• Progress in Medical Physics
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• v.33 no.4
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• pp.53-62
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• 2022

In this study, we have investigated the shielding evaluation methodology for facilities using kV energy generators. We have collected and analysis of safety evaluation criteria and methodology for overseas facilities using radiation generators. And we investigated the current status of shielding evaluation of domestic industrial radiation generators. According to the statistical data from the Radiation Safety Information System, as of 2022, a total of 7,679 organizations are using radiation generating devices. Among them, 6,299 facilities use these devices for industrial purposes, which accounts for a considerable portion of radiation. The organizations that use these devices evaluate whether the exposure dose for workers and frequent visitors is suitable as per the limit regulated by the Nuclear Safety Act. Moreover, during this process, the safety shields are evaluated at the facilities that use the radiation generating devices. However, the facilities that use radiating devices having energy less than or equal to 6 MV for industrial purposes are still mostly evaluated and analyzed according to the National Council on Radiation Protection and Measurements 49 (NCRP 49) report published in 1976. We have investigated the technical standards of safety management, including the maximum permissible dose and parameters assessment criteria for facilities using radiation generating devices, based on the NCRP 49 and the American National Standards Institute/Health Physics Society N.43.3 reports, which are the representative reports related to radiation shielding management cases overseas.

• Journal of the Korean Society of Safety
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• v.37 no.2
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• pp.35-42
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• 2022

When an accident occurs, the associated human activity is typically regarded as a "human error," or a temporal deviation. On the other hand, if the accident results in a serious loss or if it evokes a social issue, the person determined to be responsible may be punished with a "violation" of related laws or regulations. However, as Heinrich stated, it is neither appropriate nor reasonable in terms of probability theory and cognitive science to distinguish whether it is a "human error" or a "violation" with a criterion of resultant accident severity. Nonetheless, some in society get on the social climate to strengthen regulations on workers who have caused accidents, especially violations. This response can present a social issue due to the lack of systematic judgment procedure which distinguishes violations from human errors. The purpose of this study was to develop an objective and systematic procedure to assess whether workers' activities which induced industrial accidents should be categorized as violations rather than human errors. Various analysis techniques for the determination of violation procedure were investigated and compared using an analysis approach method. An appropriate technique was not found, however, for judging the culpability of intentional violations. As an alternative, this study developed the process of creating violations, based on cognitive procedure, as well as the criteria to determine and categorize an activity as a violation. In addition, the developed procedure was applied to cases of industrial accidents and nuclear power plant issues to test its practical applicability. The study demonstrated that the proposed model could be used to determine the existence of a violation even in the case of multiple workers who work simultaneously.

• Journal of Radiation Industry
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• v.17 no.2
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• pp.161-172
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• 2023

Coal-fired power plants handle large quantities of coal, one of the most prominent NORM, and the coal ash produced after the coal is burned can be tens of times more radioactive than the coal. Workers in these industries may be exposed to internal exposure by inhalation of particles while handling NORM. This study evaluated the size, concentration, particle shape and density, and radioactivity concentrations of airborne suspended particles in the main processes of a coal-fired power plant. Finally, the internal radiation dose to workers from particle inhalation was evaluated. For this purpose, airborne particles were collected by size using a multi-stage particle collector to determine the size, shape, and concentration of particles. Samples of coal and coal ash were collected to measure the density and radioactivity of particles. The dose conversion factor and annual radionuclide inhalation amount were derived based on the characteristics of the particles. Finally, the internal radiation dose due to particle inhalation was evaluated. Overall, the internal radiation dose to workers in the main processes of coalfired power plants A and B ranged from 1.47×10^-5~1.12×10^-3 mSv y^-1. Due to the effect of dust generated during loading operations, the internal radiation dose of fly ash loading processes in both coal-fired power plants A and B was higher than that of other processes. In the case of workers in the coal storage yard at power plants A and B, the characteristic values such as particle size, airborne concentration, and working time were the same, but due to the difference in radioactivity concentration and density depending on the origin of the coal, the internal radiation dose by origin was different, and the highest was found when inhaling coal imported from Australia among the five origins. In addition, the main nuclide contributing the most to the internal radiation dose from the main processes in the coal-fired power plants was thorium due to differences in dose conversion factors. However, considering the external radiation dose of workers in coal-fired power plants presented in overseas research cases, the annual effective dose of workers in the main processes of power plants A and B does not exceed 1mSv y^-1, which is the dose limit for the general public notified by the Nuclear Safety Act. The results of this study can be utilized to identify the internal exposure levels of workers in domestic coal-fired power plants and will contribute to the establishment of a data base for a differential safety management system for NORM-handling industries in the future.

• Journal of the Korean Society of Systems Engineering
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• v.14 no.1
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• pp.63-71
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• 2018

This paper discusses decommissioning procedure requirements management using requirement engineering to systematically manage the technical requirements and criteria that are required in decontamination and decommissioning activities, and the regulatory requirements that should be complied with in a decommissioning strategy for research reactors and nuclear power plants. A schema was designed to establish the traceability and change management related to the linkage between the regulatory requirements and technical criteria after classifying the procedures into four groups during the full life-cycle of the decommissioning. The results confirmed that the designed schema was successfully traced in accordance with the regulatory requirements and technical criteria required by various fields in terms of decontamination and decommissioning activities. In addition, the changes before and after the revision of the Nuclear Safety Act were also determined. The dismantling procedure requirement management system secured through this study is expected to be a useful tool in the integrated management of radioactive waste, as well as in the dismantling of research reactor and nuclear facilities.

• Corrosion Science and Technology
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• v.13 no.5
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• pp.178-185
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• 2014

Alloy 617 is considered as a candidate Ni-based superalloy for the intermediate heat exchanger (IHX) of a very high-temperature gas reactor (VHTR) because of its good creep strength and corrosion resistance at high temperatures. Helium is used as a coolant in a VHTR owing to its high thermal conductivity, inertness, and low neutron absorption. However, helium inevitably includes impurities that create an imbalance in the surface reactivity at the interface of the coolant and the exposed materials. As the Alloy 617 has been exposed to high temperatures at $950^{\circ}C$ in the impure helium environment of a VHTR, the degradation of material is accelerated and mechanical properties decreased. The high-temperature strength, creep, and corrosion properties of the structural material for an IHX are highly important to maintain the integrity in a harsh environment for a 60 year period. Therefore, an alloy superior to alloy 617 should be developed. In this study, the mechanical and high-temperature corrosion properties for Ni-Cr alloys fabricated in the laboratory were evaluated as a function of the grain boundary strengthening and alloying elements. The ductility increased and decreased by increasing the amount of Mo and Cr, respectively. Surface oxide was detached during the corrosion test, when Al was not added to alloy. However the alloy with Al showed improved oxide adhesive property without significant degradation and mechanical property. Aluminum seems to act as an anti-corrosive role in the Ni-based alloy.

• Journal of Radiation Protection and Research
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• v.39 no.4
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• pp.218-221
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• 2014

Nuclear Safety Act had described the three types of licenses on radioisotope handling, such as a general license, a special license and a supervisory license. Applicants should be qualified by careers and qualifications for the education and training to acquire the licenses. In particular, the information on the estimation for the career is notified by Nuclear Safety and Security Commission(NSSC). In this paper, we suggest an improvement by analyzing survey data at the end of the education course on a license for the supervisor of radiation handling. We applied the learning evaluation to improve the education course. The level of satisfaction with the improved curriculum was compared with the existing curriculum. The improved curriculum with the learning evaluation has shown high grades of performance, i.e. above 4.0 points (full mark: 5.0 points) on the level of satisfaction and field application. The learning evaluation should be applied to the basic education course on a general license for radioisotope handling.

• Journal of the Korean Society for Nondestructive Testing
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• v.30 no.1
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• pp.1-5
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• 2010

Acoustic emission(AE) has emerged as a powerful nondestructive tool to detect or monitor preexisting defects and leaks in the vessel structures. A Bragg grating based acoustic emission sensor system is developed. Various type of fiber Bragg grating sensor including the variable length of sensing part was fabricated and prototype sensor system was tested by using PZT pulser and pencil lead break sources. Two types of sensor attachment were used. First, the fiber Bragg grating sensor was attached fully to the surface using bonding agent. Second one is that one part of fiber was attached to the surface partly by bonding and the other part of fiber will be act as a cantilever. That is, the resonant frequency of the fiber Bragg grating sensor will depend on the length of sensing part. The final goal of the sensor system is to provide on-line monitoring of cracks or leaks in reactor vessel head penetration of nuclear power plants.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.3
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• pp.217-233
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• 2014

Securing the radiological safety is a prerequisite for the safe management of the naturally occurring radioactive materials (NORM) which cannot be reused. This becomes a crucial focus of our R&D efforts upon the implementation of the Act on Protective Action Guidelines against Radiation in the Natural Environment. To secure the safety, the establishment of technical bases and procedures for securing radiological safety related to the disposal of NORM is required. Thus, it is necessary to analyze the characteristics, to collect the data, to have the radiological safety assessment methodologies and tools, to investigate disposal methods and facilities, and to study the effects of the input data on the safety for the NORM wastes. Here, we assess the environmental impact of the NORM waste disposal with respect to the major domestic and foreign NORM characteristics. The data associated with major industries are collected/analyzed and the status of disposal facilities and methodologies relevant to the NORM wastes is investigated. We also suggest the conceptual design concept of a landfill disposal facility and the management plan with respect to the major NORM wastes characteristics. The radionuclide pathways are identified for the atmospheric transport and leachate release and the environmental impact assessment methodology for the NORM waste disposal is established using a relevant code. The assessment and analysis on the exposure doses and excessive cancer risks for the NORM waste disposal are performed using the characteristics of the representative domestic NORM wastes including flying ash, phosphor gypsum, and redmud. The results show that the exposure dose and the excessive cancer risks are very low to consider any radiation effects. This study will contribute to development in the areas of the regulatory technology for securing radiological safety relevant to NORM waste disposal and to the implementation technology for the Act.

• Journal of radiological science and technology
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• v.39 no.4
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• pp.509-516
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• 2016

A dental panoramic radiography which usually uses low level X-rays is subject to the Nuclear Safety Act when it is installed for the purpose of education. This paper measures radiation dose and spatial dose rate by usage and thereby aims to verify the effectiveness of radiation safety equipment and provide basic information for radiation safety of radiation workers and students. After glass dosimeter (GD-352M) is attached to direct exposure area, the teeth, and indirect exposure area, the eye lens and the thyroid, on the dental radiography head phantom, these exposure areas are measured. Then, after dividing the horizontal into a $45^{\circ}$, it is separated into seven directions which all includes 30, 60, 90, 120 cm distance. The paper shows that the spatial dose rate is the highest at 30 cm and declines as the distance increases. At 30 cm, the spatial dose rate around the starting area of rotation is $3,840{\mu}Sv/h$, which is four times higher than the lowest level $778{\mu}Sv/h$. Furthermore, the spatial dose rate was $408{\mu}Sv/h$ on average at the distance of 60 cm where radiation workers can be located. From a conservative point of view, It is possible to avoid needless exposure to radiation for the purpose of education. However, in case that an unintended exposure to radiation happens within a radiation controlled area, it is still necessary to educate radiation safety. But according to the current Medical Service Act, in medical institutions, even if they are not installed, the equipment such as interlock are obliged by the Nuclear Safety Law, considering that the spatial dose rate of the educational dental panoramic radiography room is low. It seems to be excessive regulation.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.3
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• pp.179-187
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• 2008

After Canada has struggled with a radioactive waste problem over for 20 years, the Canadian government finally found out that its approach by far has been lack of social acceptance, and needed a program such as public and stakeholder engagement (PSE) which involves the public in decision-making process. Therefore, the government made a special law, called Nuclear Fuel Waste Act (NFWA), to search for an appropriate nuclear waste management approach. NFWA laid out three possible approaches which were already prepared in advance by a nuclear expert group, and required Nuclear Waste Management Organization (NWMO) to be established to report a recommendation as to which of the proposed approaches should be adopted. However, NFWA allowed NWMO to consider additional management approach if the other three were not acceptable enough. Thus, NWMO studied and created a fourth management approach after it had undertaken an comparison of the benefits, risks and costs of each management approach: Adaptive Phased Management. This approach was intended to enable the implementers to accept any technological advancement or changes even in the middle of the implementation of the plan. The Canadian PSE case well shows that technological R&D are deeply connected with social acceptance. Even though the developments and technological advancement are carried out by the scientists and experts, but it is important to collect the public opinion by involving them to the decision-making process in order to achieve objective validity on the R&D programs. Moreover, in an effort to ensure the principles such as fairness, public health and safety, security, and adoptability, NWMO tried to make those abstract ideas more specific and help the public understand the meaning of each concept more in detail. Also, they utilized a variety of communication methods from face-to-face meeting to e-dialogue to encourage people to participate in the program as much as possible. Given the fact that Korea has been also having a hard time in dealing with spent nuclear fuel management, all of these efforts that Canada has made with a PSE program would give good lessons and implications to the Korean case. In conclusion, as a deliberative participation program, PSE could be a possible breakthrough approach for the Korean spent nuclear fuel management.