• Nuclear Engineering and Technology
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• v.14 no.4
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• pp.186-195
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• 1982

The steam line break accident for Kori Unit 1 is analyzed by a code SYSRAN which calculates nuclear power and heat flux using the point kinetics equation and the lumped-parameter model and calculates system transient using the mass and energy balance equation with the assumption of uniform reactor coolant system pressure. The 1.4 f $t^2$ steam line break accident is analyzed at EOL (End of Life), hot shutdown condition in which case the accident would be most severe. The steam discharge rate is assumed to follow the Moody critical flow model. The results reveal the peak heat flux of 38% of nominal full power value at 60 second after the accident initiates, which is higher than the FSAR result of 26%. Trends for the transient are in good agreement with FSAR results. A sensitivity study shows that this accident is most sensitive to the moderator density coefficient and the lower plenum mixing factor. The DNBR calculation under the assumption of $F_{{\Delta}H}$=3.66, which is used in the FSAR with all the control and the shutdown assemblies inserted except one B bank assembly and of Fz=1.55 shows that minimum DNBR reaches 1.62 at 60 second, indicating that the fuel failure is not anticipated to occur. The point kinetics equation, the lumped-parameter model and the system transient model which uses the mass and energy balance equation are verified to be effective to follow the system transient phenomena of the nuclear power plants.lear power plants.

• Journal of the Korea Safety Management & Science
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• v.7 no.4
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• pp.85-99
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• 2005

The main objective of this paper is to search whether containment vessel's best pressure may increase until how long when loss of coolant accident (LOCA) happened in containment vessel of Ulchin nuclear power plant 1 and 2. Another goal of this research is to find the influential factors that increase containment vessel pressure. Model for this research is Ulchin nuclear power plant 1 with 10 cycles. Data were collected by simulator of Ulchin nuclear power plant 1 and design of experiment was used for data analysis. For the experiment, seven factors that are going to influence in containment vessel pressure were chosen. It was found that fatter which influences in early rise of containment vessel pressure after LOCA is only explosion size. Also, containment vessel's best pressure (3.74 bar.a) was much lower than limit (4.86 bar.a) of FSAR (Final Safety Analysis Report).

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.538-543
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• 1995

영광 3,4호기 FSAR 이후에 인허가 쟁점이 되고 있는 USNRC의 GDC17에 대한 재해석의 적용과 이에 관련된 소외전원상실, 소외전원상실과 원자로정지로 인한 터빈정지사이의 3초 지연시간에 대하여 그 영향이 15장 사고해석에 미치는 영향을 고찰하여보았다. 영광 3,4호기 예비안정성 분석보고서, 최종안정성분석보고서, CESSAR-F, 영광 1,2호기, CESSAR-DC의 개정판 H 및 N의 15장에서 소외전원상실이 적용된 방법을 살펴보고 소외전원상실과 밀접히 관련된 전기계통의 설계차 이점을 살펴보았다. 각각의 접근방법의 차이점 및 타당성에 대한 검토로부터 바람직한 사고 해석 방법론을 제시하고자 하였다.

• Nuclear Engineering and Technology
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• v.17 no.4
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• pp.256-266
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• 1985

The loss of coolant flow accident is analyzed for the pressurized water reactor of Korea Nuclear Unit-1. The loss of coolant flow accident is classified into three types in accordance with its severity; partial loss of coolant flow, complete loss of coolant flow and pump locked rotor accident. Analysis has been carried out in three stages; system transient and average core analysis, DNBR calculation and hot spot analysis. The purpose of developing KTRAN is to simulate the transient fast. For the DNBR calculation, the thermal hydraulic codes, SCAN and COBRA IV-1, are adopted. And for the hot spot analysis, the fuel thermal transient code LTRAN is employed. This code system should be fast responding to the transient analysis. In case the transient occurs, severity comes within a couple of seconds. So response should be fast to accomodate the following sequence of the accident. Unfortunately this purpose could not be achieved by KTRAN. However, the calculated results are well comparable with FSAR results in range. Thereby, the effectiveness of KTRAN code analysis in this type of accident is proven.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.479-483
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• 1997

CANDU형 발전소의 사고해석 검증을 위한 계통분석 코드는 별도로 개발되어 있지 않으며, PWR 사고해석 검증용으로 널리 사용되고 있는 RELAP5 코드를 CANDU형 발전소의 사고해석 검증용으로 개발하려는 연구가 현재 진행되고 있다. CANDU형 발전소를 묘사한 RD-14 실험장치에서의 실험결과를 RELAP5 코드로 평가한 연구는 있으나, 실제 CANDU형 발전소의 사고해석에 적용한 예는 없다. 본 연구에서는 RELAP5 코드를 이용하여 CANDU형 발전소의 주증기관 파단사고를 분석하고, 그 결과를 월성 2,3,4 FSAR의 분석결과와 비교하여, CANDU형 발전소에 대한 RELAP5 코드의 적용 타당성을 평가하는데 그 목적이 있다. 연구결과, RELAP5 코드는 CANDU형 발전소의 주증기관 파단사고를 잘 모사하고 있으며, CANDU형 발전소의 사고해석 검증용 코드로서 적절함을 보여주고 있다.

• Proceedings of the KSME Conference
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• 2007.05a
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• pp.1149-1154
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• 2007

Fluctuation of control rod was experienced when plant was operating in normal operation mode in WH type NPPs. In order to cope with increased control rod fluctuation, the lead-lag controller setpoint for rod control system was optimized and resulted in increasing the margin of operation and minimizing unnecessary control rod movement. By optimization of the time constant, the margin of operation was increased by $1.5^{\circ}F$ and the control rod movement was not occurred due to mitigation of temperature fluctuation in loop. According to the mitigation of time constant, the margin of operation was increased but safety margin can be affected badly, so that the influences to FSAR design reference was evaluated. As the result of this evaluation, it satisfied the design reference of the existing safety analysis and was applied to NPP after obtaining the approval.

• Nuclear Engineering and Technology
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• v.29 no.5
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• pp.375-384
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• 1997

Relative power density distributions of the Kori Unit 1 pressurized water reactor are calculated by Monte Carlo modeling with the MCNP code. The Kori Unit 1 core is modeled on a three-dimensional representation of the one-eighth of the reactor in-vessel component with reflective boundaries at 0 and 45 degrees. The axial core model is based on half core symmetry and is divided into four axial segments. Fission reaction density in each rod is calculated by following 100 cycles with 5,000 test neutrons in each cycle after starling with a localized neutron source and ten noncontributing settle cycles. Relative assembly power distributions are calculated from fission reaction densities of rods in assembly. After 100 cycle calculations, the system converges to a k value of 1.00039 $\geq$ 0.00084. Relative assembly power distribution is nearly the same with that of the Kori Unit 1 FSAR. Applicability of the full-scope Monte Carlo simulation in the power distribution calculation is examined by the relative root moan square error of 2.159%.

• Journal of IKEEE
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• v.7 no.1 s.12
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• pp.107-118
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• 2003

In this paper, dynamic control model is formulated by considering the geometrical structure of the deaerator storage tank in nuclear power plant and input-output flow rate at steady state, and we describe fault detection and diagnosis (FDD) scheme based on the adaptive estimator. The performance and effectiveness of the proposed FDD scheme are evaluated by applying real operating data obtained from the YOUNGKWANG 3 & 4 FSAR.

• Journal of the Korean Society of Safety
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• v.21 no.5 s.77
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• pp.77-83
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• 2006

The purpose of this study is to assess the organizational personality types, employee-organization fits and the job satisfaction/involvement in a Korea standard nuclear power plant(NPP), which is a representative safety work place. First we chose 427 procedures that are related to safety out of 777 officially managed procedures referenced by 13.5 of FSAR(final safety analysis report). Next, we finally chose 70 procedures of 8 divisions for 44 employees regarding the duties for NPPs' division, experiences of operations, an operational know-how, and the indication of operational weakness. This study used OPTI(organizational personality type indicators) and the combination of 4 preference types for determining the organizational personality to produce personality types of organizations for NPPs' division. To assess the job satisfaction and involvement, we used a questionnaire and an interview, for 300 employees(83.5%) of the Korea standard NPP.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.196-199
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• 2005

The centrifuge system in liquid radwaste system(LRS) is composed of several skids including Decanter and Separator. The decanter separates the sludge over 5mm in size within liquid radwaste by centrifuge force and drops it into 55gallon drum. The separator separates the sludge over $0.1{\mu}m$ in size within the liquid radwaste processed by Decanter by centrifuge force. The process of separating the sludge from the LRS keeps the resin in Ion Exchanger from being damaged and improves the performance of Ion Exchanger, and satisfies the decontamination factor suggested in Uljin 5,6 FSAR to safety discharge into the outer environment.