시금치의 주요 해충인 흰띠명나방(Spoladea recurvalis) 유충의 살충제 5종에 대한 감수성을 검정하였다. Lufenuron EC, chromafenozide EC, chlorantraniliprole WP, tebufenozide WP, pyridalyl EW는 각각 2(12.5 ppm), 4(12.5 ppm), 8(2.5 ppm), 4(20.0 ppm), 8(12.5 ppm)배 의 희석농도에서 90% 이상의 높은 살충활성을 보였다. 추천농도로 경엽처리 후 7일이 경과된 시금치 잎에 흰띠명나방 유충이 72시간 동안 노출되 었을 경우 chromafenozide EC, chlorantraniliprole WP, tebufenozide WP, pyridalyl EW의 살충률은 각각 98.3%, 100%, 95.0%, 100%로 나타나 높은 잔효성을 보였다. 흰띠명나방에 대한 방제효과를 2개소(화성, 연천)에서 포장검정 결과, 5종의 약제 모두 2개소에서 약제처리 7일 경과 후 90% 이상의 방제효과를 보였으며 2배량에서도 약해가 없어 향후 흰띠명나방 방제약제로 시금치에 활용이 가능할 것으로 판단된다.

Neptunium (Np) is one of the daughter elements included in the decay chain of Pu. The quantitative analysis of Np isotopes is required for radioactive waste characterization, research on actinide chemistry, etc. Np-237 has a long half-life (2.144 million years), but its daughter Pa-233 has a relatively short half-life (26.975 days). For this reason, after a sufficient time elapses following the chemical preparation process of the analyte, the two nuclides are in radiation equilibrium in the sample. Np-237 emits alpha-rays while Pa-233 emits beta-rays. Both nuclides also emit gamma- and X-rays. In this study, alpha-rays were measured using liquid scintillation counting (LSC) method and alpha spectrometry. Gamma-spectrometry with a HPGe detector was used for the analysis of gammaand X-rays. In addition, we compared the radiometric results with quantitative analysis of Np using UV-Vis absorption spectrometry. The LSC method and the HPGe gamma-spectroscopy do not require extensive sample preparation procedures. Alpha spectroscopy requires a standard material spiking, separation by coprecipitation, and disk-type sample preparation procedure to obtain measurement efficiency and recovery factor. A reference material sample with a concentration of 5.8 mM was analyzed by the four analysis methods, and all of the measured results agreed well within a difference level of 4%.

We established pretreatment method of solidified cement ion-exchange resin samples generated before 2003 in nuclear power plants for measurement of non-volatile radionuclide activity. A microwave digestion system (MDS) with mixed acid (HCl-HNO3-HF-H2O2) was used to dissolve cement and to desorb non-volatile elements such as Ce, Co, Cs, Fe, Nb, Ni, Re, Sr and U from mixed ion-exchange resin. The content of Ce, Co, Fe, Nb, Ni, Re, Sr, U and Cs after pretreatment of cement plus mixed ion-exchange resin was measured by ICP-AES and ICP-MS, respectively. As iron and strontium are also present in cement, their content after dissolving a certain amount of cement was measured by ICP-AES. All elements except Nb were quantitatively recovered. Especially since the Nb recovery was low at 72.0±2.5%, the MDS following addition of the mixed acid to the resin was operated once more for desorbing Nb from it. Finally the recovery of Nb was over 95%. This sample pretreatment method will be applied to solidified cement ion-exchange resin samples generated in nuclear power plants for assessment of radionuclide inventory.

The massive amount of radioactive waste will generated during decommissioning of nuclear. Among the radioactive waste from these disposal process, 50-55 million tons of concrete waste are included. For safe disposal, it is very important to accurately analyze the concentration of radionuclides, especially 129I and 131I, contaminated concrete. 129I, a long-lived radioisotope of iodine (t1/2=1.57 × 107 y), and 131I (t1/2=8.04 d) are generated from the fission of uranium in nuclear reactors. In Korea, according to the Nuclear Safety and Security Commission (NSSC) radioactive clearance level guide, the limit for radioactive clearance level of 129I is less than 0.01 (Bq/g). Iodine can be absorbed, accumulate in organisms, and exhibit low energy emission compared with cesium, and cobalt. Therefore, it is essential to an accurately separate and analyze iodine radioactive waste. In this study, we focused on the determination of iodine in simulated cement waste form containing KI for the recovery of iodine. We performed cement waste form sieved through a different particle size (0.5 mm < ɸ < 6.35 mm). For the separation of iodine from solid samples with low iodine content, such as soil, sediment, and cement, for sample decomposition associated with solvent extraction using CHCl3 for separation of iodine from the matrix. The separation of iodine in cement waste particles was therefore carried out using an acid leaching method using KI containing cement particles. We observed that cement particle size decreased at 6.35 mm to 0.5 mm with iodine yield decrease at 0.840±0.011 to 0.582±0.010. Thus, in this study, the acid leaching method enables to determination Iodine in cement.

Concrete waste generated in the result of dismantling a concrete structure in a radiation control area and refractory brick waste generated from uranium pellet sintering furnace are surface-contaminated by uranium particle of which the enrichment is below 5%. These wastes are hard to decontaminate so it was necessary to develop the process for its disposal. So, we developed the Process Control Plan (PCP) for disposal of radioactive concrete waste describing a whole sequence of disposal and inspecting procedures based on the KNF Radioactive Waste Quality Assurance Plan (KN-WQAP) established in 2021. Based on the PCP, we crushed the concrete waste by jaw-crusher. Then we sieved the crushed concrete waste and removed the particle of which size is below 0.3 mm, using sieve-vibrator where the 0.3 mm mesh-sized sieve is installed inside. Before conducting the crush-sieving method based on the PCP, we conducted Process Control Assessment (PCA) based on the KN-WQAP. The purpose of the PCA is to check whether the output of the process satisfies the Acceptance Criteria of Korea Radioactive Waste Agency (KORAD) so that we could confirm the validity of the PCP. The evaluation item of the PCA is a particulate size verification test. The test is passed only if the component ratio of a particle size below 0.2 mm is less than 15% and the particle size below 0.01 mm is less than 1%. The very first 3 drums passed the test, so we began applying the PCP to whole target drums. In the process of conducting the crush-sieving method in earnest, qualified inspectors based on KNWQAP participated conducting sampling, measuring and checking whether a foreign material was included. They tested samples and packaged drums regarding 5 spheres of general, radiological, physical, chemical and biological characteristic. KNF disposed concrete and refractory brick waste by the crush-sieving method so that KNF could take over 100 drums to KORAD in 2021. But, it is needed to be improved that a dust size below 0.3 mm is generated as a secondary waste which needs to be solidified for the final disposal and the work environment is not good enough because of the dust.

This study introduces the licensing process carried out by the regulatory body for construction and operation of the 2nd phase low level radioactive waste disposal facility in Gyeongju. Also, this study presents the experience and lessons learned from this regulatory review for preparing the license review for the next 3rd phase landfill disposal facility. Korea Radioactive Waste Agency (KORAD) submitted a license application to Nuclear Safety and Security commission (NSSC) on December 24, 2015 to obtain permit for construction and operation of the national engineered shallow land disposal facility at Wolsong, Gyeongju. NSSC and Korea Institute of Nuclear Safety (KINS) started the regulatory review process with an initial docket review of the KORAD application including Safety Analysis Report, Radiological Environmental Report and Safety Administration Rules. After reflecting the results of the docket review, the safety review of revised 10 application documents began on November 29, 2016. Total 856 queries and requests for additional information were elicited by thorough technical review until November 16, 2021. As the Gyeongju and Pohang earthquakes occurred in September 2016 and November 2017, respectively, the seismic design of the disposal facility for vault and underground gallery was enhanced from 0.2 g to 0.3 g and the site safety evaluation including groundwater characteristics was re-investigated due to earthquake-induced fault. Also, post-closure safety assessments related to normal/abnormal/human intrusion scenarios were re-performed for reflecting the results of site and design characteristics. Finally, NSSC decided to grant a license of the 2nd phase low level radioactive waste disposal facility under the Nuclear Safety Laws in July 2022. This study introduces important issues and major improvements in terms of safety during the review process and presents the lessons learned from the experience of regulatory review process.

Se-79, a fission product of uranium, is present in spent nuclear fuel. Selenium is volatilized from the spent nuclear fuel during the pretreatment of pyroprocessing, and a filter composed of calcium oxide can capture gaseous selenium in the form of CaSeO3. Because Se-79 has a long half-life (3.27E5 years) and selenite ions have high mobility in groundwater, they must be immobilized in a chemically stable form for final disposal. This study used a composition of 50 TeO2 - 10 Al2O3 - 10 B2O3 - 10 Na2O - 10 CaO - 10 ZnO (mol%). High-purity powders of TeO2, Al2O3, H3BO3, Na2CO3, CaCO3, and ZnO were used as glass precursors. The mixed powders were placed in alumina crucibles and melted in an electric furnace under an ambient atmosphere at 800°C for 1 h before being cast on a carbon mold. The obtained glasses were ground into fine powders and then mixed with CaSeO3 powders. The powders were melted in alumina crucibles at 800°C for 1 h. To simulate a seleniumcaptured calcium filter, CaSeO3 was synthesized by a precipitation method using sodium selenite (Na2SeO3) and calcium nitrate (Ca(NO3)2) solutions. The glass samples were analyzed by an X-ray diffractometer (XRD). Retention of Se in tellurite glasses was analyzed by an X-ray fluorescence spectrometer (XRF) and inductively coupled plasma (ICP). The chemical durability of tellurite glass was evaluated through the PCT method.

Uranium-235, used in nuclear power generation, produces a lot of radioactive waste. Among radioactive waste nuclides, I-129 is problematic due to its long half-life (1.57×107 y) with high mobility in the environment. It should be captured and immobilized into a geological disposal environment through a stable waste form. In this study, various additives including Al, Bi, Pb, V, Mo and W were added to silver tellurite glass to prepare a matrix for immobilizing iodine, and its thermal and leaching properties were evaluated. To prepare glass, the glass precursor mixture was placed in alumina crucibles and heated at 800°C for 1 h. Except for aluminum, there was no significant loss of constituent elements. The loading of iodine in the matrix was approximately 11-15% by weigh, excluding oxygen. The normalized releases of all the elements obtained by PCT-A were below the order of 10-1 g/m2, which satisfies US regulation (2 g/m2). Differential scanning calorimetry was performed to evaluate the thermal properties of the glass samples. The glass transition temperature (Tg) increased by adding such as V2O5, MoO3, or WO3. The similar relative electrostatic field values of V2O5, MoO3, and WO3 could provide sufficient electro static field to the TeO2 interacting with the non-bridging oxygen forming Te-O-M (M = V, Mo, W) links. The addition of MoO3 or WO3 in the silver tellurite glass system increased glass transition temperature (Tg) and crystallization temperature (Tc) while maintaining the glass stability.

Radioactive cesium is a heat generated and semi-volitile nuclide in spent nuclear fuel (SNF). It is released gasous phase by head-end treatment which is a pretreatment of pyroprocessing. One of the capturing methods of gasous radioactive cesium is using zeolite. After ion-exchanged zeolite, it is transformed to ceramic waste form which is durable ceramic structure by heat treatment. Various ceramic wasteforms for Cs immobilization have been researched such as cesium aluminosilicate (CsAlSi2O6), cesium zirconium phosphate (CsZr2(PO4)3), cesium titanate (CsxAlxTi8-xO16, Cs2TiNb6O18) and CsZr0.5W1.5O6. The cesium pollucite is composed to aluminosilicate framework and cesium ion incorporated in matrix materials lattices. Many researchers are reported that the pollucite have high chemical durability. In this study, the Cesium pollucite was fabricated using mixtures of aluminosilicate denoted Absorbent product (AP) and Cs2CO3 by calcination and pelletized by cold pressing. The characterization of fabricated pollucite powder and pellets was analyzed by XRD, TGA, SEM, SEMEDS and XRF. The chemical durability of pollucite powder was evaulated by PCT-A and ICP-MS and OES. Thus, the optimal pressure condition without breaking the pellets which is low Cs2O/AP ratio and pelletizing pressure was selected. The long-term leaching test was performed using MCC-1 method for 28 days with the fabricated pollucite pellets. The leachate of leaching test was allard groundwaster and Deionized water and replaced 5 contact periods which is 3 hours, 3 days, 7 days, 14 days and 28 days and analyzed by ICPMS. The leaching rate was shown two stages. The first stage was rapid and relatively large amount of nuclides were leached. The leaching rate was decreased in the second stage. The fractional release rate of this study was shown same trend. These results were similar to previous studies.

The hydro-mechanical behavior of rock mass in natural barriers is a critical factor of interest, and it is mainly determined by the characteristics of the fractures distributed in the rock mass. In particular, the aperture and contact area of the fractures are important parameters directly related to the fluid flow and significantly influence the hydro-mechanical behavior of natural barriers. Therefore, it is necessary to analyze the aperture and contact area of fractures distributed in potential disposal sites to examine the long-term evolution of the natural barriers. This study aims to propose a new technique for analyzing the aperture and contact area using the natural fractures in KURT (KAERI Underground Research Tunnel), an underground research facility for the deep geological disposal of high-level radioactive waste. The proposed technique consists of a matching algorithm for the three-dimensional point cloud of the upper and lower fracture surfaces and a normal deformation algorithm that considers the fracture normal stiffness. In the matching process of upper and lower fracture surfaces, digital images obtained from compression tests with pressure films are used as input data. First, for the primary matching of the upper and lower fracture surfaces, an iterative closest point (ICP) algorithm is applied in which rotation and translation are performed to minimize the distance error. Second, an algorithm for rotation about the x, y, and z axes and translation in the normal direction is applied so that the contact area of the point cloud is as consistent as possible with the pressure film image. Finally, by applying the normal deformation algorithm considering the fracture normal stiffness, the aperture and contact area of the fracture according to the applied normal stress are derived. The applicability of the proposed technique was validated using 12 natural fractures sampled from KURT, and it was confirmed that the initial apertures were derived similarly to the empirical equation proposed in the previous study. Therefore, it was judged that the distribution of apertures and contact areas according to applied normal stress for laboratory-scale fractures could be derived through the technique proposed in this study.