• Nuclear Engineering and Technology
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• 제23권3호
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• pp.340-351
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• 1991

l/5.03 축소 원자로 모델을 이용하여 원자력 발전소 영광 3,4호기를 위한 유동시험을 수행하였다. 이 유동 시험의 목적은 ABB-CE사의 System 80과 영광 3,4호기 원자로 크기의 상대적인 차이로 인해 발생하는 원자로 용기내의 수력학적 영향을 평가하는 것이다. 유동 모델은 상사성 원리에 따라 설계하였다. 이 시험에서 얻은 결과는 노심 입구 유량 분포, 노심 출구 압력 분포, 원자로 입구 노즐에서부터 출구 노즐까지 유동로를 따른 부분 구간 및 전체 압력 손실이다. 이 데이터들은 노심의 열적 여유도 분석에 필요한 입력 자료 제공과 해석적 수력설계 방법의 검증에 이용하게 된다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 추계학술발표회논문집(1)
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• pp.395-400
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• 1996

Scoping analyses for the Safety Injection System (SIS) configuration for Korean Next Generation Reactor (KNGR) are peformed in this study. The KNGR SIS consists of four mechanically separated hydraulic trains. Each hydraulic train consisting of a High Pressure Safety Injection (HPSI) pump and a Safety Injection Tank (SIT) is connected to the Direct Vessel Injection (DVI) nozzle located above the elevation of cold leg and thus injects water into the upper portion of reactor vessel annulus. Also, the KNGR is going to adopt the advanced design feature of passive fluidic device which will be installed in the discharge line of SIT to allow more effective use of borated water during the transient of large break LOCA. To determine the feasible configuration and capacity of SIT and HPSI pump with the elimination of the Low Pressure Safety Injection (LPSI) pump for KNGR, licensing design basis evaluations are performed for the limiting large break LOCA. The study shows that the DVI injection with the fluidic device SIT enhances the SIS performance by allowing more effective use of borated water for an extended period of time during the large break LOCA.

• Nuclear Engineering and Technology
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• 제56권7호
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• pp.2458-2473
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• 2024

The turbulent flow in reactor pressure vessel (RPV) of pressurized water reactor (PWR) is important for the flow rate distribution at core inlet. Thus, it is vital to study the turbulent flow phenomena in RPV. However, the complicated fluid channel consisted of inner structures of RPV will block or refract the laser sheet of particle image velocimetry (PIV). In this work, the matched index of refraction (MIR) of sodium iodide (NaI) solution and acrylic was applied to support optical path for flow field measurements by PIV in the 1/10th scaled-down RPV model. The experimental results show detailed velocity field at different locations inside the scaled-down RPV model. Some interesting phenomena are obtained, including the non-negligible counterflow at the corner of nozzle edge, the high downward flowing stream in downcomer, large vortices above vortex suppression plate in lower plenum. And the intensity of counterflow and the strength of vortices increase as inlet flow rate increasing. Finally, the case of asymmetry flow was also studied. The turbulent flow has different pattern compared with the case of symmetrical inlet flow rate, which may affect the uniformity of flow distribution at the core inlet.

• 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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• 제6권2호
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• pp.34-41
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• 2010

This work is concerned with the analysis of stress corrosion cracking(SCC) behavior of Alloy 600 nozzle penetration mock-up according to a residual stress induced by a dissimilar metal welding(DMW) in a nuclear reactor pressure vessel. The effects of the dimension and materials of the nozzle penetration on the deformation and the residual stress induced by DMW were investigated using a finite element analysis(FEA). The inner diameter(ID) change of the nozzle by DMW and its dependance on the design variables, calculated by FEA, were well consistent with those measured from the mock-up. Accelerated SCC tests were performed for three mock-ups with different wall thicknesses in a highly acidic solution to investigate mainly the effect of the residual stress on the SCC behavior of Alloy 600 nozzle. From a destructive examination of the mock-up after the tests, the SCC behavior of the nozzle was fairly related with the residual stress induced by DMW : axial cracks were found in the ID surface of the nozzle within the J-weld region where the highest tensile hoop stress was predicted by FEA, while circumferential cracks were observed beyond both J-weld root and toe where the highest tensile axial stress was expected.

• 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.

• Nuclear Engineering and Technology
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• 제23권2호
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• pp.183-199
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• 1991

가압수형 원자로 가압기의 오동작 보조살수 과도시 용기벽에서의 온도분포에 대한 과도응답을 해석하였으며, 해석은 분무수적으로 젖게되는 용기벽면에서 나타나는 열응력에 대하여 보수적으로 수행되었다. 수적이 부딪혀서 흘러내리는 용기벽의 내부경계면에서 강제대류열전달계수를 결정하기 위하여, 분무수적들이 살수노즐을 떠나 수증기와 비응축성인 수소기체로 이루어진 혼합기로 채워져 있는 가압기 내부공간을 통하여 비행한 후에 용기내부벽면에 도달할 때의 수적들의 과도온도를 예측하였다. 용기벽에서의 과도온도분포는 유한요소법을 사용하여 구하였으며, 대표적인 결과들을 제시하였다. 열해석의 결과는 입력자료에 대한 묘사와 부합되며, 타당한 물리적 의미를 가짐 이 확인되었다.

• Nuclear Engineering and Technology
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• 제44권7호
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• pp.727-734
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• 2012

A new advanced safety feature of DVI+ (Direct Vessel Injection Plus) for the APR+ (Advanced Power Reactor Plus), to mitigate the ECC (Emergency Core Cooling) bypass fraction and to prevent switching an ECC outlet to a break flow inlet during a DVI line break, is presented for an advanced DVI system. In the current DVI system, the ECC water injected into the downcomer is easily shifted to the broken cold leg by a high steam cross flow which comes from the intact cold legs during the late reflood phase of a LBLOCA (Large Break Loss Of Coolant Accident)For the new DVI+ system, an ECBD (Emergency Core Barrel Duct) is installed on the outside of a core barrel cylinder. The ECBD has a gap (From the core barrel wall to the ECBD inner wall to the radial direction) of 3/25~7/25 of the downcomer annulus gap. The DVI nozzle and the ECBD are only connected by the ECC water jet, which is called a hydrodynamic water bridge, during the ECC injection period. Otherwise these two components are disconnected from each other without any pipes inside the downcomer. The ECBD is an ECC downward isolation flow sub-channel which protects the ECC water from the high speed steam crossflow in the downcomer annulus during a LOCA event. The injected ECC water flows downward into the lower downcomer through the ECBD without a strong entrainment to a steam cross flow. The outer downcomer annulus of the ECBD is the major steam flow zone coming from the intact cold leg during a LBLOCA. During a DVI line break, the separated DVI nozzle and ECBD have the effect of preventing the level of the cooling water from being lowered in the downcomer due to an inlet-outlet reverse phenomenon at the lowest position of the outlet of the ECBD.

• 대한용접접합학회:학술대회논문집
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• 대한용접접합학회 1996년도 특별강연 및 추계학술발표 개요집
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• pp.19-34
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• 1996

A review was made of fatigue failures of nuclear power plant components in Japan, which were experienced in service and during periodical inspection. No case has been recently reported of a service fatigue failure of a reactor pressure vessel itself, excluding nozzle corner cracks, that occurred many years ago. But, service fatigue failures have been occasionally experienced in piping systems, pumps, and valves, on which fatigue design seems to have been inadequately applied. The causes of fatigue failures can be divided into two categories: mechanical-vibration-induced fatigue and thermal-fluctuation-induced fatigue. Vibration-induced fatigue failure occurs more frequently than is generally thought. The lesson gleaned from the present survey is a recognition that a service fatigue failure may occur due to any one or a combination of the following factors: (1) lack of communication between designers and fabrication engineers, (2) lack of knowledge about a possibility of fatigue failure and poor consideration about the effects of residual stresses, (3) lack of consideration on possible vibration in the design and fabrication stages, and (4) lack of fusion or poor penetration in a welded joint.

• Nuclear Engineering and Technology
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• 제49권7호
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• pp.1547-1554
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• 2017

VESTA (Verification of Ex-vessel corium STAbilization) and VESTA-S (-small) test facilities were constructed at the Korea Atomic Energy Research Institute in 2010 to perform various corium melt experiments. Since then, several tests have been performed for the verification of an ex-vessel core catcher design for the EU-APR1400. Ablation tests of an impinging $ZrO_2$ melt jet on a sacrificial material were performed to investigate the ablation characteristics. $ZrO_2$ melt in an amount of 65-70 kg was discharged onto a sacrificial material through a well-designed nozzle, after which the ablation depths were measured. Interaction tests between the metallic melt and sacrificial material were performed to investigate the interaction kinetics of the sacrificial material. Two types of melt were used: one is a metallic corium melt with Fe 46%, U 31%, Zr 16%, and Cr 7% (maximum possible content of U and Zr for C-40), and the other is a stainless steel (SUS304) melt. Metallic melt in an amount of 1.5-2.0 kg was delivered onto the sacrificial material, and the ablation depths were measured. Penetration tube failure tests were performed for an APR1400 equipped with 61 in-core instrumentation penetration nozzles and extended tubes at the reactor lower vessel. $ZrO_2$ melt was generated in a melting crucible and delivered down into an interaction crucible where the test specimen is installed. To evaluate the tube ejection mechanism, temperature distributions of the reactor bottom head and in-core instrumentation penetration were measured by a series of thermocouples embedded along the specimen. In addition, lower vessel failure tests for the Fukushima Daiichi nuclear power plant are being performed. As a first step, the configuration of the molten core in the plant was investigated by a melting and solidification experiment. Approximately 5 kg of a mixture, whose composition in terms of weight is $UO_2$ 60%, Zr 10%, $ZrO_2$ 15%, SUS304 14%, and $B_4C$ 1%, was melted in a cold crucible using an induction heating technique.