• Journal of Radiation Protection and Research
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• v.28 no.2
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• pp.127-135
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• 2003

The primary contributors to the total occupational radiation exposure in operating nuclear power plants are operation and maintenance activities doting refueling outages. The Advanced Power Reactor 1400 (APR1400) includes a number of design improvements and plans to utilize advanced maintenance methods and robotics to minimize the annual collective dose. The major radiation exposure reduction features implemented in APR1400 are a permanent refueling pool seal, quick opening transfer tube blind flange, improved hydrogen peroxide injection at shutdown, improved permanent steam generator work platforms, and more effective temporary shielding. The estimated average annual occupational radiation exposure for APR1400 based on the reference plant experience and an engineering judgment is determined to be in the order of 0.4 man-Sv, which is well within the design goal of 1 man-Sv. The basis of this average annual occupational radiation exposure estimation is an eighteen (18) month fuel cycle with maintenance performed to steam generators and reactor coolant pumps during refueling outage. The outage duration is assumed to be 28 days. The outage work is to be performed on a 24 hour per day basis, seven (7) days a week with overlapping twelve (12) hour work shifts. The occupational radiation exposure for APR1400 is also determined by an alternate method which consists of estimating radiation exposures expected for the major activities during the refueling outage. The major outage activities that cause the majority of the total radiation exposure during refueling outage such as fuel handling, reactor coolant pump maintenance, steam generator inspection and maintenance, reactor vessel head area maintenance, decontamination, and ICI & instrumentation maintenance activities are evaluated at a task level. The calculated value using this method is in close agreement with the value of 0.4 man-Sv, that has been determined based on the experience aid engineering judgement. Therefore, with the As Low As Reasonably Achievable (ALARA) advanced design features incorporated in the design, APR1400 design is to meet its design goal with sufficient margin, that is, more than a factor of two (2), if operated on art eighteen (18) month fuel cycle.

• Nuclear Engineering and Technology
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• v.30 no.5
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• pp.435-443
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• 1998

Leak-before-break(LBB) approach has been shown to be both cost effective and risk reductive when applied to high energy Piping in nuclear Power Plants. For the Korean Next Generation Reactor (KNGR) development, LBB application is considered for the Main Steam Line(MSL) piping inside containment. Unlike the primary system leakages, the MSL leak detection systems must be based on principles other than radioactivity measurements. Among humidity, heat and acoustic noise currently being considered as indicators of leakage, we explored humidity as an effective one and developed ceramic-based humidity sensor which can be qualified for LBB applications. The ceramic material, sintered and annealed MgCr$_2$O$_4$-TiO$_2$, is shown to increase its electrical conductivity drastically upon water vapor adsorption over the entire temperature range of interest. With this ceramic sensor specimen, we suggested installation-inside-the-piping method by which we can detect leakage more rapidly and sensitively. In this paper, we describe the progress in the development and characterization of ceramic humidity sensor for the LBB application to the MSL of KNGR.

• Applied Microscopy
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• v.45 no.3
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• pp.177-182
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• 2015

The boiler tubes of X20CrMoV12.1 used in fossil-fired power plants were obtained and analyzed for the effect of water treatment on the steam corrosion-induced oxide scale in an effort to better understand the oxide formation mechanism, as well as pertinent method of maintenance and lifetime extension. The specimens were analyzed using various microscopy and microanalysis techniques, with focuses on the effect of water treatment on the characters of scale. X-ray diffraction analysis showed that the scales of specimens were composed of hematite ($Fe_2O_3$), magnetite ($Fe_3O_4$), and chromite ($FeCr_2O_4$). Electron backscatter diffraction analysis showed that the oxides were present in the following order on the matrix: outer $Fe_2O_3$, intermediate $Fe_3O_4$, and inner $FeCr_2O_4$. After all volatile treatment or oxygenated treatment, a dense protective $Fe_2O_3$ layer was formed on the $Fe_3O_4$ layer of the specimen, retarding further progression of corrosion.

• Journal of the Korean Society for Nondestructive Testing
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• v.20 no.1
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• pp.10-17
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• 2000

In boiler high-temperature pipelines such as main steam pipe, header and steam drum in fossil power plants, conventional measurement techniques(replica method, electric resistance method, and hardness test method) for measuring creep damage have such disadvantages as complex preparation and measurement procedures, too many control parameters. And also, these techniques have low practicality and applied only to component surfaces with good accessibility. It needs to apply a reliable and quantitative ultrasonic nondestructive evaluation method that can be replaced for these equipment. In this study, both artificial creep degradation test using life prediction formula and frequency analysis by ultrasonic tests for crept specimens were carried out for the purpose of nondestructive evaluation for creep damage. As a result of ultrasonic tests for crept specimens, we conformed that the high frequency side spectra decrease and central frequency components shift to low frequency band, and also their bandwidth decreases as increasing creep damage in backwall echos.

• Journal of the Korean Society for Heat Treatment
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• v.7 no.3
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• pp.155-159
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• 1994

As Cr-Mo-V steels have excellent mechanical and creep properties at elevated temperatures, they are extensively used in power plants. However, the steam turbine components are supposed to have suffered material degradation during long-term service at elevated tenperatures. Many efforts have been made to assess the safety and residual life of these components by means of non-destructive methods such as plastic replication, hardness and electric resistance techniques. Recently, a parameter correlating hardness changes during long-term heating to those during creep was introduced and it was named 'G parameter'. The electric resistivity as well as hardness are affected by damage accumulation, but there have been no efforts to correlate G parameter to resistivity changes. In this study, relationship between G parameter and changes in electric resistivity was investigated using artificially aged Cr-Mo-V steel. It is well understood that G parameter can be applied to electric: resistance techmique.

• Journal of the Korean Society for Nondestructive Testing
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• v.24 no.4
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• pp.384-389
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• 2004

As the increase of the operation year of nuclear power plants, the probabilities of the degradation of the major facilities and materials in the nuclear power plants are increased. The integrity of those facilities shall be monitored and verified by the non-destructive examination methods with the regulation codes, so called inservice inspection(ISI). The ISI of Yonggwang unit 6 was performed in four different parts, 1) non-destructive examinations for the components, piping weldments and structures, 2) automated ultrasonic examinations for pressure vessels, 3) visual examinations for the interior structures of the reactor, 4) eddy current examinations for the steam generator tubes. As the results, there was no severe indication and all detected indications were evaluated as non-relavent. Especially for the examinations of the piping weldments, PD(Performance Demonstration) was applied as a W examination method defined in the 1995 edition of ASME Code Sec. XI. The implementation of the PD for the piping weld results in an improvement of the reliability of the UT examinations.

• Journal of Radiation Protection and Research
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• v.34 no.2
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• pp.55-64
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• 2009

In a pressurized heavy water reactor (PHWR), radiation workers who have access to radiation controlled areas submit their urine samples to health physicists periodically; internal radiation exposure is evaluated by the monitoring of these urine samples. Internal radiation exposure at PHWRs accounts for approximately 20 $\sim$ 40% of total radiation exposure; most internal radiation exposure is attributed to tritium. Carbon-14 is not a dominant nuclide in the radiation exposure of workers, but it is one potential nuclide to be necessarily monitored. Carbon-14 is a low energy beta emitter and passes relatively easily into the body of workers by inhalation because its dominant chemical form is radioactive carbon dioxide ($^{14}CO_2$). Most inhaled carbon-14 is rapidly exhaled from the worker's body, but a small amount of carbon-14 remains inside the body and is excreted by urine. In this study, a method for dual analysis of tritium and carbon-14 in urine samples of workers at nuclear power plants is developed and a method for internal dose assessment using its excretion rate result is established. As a result of the developed dual analysis of tritium and carbon-14 in urine samples of radiation workers who entered the high radiation field area at a PHWR, it was found that internal exposure to carbon-14 is unlikely to occur. In addition, through the urine counting results of radiation workers who participated in the open process of steam generators, it was found that the likelihood of internal exposure to either tritium or carbon-14 is extremely low at pressurized water reactors (PWRs).

• 한국신재생에너지학회:학술대회논문집
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• 2010.06a
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• pp.113.1-113.1
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• 2010

IGCC(Integrated Coal Gasification and Combined Cycle) plants can be among the most advanced and environmental systems for electric energy generation from various feed stocks and is becoming more and more popular in new power generation fields. In this work, the performance of IGCC plants employing Shell gasification technology and a GE 7FB gas turbine engine was simulated using IPSEpro open-equation modeling environment for different operating conditions. Performance analyses and comparisons of all operating cases were performed based on the design cases. Discussions were focused on gas composition, syngas production rate and overall performance. The validation of key steady-state performance values calculated from the process models were compared with values from the provided heat and material balances for Shell coal gasification technology. The key values included in the validation included the inlet coal flow rate; the mass flow rate, heating value, and composition of major gas species (CO, H2, CH4, H2O, CO2, H2S, N2, Ar) for the syngas exiting the gasifier island; and the HP and MP steam flows exiting the gasifier island.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.32 no.1
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• pp.54-61
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• 2008

Integrated Gasification Combined Cycle (IGCC) power plant converts coal to syngas, which is mainly composed of hydrogen and carbon monoxide, by the gasification process and produces electric power by the gas and steam turbine combined cycle power plant. The purpose of this study is to investigate the influence of using syngas in a gas turbine, originally designed for natural gas fuel, on its performance. A commercial gas turbine is selected and variations of its performance characteristics due to adopting syngas is analyzed by simulating off-design gas turbine operation. Since the heating value of the syngas is lower, compared to natural gas, IGCC plants require much larger fuel flow rate. This increases the gas flow rate to the turbine and the pressure ratio, leading to far larger power output and higher thermal efficiency. Examination of using two different syngases reveals that the gas turbine performance varies much with the fuel composition.

• Journal of the Society of Naval Architects of Korea
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• v.60 no.3
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• pp.146-154
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• 2023

Efforts have been made to reduce the greenhouse gas emissions from ships by limiting the energy efficiency index, and net zero CO₂ emission was proposed recently. The most ideal measure to achieve zero emission ship is electrification, and fuel cells are considered as a practical power source of the electrified propulsion system. The electric efficiency in the electrochemical reaction of fuel cells can be achieved up to 60% practically. The remaining energy is converted to heat energy but most of them are dissipated by cooling. In the author's previous research, a hybrid propulsion system utilizing not only electricity but also heat was introduced by combining electric motor and steam turbine. In this article, long term efficiency is evaluated for the introduced hybrid propulsion system by considering a virtual 24,000 TEU class container carrier model. To reflect a more practical operating condition, the actual navigation data of a similar real ship in the real world were collected from automatic identification system data and applied. From the result, the overall efficiency of the hybrid propulsion system is expected to be higher than a conventional electric propulsion fuel cell ship by 30%.