• Nuclear Engineering and Technology
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• 제18권2호
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• pp.97-106
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• 1986

1981년 6일 9일 원자력 1호기에서 발생한 77.5% 출력상태에서의 외부전원상실사고를 열, 수력학적최적계산용 코드인 RELAP5/MODl/NSC를 사용하여 모의하였으며 해석결과는 발전소 실측자료와 잘 일치하였다. 원자로 냉각재펌프의 트립에 따른 flow coastdown후에 hot-cold leg온도차에 의하여 자연순환 유동이 형성됨이 확인되었으며 실측자료와 잘 일치하여 이와 관련된 전산코드의 열수력학 적모델의 타당성을 입증할 수 있었다. 또한 위의 사고전개가 정상운전상태인 전출력(100%)에서 재발하였을 경우를 가정하여 해석하였다. 이러한 해석을 통하여 보조급수의 공급과 더불어 증기발생기 PORV의 적절한 작동으로 원자력 1호기 노심잔열을 제거하여 안전성에 문제점을 야기하지 않음을 입증하였다. 최적 계산방법에 의한 사고해석에서는 turbine stop valve 작동시간, 증기 발생기 PORV 설정치 등 non-safety 관련요소들의 특성에 대한 정화한 모의가 필수적이다.

• 해양환경안전학회지
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• 제26권6호
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• pp.723-731
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• 2020

해양플랜트는 발주처와 선급에서 요구하는 다양한 항목들을 설계할 시에 반영하여야 한다. 특히, 해양구조물에 탑재되는 Topside Module의 경우 육상플랜트와는 다르게 공간적 제약이 크고 구조물의 움직임과 같은 해상 환경조건 및 안전과 관련된 요구사항들이 많아 그 설계 과정이 매우 까다롭다. 본 연구에서는 Topside Module에 들어가는 주요장비 중 하나인 HPU(Hydraulic Power Unit) 구조물에 작용하는 하중을 DNVGL 규칙에 따라 계산하고, 각 하중조건에 따른 구조안전성 평가를 진행하였고 개발된 제품의 구조 신뢰성을 향상하고자 하였다. 구조해석은 범용프로그램인 MSC 소프트웨어를 사용하였고, 총 5가지 하중 조건으로 구조해석을 진행하여 다양한 움직임에 대한 안전성을 검토하였다. 그 결과 선미 방향 Pitching 상태(Load Case 5)에서 최대 응력이 발생하였고, 응력 수준은 허용응력의 약 85 % 수준이고, 최대변위는 허용치의 약 5 % 수준으로 구조안전성이 확인되었으며 부재 간 간섭은 발생하지 않았다.

• Smart Structures and Systems
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• 제11권2호
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• pp.199-215
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• 2013

During the service life of a structure, design complications and unexpected events may induce unforeseen vibrations. These vibrations can be generated by malfunctioning machinery or machines that are modified or placed without considering the original structural design because of a change in the intended use of the structure. Significant vibrations occurred at a natural gas plant cooling structure during its operation due to cavitation effect within the hydraulic system. This study presents findings obtained from the in-situ vibration measurements and following finite-element analyses of the cooling structure. Comments are made on the updated performance level and damage state of the structure using the results of these measurements and corresponding numerical analyses. An attempt was also made to assess the applicability of traditional displacement-based vulnerability estimation methods in the health monitoring of structures under vibrations with a character different from those due to seismic excitations.

• 한국안전학회지
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• 제35권5호
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• pp.121-127
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• 2020

Since the Fukushima nuclear accident in 2011, the public were concerned about the safety of Nuclear Power Plants (NPPs) in extreme natural disaster situations, such as earthquakes, flooding, heavy rain and tsunami, have been increasing around the world. Accordingly, the Stress Test was conducted in Europe, Japan, Russia, and other countries by reassessing the safety and response capabilities of NPPs in extreme natural disaster situations that exceed the design basis. The extreme natural disaster can put the NPPs in beyond-design-basis conditions such as the loss of the power system and the ultimate heat sink. The behaviors and capabilities of NPPs with losing their essential safety functions should be measured to find and supplement weak areas in hardware, procedures and coping strategies. The Loss of Ultimate Heat Sink (LUHS) accident assumes impairment of the essential service water system accompanying the failure of the component cooling water system. In such conditions, residual heat removal and cooling of safety-relevant components are not possible for a long period of time. It is therefore very important to establish coping strategies considering all available equipment to mitigate the consequence of the LUHS accident and keep the NPPs safe. In this study, thermal hydraulic behavior of the LUHS event was analyzed using RELAP5/Mod3.3 code. We also performed the sensitivity analysis to identify the effects of the operator recovery actions and operation strategy for charging pumps on the results of the LUHS accident.

• Nuclear Engineering and Technology
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• 제18권1호
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• pp.9-16
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• 1986

1984년 11월 14일 원자력 1호기에서 발생된 주급수 상실사고에 대한 계통의 열수력학적인 거동을 모의·해석하고, 발전소 실측자료와의 비교를 통하여 사용된 전산코드의 신뢰도를 평가하였다. 모의된 열수력학적 변수들은 발전소 실측자료와 비교적 잘 일치하였으나 원자로 트립시에 증기발생기 증기유량과 주 냉각재 계통 평균온도에 있어서 약간의 차이를 보였다. 이는 원자로 트립시 깎은 시간에 급격한 노심 출력의 감소로 인하여 열·수력학적 변수들에 큰 변화를 야기하여 발전소 실측자료가 과도상태에서의 불학실성을 내포하기 때문으로 예측되었다. 해석에 사용된 전산코드는 RELAP5/MOD1/CY018로부터 불합리한 oscillation을 일으키는 interphase drag 및 wall heat transfer model의 수정을 통하여 개발된 RELAP5/MOD1/NSC이다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.404-410
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• 1998

In Yonggwang nuclear units 3 and 4 currently under commercial operation, the cold leg temperature was very close to the technical specification limit of 298$^{\circ}C$ during initial startup testing, which was caused by the higher-than-expected reactor coolant system flow. Accordingly, the reference temperature (Tref) program needed to be revised to allow more flexibility for plant operations. In this study, the method of a specific test performed at Yonggwang nuclear unit 4 to revise the Tref program was described and the test results were discussed. In addition, the modified Tref program was evaluated on its potential impacts on system performance and safety. The methods of changing the Tref program and the associated pressurizer level setpoint program were also explained. Finally, for Ulchin nuclear unit 3 and 4 currently under initial startup testing, the effects of reactor coolant system flow rate on the coolant temperature were evaluated from the thermal hydraulic standpoint and an optimum Tref program was recommended.

• Journal of Mechanical Science and Technology
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• 제17권9호
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• pp.1380-1387
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• 2003

When a cold HPSI (High pressure Safety Injection) fluid associated with an overcooling transient, such as SGTR (Steam Generator Tube Rupture), MSLB (Main Steam Line Break) etc., enters the cold legs of a stagnated primary coolant loop, thermal stratification phenomena will arise due to incomplete mixing. If the stratified flow enters the downcomer of the reactor pressure vessel, severe thermal stresses are created in a radiation embrittled vessel wall by local overcooling. As general thermal-hydraulic system analysis codes cannot properly predict the thermal stratification phenomena, RG 1.154 requires that a detailed thermal-mixing analysis of PTS (pressurized Thermal Shock) evaluation be performed. Also. previous PTS studies have assumed that the thermal stratification phenomena generated in the stagnated loop side of a partially stagnated primary coolant loop are neutralized in the vessel downcomer by the strong flow from the unstagnated loop. On the basis of these reasons, this paper focuses on the development of a 3-dimensional thermal-mixing analysis model using PHOENICS code which can be applied to both partial and total loop stagnated cases. In addition, this paper verifies the fact that, for partial loop stagnated cases, the cold plume generated in the vessel downcomer due to the thermal stratification phenomena of the stagnated loop is almost neutralized by the strong flow of the unstagnated loop but is not fully eliminated.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(2)
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• pp.1001-1007
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• 1995

A Comprehensive Vibration Assessment Program (CVAP) has been performed for Yonggwang Nuclear Power Plant Unit 4 (YGN 4) in order to verify the structural integrity of the reactor internals for flow induced vibrations prior to commercial operation. The theoretical evidence for the structural integrity of the reactor internals and the basis for measurement and inspection are provided by the analysis. Flow induced hydraulic loads and reactor internals vibration response data were measured during pre-core hot functional testing in YGN 4 site. Also, the critical areas in the reactor internals were inspected visually to check any existence of structural abnormality before and after the pre-core hot functional testing. Then, the measured data have been analyzed and compared with the predicted data by analysis. The measured stresses are less than the predicted values and the allowable limits. It is concluded that the vibration response of the reactor internals due to the flow induced vibration under normal operation is acceptable for long term operation.

• Nuclear Engineering and Technology
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• 제53권12호
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• pp.3892-3901
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• 2021

The pressure drop of a moisture separator in a steam generator is the important design parameter to ensure the successful performance of a nuclear power plant. The moisture separators have a wide range of operating conditions based on the arrangement of them. The prediction of the pressure drop in a moisture separator is challenging due to the complexity of the multi-dimensional two-phase vortex flow. In this study, the moisture separator test facility using the air/water two-phase flow was used to predict the pressure drop of a moisture separator in a Korean OPR-1000 reactor. The prototypical steam/water two-phase flow conditions in a steam generator were simulated as air/water two-phase flow conditions by preserving the centrifugal force and vapor quality. A series of experiments were carried out to investigate the effect of hydraulic characteristics such as the quality and liquid mass flux on the two-phase pressure drop. A new prediction model based on the scaling law was suggested and validated experimentally using the full and half scale of separators. The suggested prediction model showed good agreement with the steam/water experimental results, and it can be extended to predict the steam/water two-phase pressure drop for moisture separators.

• 한국압력기기공학회 논문집
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• 제17권2호
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• pp.75-82
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• 2021

The 3D printing technology is one of the fourth industrial revolution technology that drives innovation in the manufacturing process, and should be applied to nuclear industry for various purposes according to the manufacturing trend change. In nuclear industry, it can be applied to manufacture obsolete items and new designed parts in advanced reactors or small modular reactors (SMRs), replacing the traditional manufacturing technologies. A gate valve body was manufactured, which was obsolete in nuclear power plant, using DED(Directed Energy Deposition) metal 3D printing technology after restoring design characteristics including 3D design drawing by reverse engineering. The 3D printed valve body was assembled with commercial parts such as seat-ring, disk, stem, and actuator for performance test. For the valve assembly, including 3D printed valve body, several tests were performed, including pressure test, end-loading test, and seismic test according to KEPIC MGG and KEPIC MFC. In the pressure test, hydraulic pressure of 391kgf/cm² was applied to 3D printed valve body, and no leak was detected. Also the 3D printed valve assembly was performed well in end-loading and seismic tests.