• Nuclear Engineering and Technology
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• v.27 no.5
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• pp.670-686
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• 1995

This paper provides an improvement on our previous study ［1］ for multi-fault diagnosis in real time in large-scale systems. In the method, fault propagation probability(FPP) and fault propagation time(FPT) in a fuzzy sense are additively used to describe the fault propagation model(FPM) in more practical manner. A modified fault diagnosis procedure is also given. This method is applied for diagnosis of the primary system in the Kori nuclear power plant unit 2 under a transient condition in case of unit value of FPP on each branch of the FPM.

• KDI Journal of Economic Policy
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• v.40 no.1
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• pp.67-83
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• 2018

We have hypothesized that nuclear risk is significantly inversely related to the distance from residences to nuclear power plants and that the level of life satisfaction of residents therefore increases with the distance. We empirically explore the relationship between Ulsan citizens' life satisfaction levels and the distance between their residences and the Kori and Wolsong nuclear power plants (NPP) based on the life satisfaction approach (LSA). The dataset we used covers only Ulsan citizens from the biennial Ulsan Statistics on Citizen's Living Condition and Consciousness of 2014 and 2016. Controlling for micro-variables such as education, work satisfaction, gender, marital status, and expenditures, we found a statistically significant relationship between life satisfaction and the distance between the residences and the nuclear power plants. Nuclear negative externalities including (i) health and environmental impact, (ii) radioactive waste disposal, and (iii) the effect of severe accidents can be quantified in terms of LS units and monetary units. We were able to calculate the monetary value of NPP externalities at $277 per kilometer of distance for Kori and $280 per kilometer of distance for Wolsong at constant 2015 prices. These estimates are quite different from the traditional estimates made with the contingent valuation method, whereas they are similar to the findings of LSA studies abroad. Hence, the need to adopt the LSA in South Korea and policy implications are demonstrated.

• Nuclear Engineering and Technology
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• v.55 no.1
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• pp.45-51
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• 2023

During the operation of a nuclear power plant (NPP), the waste reactor vessel closure head (RVCH) that is replaced owing to design or manufacturing defects is buried in a designated area or temporarily stored in a radiation shielding facility within the NPP. In such cases, storing it for extended periods proves a challenge owing to space constraints in the power plant and a safety risk associated with radiation exposure; therefore, dismantling it quickly and safely is crucial. However, not much research has been done on the dismantling of the RVCH in an operational power plant. This study proposes a dismantling process based on the radioactive contamination level measured for the Kori #1 RVCH, which is currently being discarded and stored, and examines the decontamination and cutting according to this process. In addition, the amount of secondary waste and dismantling cost are evaluated, and the dismantling effect of the reactor closure head is analyzed.

• ALGAE
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• v.19 no.3
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• pp.257-270
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• 2004

In order to clarify the structure and seasonal dynamics of warm tolerant benthic marine algal community in Korea, the species composition and biomass of marine algae at the discharge canal of Uljin nuclear power plant on the East Coast of Korea were investigated seasonally from February 1992 to October 2000. 107 species of marine algae were found at the discharge canal during the past nine years. In general, the number of species observed was abundant in spring or summer and less in autumn or winter. 27 species (4 blue-green, 5 green, 6 brown and 12 red algae) of marine algae occurred more than 1/ 6 frequency and thus can be categorized as warm tolerant species. Among these, one brown (Dictyota dichotoma) and four red algae (Gelidium amansii, Anphiroa ephedraea, Hydrolithon sargassi, Marginisporum crassissimum) are recorded as warm tolerant marine algae for the first time in Korea. Padina arborescens, Anphiroa zonata and Corallina pilulifera were common species found more than 75% frequency. Seasonal fluctuations of mean biomass were 0-1,330 g dry wt m^(-2) and dominant species in biomass were Corallina pilulifera (contribution to a total biomass proportion 34%), Anphiroa zonata (23%), Padina arborescens (18%) and Sargassum micracanthum (11%). The red algae appeared as predominant algal group at the discharge canal of Uljin nuclear power plant in the qualitative and quantitative aspects. The green algae such as Enterornorpha compressa appeared rather frequently at the discharge canal, but the biomass proportion was very low, in contrast to Kori nuclear power plant where there was definite green algal dominance. Differences in algal communities developed at the discharge canals of Uljin and Kori nuclear power plant on the East Coast of Korea, particularly biomass proportions of green algae, can probably be related to local environmental factors such as water velocity through the canal and natural seawater temperatures.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.437-442
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• 1997

To determine the optimum pressurizer water level during normal operation for Kori unit 1, performance and safety analysis are peformed. The methodology is developed by evaluating "decrease in secondary heat removal" events such as Loss of Normal Feedwater accident. To demonstrate optimum pressurizer level setpoint, RETRAN-03 code is used for performance analysis. Analysis results of RETRAN following reactor trip are compared with the actual plant data to justify RETRAN code modelling, The results of performance and safety analyses show that the newly established level setpoints not only improve the performance of pressurizer during transient including reactor trip but also meet the design bases of the pressurizer volume and pressure. pressure.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2019.05a
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• pp.261-262
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• 2019

• Nuclear Engineering and Technology
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• v.26 no.1
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• pp.41-53
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• 1994

A SGTR accident postulated at Kori unit 1 is simulated with Mihama unit experience, which occurred on February 1991, to evaluate the capability of plant to cope with the transient. The system design and plant conditions of Kori Unit 1 are much similar with those of Mihama Unit 2. Therefore, special concern has been given to evaluate the sequence and the resulting consequence of the postulated SGTR accident at the Kori unit 1 An analysis is peformed as realistically as possible, with following the EOP of Kori unit 1. The result indicates that the leak through tube break terminates within about forty minutes, and the Kori unit 1 may be sufficient to cope with SGTR accident with same type of sequence. However, the reconsideration may be required for the design of Kori unit 1 which disconnects non-safety AC power from off-site power on SI signal generation. It may be pointed out that the content of EOP for SGTR accident is not enough to require operator's proper judgements. An analysis of SGTR accident tested in the LSTF which simulated the SGTR accident at the Mihama Unit 2 is peformed using the RELAP5/MOD3. The results indicates that the code yields in general good agreement with the test, except the break flowrate at the early stage of the event.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.02a
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• pp.12-27
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• 2004

Pretreatment process consisted of submerged hollow-fiber microfiltration(HMF) membrane and spiral-wound nanofiltration(SNF) membrane has been developed by NETEC, KHNP for the purpose of improving the impurities of liquid radioactive waste before entering Selective Ion Exchange System(SIES). The lab-scale combined system was installed at Kori NPP #2 nuclear power plant and demonstration tests using actual liquid radioactive waste were carried out to verify the performance of the combined system. The submerged HMF membrane was adopted for removal of suspended solid in liquid radioactive waste and the SNF membrane was used for removal of particulate radioisotope such as, Ag-l10m and oily waste because ion exchange resin can not remove particulate radioisotopes. The liquid waste in Waste Holdup Tank (WHT) was processed with HMF and SNF membrane, and SIES. The initial SS concentration and total activity of actual waste were 38,000ppb and $1.534{\times}10_{-3}{\mu}Ci/cc$, respectively. The SS concentration and total activity of permeate were 30ppb and lower than LLD(Lower Limit of Detection), respectively.

• Journal of Nuclear Fuel Cycle and Waste Technology
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• v.1 no.1
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• pp.37-47
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• 2013

This paper describes basic plans for the development of a radioactive waste disposal facility with the introduction of Nuclear Power Plants (NPPs) for Kenya. The specific objective of this study was to estimate the total projected waste volumes of low- and intermediate-level radioactive waste (LILW) expected to be generated from the Kenyan nuclear power programme. The facility is expected to accommodate LILW to be generated from operation and decommissioning of nuclear power plants for a period of 50 years. An on-site storage capacity of 700 $m^3$ at nuclear power plant sites and a final disposal repository facility of more than 7,000 $m^3$ capacity were derived by considering Korean nuclear power programme radioactive waste generation data, including Kori, Hanbit, and APR 1400 nuclear reactor data. The repository program is best suited to be introduced roughly 10 years after reactor operation. This study is important as an initial implementation of a national LILW disposal program for Kenya and other newcomer countries interested in nuclear power technology.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.21 no.7
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• pp.1089-1096
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• 1997

In this paper, Rancho Seco transient which is reported as a typical pressurized thermal shock event is postulated to be occuring in the Kori unit 1 plant, the oldest nuclear power plant in Korea. For the given material properties, transient history such as temperature and pressure, and postulated flaw, the stress distribution is obtained to calculate stress intensities for a wide range of assumed crack sizes. The stress intensities are compared with the fracture toughness, which is determined using the material properties and the distribution of the nil ductility transition temperature, to determine if cracking is expected to occur during the transient. The allowable operating year for the transient is determined and the evaluation results are discussed.