO2 changes with Gd incorporation in U lattice. Our study indicates that Gd remarkably enhances the thermodynamic stability of pristine UO2 against surface oxidation via three underlying mechanisms: (i) weakens the chemical bonding of adsorbed oxygen atom (O) with U, (ii) reduces active sites (U) for oxygen adsorption, and (iii) suppresses the subsurface diffusion of adsorbed O delaying the growth of the oxide layers on the UO2. Electronic and lattice structure analyses for Gd-doped UO2 indicate that amount of charge transfer from U to O is critically reduced and the lattice of the UO2 surface is contracted. Our results provide useful information for understanding long-term stability and improving the structural integrity of UO2 through the chemical doping process.

Efficient capture and storage of radioactive iodine (consisting of two isotopes: 129I and 131I), produced or released from nuclear activities, are of paramount importance for sustainable development of nuclear energy due to their volatility and long half-life. Therefore, it is very important to develop new adsorbents for efficient utilization of radioactive iodine from nuclear waste. Various methods and materials are used for I2 capturing and removing, including MOFs due to their high porosity and fast adsorption kinetics, which are rightfully considered effective sorbents for removing I2. Metal–organic frameworks (MOFs) are porous crystalline materials which have diverse pore geometry and unique physicochemical properties, have attracted enormous attention for use in gas storage, separation and catalysis. The ability of MOFs to adsorb volatile products at room temperature can significantly improve the cost-effectiveness of the utilization process. This work describes the synthesis and characterization of three new metal-organic frameworks based on pyrazine (pyz), 44’bipyridine (bpy), 1,2 -bis(4 - pyridyl) – ethane (bpe) and copper (II) hexafluorozironate, as potential adsorbents for I2 capture. All of these three MOFs exhibit a two - dimensional (2D) crystal structure consisting from infinity non-crossing linear chains. The crystal structure of [Cu(pyz)2(ZrF6)2(H2O)2], [Cu(bpy)4(H2O)2ZrF6] and [Cu(bpe)4(H2O)2ZrF6] were characterized using powder X-ray diffraction (PXRD), single crystal X-ray diffraction (SC-XRD). Comparative characteristics of synthesized MOFs, including Fourier-transform infrared spectroscopy (FTIR) and thermogravimetric analysis (TGA) were also performed. The I2 sorption experiments were examined by UV-vis spectroscopy.

As the design life of nuclear power plants are coming to the end, starting with Kori unit 1, nuclear power related organizations have been actively conducted research on the treatment of nuclear power plant decommissioning waste. In this study, among various types of radioactive waste, stabilization and volume reduction experiments were conducted on radioactive contaminated soil waste. Korea has no experience in decommissioning nuclear power plants, but a large amount of radioactively contaminated soil waste was generated during the decommissioning of the KAERI research reactor (TRIGA Mark- II) and the uranium conversion facility. This case shows the possibility of generating radioactive soil waste from nuclear power plants and nuclear-related facilities sites. Soil waste should be solidified, because its fluidity and dispersibility wastes specified in the notification of the Korea Nuclear Safety and Security Commission. In addition, the solidified waste forms should have sufficient mechanical strength and water resistance. Numerous minerals in the soil are components that can make glass and ceramics, for this reason, glass-ceramic sintered body can be made by appropriate heat and pressure. The sintering conditions of soil were optimized, in order to make better economical and more stable sintered body, some additives (such as additives for glass were mixed) with the soil and sintering experiments were conducted. Uncontaminated natural soil was collected and used for the experiment after air drying. Moisture content, pH, bulk density, and organic content were measured to understand the basic properties of soil, and physicochemical properties of the soil were identified by XRD, XRF, TG, and SEM-EDS analysis. In order to understand the distribution by particle size of the soil, it was divided into Sand (0.05–2 mm) and Fines (< 0.05 mm). The green body was manufactured in the form of a cylinder with a diameter of 13mm and a height of about 10mm. Appropriate pressure (> 150 MPa) was applied to the soil to make a green body, and appropriate heat (> 800°C) was applied to the sintered body to make a sintered body. The sintering was conducted in a muffle furnace in air conditions. The volume reduction and compressive strength of the sintered body for each condition were evaluated.

For producing radionuclides which were mostly used in medical purposes, for instance, Positron Emission Tomography (PET), there were 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) of cyclotrons, there will be demands for decommissioning cyclotron facilities in the near future. PET cyclotron produces radionuclides by irradiating charged particles to the targets. During this phase, nuclear reactions (18O(p,n)18F, 14N(d,n)15O etc.) produce secondary neutrons which induce neutron activation of accelerator itself as well as surrounding infrastructures (the ancillary subsystems, peripheral equipment, concrete walls etc.). Most of the ancillary systems including peripheral equipment can be neutron activated, since, most of them were made of steels. Steels like stainless steel or carbon steel may contain some impurities, typically cobalt. Although, there were several researches evaluating activation of concrete walls and accelerator components, estimating the activation and influence on neutron interaction of the other surrounding infrastructures were insufficient. In this study, by using computational neutron transport code (MCNP 6.2), and source term calculation code (FISPACT- II), we estimated neutron distribution in cyclotron vault and activation of ancillary subsystems including some peripheral equipment. Also, using Au foil and Cd cover, we measured thermal neutron distribution at 16 points on the concrete wall, and compared it to calculated results (MCNP). Even though, the compared results matches well, there was a discrepancy of neutron distributions between presence and absence of those equipment. Additionally, in estimating activation distributions by calculating, most of the steel-based subsystems including peripheral equipment should be managed by radioactive wastes after 20 years of operation. Throughout this study, we could find that influence on neutron interaction of those equipment can affect neutron distribution in concrete walls. This results vary the activation depth as well as location of the hot contaminated spot in concrete wall. For estimating or evaluating activation distributions in cyclotron facilities, there was need to consider some equipment located in cyclotron vault.

Mechanism and kinetics of Rhenium complexes as a surrogate of Technetium-99 (99Tc) is worthy of study from radioactive waste safe disposal perspective. Re(IV)-EDTA was synthesized via the reduction of Re(VII) with Sn(II) in the presence of Ethylenediaminetetracetic acid (EDTA). The Re(IV)-EDTA was then degraded by H2O2 (7–30%) at pH of 3–11 in ionic strength I = 0–2 M solution. The Re- EDTA was observed to degrade more rapidly at pH of ≤ 3–4 than one of ≥ 10–11 and remained stable at pH = 7–9. At a low acidic pH, the complex degradation process was facilitated by protonation and corresponded to the exponential model (y = k. e–nt). In contrast, at an alkaline pH, the degradation was facilitated OH– complexation with Re(IV) and corresponded to a linear model (y = –mt + C). Complex degradation followed the zero-order rate kinetics for the H+ and Re-EDTA parameters, apart from a pH of 3, for which degradation was a better fit to first order kinetics. A higher Re(IV)-EDTA stability at a pH of 7–9 demonstrated that Re(IV)-EDTA (or 99Tc(IV)-EDTA) tends to be more persistent in natural environmental conditions.