• 응용통계연구
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• 제29권7호
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• pp.1385-1398
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• 2016

원자력 발전소의 가압기는 1차 계통의 냉각재가 고온에서도 기화되지 않도록 압력을 가해주는 장치이다. 즉, 가압기의 고장은 원자력 발전소에 큰 영향을 미칠 수 있으며, 따라서, 가압기의 조기 고장 감지는 원자력 발전소의 안전에 매우 중요하다. 이를 위해, 본 연구에서는 마할라노비스 거리 개념과 다구찌 품질 공학 이론에 기반한 패턴 분류 인식 알고리즘 중 하나인 마할라노비스 다구찌 시스템(MTS)을 가압기 압력 제어 계통의 조기 고장 감지에 적용하였다. MTS의 고장 감지 성능을 검증하기 위해, 실제 원자력 발전소에서 발생하고 있는 가압기 압력전송기 고장 시나리오를 대상으로 하여, Full Scope 시뮬레이터를 통해 모사된 데이터에 적용하였다. 실험 결과, MTS는 단일 센서모니터링 기반의 전통적인 고장 감지에 비하여 매우 빠르게 고장을 감지할 수 있었다.

• 제어로봇시스템학회:학술대회논문집
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• 제어로봇시스템학회 1997년도 한국자동제어학술회의논문집; 한국전력공사 서울연수원; 17-18 Oct. 1997
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• pp.1696-1699
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• 1997

The robot mainpulator for inspection of pressurizer in the nuclear power plant has been developed, which consists of four parts : 2 arms, movable gripper, base frame, contorl console. To extract the damaged electric heating rod inside pressurizer, the gripper has been developed using wire lope and self-locking mechanism. for the examination of the structural stability of the robot manipulator, stress analysis is performed by using the ANSYS code.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.310-315
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• 1997

System-Integrated Modular Advanced Reactor (SMART), where major primary components such as modular helical steam generator and self regulating pressurizer are integrated into reactor vessel, is currently under development. The pressurizer is designed to control the primary pressure mainly with partial pressure of nitrogen gas and to maintain the fluid temperature as low as possible for the purpose of minimizing steam contribution. The steam generator (SG) is designed to produce super-heated steam inside tube at power operation. Because the in-vessel pressurizer and in-vessel SG are classified as the characteristic components of SMART, it is important to perform a steady state calculation of these components in order to evaluate the adoption of these components. A steady state analysis of the in-vessel pressurizer and in-vessel SG has been performed under normal power operation and the results show an acceptable performance of the components.

• 한국에너지공학회:학술대회논문집
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• 한국에너지공학회 1993년도 추계학술발표회 초록집
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• pp.103-108
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• 1993

In nuclear power plants using pressurized water as the main coolant, it is necessary to maintain system pressure within operational range. During transients, the coolant shrinks and expands causing insurge and outsurge of coolant in the pressurizer. In CANDU system, the pressure is controlled mainly by the pressurizer/degasser-condenser system. In CANDU system, the control of heat transport system pressure is achieved by giving heat to the pressurizer by activating the heaters to compensate a diminution in pressure or by removing heat from the pressurizer by bleeding steam to the degasser-condenser to compensate an increase in pressure. This study aims at developing a theoretical model capable to simulate various operational transients in the CANDU primary heat transport system (PHTS), applicable to CANDU engineering simulator on real time basis.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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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.

• 제어로봇시스템학회:학술대회논문집
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• 제어로봇시스템학회 2005년도 ICCAS
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• pp.2307-2310
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• 2005

A pressurizer is one of the major safety related equipment of nuclear power plants. In order to inspect and maintain the internal structures of a pressurizer, jumpers have to enter the pressurizer, in spite of the high dose exposure. Therefore, a tele-operated robotic manipulator has been developed, which consists of four parts with 2DOFs arms, a gripper, base frame, and control console. The task of this robotic manipulator is to extract the damaged electric heaters and inspect the internal structure of the pressurizer. The gripper hanging from the manipulator approaches the heaters and extracts the damaged heater by using a self-locking mechanism. In order to investigate the structural stability of the robotic manipulator, a stress analysis has been performed by using the ANSYS code. The results of this paper include the position control and vibration control of robotic gripper and the development of processing visual information for a vision sensor.

• Nuclear Engineering and Technology
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• 제45권3호
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• pp.347-360
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• 2013

This paper presents an experimental investigation of the small-break loss-of-coolant accident (SBLOCA) and the loss-of-feedwater accident (LOFW) in a scaled integral test facility of REX-10. REX-10 is a small integral-type PWR in which the coolant flow is driven by natural circulation, and the RCS is pressurized by the steam-gas pressurizer. The postulated accidents of REX-10 include the system depressurization initiated by the break of a nitrogen injection line connected to the steam-gas pressurizer and the complete loss of normal feedwater flow by the malfunction of control systems. The integral effect tests on SBLOCA and LOFW are conducted at the REX-10 Test Facility (RTF), a full-height full-pressure facility with reduced power by 1/50. The SBLOCA experiment is initiated by opening a flow passage out of the pressurizer vessel, and the LOFW experiment begins with the termination of the feedwater supply into the helical-coil steam generator. The experimental results reveal that the RTF can assure sufficient cooldown capability with the simulated PRHRS flow during these DBAs. In particular, the RTF exhibits faster pressurization during the LOFW test when employing the steam-gas pressurizer than the steam pressurizer. This experimental study can provide unique data to validate the thermal-hydraulic analysis code for REX-10.

• 대한기계학회논문집A
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• 제21권8호
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• pp.1259-1269
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• 1997

The procedure for the determination of the residual life of the pressurizer surge line nozzle in the nuclear plant is developed. The design fatigue life for the 1800 $ft^3$ pressurizer surge line nozzle in cast head design is compared with that of Westinghouse stress report, and the percentage difference between two results is less than 9%. The design fatigue life evaluation of the 1000 $ft^3$ pressurizer surge line nozzle in fabricated head design is carried out, and the consuming rate and residual life are estimated using the operating data.

• 한국안전학회지
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• 제33권5호
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• pp.164-169
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• 2018

A detailed fatigue evaluation procedure was developed to mitigate the excessive conservativeness of the conventional environmental fatigue evaluation method for the pressurizer spray line elbow of domestic new nuclear power plants. The pressurizer spray line is made of austenitic stainless steel, which is relatively sensitive to the environmentally assisted fatigue, and has a low degree of design margin in terms of environmentally assisted fatigue due to the thermal stratification phenomenon on the pipe cross section as a whole or locally. In this study, to meet the environmental fatigue design requirements of the pressurizer spray line elbow, the new environmental fatigue evaluation has been performed, which used the ASME Code NB-3200-based detailed fatigue analysis and the environmental fatigue correction factor instead of the existing NB-3600 evaluation method. As a result, the design requirements for environmentally assisted fatigue were met in all parts of the pressurizer spray line elbow including the fatigue weakened zones by thermal stratification.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2736-2747
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• 2022

The operations in the design lifecycle of a nuclear power plant targeted to be decommissioned lead to neutron activation. Operations in the decommissioning process include cutting, decontamination, disposal, and processing. Among these, cutting is done close to the target material, and thus workers are exposed to radiation. As there are only a few studies on pressurizers, there arises the need for further research to assess the radiation exposure dose. This study obtained the specifications of the AP1000 pressurizer of Westinghouse and the distribution of radionuclide inventory of a pressurizer in a pressurised water reactor for evaluation based on literature studies. A cutting scenario was created to develop an optimal method so that the cut pieces fill a radioactive solid waste drum with dimensions 0.571 m × 0.834 m. The estimated exposure dose, estimated using the tool VISIPLAN SW, in terms of the decontamination factor (DF) ranged from DF-0 to DF-100, indicating that DF-90 and DF-100 meet the ICRP recommendation on exposure dose 0.0057 mSv/h. At the end of the study, although flame cutting was considered the most efficient method in terms of cutting speed, laser cutting was the most reasonable one in terms of the financial aspects and secondary waste.