• Nuclear Engineering and Technology
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• 제50권8호
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• pp.1412-1420
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• 2018

An experiment was performed using the large-scale test facility (LSTF), which simulated a 1% vessel upper head small-break loss-of-coolant accident with an accident management (AM) measure under an assumption of total-failure of high-pressure injection (HPI) system in a pressurized water reactor (PWR). In the LSTF test, liquid level in the upper head affected break flow rate. Coolant was manually injected from the HPI system into cold legs as the AM measure when the maximum core exit temperature reached 623 K. The cladding surface temperature largely increased due to late and slow response of the core exit thermocouples. The AM measure was confirmed to be effective for the core cooling. The RELAP5/MOD3.3 code indicated insufficient prediction of primary coolant distribution. The author conducted uncertainty analysis for the LSTF test employing created phenomena identification and ranking table for each component. The author clarified that peak cladding temperature was largely dependent on the combination of multiple uncertain parameters within the defined uncertain ranges.

• Nuclear Engineering and Technology
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• 제43권2호
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• pp.167-174
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• 2011

A pressurizer in a small integral type pressurized water reactor is located inside the upper region of the reactor vessel, and uses a space between the upper head of the reactor vessel and the upper region of the upper guide structure which is partially filled with fluid depending on the operating power. This new design requires a comprehensive investigation of vibration characteristics. This study investigates the modal characteristics of a pressurizer which uses a simplified cylindrical shell model, focusing on how having fluid in the shell affects vibration and response characteristics. In addition, an analysis of sloshing is performed and the response characteristics are addressed.

• 대한기계학회논문집A
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• 제36권10호
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• pp.1235-1239
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• 2012

원자력발전소의 원자로압력용기 상부헤드에는 출력제어 및 정지용 제어봉이 통과하는 노즐이 있으며 이 노즐은 상부헤드 노즐과 J 형태의 홈으로서 용접 되어 있다. 최근 외국의 원자력발전소에서 이 용접영역 주변의 노즐 및 용접부에서 일차수응력부식 균열이 발생한 사례가 보고되고 있다. 본 논문에서는 이 용접부의 용접잔류응력과 운전 중에서의 응력상태를 유한요소해석을 이용하여 평가함으로써 고응력 위치를 확인하고 응력관점에서 균열발생 가능성이 높은 지역을 예측하고자 하였다. 해석결과 용접에 의해서 형성된 잔류응력이 수압시험과 운전조건에 의해 다소 변동되기는 하나 응력분포형태는 큰 변화가 없었다. 전반적으로 노즐내면에서는 용접이 시작되는 지점 주변에서 최대 인장응력이 형성되고 노즐외면에서는 용접이 끝나는 지점 주변에서 최대인장응력이 형성되는 것을 확인하였다.

• 대한기계학회논문집A
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• 제38권6호
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• pp.637-647
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• 2014

국내 가압경수로형 원자로의 압력용기 상부헤드에는 많은 제어봉구동장치(CRDM) 노즐이 분포한다. 이들 노즐은 억지끼워맞춤(Shrink fitting) 방식으로 결합되어 용접 처리 된다. 용접에 의해 발생되는 인장잔류응력은 일차수응력부식균열을 발생시키는 주요 요인이다. 이러한 이유로 최근 15 여 년 동안 관통노즐 용접부 부위에서 균열 발생 사례가 증가하고 있으며, 이를 극복하기 위해 다양한 방안이 모색되고 있다. 또한 용접과정에서 발생되는 불필요한 결함은 일차수응력부식균열(PWSCC)을 가속화 시키는 원인이 되기도 한다. 원자로 제작과정에서 용접에 의한 결함은 보수용접에 의해 즉시 수리가 이루어 진다. 기존의 연구에서는 정상적인 용접과정에서 발생되는 잔류응력을 예측하였으나, 본 연구에서는 용접과정에서 발생되는 결함을 보수하기 위해 실시되는 보수용접이 용접잔류응력에 미치는 영향을 분석하였다.

• 한국압력기기공학회 논문집
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• 제9권1호
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• pp.56-61
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• 2013

This paper presents the results of integrity assessment for the cracks happened in reactor vessel upper head penetration nozzles. The crack morphology for a boat sample from crack area was analyzed through microscope. The stress condition including weld residual stress around crack was analyzed using finite element analysis. From the results of crack morphology and stress condition, the crack was concluded as primary water stress corrosion cracking. The integrity of the cracked nozzle was assessed by the methodology provided in ASME Section XI. According to the assessment results, the remaining life of the cracked nozzle was 1.43 yrs. and the plant decided to repair it.

• 대한기계학회논문집A
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• 제37권9호
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• pp.1159-1168
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• 2013

가압형 경수로 원자로의 압력용기 상부헤드 관통노즐 J-groove 용접부 주변에서 일차수응력부식균열(PWSCC)로 인한 냉각수 누설사례가 발생하고 있다. 본 연구에서는 PWSCC 의 주요 원인 중 하나인 용접 잔류응력을 유한요소 해석을 이용해 평가하고 원자력 발전소의 정상가동 조건을 해석에 반영하는 방법이 용접잔류응력 분포에 미치는 영향에 대한 분석을 수행하였다. 또한 반복되는 원자력 발전소의 가동 주기가 용접잔류응력 분포에 미치는 영향을 확인하여 정상가동조건에서의 정확한 용접 잔류응력을 예측할 수 있는 방법을 분석하였다.

• Nuclear Engineering and Technology
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• 제49권7호
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• pp.1524-1536
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• 2017

During the fabrication process of reactor vessel head penetration (RVHP), the grain size of the tube material can be changed by hot or cold work and the inner side of the tube can also be shrunk due to welding outside of the tube. Several nonregular time-of-flight diffraction (TOFD) signals were found because of deformed grains. In this paper, an investigation of nonregular TOFD indications acquired from RVHP tubes using experiments and computer simulation was performed in order to identify and distinguish TOFD signals by coarse grains from those by Primary Water Stress Corrosion Crack (PWSCC). For proper understanding of the nonregular TOFD indications, microstructural analysis of the RVHP tubes and prediction of signals scattered from the grains using Finite Element Method (FEM) simulation were performed. Prediction of ultrasonic signals from the various sizes of side drilled holes to find equivalent flaws, determination of the size of the nonregular TOFD indications from the coarse grains, and experimental investigation of TOFD signals from coarse grain and shrinkage geometry to identify PWSCC signals were performed. From the computer simulation and experimental investigation results, it was possible to obtain the nonregular TOFD indications from the coarse grains in the alloy 690 penetration tube of RVHP; these nonregular indications may be classified as PWSCC. By comparing the computer simulation and experimental results, we were able to confirm a clear difference between the coarse grain signal and the PWSCC signal.

• Nuclear Engineering and Technology
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• 제56권6호
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• pp.2002-2010
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• 2024

Korea Atomic Energy Research Institute (KAERI) has operated an integral effect test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), with reference to the APR1400 (Advanced Power Reactor 1400) for tests for transient and design basis accidents simulation. A test for a loss of coolant accident (LOCA) at the top of the reactor pressure vessel (RPV) had been conducted at ATLAS to address the impact of the loss of safety injections (LSI) and to evaluate accident management (AM) actions during the postulated accident. The experimental data has been utilized to validate system analysis codes within a framework of the domestic standard problem program organized by KAERI in collaboration with Korea Institute of Nuclear Safety. In this study, the test has been analyzed by using thermal-hydraulic system analysis codes, MARS-KS 1.5 and TRACE 5.0 Patch 6, and a comparative analysis with experimental and calculation results has been performed. The main objective of this study is the investigation of the thermal-hydraulic phenomena during a small break LOCA at the RPV upper head with the LSI as well as the predictability of the system analysis codes after the AM actions during the test. The results from both codes reveal that overall physical behaviors during the accident are predicted by the codes, appropriately, including the excursion of the peak cladding temperature because of the LSI. It is also confirmed that the core integrity is maintained with the proposed AM action. Considering the break location, a sensitivity analysis for the nodalization of the upper head has been conducted. The sensitivity analysis indicates that the nodalization gave a significant impact on the analysis result. The result emphasizes the importance of the nodalization which should be performed with a consideration of the physical phenomena occurs during the transient.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(2)
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• pp.574-579
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• 1995

Saturated boiling heat transfer experiment in a hemispherical narrow space is conducted using Freon-113 to investigate an additional heat removal capability through a hypothetical gap between lower head and degraded core. The narrow space of 1mm consists of a 124mm diameter heated stainless steel hemisphere and a glass outer vessel. Within the hemispherical narrow space large coalesced bubbles are produced and these bubbles rise in random direction, causing liquid flow in from the opposite side to fill the region. Such flow in random direction makes the flow field in the narrow space very chaotic and thus enhance heat transfer. The heat transfer coefficient is higher at lower angle and at higher heat flux. The present study shows that the liquid from upper region can effectively penetrate into the gap and augment the heat removal capability through tile gap.

• 대한기계학회논문집A
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• 제35권10호
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• pp.1337-1345
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• 2011

가압경수로형 원자로의 원자로압력용기 상부헤드에는 많은 제어봉구동장치(CRDM) 노즐이 분포한다. 최근 10 여 년 동안 제어봉구동장치 alloy 600 CRDM 노즐에서 균열 발생 사례가 증가하고 있으며, 이는 용접과 연관성이 매우 깊은 것으로 알려져 있다. CRDM 노즐에서 발생하는 축 및 원주방향 균열은 유럽과 미국의 원자력 발전소에서 발견되었으며, 사고의 원인은 용접 잔류응력 및 작용하중에 기인하는 일차수응력부식균열(PWSCC)임이 확인되었다. 이러한 이유로 본 연구에서는 유한요소해석을 통해 한국형 원자로의 CRDM 관통 노즐 용접부를 대상으로 용접 잔류응력을 예측하였으며, 특히, 관통노즐의 위치와 형상, 용접부 필렛 형상 및 인접노즐 용접에 의한 영향을 분석하였다.