The optimization of deacetylation process parameters for producing chitosan from isolated chitin shrimp shell waste was investigated using response surface methodology with central composite design (RSM-CCD). Three independent variables viz, NaOH concentration (X1), radiation power (X2), and reaction time (X3) were examined to determine their respective effects on the degree of deacetylation (DD). The DD of chitosan was also calculated using the baseline approach of the Fourier Transform Infrared (FTIR) spectra of the yields. RSM-CCD analysis showed that the optimal chitosan DD value of 96.45 % was obtained at an optimized condition of 63.41 % (w/v) NaOH concentration, 227.28 W radiation power, and 3.34 min deacetylation reaction. The DD was strongly controlled by NaOH concentration, irradiation power, and reaction duration. The coefficients of correlation were 0.257, 0.680, and 0.390, respectively. Because the procedure used microwave radiation absorption, radiation power had a substantial correlation of 0.600~0.800 compared to the two low variables, which were 0.200~0.400. This independently predicted robust quadratic model interaction has been validated for predicting the DD of chitin.

본 연구에서는 단기간의 UV-A 조사가 시금치(Spinacia oleracea L.)의 생장과 생리활성물질에 미치는 영향을 평가 하였다. 시금치 묘는 200μmol·m-2·s-1 PPFD, white LED, 광 주기 12시간, 온도 20°C, 상대습도 70%, 이산화탄소 농도 500μmol·mol-1의 수직농장 모듈에서 재배되었다. 파종 후 5 주된 묘는 7일 동안 20W·m-2와 40W·m-2의 두 가지 에너지 수 준에서 연속적으로 UV-A(피크파장: 385nm) 조사한 후 생육 특성, 광합성 파라미터, 이미지 형광, 총 페놀 함량, 항산화도, 그리고 총 플라보노이드 함량을 분석하였다. 결과적으로, UV-A20W 처리는 시금치의 생체중과 건물중을 증가시켰다. 하지만, UV-A 처리구와 대조구 사이의 광합성 파라미터에는 유의한 차이가 나타나지 않았다. 광계Ⅱ의 최대양자수율 (Fv/Fm)은 모든 UV-A 처리에서 7일동안 지속적으로 감소했 다. 또한, UV-A20W 처리에서 식물체당 총 페놀 함량과 항산화 도는 처리 7일째 증대되었으며, 총 플라보노이드 함량은 처리 5일째부터 유의적으로 증가하였다. 이러한 결과는 UV-A LED 보광이 수직농장과 같은 폐쇄형 식물 생산 시스템에서 재배되는 시금치의 생장과 품질을 향상시킬 수 있음을 시사 한다.

Currently, Japan is undertaking a nationwide project to measure and map radioactive contamination around Fukushima, as part of the efforts to restore normalcy following the nuclear accident. The Japan Atomic Energy Agency (JAEA) manages the Fukushima Environmental Safety Center, located approximately 20 km north of the Fukushima Daiichi nuclear power plant in Minamisōma City, Fukushima Prefecture. In collaboration with the JAEA, this study involved conducting comparison experiments and analyses with radiation detectors in high radiation environments, a challenging task in Korean environments. Environmental radiation surveys were conducted using three types of detectors: CZT, NaI(Tl), and LaBr3(Ce), across two contaminated areas. Dose rate values were converted using dose rate conversion factors for each detector type, and dose rate maps were subsequently created and compared. The detectors yielded similar results, demonstrating their feasibility and reliability in high radiation environments. The findings of this study are expected to be a crucial reference for enhancing the verification and supplementation of procedures and methods in future radiation measurements and mobile surveys in high-radiation environments, using these three types of radiation instruments.

국내 큰느타리버섯은 매년 수출이 증가하고 있는 주요 신선 농산물 중 하나이다. 긴수염버섯파리는 농업, 특히 버섯생산에 피해를 주는 악 명높은 해충이다. 긴수염버섯파리의 유충은 주로 농작물에 직접 피해를 유발하고 성충은 몇몇 위험한 진균 병원체의 매개체 역할을 한다. 본 연구 에서는 전자빔, 엑스선, 그리고 감마선의 조사선량에 따른 긴수염버섯파리의 발육 및 생식에 미치는 영향을 평가했다. 또한 큰느타리버섯이 채워 진 박스 안에서 긴수염버섯파리를 제어할 수 있는 최적선량을 찾기 위해 방사선 및 에너지량에 따른 실증실험을 수행하였다. 그 결과 전자빔, 엑 스선, 감마선 모두 50 Gy에서 긴수염버섯파리의 발육 및 생식이 억제되었다. 또한 큰느타리버섯이 채워진 수출용 박스 상, 중, 하 위치에서 긴수 염버섯파리는 전자빔 150 Gy, 엑스선 100 Gy, 그리고 감마선 50 Gy에서 발육 및 생식을 억제하는 것으로 나타났다. 이러한 결과들은 수출 검역 통합관리 시스템 구축의 기초자료로 제공될 수 있다. 또한 농산물의 안전성 확보와 수출경쟁력 강화에 기여하리라 사료된다.

Owing to the rapid rise of global energy demands, the operation of nuclear power plants is still indispensable. However, following the nuclear accident at Fukushima-Daiichi in 2011, the secure sequestration of radioactive waste has become critical for ensuring safe operations. Among various forms of nuclear wastes, capturing radioactive organic iodide (ROIs, e.g., methyl iodide, ethyl iodide, and propyl iodide) as one of the important species in gas phase waste has been challenged owing to the insufficient sorbent materials. The environmental release of ROIs with high volatility can give rise to adverse effects, including the accumulation of these substances in the thyroid and the development of conditions such as hypothyroidism and thyroid cancer. Compared to an iodine molecule, ROIs exhibit low affinity for conventional sorbents such as Ag@mordenite zeolite and triethylenediamine-impregnated activated carbon (TED@AC), resulting in lower sorption rates and capacities. Furthermore, in conditions resembling practical adsorption environments with high humidity, the presence of H2O significantly impedes the adsorption process, leading to a nearly complete cessation of adsorption. To address these issues, metal-organic frameworks (MOFs) can be effective alternative sorbents owing to their high surface area and designable and tailorable pore properties. In addition, the wellfined crystalline structures of MOFs render in-depth study on the structure-properties relationship. However, there has been limited research on the adsorption of ROIs using MOFs, with the majority of adsorption processes relying on highly reversible physisorption. This type of ROIs adsorption not only exists in a precarious state that is susceptible to volatilization but also exhibits significantly reduced adsorption capabilities in humid environments. Thus, for the secure adsorption of the volatile ROIs, the development of sorbents capable of chemisorption is highly desirable. In this study, we focused on ROIs adsorption by electrophilic aromatic substitution with the electron-rich m-DOBDC4− (m-DOBDC4− = 4,6-dioxo-1,3-benzenedicarboxylate) present in Co2(m -DOBDC). The chemisorption of ROIs via electrophilic aromatic substitution not only leads to the formation of C-C bonds, ensuring stability but also triggers color changes in the crystal by interacting with open-metal sites and iodide ions. Leveraging these advantages, we developed an infrared radiation-based sensing method that demonstrates superior performance, exhibiting high adsorption capacities and rates, even under the challenging conditions of high-humidity practical environments.

The first commercial operation of Kori-1, which commenced in April 1978, was permanently shut down in June 2017, with plans for immediate dismantling. The decommissioning process of nuclear power plants generates a substantial amount of radioactive waste and poses significant radiation exposure risks to workers. Radioactivity is widely distributed throughout the primary coolant system of the reactor, including the reactor pressure vessel (RPV), steam generator (SG), and pressurizer. In particular, the SG has a considerable size and complex geometry, weighing approximately 326 tons and having a volume of 400 m3. The SG tubes are known to contain high levels of radioactivity, leading to significant radiation exposure to workers during the dismantling process. Therefore, this study aims to evaluate the workers’ radiation exposure during the cutting of SG tubes, which account for approximately 95% of the total radiation dose in the SG. Firstly, the CRUDTRAN code, developed to predict the behavior of soluble and particulate corrosion products in a PWR primary coolant system, is used to estimate the radioactive inventory in the SG tubes. Based on decontamination factors (DF) obtained in the SG tubes through overseas experience, the expected reduction in radioactivity during the Kori-1 reactor’s full-system decontamination (FSD) process is considered in the CRUDTRAN results. These results are then processed to estimate the radioactivity in both the straight and bent sections of the tubes. Subsequently, these radioactivity values are used as inputs for the MicroShield code to calculate the worker radiation exposure during the cutting of both straight and bent sections of the tubes. The cutting process assumes that each SG tube section is cut in a separate, shielded area, and the radiation exposure is assessed, taking into account factors such as cutting equipment, cutting length, working hours, and working distance. This study evaluates the worker radiation exposure during the cutting of SG tubes, which are expected to have a significantly high radioactivity due to chalk river unidentified deposit (CRUD). This assessment also considers the reduction in radioactivity within the steam generator tubes resulting from the FSD process. Consequently, it enables an examination of how worker radiation exposure varies based on the extent of FSD. This study may provide valuable insights for determining the scope and extent of the FSD process and the development of shielding methods during the dismantling of SG tubes in the future.

Korea Atomic Energy Research Institute (“KAERI”) has been developing pyroprocess technology for the sustainable use of nuclear energy and radioactive waste reduction, and is conducting design studies for a Pyroprocess Commercializing Research Facility (PCRF). High-level radioactive materials such as spent nuclear fuel, which are handled in the hot cell of the PCRF, physically change materials directly or cause chemical changes through ionization or excitation depending on the energy and types of radiation. Therefore, all facilities, including process equipment and remote handling equipment, installed into the hot cell must be evaluated for radiation hardness to be maintained in the radiological environmfent so that processes can proceed throughout the design life of the facility. In addition, as the maintenance paradigm has recently shifted from corrective maintenance to predictive maintenance, it is necessary to know in advance the condition of the equipment or facility in the radiological environment. In this study, an analysis of the radiation environment of the hot cell in the PCRF was conducted through source term, and the radiological dose impact was evaluated through the results of irradiation experiments of major components by reference data. Then, the actual dose contribution was identified through dose assessment using the MCNP code based on the pyroprocess equipment, and the radiation hardness requirements for the facility and equipment in the hot cell were derived by the above results.

According to IAEA GSR Part.6, Decommissioning is carried out on the basis of planning and evaluation to ensure safety, protection of workers, public, and environment. Then, the decommissioning project of nuclear facility includes a radiation protection plan that reflects the regulatory requirements and international recommendations of each country and the internal regulations of the licensee. The scope of the radiation protection plan covers all radiation activities related to the dismantling and disposal of contaminated facilities subject to decommissioning. Radiation protection applications in the United States, a country with previous experience in decommissioning nuclear facilities, include 10 CFR 20 for NRC management facilities and 10 CFR 835 for facilities under DOE. In this study, we analyzed two cases of decommissioning plans to which NRC regulations are applied. In 1992, Yankee Atomic Electric Company (YAEC), the licensee of Yankee Nuclear Power Station (YNPS), notified NRC of the permanent shutdown of YNPS and submitted decommissioning plan accordingly. This decommissioning plan consists of a total of 9 chapters, and section 3.2 describes the radiation protection of decommissioning workers. The contents of the radiation protection program consist of 16 subsections. Another case is the decommissioning work plan of U.S. Navy Surface Ship Support Barge (SSSB), which used in Virginia to support the refueling of the U.S. Navy’s reactor vessel. This document was developed based on the NUREG-1757 and was revised in 2021 after receiving NRC comment. SSSB’s project radiation protection plan is described in appendix 1, and the contents consist of a total of 28 sections except for reference. In Korea, decommissioning plan is developed in accordance with “Standard Format and Content of the Decommissioning Plan for Nuclear Facilities”. According to this regulation, the radiation protection plan for licensing documents submitted at the time of application for approval of decommissioning execution shall describe the organization and functions for implementing of plan, methods, cycles and procedures for performing radiation protection and radiological monitoring. Also, the safety review guidelines of regulatory body also require radiation protection plans and procedures to ensure ALARA activities during decommissioning. In the case of the final decommissioning plan of Kori-1, which is currently submitted to regulatory body for licensing review, the decommissioning radiation protection plan is divided into 8 sections. Although the classification criteria for the radiation protection plan categories described above facilities are different, it could be seen that the following 7 contents are included in common: (a) ALARA application and organization for implementation, (b) Management of radiation control area, (c) Process of radiation work, (d) Radiation and contamination control, (e) Personnel radiation exposure monitoring, (f) Radioactive material management, (g) Radiation protection training.

If radioactive plumes are released outside due to loss of containment building integrity during a nuclear power plant accident, these materials might travel with the wind, affecting both the surrounding environment and neighboring countries. In China, most nuclear power plants are located on the eastern coast. Consequently, a radioactive plume generated during an accident could negatively impact even the western part of the Korean Peninsula due to westerly winds. To detect such problems early, respond quickly, and protect residents, a system that can monitor aerial radiation under normal conditions is needed. Additionally, a detection system that can operate in real-time in an emergencies conditions is required. The current method for aerial radiation measurement takes environmental radiation data from a monitoring post 1.5 m above the ground and converts it to altitude. To measure actual aerial radiation, an expansive area is surveyed by aircraft. However, this approach is both time-consuming and expensive. Thus, to monitor radioactive plumes influenced by environmental factors like wind, we need a radiation detector that can gauge both radioactivity and directionality. In this study, we developed a radiation detector capable of assessing both the radioactivity and directionality of a radioactive plume and conducted its performance evaluation. We miniaturized the radiation detector using a CZT (Cadmium Zinc Telluride) sensor, enabling its mounting on unmanned aerial vehicles like drones. It is configured with multi-channels to measure directionality of a radioactive plumes. For performance evaluation, we positioned two-channel CZT sensors at 90 degrees and measured the energy spectrum for angle and distance using a disk-type radioactive isotope. Using this method, we compared and analyzed the directionality performance of the multi-channel radiation detector. We also confirmed its capability to discern specific radioactivity information and nuclide types in actual radioactive plumes. Our future research direction involves mounting the multi-channel radiation detector on a drone. We aim to gather actual aerial radiation data from sensors positioned in various directions.

At the end of 2022 there were 439 nuclear power reactors in operating around the world, including 25 nuclear power reactors of Korea. Domestic nuclear power plants (NPPs) continuously produce low and intermediate-level radioactive waste (LILW) and spent nuclear fuel (SNF). As amount of radioactive waste is increasing and interim storage facilities meet limitation of their capacity, radioactive waste need to be transported. Consequently, the demand for radioactive waste transportation is increasing and the importance of Radiation Risk Assessment Codes (RRACs) for radioactive waste transportation is also on the rise. Considering the domestic situation where all NPPs are located on seaside, the radioactive waste transportation by ship is inevitable and the its risk assessment is very important for safety. Although various researches on radioactive waste transportation risk assessment is being actively conducted, research on domestic radioactive waste maritime transportation is insufficient. In this study, MARINRAD and KM-RAD were used to review on the radioactive waste transportation risk assessment. The result of reviewing shows that MARINRAD used SNF as transporting radioactive materials and ‘SAND87-7067 (1987)’ as nuclide database, whereas KMRAD used LILW and ‘IAEA Technical Report Series-422 (2004)’. To complement these limitations, we present an modernized integrated database by updating data and covering the radioactive materials from LILW to SNF. These results are expected to contribute to the development of RRACs for domestic radioactive waste maritime transportation.

This study focuses on the development of coatings designed for storage containers used in the management of radioactive waste. The primary objective is to enhance the shielding performance of these containers against either gamma or neutron radiation. Shielding against these types of radiation is essential to ensure the safety of personnel and the environment. In this study, tungsten and boron cabide coating specimens were manufactured using the HVOF (High-Velocity Oxy Fuel) technuqe. These coatings act as an additional layer of protection for the storage containers, effectively absorbing and attenuating gamma and neutron radiation. The fabricated tungsten and boron carbide coating specimens were evaluated using two different testing methods. The first experiment evaluates the effectiveness of a radiation shielding coating on cold-rolled steel surfaces, achieved by applying a mixture of WC (Tungsten Carbide) powders. WC-based coating specimens, featuring different ratios, were prepared and preliminarily assessed for their radiation shielding capabilities. In the gamma-ray shielding test, Cs-137 was utilized as the radiation source. The coating thickness remained constant at 250 μm. Based on the test results, the attenuation ratio and shielding rate for each coated specimen were calculated. It was observed that the gammaray shielding rate exhibited relatively higher shielding performance as the WC content increased. This observation aligns with our findings from the gamma-ray shielding test and underscores the potential benefits of increasing the tungsten content in the coating. In the second experiment, a neutron shielding material was created by applying a 100 μm-thick layer of B4C (Boron Carbide) onto 316SS. The thermal neutron (AmBe) shielding test results demonstrated an approximate shielding rate of 27%. The thermal neutron shielding rate was confirmed to exceed 99.9% in the 1.5 cm thick SiC+B4C bulk plate. This indicates a significant reduction in required volume. This study establishes that these coatings enhance the gamma-ray and neutron shielding effectiveness of storage containers designed for managing radioactive waste. In the future, we plan to conduct a comparative evaluation of the radiation shielding properties to optimize the coating conditions and ensure optimal shielding effectiveness.

Structural stability of a waste form can be provided by the waste form itself (steel components, etc.), by processing the waste to a stable form (solidification, etc.), or by emplacing the waste in a container or structure that provides stability (HICs or engineered structure, etc.). The waste or container should be resistant to degradation caused by radiation effects. In accordance with the requirements for the domestic waste acceptance criteria, irradiation testing of solidified waste forms containing spent resin should be conducted on specimens exposed to a dose of 1.0E+6 Gy and other material 1.0E+7 Gy. Expected cumulative dose over 300 years is about 1.770E+6 Gy for spent resin and 0.770E+6 Gy for dried concentrated waste generated from NPPs generally. According to NRC Waste Form Technical Position, to ensure that spent resins will not undergo adverse degradation effects from radiation, resins should not be generated having loadings that will produce greater than 1E+6 Gy total accumulated dose. If it necessary to load resins higher than 1E+6 Gy, it should be demonstrated that the resin will not undergo radiation degradation at the proposed higher loading. This is the recommended maximum activity level for organic resins based on evidence that while a measurable amount of damage to the resin will occur at 1E+6 Gy, the amount of damage will have negligible effect on disposal site safety. Cementitious materials are not affected by gamma radiation to in excess of 1E+6 Gy. Therefore, for cement-stabilized waste forms, irradiation qualification testing need not be conducted unless the waste forms contain spent resins or other organic media or the expected cumulative dose on waste forms containing other materials is greater than 1E+7 Gy. Testing should be performed on specimens exposed to IE+6 Gy or the expected maximum dose greater than 1E+6 Gy for waste forms that contain ion exchange resins or other organic media or the expected maximum dose greater than 1E+7 Gy for other waste forms. This is suggestion as a review result that requirement for irradiation testing of solidified waste forms has something to be revise in detail and definitively.

The high-level nuclear waste (HLW) repository disposes of high-level nuclear waste at a depth of 500 m to 1,000 m underground. Structural health monitoring must be accompanied by the complex environmental conditions of high temperature, high humidity, radiation, and mechanical stress. A thermocouple for measuring temperature, total stress meter and pore pressure meter for measuring stress and water pressure, relative hygrometer and electrical resistivity sensor (TDR or SUS) for measuring humidity, accelerometer for measuring crack signals, and strain gauge for measuring displacement are used. For safety, after disposing of HLW in the HLW repository, access to the disposal tunnel gets blocked, making it impossible to replace or remove the monitoring sensors. So, it is necessary to evaluate the effect of the HLW repository’s environmental conditions on the monitoring sensors and enhance their durability through quantitative life evaluation and shielding. Before evaluating the life of accelerometers and strain gauges used in the HLW repository, an experimental study is conducted to determine failure modes and failure mechanisms under radiation conditions, which are unique environmental conditions of the HLW repository.

In the nuclear environment, sensors ensure safety, monitoring, and operational efficiency under various operating conditions. These sensors come in various forms, each tailored to specific purposes, including nuclear safety and security, waste treatment and storage, gas leak detection, temperature and humidity monitoring, and corrosion detection. Ensuring the longevity of sensors without the need for frequent replacements is a vital goal for researchers in this field. This paper explores materials that can act as shields to protect sensors from harsh environmental conditions (high radiation and temperatures) to enhance their lifetime. The types of material that had been explored were divided into categories: metal and non-metal. Fourteen types of metal and seven different plastic materials were studied and focused on their characteristics and current applications. Considering properties like melting point, intensity, and conductivity, plastic materials are chosen to be examined as sensor shielding material. A preliminary experiment was conducted to verify signal characteristics changes by shielding material. Metal material and plastic material each were placed in the middle of the granite and the target sensor. The result showed that when metal is between the granite and the sensor, the density and impedance are higher in granite than in the metal. This leads to signal attenuation and a shift in resonance frequency, while plastic does not. Therefore, PPS (Polyphenylene sulfide) and PAI (Polyamide-imide) have lower density and impedance than granite while also possessing heat, moisture, and radiation resistance for effective shielding.

The objective of this study is development of graphite-boron composite material as a replacement for metal canisters to Improve the heat dissipation and radiation shielding performance of dry spent nuclear fuel storage system and reduce the volume of waste storage system. KEARI research team plan to use the graphite matrix manufacturing technology to pelletize the graphite matrix and adjust the content of phenolic resin binder to minimize pore formation. Specifically, we plan to adjust the ratio of natural and synthetic graphite powder and use uniaxial pressing technology to manufacture black graphite matrix with extremely high radial thermal conductivity. After optimizing the thermal conductivity of the graphite matrix, we plan to mix it with selected boron compounds, shape it, and perform sintering and purification heat treatments at high temperatures to manufacture standard composite materials.

After the Fukushima disaster, overseas nuclear power plants have established conditions for issuing a red alert in the event of fuel damage within the spent fuel pool and they have already implemented conditions for issuing a blue alert when fuel is exposed above the water surface. In South Korean nuclear power plants, a real-time monitoring system is in place to oversee the exposure of spent fuel to the surface within the spent fuel pool. To achieve this, a water level indicator gauge is installed within the spent fuel pool, allowing for continuous real-time monitoring. This paper conducted a comparative assessment of radiation levels from water level monitoring system in two units’ spent fuel pools based on the low water levels (1 feet from the storage rack), utilizing the radiation analysis code (MCNP).

Graphene oxide (GO) and ultrafine slag (UFS) have been applied to reinforce cement mortar cubes (CMC) in this research. The consequences of GO and UFS on the mechanical attributes of the CMC were explored through experimental investigations. Established on the results, at the 28 days of hydration, the CMC compressive and flexural strength with 0.03% of GO and 10% UFS were 89.8 N/mm2 and 9.1 N/mm2, respectively. Furthermore, the structural changes of CMC with GO and UFS were qualitatively analysed with instrumental techniques such as scanning electron microscope (SEM), X-ray fluorescence (XRF), thermogravimetric analysis (TGA), Fourier transform infrared spectroscopy (FT-IR), FT Raman spectroscopy, atomic force microscopy (AFM), and 27Al, 29Si-Nuclear magnetic resonance spectroscopy (NMR). SEM results reported that GO and UFS formed an aggregated nanostructure that improved the microstructural properties of the CMC. TGA analysis revealed the quantum of calcium hydrate and bound water accomplished by supplementing GO bound to the UFS aggregates. FT-IR analysis of the CMC samples confirmed the ‘O-’comprising functional groups of GO which expedited the formation of complexes between calcium carbonate ( CaCO3) and UFS. 0.03% GO was the optimum dosage that enhanced the compressive and flexural attributes when combined with 10% UFS in CMC.

To harvest marketable cucumbers, high quality seedlings must be used. Producing seedlings in the greenhouse during the low radiation period decreases marketability due to insufficient light for growth. Supplemental lighting with artificial light of different quality can be used to improve low light conditions and produce high quality seedlings. Therefore, this study was conducted to select the appropriate supplemental light sources on the growth and seedling quality of grafted cucumber seedlings during the low radiation period. Three cultivars of cucumber were used as scions for grafting; ‘NakWonSeongcheongjang’, ‘Sinsedae’, and ‘Goodmorning baekdadagi’. Figleaf gourd (Cucurbita ficifolia) ‘Heukjong’ was used as the rootstock. The seeds were sown on January 26, 2023, and grafted on February 9, 2023. After graft-taking, cucumbers in plug trays were treated with RB light-emitting diodes (LED, red and blue LED, red:blue = 8:2), W LED (white LED, R:G:B = 5:3:2), and HPS (high-pressure sodium lamp), respectively. Non-treatment was used as the control. Supplemental lighting was applied 2 hours before sunrise and 2 hours after sunset for 19 days. The stem diameter and fresh and dry weights of roots did not differ significantly by supplemental light sources. The plant height and hypocotyl length were decreased in W LED. However, the leaf length, leaf width, leaf area, and fresh and dry weights of shoots were the highest in the RB LED. Seedling qualities such as crop growth rate, net assimilation rate, and compactness were also increased in RB LED and W LED. After transplanting, most of the growth was not significant, but early yield of cucumber was higher in LED than non-treatment. In conclusion, using RB LED, W LED for supplemental light source during low radiation period in grafted cucumber seedlings improved growth, seedling quality, and early yield of cucumber.