• Journal of Korea Society of Industrial Information Systems
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• v.25 no.5
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• pp.31-39
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• 2020

According to lessons learned from an accident of Fukushima Daiichi nuclear power plant, it is advisable to make a comprehensive radiation survey by the accident phase for efficient response and risk management using diverse survey platforms. This study focuses on the technical status of environmental radiation monitoring using a UAV (Unmanned aerial vehicle) and the performance test of developed aerial survey system based on two detectors with an high energy resolution through the field application to contaminated areas. Finally, the performance of aerial survey at diverse flight heights was successfully achieved by introducing the correction factor to represent the results into ambient dose rate at 1m above the ground.

• Journal of the Korean Society for Nondestructive Testing
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• v.24 no.4
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• pp.325-330
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• 2004

Several kinds of low voltage cables have been used in nuclear power plants for the supply of electric power, supervision, and the propagation of control signals. These low voltage tables must be inspected for safe and stable operation of nuclear power plants. In particular, the degradation diagnosis to estimate the integrity of low voltage rabies has recently been emphasized according to the long use of nuclear power plants. In order to evaluate their degradation, the surrounding temperature, hardness of insulation material, elongation at breaking point (EAB), etc. have been used. However, the measurement of temperature or hardness is not useful because of the absence of quantitative criteria; the inspection of a sample requires turning off of the power plant power; and, the electrical inspection method is not sufficiently sensitive from the initial through the middle stage of degradation. In this research, based on the theory that the ultrasonic velocity changes with relation to the degradation of the material, we measured the ultrasonic velocity as low voltage cables were degraded. To this end, an ultrasonic degradation diagnosis device was developed and used to measure the ultrasonic velocity with the clothing on the cable, and it was confirmed that the ultrasonic velocity changes according to the degradation of low voltage cables. The low voltage cables used in nuclear power plants were degraded at an accelerated rate, and EAB was measured in a tensile test conducted after the measurement of ultrasonic velocity. With the increasing degradation degree, the ultrasonic velocity decreased, whose potential as a useful parameter for the quantitative degradation evaluation was thus confirmed.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.4
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• pp.375-387
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• 2019

In this study, the radiation dose rates for the design basis fuel of 360 assemblies CANDU spent nuclear fuel transportation cask were evaluated, by measuring radiation source terms for the design basis fuel of a pressurized heavy water reactor. Additionally, radiological safety evaluation was carried out and the validity of the results was determined by radiological technical standards. To select the design basis fuel, which was the radiation source term for the spent fuel transportation cask, the design basis fuels from two spent fuel storage facilities were stored in a spent fuel transportation cask operating in Wolsung NPP. The design basis fuel for each transportation and storage system was based on the burnup of spent fuel, minimum cooling period, and time of transportation to the intermediate storage facility. A burnup of 7,800 MWD/MTU and a minimum cooling period of 6 years were set as the design basis fuel. The radiation source terms of the design basis fuel were evaluated using the ORIGEN-ARP computer module of SCALE computer code. The radiation shielding of the cask was evaluated using the MCNP6 computer code. In addition, the evaluation of the radiation dose rate outside the transport cask required by the technical standard was classified into normal and accident conditions. Thus, the maximum radiation dose rates calculated at the surface of the cask and at a point 2 m from the surface of the cask under normal transportation conditions were respectively 0.330 mSv·h^-1 and 0.065 mSv·h^-1. The maximum radiation dose rate 1 m from the surface of the cask under accident conditions was calculated as 0.321 mSv·h^-1. Thus, it was confirmed that the spent fuel cask of the large capacity heavy water reactor had secured the radiation safety.

• Tuberculosis and Respiratory Diseases
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• v.44 no.5
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• pp.1001-1010
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• 1997

Background : PCR technique is useful in diagnosis of pulmonary tuberculosis. But, its sensitivity and specificity is some different among several studies. Our aim is compare our PCR results with other's previous PCR results in AFB smear negative patients. Methods : PCR were performed in patients that their disease were suspected as active pulmonary tuberculosis and that their initial serial sputum AFB smear results were negative. Total number of patients studied by PCR technique was 177. Also, we analyzed the data only in patients whose bronchial washing fluid AFB smear was negative. And the primer had been used was IS 6110. Results : In our retrograde study, the number of patients who are diagnosed as having active pulmonary tuberculosis, inactive pulmonary tuberculosis and nontuberculous pulmonary disease was 99, 28, 50, respectively. In the sputum study, the sensitivity of PCR is 41.5% (27 PCR positive cases/65 active TBc cases). And the sensitivity of TB culture is 53.8% (35 TB culture positive cases/65 active TBc cases). In the bronchial washing specimen study, the sensitivity of PCR is 53.8% (21 PCR positive cases/39 active TBc cases). And the sensitivity of TB culture is 43.6% (17 TB culture Positive cases/39 active TBc cases). The specificity of PCR in our study is 94.9%. (74 PCR negative cases/78 inactive TBc or nontubereulous cases) In the cases of patients who were never takened anti-TBc medication, the sensitivity of PCR (45.6%--25 positive cases/55 cases) is some lower than culture (58.2%--32 positive cases/55 cases). In the cases of patients who had been takened anti-TBC medication. the sensitivity of PCR (60%--18 positive cases/30 cases) is some superior than culture (50%--15 positive cases/30 cases). Conclusion : We think that PCR results in cases of sputum AFB smear negative patients is nearly same as culture. And PCR is especially useful in patients who had been takened anti-TBc medication on admission.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.34 no.9
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• pp.1273-1279
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• 2010

Pressure tubes at the CANada Deuterium Uranium (CANDU) nuclear power plants are periodically inspected in accordance with the CSA N285.4 code. If flaws that do not satisfy the criteria given in CSA N285.4 are detected, the code permits a fitness-for-service assessment to determine the acceptability of the flawed pressure tubes. In this paper, the Wolsong In-service Evaluation System (WIES) is introduced; this system has been developed for the assessment of the flawed pressure tubes and is based on CSA N285.8. Since the system evaluates the integrity of flawed pressure tubes exactly and promptly during an in-service inspection, it will help in operating the Wolsong nuclear power plants without prolonging the outage period.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.2
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• pp.143-150
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• 2010

Vitrification technology has been gradually recognized as one of effective solidification methods for concentrated boric acid wastes generated in PWR. Vitrification for low- and intermediate-level radioactive wastes has a large volume reduction and good durability for the final products. A feasibility study for the vitrification of concentrated boric acid wastes has been performed with developing the pre-treatment methods of powdered wastes, glass compositions using glass formulation and demonstration test. The pre-treatment method is pelletizing the powder type for stable feeding within cold crucible melter. The glass compositions should be developed considering molten glass are related with wastes reduction. High contents of sodium and boron within borate wastes give influence to waste loading. A variety of factors obtained from the demonstration test are reviewed, which is wastes feeding rate, off-gas characteristics on stack and glass characteristics of final products such as durability for implementing the wastes disposal requirement. The aim of this paper is to present the feasibility of vitrification and review the solidification method for concentrated boric acid wastes and obtain the physicochemical characteristics of solidified glass.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.4
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• pp.247-254
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• 2012

Vitrification technology has been widely applied as one of effective processing methods for wastes generated in nuclear power plants. The advantage of vitrifying for low- and intermediate-level radioactive wastes has a large volume reduction and good durability for the final products. Recently, a filter using on HVAC(Heating Ventilating & Air Conditioning System) is composed with media (glass fiber) and separator (aluminum film) has been studied the proper treatment technology for meeting the waste disposal requirement. Present paper is a feasibility study for the filter vitrification that developing of the glass compositions for filter melting and melting test for physicochemical characteristic evaluation. The aluminum metal of film type is preparing with 0.5 cm size for proper mixing with glass frit, glass fiber is also preparing with 1 cm size within crucible. The glass compositions should be developed considering molten glass are related with wastes reduction. Glass compositions obtained from developing on glass formulation are mainly composed of $SiO_2$ and $B_2O_3$ for aluminum metal. A variety of factors obtained from the glass formulation and melting test are reviewed, which is feeding rate and glass characteristics of final products such as durability for implementing the wastes disposal requirement.

• Journal of Energy Engineering
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• v.18 no.3
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• pp.191-202
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• 2009

South Korea, ranks 10th largest emitter of carbon dioxide in the world, will probably be under the obligation to reduce GHG emission from 2013. It is very important to reduce the electrical energy consumption since 30% of GHG emission in South Korea is made during electricity generation. In this study, based on "the 1st national energy master plan", the GHG emission reduction potential and the feasibility of the scenario in the electricity generation have been analyzed using LEAP(Long-range Energy Alternative Planning system). The scenario of the mater plan contains the 41% expansion of nuclear power plant facilities and the 11% diffusion of renewable energy until 2030. In result, total $CO_2$ emission reduction rate is 28.8% in 2030. Also $CO_2$ emission of unit electricity generation of bituminous coal power plant is $0.85kgCO_2/kWh$ and its LNG power plant is $0.51kgCO_2/kWh$ in BAU scenario. Therefore when existing facilities is exchanged for nuclear or renewable energy power plant, substitute of bituminous power plant is more effective than LNG power.

• Fire Science and Engineering
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• v.31 no.2
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• pp.52-60
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• 2017

Numerical simulations were performed to evaluate the habitability of an operator for a cabinet fire in the main control room of a nuclear power plant presented in NUREG-1934. To this end, a Fire Dynamics Simulator (FDS), as a representative fire model, was used. As the criteria for determining the habitability of operator, toxic products, such as CO, were also considered, as well as radiative heat flux, upper layer temperature, smoke layer height, and optical density of smoke. As a result, the probabilities of exceeding the criteria for habitability were evaluated through the sensitivity analysis of the major input parameters and the uncertainty analysis of fire model for various fire scenarios, based on V&V (Verification and Validation). Sensitivity analyses of the maximum heat release rate, CO and soot yields, showed that the habitable time and the limit criterion, which determined the habitability, could be changed. The present methodology will be a realistic alternative to enhancing the reliability for a habitability evaluation in the main control room using uncertain information of cabinet fires.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.38 no.6
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• pp.655-662
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• 2014

This study conducted tensile tests on an Alloy 690TT tube at room temperature (RT) and at $343^{\circ}C$ using tube- and ring-type specimens to investigate the stress-strain behavior and tensile properties of a steam generator (SG) tube in the axial and circumferential directions at RT and at the design temperature of a nuclear power plant (NPP). The results of the axial tensile test showed that yield point phenomena appeared at both RT and $343^{\circ}C$, and serrated flow in the stress-strain curve appeared at $343^{\circ}C$. Yield and tensile strengths for both directions were clearly lower at $343^{\circ}C$ compared to RT; however, the elongations were approximately the same at both test temperatures. Regardless of the test temperature, the strengths in the circumferential direction were lower by approximately 5~10 % than those in the axial direction. In addition, the test data revealed that the reduction in the yield and tensile strengths of the Alloy 690TT SG tube with the test temperature was more significant than that estimated by the temperature correction factor of ASME Sec.II.