This study is about the production of radiation sources of simulated concrete and soil reference materials to verify the validity of the quality establishment and measurement of the detector (HPGe) of the radioactive soil and concrete waste classification system, which is being developed to quickly and accurately classify nuclear decommissioning waste. Specific activity of gamma nucleus among radioactive wastes is evaluated using gamma spectroscopy. At this time, in order to verify the validity and reliability of measuring equipment, it shall be a standardized substance of the same medium as nuclear decommissioning waste (chemical ingredients, particles, density, etc.) in order to correct the energy and efficiency of gamma nuclide analysis equipment. The CRM used for the detector’s energy correction used a 1 L Marinelli beaker standard correctional radiation source consisting of 10 radioactive isotopes. In order to correct efficiency, in accordance with the production and certification process of the Korea Standards and Research Institute, it has produced artificial simulated radioactive concrete similar to nuclear decommissioning waste (30% for cement, 60% for regulation and 10% for bentonite). The radioactive homogeneity of the simulated concrete reference materials was evaluated using dispersion analysis (ANOVA) in accordance with ISO Guide 35, while 137Cs and 60Co of concrete reference materials were able to obtain homogeneous measurements both in and between bottles. The self-absorption rate of the simulated concrete reference material was determined by the MCNP computer simulation measurement method, and the self-absorption correction coefficients of 137Cs and 60Co were assessed at 0.995 and 0.996, respectively, and the standard value for the radiation of the simulated concrete reference material was calculated on the weighted average of the measurements of 20 samples. The uncertainty about the reference value was calculated by combining measurement uncertainty (Type B evaluation), bottle to bottle standard deviation, and uncertainty within bottle by modifying the formula suggested in ISO Guide 35. The concentration of 137Cs and 60Co of reference materials was divided into high-speed measurement mode and precision measurement mode in consideration of the self-disposal standard. The reference value and uncertainty of expansion among reference materials for high-speed measurement mode were rated at 1,032.7 ± 64.0 Bq·kg−1and 1,083.7 Bq·kg−1, respectively. The standard value and expansion uncertainty for 137Cs and 60Co among reference materials for precision measurement mode were rated at 113.7 ± 10.0 Bq·kg−1 and 122.3 ± 10.3 Bq·kg−1, respectively.

As Kori-1 permanently shut down in Korea, it is expected that a large amount of radioactive waste will be generated during decommissioning of nuclear power plants. Radioactive concrete waste is contaminated up to depth of 100 mm with radionuclides such as 137Cs and 60Co. The radioactive waste should be accurately classified to reduce the cost of disposing of radioactive waste. Therefore, the specific radioactivity of waste must be precisely evaluated by gamma-ray measurements emitted from the radionuclides. In general, the effectiveness of the radioactivity measurement and process is confirmed using certified reference material (CRM) composed of water or agar. However, the decommissioning waste differs from this CRM in apparent density and chemical elements, so the specific radioactivity is underestimated or overestimated. Therefore, reference material composed of the same apparent density and chemical elements as the sample is required to improve the quality of radioactivity measurement. The purpose of this study is to develop a concrete reference material for the nuclear decommissioning waste. The concrete reference material composed of SiO2, CaO, and Al2O3 were manufactured in compliance with ISO Guide 35. 10 bottles were randomly selected for homogeneity test, and 2 samples for analysis were taken from each bottle. The specific radioactivity was measured using an HPGe detector with an efficiency of 30%. The results of the homogeneity test of 137Cs and 60Co satisfied the requirements of ISO Guide 35. Coincidence summing and selfabsorption effects were corrected using the Monte Carlo efficiency transfer code and Monte Carlo NParticle transport code. The reference values of 137Cs and 60Co in the concrete reference material were evaluated in the range of 1,000–1,100 Bq·kg−1 and extended uncertainty was around 7%.

As the plan for the nuclear dismantlement due to the permanent shutdown of Kori-1 and Wolseong- 1 nuclear power plants has been concretized, a “movable radionuclide analysis system” is being developed that can quickly and accurately analyze large amounts of radioactive waste generated on the sites during dismantling. This system has various advantages from the perspective of strict regulations on the radioactive waste movement and social acceptability, such as preventing unexpected accidents while moving on the national highway or expressway, reducing various documents and immediate response to dismantling plans. Currently the system is being developed to be equipped with previously developed sample pretreatment and radioactivity measuring equipment and automated volatile and nonvolatile nuclide separation equipments, but to ensure mobile stability, it needs to analyze factors and establish stability standards. In the KS Q ISO/IEC 17025:2017 standard, the requirements for “facilities and environmental conditions” are a very important factor in building reliability for consumers as part of the quality guarantee for this facility. In order to meet the requirements, the technical standards of various test equipment to be installed in this facility were investigated. The physical, chemical, and radiological hazards that could affect the safety of the equipment and workers in the process of moving the equipment between nuclear power plants or between nuclear dismantling sites were derived from vibrations, rapid changes in temperature and humidity, and the spread of contamination from radioactive waste samples. Therefore, the scope of application of the law, which is the basis for securing stability during movement, was classified into two situations: movement from facility manufacturer to installation site (non-contaminated) and movement from primary to secondary use (contaminated). And in order to investigate the Nuclear Safety Act, enforcement ordinances, and radiation safety management, and to establish standards for packaging and transportation of radioactive materials, the results of transportation tests and transport details were compared and analyzed. Finally, the air suspension systems and the automatic temperature and humidity control devices were analyzed to establish standards for securing stability against the vibration and the sharp changes in the temperature and humidity, and countermeasures such as accident measures in accordance with the Enforcement Decree of the Nuclear Safety Act were also investigated.

The design of nuclear fuel storage and handling area includes the activities related to the storage and inspection before fuel loading, transfer into the reactor, removal of irradiated fuel to the spent fuel storage rack, underwater handling and storage, and handling into a shipping cask. The purpose of this study is to provide the design requirements for the spent fuel pool to be prevented from the loss of cooling water and for heavy load control to prevent any load drop resulting in damage to safetyrelated systems during heavy load handling in accordance with the regulatory guidelines. And another purpose is to review the sizing of minimum wet storage capacity in the spent fuel pool based on the maximum refueling batch from the core during refueling plus a full core off-load of fuel assemblies and the minimum discharge burnup spent fuel storage during the design life of plant requested by the utility. As the results of this study, the current general arrangement for the spent fuel storage and handling area and the minimum storage capacity are evaluated. These can be good recommendations to enhance more safe and efficient if implemented to the new nuclear power plants.

This study provides technical information about the nuclear fuel handling process, which consists of various subprocesses starting from new fuel receipt to spent fuel shipment at a nuclear power plant and the design requirements of fuel handling equipment. The fuel handling system is an integrated system of equipment, tools, and procedures that allow refueling, handling and storage of fuel assemblies, which comprise the fuel handling process. The understanding and reaffirming of detailed code requirements are requested for application to the design of the fuel handling and storage facility. We reviewed the design requirements of the fuel handling equipment for its adequate cooling, prevention of criticality, its operability and maintainability, and for the prevention of fuel damage and radiological release. Furthermore, we discussed additional technical issues related to upgrading the current code requirements based on the modification of the fuel handling equipment. The suggested information provided in this paper would be beneficial to enhance the safety and the reliability of the fuel handling equipment during the handling of new and spent fuel.

This study evaluates the gastroprotective effect of cabbage extract with sulforaphane content of 5.19 mg/L and Smethylmethionine content of 469.28 μg/L. In vitro, the lipopolysaccharide (LPS)-treated group had an increased NO activity compared to the normal group, and the concentration of NO was reduced when the cabbage extract was treated in the dose manner. The level of IL-6 induced by LPS was dose-dependently reduced when the extract was treated. The cabbage extract concentration was orally administered in rats at 5.75 mg/kg, 11.5 mg/kg, and 23 mg/ kg, and the inhibitory effect on gastric damage by HCl-ethanol was observed. Histological analysis exhibited mucosal erosion in the gastritis model compared to the normal group, while the ameliorating effect of the generated erosion was observed in the cabbage-treated group. The histamine concentration was significantly increased in the gastritis-induced animal model, and the histamine concentration was decreased in the 23 mg/L-treated group of cabbage extract. In conclusion, the results of this study suggest that cabbage extract not only down-regulates cytokines in vitro, but is also directly involved in histamine secretion in an animal model of gastritis; therefore, cabbage extract can help inhibit gastrointestinal disorders by improving the protective barrier.

프리지아 ‘Sunny Gold’는 농촌진흥청 국립원예특작과학원 에서 2010년 노랑색 반겹꽃 프리지아 육성계통 ‘036010’을 모본으로 진노란색 홑꽃 ‘Golden Flame’을 부본으로 교배하여 획득한 종자로부터 2011년 진노란색 겹꽃의 향기가 강한 프리지아 계통을 선발하여 품종화 하였다. 2011년부터 2016년까지 개화 생육특성검을 수행하였으며 핵심수요자의 기호도 평가를 통해 선발되어 2017년 ‘Sunny Gold’ 로 명명되었다. ‘Sunny Gold’는 RHS color chart YO17B의 노란색 겹꽃 프리지아 품종으로 화폭은 6.7cm로 대조품종 ‘Golden Flame’ 6.1cm에 비해 크고, 분지수는 6.5로 다수확성 품종이다. 초장이 101.9cm로 초세가 강하다. ‘Sunny Gold’의 소화수 및 소화장은 각각 13.0개, 9.3cm이며 개화소요일수는 137.7일이다. 이 품종의 절화수명은 약 9일이며 자구번식력은 5.3배로 대조 품종 ‘Golden Flame’ 4.3배에 비해 우수하다. 전자코를 이용한 PCA분석결과 PC1과 PC2는 각각 99.3%와 0.6%로 전체 변이량의 99.9%를 반영하고 있다. Rader plot 분석결과 총 6개 센서에서 모두 ‘Sunny Gold’의 센서값이 향기가 강한 상용품종 ‘Yvonne’의 값에 비해 높게 나타나 ‘Sunny Gold’의 향기가 더 강한 것으로 나타났다.

포인세티아(Euphorbia pulcherrima. Willd. Ex Klotzch) ‘Red Ball’은 국립원예특작과학원에서 2019년에 육성한 품종이다. ‘Red Ball’은 긴 관상 기간을 가지는 ‘Christmas Eve’를 모본으로 빨간색 주름진 덮개잎이 볼 타입 꽃을 만드는 ‘Winter Rose Early Red’를 부본으로 교배하여 획득한 실생 계통을 선발하여 육성하였다. 2016년부터 2018년 생육, 개화, 균일성 등에 대하여 1,2차 특성 검정을 실시하였다. 이후 2019년 3차 특성검정을 실시하여 최종선발한 후 직무육성품종심의회에 상정하여 ‘Red Ball’로 명명하였다. ‘Red Ball’품종은 전체적으로 짧은 적색 덮개잎을 가지며 꽃 모양이 납작하지 않아 입체감이 높다. 또한 단일 처리후 약 5.5~6주가 경과하면 완전히 착색되어 출하가 가능할 만큼 착색소요기간이 짧다. ‘Red Ball’은 2021년 4월 12일에 국립종자원에 품종등록(제8497호) 되었다.