• 한국압력기기공학회 논문집
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• 제14권2호
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• pp.16-23
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• 2018

In order to investigate the ways to enhance the seismic performance of APR1400 seismic fragile equipment by direct design changes, four equipment such as Reactor Vessel Support, Integrated Head Assembly, Remote Shutdown Console, and Pressurizer are reviewed using information of the main dimensions, seismic stress evaluation results, design FRS, etc. in this paper. In addition to the direct reinforcement of equipments, the feasibility of seismic isolation for the safety related cabinet is also investigated and the actual adaption plan of a commercial spring-damper system is briefly reviewed.

• 한국압력기기공학회 논문집
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• 제16권1호
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• pp.84-91
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• 2020

The APR1400 reactor stud holes can be stuck due to high temperatures, high pressure, prolonged engagement, and load changes according to pressure changes in the reactor. Threaded surfaces of a stud hole should be cleaned for the sealing of pressure in reactor vessel by removing any foreign materials which may exist in the stud holes. Human workers can access to the stud hole for the cleaning of stud holes manually, but the radiation exposure of human workers is increased. Robot is an effective way to work in hazardous area. So we introduced robot for the cleaning of stud holes. Localization of mobile robots is generally based on odometry, but with increased mileage, position errors can be accumulated. In order to eliminate cumulative error and to ensure stability of its driving, laser sensors and new control algorithm were utilized. The distance between the robot and the wall was measured by laser sensors, and the control algorithm was implemented so as to travel the desired trajectory by using the measured values from sensors. The performance of driving and hole sensing were verified through field application, and mobile robot was confirmed to be applicable to the APR 1400 NPP.

• 한국압력기기공학회 논문집
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• 제4권2호
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• pp.13-19
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• 2008

In order to predict a remaining life of a plant, it is necessary to select the components that are critical to the plant life. The remaining life of those components shall be evaluated by considering the aging effect of materials used as well as numerous factors. However, when evaluating reliability of nuclear structural components, some problems are quite formidable because of lack of information such as operating history, material property change and uncertainty in damage models. Accordingly, if structural integrity and safety are evaluated by the deterministic fracture mechanics approach, it is expected that the results obtained are too conservative to perform a rational evaluation of plant life. The probabilistic fracture mechanics approaches are regarded as appropriate methods to rationally evaluate the plant life since they can consider various uncertainties such as sizes and shapes of cracks and degradation of material strength due to the aging effects. The objective of this study is to evaluate the structural integrity for a reactor pressure vessel under the small break loss of coolant accident by applying the deterministic and probabilistic fracture mechanics. The deterministic fracture mechanics analysis was performed using the three dimensional finite element model. The probabilistic integrity analysis was based on the Monte Carlo simulation. The selected random variables are the neutron fluence on the vessel inside surface, the content of copper, nickel, and phosphorus in the reactor pressure vessel material, and initial RTNDT.

• 한국압력기기공학회 논문집
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• 제15권2호
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• pp.58-63
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• 2019

This paper presents preliminary analysis and results on IASCC sensitivity of a core shroud in the reactor pressure vessel. First, neutron irradiation flux distribution of the reactor internals was calculated by using the Monte Carlo simulation code, MCNP6.1 and the nuclear data library, ENDF/B-VII.1. Second, based on the neutron irradiation flux distribution, temperature and stress distributions of the core shroud during normal operation were determined by performing finite element analysis using the commercial finite element analysis program, ABAQUS, considering irradiation aging-related degradation mechanisms. Last, IASCC sensitivity of the core shroud was assessed by using the IASCC sensitivity definition of EPRI MRP-211 and the finite element analysis results. As a result of the preliminary analysis, it was found that the point at which the maximum IASCC sensitivity is derived varies over operating time, initially moving from the shroud plate located in the center of the core to the top shroud plate-ring connection brace over operating time. In addition, it was concluded that IASCC will not occur on the core shroud even after 60 years of operation (40EFPYs) because the maximum IASCC sensitivity is less than 0.5.

• 한국압력기기공학회 논문집
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• 제15권2호
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• pp.1-8
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• 2019

The integrity of the Reactor Pressure Vessel (RPV) is affected by the neutrons bombarding the vessel wall leading to embrittlement. This irradiation-induced embrittlement leads to reduction in the fracture toughness of RPV materials. This paper presents a comparative study of typical Optimized Power Reactor (OPR)1000 reactor pressure-temperature (P-T) limit curves using the pre-2006 American Society of Mechanical Engineers (ASME) editions used in the power plant and the current ASME edition of 2010. The current ASME Code utilizes critical reference stress intensity factor based on the lower bound of static, while the Pre-2006 ASME editions are based the critical reference stress intensity factor based on the lower bound of static, dynamic and crack arrest. Model-Based Systems Engineering approach was used to evaluate ASME Code Section XI Appendix G for generating the P-T limit curves. The results obtained from this analysis indicate decrease in conservatism in P-T limit curves constructed using the current 2017 ASME code, which can potentially increase operational flexibility and plant safety. Hence it is recommended to use ASME code edition after 2006 be used in all operating nuclear power plants (NPPs) to establish P-T limit curve.

• 한국압력기기공학회 논문집
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• 제13권1호
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• pp.61-68
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• 2017

Reactor pressure vessel (RPV) under severe accident conditions accompanied by core melting is exposed to direct high-temperature thermal loads. Understanding the creep behavior of the material is one of the most important factors for evaluating the structural integrity at these conditions. While damage evaluation studies have been conducted on critical structures of nuclear power plants through finite element (FE) analyses considering creep behavior, for accurate creep damage evaluation, constitutive equations considered in the FE analyses may have different results depending on the time hardening and strain hardening models as well as the tertiary creep consideration. The purpose of this study is to evaluate the creep damage under severe accident conditions by using FE method for a representative domestic RPV material, SA508 Gr.3. The effect of material hardening models and constitutive equations which are the main variables were also investigated.

• 한국압력기기공학회 논문집
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• 제11권1호
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• pp.53-60
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• 2015

This study was investigated to develop process technology of laser cladding with austenite stainless steel for Weld Inlay repair of dissimilar metal weld in reactor vessel in/outlet nozzles. Weld Inlay experiments were performed by laser cladding repair system consisting of common manipulator, laser apparatus and welding process scheduler, etc. Single pass welding experiments were conducted in order to obtain the optimum welding process parameters for filler wires of ER309L and Alloy 52M before multi-layer laser cladding. Based on the above obtained results, multi-layer laser cladding experiments were carried out, and welding qualities for weld specimens were estimated by PT, OM, SEM and EDS analysis. Consequently, it was revealed that multi-layer laser cladding on austenite stainless steel using filler wires of ER309L and Alloy 52M could be possible to meet ASME Code standard without any weld defect.

• 비파괴검사학회지
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• 제17권1호
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• pp.11-22
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• 1997

원자력발전소 압력용기 및 배관은 많은 용접부를 포함하고 있으며 용접부내 결함은 크기, 위치 및 형태에 따라 압력용기 및 배관의 건전성에 커다란 영향을 미친다. 따라서 주요 압력용기와 배관의 용접부에 대해서는 가동 전 중 검사시 초음파 탐상시험을 실시하여 그 건전성을 확인하고 있다. 초음파 결함 신호로부터의 결함 분류는 비파괴 평가에 있어 매우 중요하며 초음파 형상 인식 방법이 적당하다. 본 논문에서는 탄소강 압력용기 용접부에 내재하는 결함으로부터 얻어진 초음파 결함 신호의 형상 인식을 위한 절차로써 데이터 수집, 특징 추출, 특징 선택 및 결함 분류를 하였으며, 결함 분류에 있어 결함의 종류를 크게 선형(linear)과 체적(volumetric)의 두 종류로 분류함에 있어 퍼지이론을 적용하여 퍼지이론을 적용한 초음파 형상 인식 기법의 가능성 및 효율성을 제시하였고 그 결과 기존의 분류기(classifier)들에 비해 보다 우수한 결과를 얻을 수 있었다.

• 한국압력기기공학회 논문집
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• 제5권1호
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• pp.9-13
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• 2009

The containment local leakage rate testing in nuclear power plants is performed in accordance with ANSI/ANS-56.8(1994) in Korea. Two methods, the make-up flow rate and the pressure decay, are used for LLRT. Though ANSI/ANS-56.8 does not define clearly the minimum test duration for the make-up flow rate method, it requires obtaining the data after reaching the stable condition. Thus the prerequisite stable condition for data acquisition and the test period for type "C" LLRT is differently applied to each NPPs. Therefore, this study presents a unified test criteria for data stabilization and test duration through experiments to improve the test reliability for type "C".

• 한국압력기기공학회 논문집
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• 제8권2호
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• pp.42-47
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• 2012

The containment local leakage rate testing in nuclear power plants is performed in accordance with ANSI/ANS 56.8-1994 in Korea. Two methods, the make-up flow rate and the pressure decay, are used for local leakage rate testing. Though ANSI/ANS 56.8-1994 does not define clearly the minimum test duration for the make-up flow rate method, it requires obtaining the data after reaching the stable condition. Thus the prerequisite stable condition for data acquisition and the testing time is differently applied to each NPPs. Therefore, this study presents a standardized test procedure for data stabilization and testing time through experiments to improve the test reliability.