• Nuclear Engineering and Technology
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• v.52 no.11
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• pp.2511-2521
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• 2020

The Fukushima accident is characterized by the fact that three reactors at the same site experienced reactor vessel failure and the accident resulted in significant radiological release to the environment, which was about 1/10 of the Chernobyl releases. The safe removal of fuel debris in the reactor vessel and Primary Containment Vessel (PCV) and treatment of huge amount of contaminated water are the major issues for the decommissioning in coming decades. Discussions on the new researches efforts being carried out in the area of investigation of the end state of fuel debris and Boling Water reactor (BWR) specific core melt progression, development of technologies for the mitigation of radiological releases to comply with the strengthened safety requirement set after the Fukushima accident are discussed.

• Journal of Soil and Groundwater Environment
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• v.23 no.4
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• pp.1-15
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• 2018

Despite of safety issues related to radiological hazards, 31 countries around the world are operating more than 450 nuclear power plants (NPPs). To operate NPPs safely, safety regulations from radiation protection organizations were developed and adopted in many countries. However, many cases of radionuclide releases at foreign NPPs have been reported. Almost all commercial NPPs routinely release radioactive materials to the surrounding environments as liquid and gas phases under control. These releases are called 'planned releases' which are planned, regularly monitored, and well documented. Meanwhile, the releases focused in this review, called 'unplanned releases', are neither planned nor monitored by regulatory and/or protection organizations. NPPs are generally composed of various structures, systems and components (SSCs) for safety. Among them, the SSCs near reactors are closely related to safety of NPPs, and typically fabricated to comply with stringent requirements. However, some non-safety related SSCs such as underground pipes may be constructed only according to commercial standards, causing the leakage of radioactive fluids usually containing tritium ($^3H$). This paper discusses SSCs of NPPs and introduces several cases of unplanned releases at foreign NPPs. The current regulation on the environmental radiological surveillance and assessment around the NPPs in South Korea are also examined.

• Journal of Radiation Protection and Research
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• v.28 no.3
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• pp.247-254
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• 2003

To prevent the potential health detriment to the public from radioactive effluents, radiological dose assessments due to the operation of nuclear facilities located at Korea Atomic Energy Research Institute (KAERI) site has been performed semiannually in compliance with the Minister of Science and Technology (MOST)'s Notice in Korea. Radiological dose assessment based on the new recommendation of the International Committee on Radiation Protection (ICRP-60) has been conducted since 1998. In this manuscript, a serial activities at KAERI site to meet the regulatory standards for routine releases of radioactive effluents are introduced and discussed including technical approaches. It is clear that each nuclear facility has been operated in compliance with regulatory standards. Furthermore, it is identified that the radiation induced health effects for residents around the site are neglectable.

• Journal of Radiation Protection and Research
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• v.9 no.2
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• pp.118-128
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• 1984

The analysis of radiological protection at Kori Nuclear Unit #1, which started commercial operation on April 29, 1978, has been carried out in three areas, namely, radiation exposure, radioactive waste disposal and environmental management. The analysis results for radiation exposure and radioactive effluent releases appear to increase gradually with plant operating years. On the other hand, the apex of radiation exposure and radioactive effluent releases was in 1983 and 1982 respectively and then decreased or stabilized. Also, the operation environmental radioactivity seems to be no higher than preoperation environmental radioactivity.

• Nuclear Engineering and Technology
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• v.55 no.2
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• pp.696-706
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• 2023

Consequences of an anticipated Beyond Design Basis Accident (BDBA) Long-Term Station Blackout (LTSBO) event with complete loss of grid power in the VVER-1200 reactor of Rooppur Nuclear Power Plant (NPP) of Unit-1 are assessed using the RASCAL 4.3 code. This study estimated the released radionuclides, received public radiological dose, and ground surface concentration considering 3 accident scenarios of International Nuclear and Radiological Event Scale (INES) level 7 and two meteorological conditions. Atmospheric transport, dispersion, and deposition processes of released radionuclides are simulated using a straight-line trajectory Gaussian plume model for short distances and a Gaussian puff model for long distances. Total Effective Dose Equivalent (TEDE) to the public within 40 km and radionuclides contribution for three-dose pathways of inhalation, cloudshine, and groundshine owing to airborne releases are evaluated considering with and without passive safety Emergency Core Cooling System (ECCS) in dry (winter) and wet (monsoon) seasons. Source term and their release rates are varied with the functional duration of passive safety ECCS. In three accident scenarios, the TEDE of 10 mSv and above are confined to 8 km and 2 km for the wet and dry seasons, respectively in the downwind direction. The groundshine dose is the most dominating in the wet season while the inhalation dose is in the dry season. Total received doses and surface concentration in the wet season near the plant are higher than those in the dry season due to the deposition effect of rain on the radioactive substances.

• Journal of Radiation Protection and Research
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• v.44 no.4
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• pp.161-165
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• 2019

Background: Not only regulatory framework including radiation protection quantities and regulatory standards, but also methodology for regulatory compliance may be different in each country due to inherent philosophy for radiation protection. Materials and Methods: Based on the Korean regulatory models, off-site radiological dose resulting from the routine releases of gaseous radioactive effluents was calculated by applying the parameter values and assumptions recommended in the Korean and Japanese regulations. Results and Discussion: Effective dose for adult based on the Korean recommendation were 17.5 and 1.6 times higher than those of Japanese recommendation for ¹³¹I and ¹³³I, respectively, for the same atmosphere dispersion and ground deposition factors. Conclusion: It was due to different parameter values and assumptions recommended for the purpose of evaluating compliance with dose criteria for the radiation protection of the public in each country.

• Biomedical Science Letters
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• v.27 no.3
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• pp.170-176
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• 2021

Thapsigargin (TG), a sarco/endoplasmic reticulum (ER) Ca²⁺-ATPase (SERCA) inhibitor, has been widely used as an agonist for platelet aggregation for decades. In this study, we investigated the effect of TG on the release of lactate dehydrogenase (LDH) for platelets and elucidated its mechanism. Platelet LDH release and platelet aggregation were increased by TG treatment; 1,000 nM of TG induced the complete lysis of platelets. Other agonists such as collagen (2.5 ㎍/mL), thrombin (0.1 U/mL), and ADP (10 mM) did not induce significant platelet LDH release despite platelet aggregation. Finally, we investigated the effects of pharmacological inhibitors on TG-induced platelet aggregation and LDH release. SP600125, a JNK inhibitor, and LY294002, a PI-3K inhibitor, inhibited TG-induced platelet LDH release but not platelet aggregation. Forskolin, an adenylyl cyclase activator, also inhibited LDH release without affecting platelet aggregation by TG. These results suggest that the TG-induced platelet aggregation was accompanied by LDH release but regulated by a different signaling pathway.

• Journal of Radiation Protection and Research
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• v.29 no.1
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• pp.41-48
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• 2004

A safety assessment code, called SAGE (Safety Assessment Groundwater Evaluation), has been developed to describe post-closure radionuclide releases and potential radiological doses for low- and intermediate-level radioactive waste (LILW) disposal in an engineered vault facility in Korea. The conceptual model implemented in the code is focused on the release of radionuclide from a gradually degrading engineered barrier system to an underlying unsaturated zone, thence to a saturated groundwater zone. The radionuclide transport equations are solved by spatially discretizing the disposal system into a series of compartments. Mass transfer between compartments is by diffusion/dispersion and advection. In all compartments, radionuclides ate decayed either as a single-member chain or as multi-member chains. The biosphere is represented as a set of steady-state, radionuclide-specific pathway dose conversion factors that are multiplied by the appropriate release rate from the far field for each pathway. The code has the capability to treat input parameters either deterministically or probabilistically. Parameter input is achieved through a user-friendly Graphical User Interface. An application is presented, which is compared against safety assessment results from the other computer codes, to benchmark the reliability of system-level conceptual modeling of the code.

• Nuclear Engineering and Technology
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• v.42 no.1
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• pp.1-8
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• 2010

Geological disposal is designed to provide safe containment of radioactive waste for very long times, with the containment provided by a combination of engineered and geological barriers. In the extreme long term, after many thousands of years or longer, residual amounts of long-lived radionulides such as Cl-36, but also radionuclides in the natural decay chains, may be released into the environment normally accessed and used by humans, termed here, the biosphere. It is necessary to ensure that any such releases meet radiation protection objectives through the development of a safety case, which will include assessment of radiation doses to humans. The design of such dose calculations over such long timeframes is not straightforward, because of the range of potentially relevant assumptions which could be made, concerning environmental change and changes in human behavior. These conceptual uncertainties are additional to those that more typically arise, for example, in the assessment of present day situations, but which also have to be addressed. The issue has therefore been subject to international cooperation for many years. This paper summarizes the evolution and results of that collaboration leading up to the present day, taking account of developments in international recommendations on radiation protection objectives and the more recent greater focus on preparation of site specific safety cases.

• Journal of Radiation Protection and Research
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• v.23 no.2
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• pp.109-114
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• 1998

Radiological safety for the conventional operation of Demonstration-Scale Incineration Plant (DSIP) was assessed on the basis of the results of trial burns using the simulated and real radioactive wastes. Radiation dose assessments for routine releases on an annual basis as well a several severe accidental releases on a short-term basis (2h) revealed that there would be no significant environmental impact when low-level waste Is incinerated in DSIP. For semivolatile radioactive cesium species, expected emission concentrations slightly exceeded 10% of maximum permissible concentration. Removal characteristics of the bag filter for condensed-phase cesium species was investigated by the trial burns of simulated waste with inactive cesium tracer. In the off-gas before passing through bag filter, distributions of condensed cesium species in the transition size ranging between the diffusional and inertial region are less than 5%. The overall collection efficiency of the bag filter for cesium species was higher than 99.9%, showing enough decontamination capability as a primary filter for the low-temperature dry off-gas system in radwaste incineration plant.