• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.4
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• pp.537-548
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• 2020

Currently, radioactive waste for disposal has been restricted to low and intermediate level radioactive waste generated during operation of nuclear power plants, and these radioactive wastes were managed and disposed of the 200 L and 320 L of steel drums. However, it is expected that it will be difficult to manage a large amount of decommissioning waste of the Kori unit 1 with the existing drums and transportation containers. Accordingly, the KORAD is currently developing various and large-sized containers for packaging, transportation, and disposal of decommissioning waste. In this study, the radiation exposure doses of workers and the public were evaluated using RADTRAN computational analysis code in case of the domestic on-road transportation of new package and transportation containers under development. The results were compared with the domestic annual dose limit. In addition, the sensitivity of the expected exposure dose according to the change in the leakage rate of radionuclides in the waste packaging was evaluated. As a result of the evaluation, it was confirmed that the exposure dose under normal and accident condition was less than the domestic annual exposure dose limit. However, in the case of a number of loading and unloading operations, working systems should be prepared to reduce the exposure of workers.

• 제어로봇시스템학회:학술대회논문집
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• 1997.10a
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• pp.1371-1374
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• 1997

We had developed the on-line environmental monitoring system which has installed around Kori Nuclear Power Plants and will be taken the place of the existing system. The system consists of a main computer and 11 sets of radiation monitoring post equipments. Nal(Tl) scintillation detectro was adopted in addition to ion-chamber detector and implemented with DCU(Dose Conversion Unit) and SCA(Single Channel Analyzer). Compared with the existing system, it has revised feature in the radiation measurements which are detection of artificial radioactivity and 2-ways of the radiatiion detectors. The field test trsults show that the developed radiation detecting equipments can measure environmental radiation withn 5.0% of the theoretical value.

• Nuclear Engineering and Technology
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• v.25 no.3
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• pp.413-423
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• 1993

This paper reports an experimental study of direct containment heating (DCH) which would occur if the primary system pressure is still high at the time of vessel breach during a light water reactor core melt accident. The experiments were conducted in 1/30-scale cavity models of Kori unit 1 and 2 and Young Kwang unit 3 and 4 nuclear power plants. One 1/20-scale model of the Kori plant was also used to investigate the scaling effect. The primary variables in the experiments were initial vessel pressure, vessel breach size and cavity geometry. It is observed that higher initial pressure and larger breach size enhance the melt dispersal fraction. Also, the cavity geometry appears to affect the dispersal rate greatly. A simple correlation of melt dispersal fraction is proposed in terms of nondimensional effective period. This correlation shows good agreement with the present experimental data, the KAIST data and the BNL data.

• Nuclear Engineering and Technology
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• v.22 no.1
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• pp.12-19
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• 1990

The Loss of Offsite Power Transient at 77.5% power which occurred on June 9, 1981 at the Kori Unit 1 PWR (Pressurized Water Reactor) is simulated using the RELAP5/MOD2 system thermal-hydraulics computer code. Major thermal-hydraulic parameters are compared with the available plant data. The comparison of the analysis results with the plant data demonstrates that the RELAP5/MOD2 code has the capability to simulate the thermal-hydraulic behaviour of PWRs under accident conditions of this type with accuracy, except the pressurizer pressure and level. The pressurizer pressure increase is sensitive to the in surge now it is believed that the interracial heat transfer in a horizontal stratified flow regime may be estimated low and the compression effect due to insurge flow may be high. In the nodalization sensitivity study it is found that S/G noding with junctions between bypass plenum and steam dome is preferred to simulate the S/G water level decreasing and avoid the spurious level peak at trubine trip.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.1
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• pp.69-75
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• 2004

DORT and ORIGEN2 code were used for calculation of neutron flux and inventory in reactor pressure vessel(RPV) of Kori unit-1, To calculate neutron flux using DORT code, the reactor was divided into 94 mesh from the center of core to RPV and from 0 to 45 degree along the azimuth. The cross-sections of main nuclides were recalculated using neutron flux in the RPV region. The results showed that 95% of the total activity in RPV came from the nuclides of $^{55}$ Fe, $^{60}$ Co, $^{59}$ Ni and $^{63}$ Ni. And the total activity with cooling of more than 50 years after decommissioning was no more than 0.2% of at the time of shutdown. Considering the weight of RPV is 210 tons, the initial total activity of RPV reached 5.25${\times}$10$^{6}$ GBq. To verify results of ORIGEN2 calculation, comparison between calculated and measured value at RPV of Kori unit-1 was peformed. The comparison results showed a good agreement.

• Nuclear Engineering and Technology
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• v.14 no.3
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• pp.122-137
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• 1982

Analyzed is the sensitivity of reactor transient behavior to various reactor parameters during the reactivity induced accidents (RIA) of the Kori Unit 1. Included in the analysis is a partial spectrum of RIAs with relatively fast transients such as uncontrolled rod cluster control assembly bank withdrawl from a subcritical or low power startup condition and rod ejection accidents. The analysis can be performed generally in three steps: calculation of an average core power change, hot spot heat transfer calculation and DNBR (departure from nucleate boiling ratio) calculation. The computer codes used for the analysis are either developed based on the codes relevent to it. These codes are evaluated to be highly reliable. An extensive sensitivity analysis is performed to study the effects of various reactor design and operating parameters on the reactor transient behavior during the accidents. The assumptions and initial conditions used for the RIA analysis in the Kori Unit 1 FSAR (Final Safety Analysis Report) are reexamined, and the corresponding analysis results are reassessed, based on the sensitivity analysis results, to be conservative and reliable.

• 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.

• Nuclear Engineering and Technology
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• v.54 no.9
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• pp.3242-3249
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• 2022

As Kori unit 1 is undergoing the decommissioning process, estimating the disposal amount of waste from the decommissioned nuclear reactor has become one of the challenging issues. Since the waste disposal amount estimation depends on the packing of the waste, it is highly desirable to optimize the waste packing plan. In this study, we developed an efficient scheme for packing waste component segments. The scheme consists of 1) preparing three-dimensional models of segments, 2) orienting each segment in such a way to minimize the bounding box volume, and 3) applying hybrid genetic algorithm to pack the segments in the disposal containers. When the packing solution converges in the algorithm, it comes up with the number of containers used and the placement of segments in each container. The scheme was applied to Kori-1 reactor pressure vessel. The required number of containers calculated by the developed scheme was 24 compared to 42 that was the estimation of the prior packing plan, resulting in disposal volume savings by more than 40%. The developed method is flexible for applications to various packing problems with waste segments from different cutting options and different sizes of containers.

• 전기의세계
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• v.24 no.2
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• pp.71-77
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• 1975

This paper describes the results of a study for the stability of power system when the Kori Nuclear #1 P/P is operated with existing system. First, a transient disturbance, which effects the stability of entire power system, was analysed and to cope with the problem a load shedding method was studied to recover the fluctuation of the power system. Second, transient stability problem was studied when three phase fault occurs in 345 Extra High Voltage power System, and from this result, it was found to be highly effective that high speed protecting device should be provided and operated to recover the fault.

• Journal of computational fluids engineering
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• v.7 no.2
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• pp.43-52
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• 2002

This paper presents the numerical methodology of ATHOS3 code for thermal hydraulic analysis of steam generators in nuclear power plant. Topics include porous media approach, governing equations, physical models and correlations for solid-to-fluid interaction and heat transfer, and numerical solution scheme. The ATHOS3 code is applied to the thermal hydraulic analysis of steam generator in the Korea Kori Unit-1 nuclear power plant and the computed results are presented