• 한국압력기기공학회 논문집
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• 제17권2호
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• pp.90-100
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• 2021

Molten corium-concrete interaction (MCCI) is one of the most important phenomena that can lead to the potential hazard of late containment failure due to basemat penetration during a severe accident. In this study, MCCI analytical models of the CORQUENCH code were prepared through verification calculations of several experiments, which had been performed using concrete types similar to those of the calandria vault floor in CANDU-6 plants. The behaviors of thermal-hydraulic variables related to MCCI phenomena were analyzed under the conditions of dry floor and water flooding during the severe accident stemming from a hypothetic station blackout. Uncertainty analyses on the ablation depth were also carried out. It was estimated that the concrete ablation was not interrupted due to the continuous MCCI process under the dry condition but was terminated within 24 hours under the water flooding condition. It was confirmed that the water flooding as a mitigating action was effective to achieve the quenching and thermal stabilization of the melt discharged from the calandria vessel, showing that the present models are capable of reasonably simulating MCCI phenomena in CANDU-6 plants. This study is expected to provide the technical bases to the accident management strategy during the late-phase severe accidents.

• 한국압력기기공학회 논문집
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• 제17권2호
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• pp.137-144
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• 2021

The mode superposition time history analysis method is commonly used in a seismic analysis. The maximum response in the time history analysis can be derived by combining the responses of individual modes. The residual mode response is the response of the modes which are not considered in the time history analysis. In this paper, the residual vector method to consider the residual mode response in the time history analysis is introduced and evaluated. Seismic analyses for a sample structure model and a reactor vessel model are performed to evaluate the residual vector method. The analysis results show that residual mode response is well calculated when the residual vector method is used. It is confirmed that the residual vector method is useful and acceptable to consider the residual mode response in a seismic analysis of the nuclear power plant equipment.

• 한국압력기기공학회 논문집
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• 제12권1호
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• pp.93-100
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• 2016

Since the sodium-cooled fast reactor is operated in a hostile environment due to the use of liquid sodium as its coolant, advanced techniques for in-service inspection are required to periodically verify the integrity of the reactor. This paper presents the development of in-service inspection techniques for Proto-type Generation IV Sodium-cooled Fast Reactor. First, the 10 m long plate-type ultrasonic waveguide sensor has been developed for in-service inspection of reactor internals, and its feasibility was verified through several under-water and under-sodium experiments. Second, the combined inspection system for in-service inspection of ferromagnetic steam generator tubes has been developed. The remote field eddy current testing and magnetic flux leakage testing can be conducted simultaneously by using the developed inspection system, and the detectability was demonstrated through several damage detection experiments. Finally, the electro-magnetic acoustic transducer which can withstand high temperature and be installable in the remote operated vehicle has been developed for in-service inspection of the reactor vessel, and its detectability was investigated through damage detection experiments.

• 비파괴검사학회지
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• 제5권2호
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• pp.22-33
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• 1986

The application and development of on-line monitoring technology of AE to surveillance of crack propagation will contribute to the structural integrity of reactor pressure vessel and piping system. This research has been performed in order to obtain the evaluation technology for source location of AE and the analysis for the AE signal of the welded specimen. AE is detected by 4-channels AE system during pressurization in small pressure vessels. The cracking of artificial defects can be accurately located and categorized in real time. The welded specimens have more events rate and higher amplitude than the weldless less specimens, and the events rate have a peak around the yield point and just before the failure under tensile test.

• 한국기계가공학회지
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• 제17권4호
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• pp.23-31
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• 2018

Stress of a steam separator, equipment of the high-pressure (HP) evaporator for a HRSG, was analyzed and evaluated according to ASME Boiler & Pressure Vessel Code Section VIII Division 2. First, from the analysis results of the piping system model of the HP evaporator, reaction forces of the riser tubes connected to the steam separator, i.e., nozzle loads, were derived. Next, a finite element model of the steam separator was constructed and analyzed for the design pressure and the nozzle loads. The results show that the maximum stress occurred at the bore of the riser nozzle. The primary membrane stresses at the shell and nozzle were found to be less than the allowable stress. Next, the steam separator was analyzed for the steady-state operating conditions of operating pressure, operating temperature, and nozzle loads. The maximum stress occurred at the bore of the riser nozzle. The primary plus secondary membrane plus bending stress at the shell and nozzle was found to be less than the allowable stress.

• 한국항해항만학회지
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• 제46권6호
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• pp.562-569
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• 2022

기존 운항선박에 적용되어 있는 알람 모니터링 기술은 온도, 압력 등의 데이터 항목을 AMS(Alarm Monitoring System)으로 관리하고 해당 센싱 데이터가 정상 수준 범위를 초과할 경우만 선원에게 알람을 제공한다. 또한 기존 선박의 정비는 PMS(Planned Maintenance System)를 따른다. 이는 장비로부터 측정된 센싱 데이터가 설정범위 이상으로 측정되어 이에 따른 알람을 통해 정비하거나, 대상 기기의 고장 유무에 관계없이 일정 시간 사용 후 해당 부품을 사전에 교체하는 방식으로 운영되고 있다. 하지만 선박 기관운영의 신뢰성과 운항 안전성을 확보하기 위해서는 실시간 상태 모니터링 데이터 기반의 사전적 진단 및 예측이 가능해야 한다. 그러기 위해서 실선 데이터를 종합적으로측정하여 데이터베이스화 하고 이를 선박의 보조기기와 배관의 상태기반 예지보전을 위한 상태 진단 모니터링 시스템을 구현하고자 한다. 특히 반응형 웹 기반으로 선박의 보조기기와 배관 상태 정보를 관리할 수 있도록 하였으며, 선내 개인용 컴퓨터(Personal Computer, PC)에서 보는 용도뿐만 아니라 스마트폰 등 다양한 모바일 기기의 접근 및 활용이 가능하도록 화면과 해상도에 맞춰 최적화된 상태 관리가 가능하도록 하여 업데이트 비용이 적게 들며, 관리 방법도 쉽다. 본 논문에서는 자율운항선박 핵심 기술인 상태기반정비(Condition Based Management, CBM) 기술력을 확보하기 위해 선박의 보조기기 중 펌프와 청정기, 그리고 배관 중 해수 및 스팀 배관의 상태 진단 모니터링을 통해 이상 현상을 파악하고, 이를 통해 융합 분석할 수 있도록 선박 보조기기 및 배관의 성능 진단 및 고장 예측에 활용하여 예방정비 의사결정을 지원하고자 한다.

• 한국압력기기공학회 논문집
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• 제12권1호
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• pp.62-69
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• 2016

Current guidance considers that uniaxially loaded specimen with a deep crack is used for the determination of the ductile-to-brittle transition temperature. However, reactor pressure vessel is under biaxial loading in real and the existence of deep crack is not probable through periodic in-service-inspection. The elastic stress intensity factor and the elastic-plastic J-integral which were used for crack-tip stress field and fracture mechanics assessment parameters. The difference of the loading condition and crack geometry can significantly influence on these parameters. Thus, a constraint effect caused by differences between standard specimens and a real structure can over/underestimate the fracture toughness, and it affects the results of the structural integrity assessment, consequentially. The present paper investigates the constraint effects by evaluating the master curve $T_0$ reference temperature of PCVN (Pre-cracked Charpy V-Notch) and small scale cruciform specimens which was designed to simulate biaxial loading condition with shallow crack through the fracture toughness tests and 3-dimensional elastic-plastic finite element analyses. Based on the finite element analysis results, the fracture toughness values of a small scale cruciform specimen were estimated, and the geometry-dependent factors of the cruciform specimen considered in the present study were determined. Finally, the transferability of the test results of these specimens was discussed.

• 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.

• 한국압력기기공학회 논문집
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• 제15권1호
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• pp.62-70
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• 2019

The APR1400 reactor may be operated for a long time under high temperature and pressure conditions, causing damage to the stud holes and causing stud bolts and holes to stick. The present practice is to manually remove the anti-sticking agent and foreign matter remaining in the APR1400 reactor stud hole and to visually check the surface condition of the thread to check the damage status of the threads. In the case of the APR1400 reactor stud holes, manually cleaning the threads increases the risk of radiation exposure and operator's fatigue. To avoid this, the autonomous mobile robot is used to automatically clean the reactor stud holes. The purpose of this study is to optimize the cleaning performance of the mobile robot by looking at the behavior of the surface roughness of the stud surface cleaned by the brush attached to the mobile robot due to changes in brush material, thickness of wire, and rotation speed. A microscopic approach to the surface roughness of the flank is needed to investigate the effects of the newly proposed brush of the autonomous mobile robot on the thread holes. According to this experiment, it is reasonable to use STS brush rather than Carbon one. Optimal operating conditions are derived and the safety of APR1400 reactor stud holes maintenance can be improved.

• 한국압력기기공학회 논문집
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• 제20권1호
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• pp.25-31
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• 2024

The core shroud of rector vessel internals (RVI) of OPR1000 and ARP1400 is made of Type 304 stainless steel (SS) by bending and welding process that may induce high deformation and residual stress in manufacturing. This work aims to evaluate the susceptibility of stress corrosion crack (SCC) initiation of bent parts of RVI in high temperature primary water environment. For SCC initiation test, tensile specimens were fabricated from the 90 degree bent plate of Type 304 SS (DT specimen), that is an archived part of a Korean APR1400. After the SCC initiation test, the specimen surface was thoroughly examined by optical and scanning electron microscopy, and compared to the specimen fabricated from the as-received plate of Type 304 SS (AR specimen). The surface observation revealed that SCC initiated on the AR specimen surface in typical intergranular (IG) mode, while SCC on the DT specimen occurred in transgrannular mode as well as IG mode. It was also found that the size and number of SCC on the DT specimen were larger than that on the AR specimen. This was attributable to a strain-hardening during the bending process. To compare the susceptibility of SCC initiation, total crack density (TCD) was calculated from the total crack length divided by the measured area of AR and DT specimens. TCD of DT specimen was 4.6 times higher than AR specimen in average, indicating that higher possibility of degradation of bent parts of RVI for a long-term operation.