In this research, a detailed analysis of the decay heat contributions of both actinides and non-actinides (fission fragments) from spent nuclear fuel (SNF) was made after 50 GWd·tHM−1 burnup of fresh uranium fuel with 4.5% enrichment lasted for 1,350 days. The calculations were made for a long storage period of 300 years divided into four sections 1, 10, 100, and 300 years so that we could study the decay heat and physical disposal ratios of radioactive waste in medium- and long-term storage periods. Fresh fuel burnup calculations were made using the code MCNP, while isotopic content and then decay heat were calculated using the built-in stiff equation solver in the MATLAB code. It is noted that only around 12 isotopes contribute more than 90% of the decay heat at all times. It is also noted that the contribution of actinides persists and is the dominant ether despite decreasing decay heat, while the effect of fission products decreases at a very rapid rate after about 40 years of storage.

To investigate the mechanical integrity of spent nuclear fuel, the failure behavior of the cladding tube was examined under accident conditions. According to the SNL report, the failure behavior of cladding can be broadly classified into two types. The first is failure due to bending load caused by falling. The second is failure due to pinch load caused by space grid. In this study, mechanical integrity was evaluated through the stress intensity factor applied to the crack in failure behavior due to bending load. Since the exact value of the impact load due to fall was unknown, the load was applied by increasing the value up to 200 G in 20 G increments. The size of the crack is an important input variable, and 300 um was given by referring to the EPRI report, and the elastic modulus, a material property that determines the stress field, was given 75.22 GPa by referring to the FRAPCON code. Since the relationship between the direction of stress and the direction of the crack is also a major variable, simulations were conducted for both cracks perpendicular to and parallel to the stress direction. It was confirmed that at a load of 200 G, when the crack was parallel to the stress direction, stress concentration did not occur and had a very low stress intensity factor 0.01 􀜯􀜲􀜽√􀝉. When perpendicular to the direction of stress, the stress intensity factor showed a value of 1 􀜯􀜲􀜽√􀝉. However, considering that the critical value of the stress intensity factor due to hydride is 5 􀜯􀜲􀜽√􀝉, it can be seen that perpendicular result also ensures the mechanical integrity of the cladding.

Commercial operation of KORI Unit 1 ended in 2017, and the final decommissioning plan is currently under approval from the KINS. In order for the dismantling waste to go to the repository, it is judged that the radioactive waste generated during the commercial operation should be treated and disposed in advance. Among these radioactive wastes, spent filters contain various radionuclides. The radiation dose rate from the radiation coming out of the filters ranges from a low dose rate to high dose rate. Therefore, in order to handle the spent filters, a remote processing system is required to reduce the radiation exposure of workers. This paper evaluates the radioactive inventory of filters that are stored in the filter room at the KORI unit #1. For this purpose, a method for predicting the radioactivity of each nuclide in the filter, based on the radiation dose rate, has been described using the MicroShield code, which is a commercial shielding code. The information on the filters in the field has only the creation date, type, size, and surface dose rate. In order to evaluate the radioactivity inventory using such limited data, it is possible to know the nuclide radioactivity ratio in the filter. We took out some of the filters stored on site and measured from using the ISCOS system, a gamma nuclide analyzer. The radioactivity of each nuclide in the filter was inferred by modeling with the MicroShield code, based on the radiation dose rate and the radioactivity value of each nuclide measured in the field.

Research on the safety of nuclear spent fuel has been heavily experimented and modelled from a mechanical perspective. The issues of corrosion, irradiation creep, hydride and hydrogen embrittlement have been addressed more than two decades since the early 2000s. Among these degradation behavior, hydrogen embrittlement and hydride reorientation have been the most important topics for establishing the integrity of nuclear spent fuel and have been studied in depth. In order to assess the safety of spent nuclear fuel, firstly, it is necessary to establish the safety criteria in all nuclear cycle, i.e., the failure criteria guidelines for nuclear fuel assemblies and nuclear fuel rods, and then examine the safety analysis. The contents of U.S.NRC Regulations, Title 10 General, Chapter 1 Code of Federal Regulation (CFR), Part 50, 71 and 72, describe the safety criteria for the safety assessment of nuclear fuel assemblies and nuclear fuel rods. In this study, technically important points in safety analysis on nuclear fuel are checked through the reference of those NRC regulation. As result, we confirmed that the safety assessment of nuclear fuel after 20 years of interim storage is now being tested by ORNL and PNNL. There are not quantitative criteria related to material safety. However qualitative criteria which is dependent on environmentally condition describe the safety analysis. There is some literature study about DBTT, yield stress, ultimate tensile strength, flexural rigidity data. In FRAPCON code Modelling of yield strength and creep had been established, but radial hydride or hydride reorientation has not considered.

On-site storage facility using concrete silo dry storage systems for spent nuclear fuel at Wolsong NPP site came into operation in 1992 and was expanded four times, and a total of 300 silo dry storage systems are currently in operation. The design lifetime of silo dry storage systems has been licensed for 50 years. As the dry storage systems are subject to time constraints for a limited lifetime, countries operating the dry storage systems are working to ensure the long-term integrity of dry storage systems and IAEA also recommends that the dry storage systems be assessed for long-term storage. To demonstrate the long-term integrity due to material degradation during the licensed design lifetime, the structural integrity of silo dry storage systems was evaluated by considering the material degradation characteristics of concrete. The concrete compressive strength results measured so far by the rebound hammer method, which is an internationally standardized nondestructive test method for converting hardness into compressive strength using the correlation between rebound number and strength at the time of a Schmidt hammer strike, were analyzed in accordance with Wolsong NPP’s procedure to quantify the degradation characteristics, and the prediction of concrete strengths for 20 years and 50 years after construction of the silo dry storage systems was determined, respectively. Based on these residual compressive strengths, structural analyses of the silo dry storage systems were carried out under normal, off-normal and accident conditions of the related regulations, and the structural integrity of silo dry storage systems was reevaluated. It was confirmed the silo dry storage systems are able to maintain structural integrity up to the design lifetime of 50 years even if the concrete is deteriorated.

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.

To reduce the environmental burden caused by the disposal of spent nuclear fuel and maximize the utilization of the repository facility, waste burden minimization technology is currently being developed at the Korea Atomic Energy Research Institute (KEARI). The technology includes a nuclide management process that can maximize disposal efficiency by selectively separating and collecting major nuclides in spent nuclear fuel. In addition, for efficient storage facility utilization, the short-term decay heat generated by spent nuclear fuel must be removed from the waste stream. To minimize the short-term thermal load on the repository facility, it is necessary to separate heat generating nuclides such as Cs-137 and Sr-90 from the spent fuel. In particular, Sr-90 must be separated because it generates high heat during the decay process. KAERI has developed a technology for separating Sr nuclides from Group II nuclides separated through the nuclide management process. In this study, we prepared Sr ceramic waste form, SrTiO3, by using the solid-state reaction method for long-term storage for the decay of separated Sr nuclides and evaluated the physicochemical properties of the waste form. Also, the radiological and thermal characteristics of the Sr waste form were evaluated by predicting the composition of Sr nuclides separated through the nuclide management process, and the estimation of centerline temperature was carried out using the experimental thermal data and steady state conduction equation in a long and solid cylinder type waste form. These results provided fundamental data for long-term storage and management of Sr waste.

The conventional research trend on spent fuel was safety analysis based on mechanical perspective. Analysis of spent fuel cladding is based on the temperature of cladding and pressure inside cladding. To improve fuel cladding analysis, precise and accurate thermal safety evaluation is required. In this study a database which is about thermal conductivity and emissivity for the thermal modeling was established for a long-term safety analysis of spent fuel. As a result, we confirmed that the thermal conductivity of zirconium hydride was not accounted in conventional model such as FRAPCON and MATPRO. The conductivity of zirconium and its oxide was evaluated only as a function of temperature. However, the behavior of heat conductivity and emissivity is determined by the change of the material properties. The material properties depend on the microstructural characteristic. It can be seen that this conventional approach does not consider the microstructure change behavior according to vacuum drying process or burn-up induced degradation phenomena. To improve the thermal properties of spent nuclear fuel cladding, the measurement experiments of heat conduction and emissivity are required according to spent fuel experience and status such as the number of vacuum drying, cooling rate, burn up, hydrogen concentration and oxidation degree. In previous domestic reports and papers, we found that relative data between thermal properties and spent fuel experience and status does not exist. Recently, in order to understand the failure mechanism of hydrogen embrittlement, many studies have been conducted by accounting and spent fuel experience and status in a mechanical perspective. If microstructure information could be obtained from these studies, the modeling of thermal conductivity and emissivity will be possible indirectly. According to a recent abroad paper, it was confirmed that the thermal conductivity decreased by about 30% due to irradiation damage. The radiation damage effects on thermal conductivity also has not been studied in zirconium oxide and hydride. These un-revealed phenomena will be considered for the thermal safety model of spent fuel.

Currently, the interim storage pools of spent fuels in South Korea are expected to become saturated from 2024. It is required to prepare an operation plan of a domestic dry storage facility during a long-term period, with the researches on safety evaluation methods. This study modified the FRAPCON code to predict the spent fuel integrity evaluation such as the axial cladding temperature, the hoop stress and hydrogen distribution in dry storage. The cladding temperature in dry storage was calculated using the COBRA-SFS code with the burnup information which was calculated using the FRAPCON code. The hoop stress was calculated using the ideal gas equation with spent fuel information such as rod internal pressure. Numerical analysis method was used to calculate the degree of hydrogen diffusion according to the hydrogen concentration and temperature distribution during a dry storage period. Before 50 years of dry storage, the cladding temperature and hoop stress decreased rapidly. However, after 50 years, they decreased gradually and the cladding temperature was below 400 K. The initial temperature distribution and hydrogen concentration showed a parabolic line, but hydrogen was transferred by the hydrogen concentration and temperature gradient over time.

In this study, an appropriate modified atmosphere packaging (MAP) condition to minimize physiological disorders while lowering weight loss was sought. To reduce weight loss during storage, kimchi cabbages packed with 0, 32, 40, 48 perforated low-density polyethylene (LDPE) films, with a diameter of 14 mm, were stored in pallet units for 90 days at 1-2oC, and their loss rate, physiological disorders, total bacteria count, pH, and solid content were analyzed. It was found that as the number of holes increased, the weight loss ratio increased proportionally. However, the difference between the perforations was relatively small compared with the sample without film packaging. On the other hand, it was also observed that the lower the number of holes was, the lower the incidence of physiological disorder was because the cold air penetrated through the perforated hole while inhibiting physiological effects, releasing heat and carbon dioxide generated by respiration. Considering the weight loss rates and physiological disorders such as black speck and soft rot, the kimchi cabbage packed with 48 perforated films (73.9 cm2) exhibited the most satisfactory condition. Using this storage condition, along with 2-3oC temperature and 91-95% relative humidity inside the pallet, a highly suitable condition for kimchi processing was obtained to secure kimchi cabbage.

사용후핵연료를 저장하는 볼트체결 저장용기의 격납경계를 형성하는 주된 구성요소는 금속 밀봉재이다. 이러한 금속 밀봉 재는 열과 방사선에 의해 그 성질이 저하된다. 또한, 금속 밀봉재가 강한 열에 장기간 노출되면 크리프 현상이 발생한다. 이러한 크리프는 밀봉시스템에 응력 이완을 가져와서, 결국 밀봉재의 건전성을 해치게 된다. 이러한 응력 이완은 금속 밀봉재의 밀봉성능 저하로 이어지며, 저하의 정도에 따라 저장용기의 누설을 야기할 수 있다. 또한, 볼트 체결력의 감소도 밀봉성능 저하에 영향을 미친다. 본 논문에서는 금속 밀봉재의 격납건전성과 볼트체결력 감소를 평가하기 위해 수행한 가속화 시 험의 결과에 대하여 기술한다. 전 시험기간 동안 각 시편에서의 누설률, 볼트 변형률, 금속 밀봉재 주변 온도를 계측하여 분석하였고, 금속 밀봉재는 저장기간 50년 동안 격납건전성을 유지함을 입증하였다. 또한, 가속화 시험의 타당성에 대해서 기술하였다.

The exosomes are the most studied nanometer-sized extracellular vesicles (40-100 nm) in eukaryotic cell. Exosomeharbors various molecular components of their cell origin, including nucleic acids and proteins, and it is involved in intraand intercellular communication. These characteristics make exosomes to be a prospective biomarkers, therapeutic agents,or drug delivery vehicles. Edible insect industry is rapidly growing up in Korea. The insect exosomes were isolated fromthe larvae of Korean rhinoceros beetle, Allomyrina dichotoma, and White-spotted flower chafer, Protaetia brevitarsis, sothat they can be used for diagnosis of insect disease. The stable preservation of exosome is very important for diagnosisand research, especially for long term storage. Here, the stable recovery of exosome isolated from hemolymph of A. dichotomalarva was evaluated by analyzing exosomal protein and RNA after storage in -70°C for three months.

Quality assessment of sweet persimmon or “Fuyu” was evaluated over room temperature distribution periods after 40-day storage at 0oC cold room. All Hunter’s values (L, a, and b) of sweet persimmon were significantly decreased after 4-day room temperature distribution while the flesh firmness was drastically decreased after 3-day room temperature distribution. Glucose and fructose contents were significantly increased by the degradation of sucrose after 3-day room temperature distribution. The highest extraction yield was obtained in the flesh of persimmon. Extraction yields of each part of “Fuyu” persimmon were 15.53% for flesh, 10.53 for seed, and 13.83 for flower bud. However, the total phenolic content of flesh was 4.8 μg/mg which was extremely lower than that of the seed (175.5 μg/ mg) and flower bud (178.2 μg/mg) of sweet persimmon. With high phenolic content, both the flower bud and seed of sweet persimmon showed excellent antioxidant activities by 2,2-diphenyl-1-picrylhydrazyl (DPPH) and 2,2'-azinobis( 3-ethylbenzothiazoline-6-sulfonic acid) (ABTS) free radical scavenging test. The results showed a great potential for byproducts (seed and flower bud) of long-term, cold-stored sweet persimmon “Fuyu” as a good antioxidant material for novel functional foods.

본 실험은 1-MCP처리와 저장방법이 아스파라거스의 신선도 유지에 미치는 영향을 알아보고자 수행하였다. 1-MCP 처리는 MA 조건과 유공필름(관행조건)에 처리 되었으며 무처리구를 대조구로 사용하여 저장하였다. 처리된 아스파라거스는 40일 동안 4oC에 저장되었다. 저장 중 생체중 감소율은 1-MCP 처리를 포함한, MA저장 처 리구에서 0.2% 이하였고 유공필름에 저장된 1-MCP 처리는 생체중 감소율이 저장 20일 후부터 다소 낮아졌다. 포장 내 이산화탄소 농도는 MA저장 처리구 모두 최적 CA 조건인 5-12%(CO2)의 농도를 유지하였고, 산소농도는 5-15%를 유지하였다. 에틸렌 농도는 MA저장 1- MCP처리구가 가장 낮았다. 저장 최종일의 외관상 품질, 이취, 경도, 당도, 그리고 Hue angle 값은 MA저장에서 관행저장보다 우수하였으나 1-MCP처리간 차이는 없었다. 이상의 결과로 1-MCP처리는 아스파라거스의 품질 및 저장성에 영향을 미치지 않았다.

본 연구는 다양한 조건에서 장기 보존한 느티만가닥버섯 균주의 균사 생존력과 재배성능 검증을 수행하였다. 균사 생존력은 장기 보존된 균주를 9, 21,33, 45개월 단위로 해동하여 PDA 평판배지에 접종한 다음 1, 2, 3, 4, 5차 계대배양에 따른 균사 생장길이를 측정하여 검증하였다. 액체질소에 보존한 균사를 해동 직후 실시한 1차 계대배양에서는 동결보호제로써 Glycerol 10%를 단용으로 처리한 실험구에서균사생장이 저조한 결과를 나타내었으나, 5차 계대배양에서는 버섯 균사의 생장력이 다른 실험구와 동일한 수준으로 회복되어, 저장방법에 따른 균사의 생존력에는 큰 영향이 없는것으로 판단된다. 그러나 느티만가닥버섯을 액체 질소에 장기간 보존하고자 할 때는 동결보호제로 사용되는 Glycerol 10%에 Trehalose5%를 첨가하여 보존해야 균사의 생존력을 유지하는데 도움이 된다.장기 저장에 따른 느티만가닥버섯 균주의 재배 성능 시험에서는 4oC 사면배지에 보호제로써 mineraloil을 사용하는 것이 사면배지 자체로 보존하는 것에비해 수확량이 감소하는 결과를 나타내므로 장기 저장시에 mineral oil을 사용하는 것은 지양해야 할 것으로 판단된다.

본 연구는 기존의 농가 보급형 고구마 저장고에 가열기(본체), 환기팬, 배풍구 팬, 습도 조절을 위한 관으로 구성된 간이 큐어링 설비를 부가시킨 후, 그 효과를 검증하기 위하여 수행되었다. 6일간의 큐어링 처리 기간 동안 저장고 내 기온 및 상대습도는 32℃, 90%로 설정되었다. 고구마의 저장 기간(6 개월) 동안 1개월에 1회씩 표본을 수집하여 물리화학적 품질을 측정한 결과, 부패율과 부패 정도를 함께 보여주는 McKinney Index는 큐어링 처리구의 경우 0.83%였으며, 큐어링 처리를 하지 않은 대조구의 경우 5.08%였다. 큐어링 처리구의 경도 및 당도, 건물율은 모두 대조구보다 높았으며, 대조구의 표피색이 큐어링 처리구보다 빠르게 변하여 큐어링 처리가 저장기간 동안 고구마의 변색을 지연시켰음을 알 수 있었다. 본 연구를 통하여 간이 큐어링 설비를 이용한 큐어링 처리는 초기 비용이 많이 소모되는 일반 큐어링 설비를 대체하여 고구마의 물리화학적 품질을 증진시킬 수 있으며, 저장 기간을 늘릴 수 있음을 알 수 있었다.