Collagen peptides have garnered significant attention as functional foods across multiple fields due to their capacity to regulate physiological and hormonal processes, offering numerous advantages. However, despite their broad range of applications, comprehensive research on the potential toxicity of these substances remains lacking. Therefore, this study sought to assess the acute oral toxicity of a collagen peptide derived from skate (Raja kenojei) skin (CPSS) in both rats and dogs. In the rat model, CPSS was orally administered at doses of 300 and 2,000 mg/kg to Sprague-Dawley rats. An escalating single-dose oral toxicity assessment at doses of 500, 1,000, and 2,000 mg/kg was carried out in beagle dogs with 3-day intervals between doses. Throughout the 14-day post-administration assessment period, clinical signs, mortality rates, changes in body weight, and necropsy observations were closely monitored. After oral administration, no signs of toxicity associated with CPSS were observed in either rats or dogs. Therefore, the oral LD50 (approximate lethal dose for 50% mortality) for CPSS in rats was determined to exceed 5,000 mg/kg, and the maximum tolerated dose for dogs was estimated to be above 2,000 mg/kg. Consequently, this study offers safety data on the use of CPSS in functional foods and medicinal applications.

박물관, 기록원 등의 주요 소독약제인 메틸브로마이드(MB)와 Ethylene Oxide(EO)는 오존층 파괴 물질 및 1급 발암물질로 규제가 되고 있어 대체 훈증제 개발이 필요한 실정이다. 따라서, 친환경 약제인 검역훈증제 에틸포메이트 (베이퍼메이트®)의 적용가능성을 확인하고자 주요 해충인 독일바퀴(Blatella germanica) 및 흰개미(Reticulitermes speratus)를 대상으로 Dose response 실험을 실시하였다. 공시충은 팜한농 작물보호연구센터 곤충사육실에서 누대사육중인 개체를 사용하였으며, 데시게이터(6.9L)에 농도별로 24시간동안 밀폐 훈증처리하여 약효를 조사하였다. 통계분석은 Probit analysis 통해 L(Ct) (Lethal Concentration x Time, gꞏhꞏm3)값을 산출하였다. 시험결과 L(Ct)50 및 L(Ct)99 값은 독일바퀴(혼합태)의 경우 48.38 및 346.34 gꞏhꞏm3 흰개미(성충)은 14.91및 660.94 gꞏhꞏm3로 나타났다. 모두 방제가 가능한 L(Ct)99 값은 660.94 gꞏhꞏm3으로 이는 에틸포메이 트 28.2 g/m3(베이퍼메이트® 170 g/m3)를 24시간 처리시 완전 방제가 가능할 것으로 판단된다.

본 연구는 인천광역시내 유통되고 있는 기타가공품에 대하여 중금속 (납, 카드뮴, 수은), 부정유해물질 (발기부전치료제와 그 유사물질 36종, 비만치료제 3종, 스테로이드 28종)을 조사하여 안전성을 평가하고, 기타가공품에 대한 기준규격 제정 시 기초자료로서 제공하고자 실시하였다. 그결과 부정유해물질은 검출되지 않았고, 중금속은 납(Pb) 0.001~13.39(mg/kg), 카드뮴(Cd) 0.003~1.231(mg/kg), 수은(Hg)은 0.001~0.650(mg/kg)의 결과를 얻었다. 기타가 공품은 중금속에 대한 규격 기준이 없으므로, 중금속의 국내·외 기준을 비교하였을 때 중금속 최고기준을 적용하 면 납(Pb)과 수은(Hg)에서 각각 2건씩 4건이 초과되는 것으로 확인되었다. 검사결과를 바탕으로 한 중금속의 위해성 평가는 FAO/WHO (Codex)의 잠정주간섭취허용량(PTWI)과 비교결과, 해물파전믹스, 들깨가루는 Pb의 PTWI의 기준 0.214를 초과하였다. 결론적으로 분말류나 환제품을 유사식품 기준으로 적용 하면 부적합으로 판정 될 수 있기 때문에 규격검사를 피할 목적으로 기타가공품으로의 등록을 지향한다. 이를 악용할 경우 검사의 사각지대로 인해 시민의 건강이 위협 받을 수 있다. 따라서 품목제조허가 신고시 규격외 일반가공식품에서 기타가공품으로의 등록을 지양하고, 기타가공품의 세부적인 검사규격을 제정하여 관리해야 함을 제언한다.

Aphids are common pests frequently found in imported and exported fruits and vegetables. Methyl bromide(MB), a Quarantine and Pre Shipment(QPS) fumigant, could offer eradication of aphids within short period. However, MB is limited in use because of poor gas evaporation at low temperature(<5℃) and there is phytotoxic effect or damage on quality in post-harvest vegetables and fruits even at >5℃. Two candidates of MB alternative, ethyl formate(EF) and phosphine(PH3), are used and being investig at edonvarious fruits and vegetables fumigation to replace MB. Aphids are known as quarantine pest that are hard to control when conduct short period fumigation with PH3 and low dosage of EF. In this paper, dose response assessment of EF and PH3 are presented for three different aphid pecies : cotton aphid(Aphisgossypii), green peach aphid(Myzuspersicae) and turnip aphid (Lipaphiserysimi). The LCt99% values of EF at room temp. and low temp. (5℃) were 4.42 and 4.45 g·h·m-3 for cotton aphid, 3.23 and 5.58 g·h·m-3 for turnip aphid, 3.23 and 5.58 g·h·m-3 for green peach aphid when 2-hours fumigation. PH3 showed 0% efficacy on all species when 2-hours fumigation.

Manchester system 타입의 장착기중 상, 하부에 차폐체가 장착되어 있는 Henschke 장착기를 이용하여 자궁암 근접치료시 자궁 및 주변장기의 선량분포를 평가하기 위하여 치료계획수립에 사용되는 실용프로그램 결과와 몬테칼로 모의계산 결과를 비교하였다. 또한 자궁 및 주변 정상조직이 받은 선량을 계산하기 위해 ORNL(Oak Ridge National Laboratory)에서 수립한 여성의 MIRD (Medical Internal Radiati

미생물 위해성 평가의 용량-반응 모델은 생물학적 모델과 경험적 모델로 나눌 수 있다. 생물학적 모델은 미생물의 분포형태, 미생물에 대한 숙주의 감수성, 감염을 일으킬 수 있는 미생물 수에 대한 가정을 바탕으로 성립된 모델로서, 대표적으로 Exponential model과 β-Poisson model이 있다. 경험적 모델은 주로 화학물질의 독성을 나타내는데 이용되어 온 모델로, Weibull-Gamma model 등이 있다. 여러 용량-반응 모델 중에서 실험 데이터에 적합한 모델을 선정하는 데에는 deviance function(Y)을 이용하며, 현재 일부 식중독균에 대해서는 사람과 실험동물에서의 용량-반응 모델이 연구되어 있다.

This study was designed to predict the risk of a hazard chemical, vinyl chloride, by applying dose-response assessment that are one of the major process in practicing risk assessment. After extrapolating from the high dose exposure of vinyl chloride based upon animal carcinogenic data to the low dose exposed to human using several mathematical models, we calculated the cancer potency factors as well as virtually safe dose and the resulted values were compared. This process will provide the new insight to assess the risk of a chemical accurately imposed to human in the future.

The Korea Atomic Energy Research Institute (KAERI) has facilities that are operated for the purpose of treating radioactive wastes and storing drums before sending them to a disposal site. Domestic regulations related to nuclear facility require radiological dose assessment resulting from release of gaseous radioactive effluent of nuclear facilities. In this study, ICRP-60-based dose conversion factors were applied to evaluate the radiation dose to residents in the event of operation and accident for the radioactive waste management facilities in KAERI. The radioactive gaseous effluent generated from each facility diffuse outside the exclusion area boundary (EAB), causing radiation exposure to residents. To evaluate the external exposure dose, the exposure pathways of cloudshine and radioactive contaminated soil were analyzed. The internal exposure dose was estimated by considering the exposure from respiration and ingestion of agricultural and livestock products. The maximum individual exposure dose was evaluated to be 1.71% compared to the dose limit. The assumed situation used for accidental scenarios are as follows; A fire inside the facility and falling of radioactive waste drum. It was a fire accident that caused the maximum exposure dose to individual and population living within an 80 km radius of the site. At the outer boundary of the low population zone (LPZ), the maximum effective dose and thyroid equivalent dose were estimated as 8.92 E-06% and 5.29 E-06%, respectively, compared to the dose limit. As a result of evaluating the radiological exposure dose from gaseous emissions, the radioactive waste treatment facilities and its supplementary facilities meet the regulations related to nuclear facility, and are operated safely in terms of radiological environmental impact assessment.

To construct and operate nuclear power plants (NPPs), it is mandatory to submit a radiation environmental impact assessment report in accordance with Article 10 and Article 20 of the Nuclear Safety Act. Additionally, in compliance with Article 136 of the Enforcement Regulations of the same law, KHNP (Korea Hydro & Nuclear Power) annually assesses radiation environmental effects and publishes the results for operating NPPs. Furthermore, since the legalization of emission plans submission in 2015, KHNP has been submitting emission plans for individual NPPs, starting with the Shin-Hanul 1 and 2 units in 2018. These emission plans specify the emission quantities that meet the dose criteria specified by the Nuclear Safety and Security Commission. Before 2002, KHNP used programs developed in the United States, such as GASPAR and LADTAP, for nearby radiation environmental impact assessments. Since then, KHNP has been using K-DOSE60, developed internally. K-DOSE60 incorporates environmental transport analysis models in line with U.S. regulatory guidance Regulatory Guide 1.109 and dose assessment models reflecting ICRP-60 recommendations. K-DOSE60 is a stand-alone program installed on individual user PCs, making it difficult to manage comprehensively when program revisions are needed. Additionally, during the preparation of emission plans and the licensing phase, improvements to KDOSE60’ s dose assessment methodology were identified. Furthermore, in 2022, regulatory guidelines regarding resident dose assessments were revised, leading to additional improvement requirements. Currently, E-DOSE60, being developed by KHNP, is a network-based program allowing for integrated configuration management within the KHNP network. E-DOSE60 is expected to be developed while incorporating the identified improvements from K-DOSE60, in response to emission plan licensing and regulatory guideline revisions. Key improvements include revisions to dose assessment methodologies for H-13 and C-14 following IAEA TRS-472, expansion of dose assessment points, and changes in socio-environmental factors. Furthermore, data such as site meteorological information and releases of radioactive substances in liquid and gaseous forms can be linked through a network, reducing the potential for human errors caused by manual data entry. Ultimately, E-DOSE60 is expected to optimize resident exposure dose assessment and enhance public trust in NPP operation.

For the release of the nuclear power plant site after the decommissioning, a reliable exposure dose assessment considering the environmental impact of residual radionuclides is essentially required. In this study, the Derived Concentration Guideline Level (DCGL) for the hypothetically contaminated surface soil at the Wolsong nuclear power plant (NPP) unit 1 site was preliminarily calculated by using the RESRAD-OFFSITE computational code and compared with the other case studies. Moreover, radiation exposure dose for local residents and relevant exposure pathways were quantitatively analyzed based on the calculation model established through this work. For the target site modeling, the source term was determined by referring to the previous case studies regarding the nuclear power plant decommissioning, quantification analysis data of pressure tubes of Wolsong NPP unit 1, and radionuclide data estimated by using the MCNP/ORIGEN-2 code. In total, 14 different radioisotopes such as Ag-108m, C-14, Co-60, Cs-134/137, Fe-55, H-3, Nb-93m/94, Ni-63, Sb-125, Sn-121m, Sr-90, and Zr-93 were considered as target radionuclides. In addition, the geological structure model of the Wolsong NPP site was established based on the final safety analysis report of Wolsong NPP unit 1. The distribution coefficients (Kd) were taken from the JAEA-SDB to estimate the migration/retardation behavior of various radionuclides under the groundwater condition of the Wolsong NPP site. In the present work, the DCGL values were calculated according to the site release criterion of 0.1 mSv/yr, which indicates the radiation protection standard for the site release. Moreover, the exposure pathway and sensitivity analyses were conducted to assess the sensitive input parameters remarkably influencing the calculation result. For the evaluation of exposure dose for local residents, a site layout centered around Wolsong NPP unit 4, located in the closest proximity to the residents’ habitation area, was alternatively established and all potential exposure pathways were considered as a comprehensive resident farmer scenario. The results obtained from this study are expected to serve as a preliminary case study for the DCGL values regarding the surface soil at the Wolsong NPP unit 1 site and for evaluating the radiation exposure dose to local residents resulting from the residual radioactivity at the site after the decommissioning.

The decommissioning of Korea Research Reactor Units 1 and 2 (KRR 1&2), the first research reactors in South Korea, began in 1997 and the decommissioning status is currently proceeding with phase 3. It is expected that more than 5,000 tons of dismantled wastes will be generated as the contaminated building is demolished. Since these dismantled wastes must be disposed of in an efficient method considering economic feasibility, it is desirable to clearance extremely low-level wastes whose contamination is so minimal that the radiological risk is negligible. In Korea, in order to approve the clearance of radioactive waste, it must be proven that the nuclide concentration standards are met or that the dose to individuals and collectives is below the allowable dose value. At the KRR 1&2 decommissioning site, dismantled wastes have been steadily being disposed of through clearance procedure since 2021. Clearance was approved by the Korean Institute of Nuclear Safety (KINS) for one case of concrete waste in 2021 and two cases of metal waste in 2022. In 2023, the clearance of metal waste and asbestos waste has been approved so far, and in particular, this is the first case in Korea for asbestos waste. In this study, we compared the dose assessment methods and results of clearance wastes at the KRR 1&2 decommissioning site from 2021 to present. Dose assessment was conducted by applying the landfill scenario for concrete and asbestos and the recycling scenario for metal waste. The calculation codes used were RESRAD-onsite 7.2 and RESRAD-recycle 3.10. The dose conversion factors (DCF) for each age group (infant, 1y, 5y, 10y, 15y, adult) of the target nuclide used the values presented in ICRP-72, and in particular, geo-hydrological data of the actual landfill site was used as an input factor when evaluating landfill scenarios. As a result of the dose assessment, when landfilling concrete wastes in 2020, the personal dose and collective dose were evaluated the most at 2.80E+00 μSv/y and 4.83E-02 man·Sv/y, respectively.

The Korea Atomic Energy Research Institute (KAERI) is currently developing a process-based performance assessment model known as APro. Distinguished from the previous system-level safety assessment model developed by KAERI, APro exhibits the capacity to encompass a threedimensional biosphere domain, evolving over the long term. In this study, we elucidate the methodology employed in developing the dose assessment module of APro and present the module’s functionalities. The procedural steps underlying radiation dose calculations within the APro framework can be succinctly outlined as follows: 1) Definition of a landscape model, utilizing information derived from a specified snapshot period provided by the APro biosphere transport module; 2) Generation of unit biotope objects spanning the landscape; 3) Evaluation of radionuclide transfer within the soil medium; 4) Calculation of activity concentration for flora and fauna groups; 5) Assessment of the distribution of effective dose among representative human groups; 6) Progressing through successive time steps. The APro dose calculation module exhibits notable capabilities that encompass: 1) Accounting for radionuclide decay and ingrowth; 2) Facilitating transfer through unsaturated porous media; 3) Considering sorption effects; 4) Addressing the inheritance of radioactivity between various landscape models; 5) Offering customizable ecosystem parameters; 6) Providing flexibility for user-defined exposure pathways. Leveraging these functionalities of the dose assessment module, APro is proficient in evaluating the distribution of radiological doses and associated risks for representative population groups, all while accounting for the dynamic, long-term evolution of the biosphere, including alterations in land cover.

In this research, the dose rate was measured using a backpack-type scan survey device at 4 sites in sites around Nuclear Power Plants (Kori, Wolsong, Hanbit, Hanul), and the radioactivity ratio for each nuclide was evaluated using an high-purity germanium (HPGe) detector. Kori, Wolsong and Hanul power plants were measured within 2 km of the power plant, and Hanbit power plants were measured about 6.7 km from the power plant. As a result of measuring the dose rate with a backpacktype scan survey device, the average dose rate was the lowest in the measurement site 1 at 0.090 μSv/h, and the highest in the measurement site 4 at 0.145 μSv/h. All measurement points showed the domestic environmental dose rate level. The data obtained by the scan survey was visualized using the classed post and gridding functions of the surfer program. As a result of measurement with the HPGe detector, 137Cs was not detected, and only natural nuclides were detected. Among the detected natural nuclides, the radioactivity ratio was the highest for 40K with an average of 94.56%, and the lowest for 214Pb with an average of 0.26%. The results of this research can be used as basic data for radiation environment surveys around nuclear power plants. Further studies are needed to evaluate the radiation impacts by region and environment through periodic measurements.

KHNP is carrying out international technical cooperation and joint research projects to decommission Wolsong unit 1 reactor. Construction data of the reactor structures, experience data on the pressure tube replacement projects, and the operation history were reviewed, and the amount of dismantled waste was calculated and waste was classified through activation analysis. By reviewing COG (CANDU owners Group) technical cooperation and experience in refurbishment projects, KHNP’s unique Wolsong unit 1 reactor decommissioning process was established, and basic design of a number of decommissioning equipment was carried out. Based on this, a study is being conducted to estimate the worker dose of dismantling workers. In order to evaluate the dose of external exposure of dismantling workers, detailed preparation and dismantling processes and radiation field evaluation of activated structures are required. The preparation process can be divided into dismantlement of existing facilities that interfere with the reactor dismantling work and construction of various facilities for the dismantlement process. Through process details, the work time, manpower, and location required for each process will be calculated. Radiation field evaluation takes into account changes in the shape of structures by process and calculates millions of areas by process, so integrated scripts are developed and utilized to integrate input text data. If the radiation field evaluation confirms that the radiation risk of workers is high, mutual feedback will be exchanged so that the process can be improved, such as the installation of temporary shields. The results of this study will be used as basic data for the final decommissioning plan for Wolsong unit 1. By reasonably estimating the dose of workers through computer analysis, safety will be the top priority when decommissioning.

RUCAS (Recycling-Underlying Computational Dose Assessment System), a dose assessment program based on the RESRAD-RECYCLE framework, is designed to evaluate dose for recycling scenarios of radioactive waste in metals and concrete. To confirm the validity of the recycling scenarios provided by RUCAS, comparative evaluations will be conducted with RESRAD-RECYCLE for metal radioactive waste recycling scenarios and with MicroShield® for concrete radioactive waste recycling scenarios. In the evaluation of metal recycling scenarios without shielding, RUCAS showed similar results when compared to both MicroShield® and RESRAD-RECYCLE. This validates the function of dose assessments using RUCAS for metal recycling scenarios. However, when shielding was present, RUCAS produced results that were comparable to MicroShield®, but differed from those of RESRAD-RECYCLE. The underestimation of dose values up to 1.66E+08 times difference by RESRAD-RECYCLE could potentially decrease reliability and safety in evaluated doses, further emphasizing the importance of RUCAS. Because validation is also necessary for the expanded calculation capabilities resulting from methodological changes of RUCAS (i.e., various radiation source geometries), based on prior validations, it was determined that additional validations are required for different radiation source materials and shielding conditions. In case where the radiation source and shielding materials were identical, RUCAS and MicroShield® produced similar results according to both the Kalos et al. (1974) and Lin and Jiang (1996) methodologies. This demonstrates that the that differences in methodology are inconsequential when considering the same source and shielding materials. However, when the atomic number of the radiation source materials was larger than that of shielding material (HZ-LZ condition), RUCAS obtained results similar to MicroShield® only for the Kalos et al. (1974) methodology. While Lin and Jiang (1996) methodology yield higher results than MicroShield®. Lastly, in case where the atomic number of the radiation source material was smaller than that of the shielding material (LZ-HZ condition,) both methodologies yielded results comparable to MicroShield®. In conclusion, the validity of RUCAS’s shielding calculations has been verified, confirming improvements in dose assessment compared to RESRAD-RECYCLE. Additionally, we observed that shielding effectiveness calculations differ depending on the methodology of build-up effect. If the validity of these methodologies is confirmed, it is expected that selecting the most advantageous methodology for each condition will enable more rational dose assessments. Consequently, in future research, we plan to evaluate the validity of Lin and Jiang (1996) methodology using particle transport codes based on the Monte Carlo method, such as MCNP and Geant 4, rather than MicroShield®.