• Nuclear Engineering and Technology
• /
• 제54권4호
• /
• pp.1374-1381
• /
• 2022

Under the influence of nuclear radiation, the reliability of steam generators (SGs) is an important factor in the efficiency and safety of nuclear power plant (NPP) reactors. Motion planning that remotely manipulates an SG mobile tube-inspection robot to inspect SG heat transfer tubes is the mainstream trend of NPP robot development. To achieve motion planning, conditional traversal is usually used for base position optimization, and then the A^* algorithm is used for path planning. However, the proposed approach requires considerable processing time and has a single expansion during path planning and plan paths with many turns, which decreases the working speed of the robot. Therefore, to reduce the calculation time and improve the efficiency of motion planning, modifications such as the matrix method, improved parent node, turning cost, and improved expanded node were proposed in this study. We also present a comprehensive evaluation index to evaluate the performance of the improved algorithm. We validated the efficiency of the proposed method by planning on a tube sheet with square-type tube arrays and experimenting with Model SG.

• 한국압력기기공학회 논문집
• /
• 제4권2호
• /
• pp.34-39
• /
• 2008

Steam Generator(SG) tubes are classified as one of the key components in nuclear power plants, and they should be periodically examined by the intensified management program for the assurance and diagnosis of their structural integrity. In this study, we use the optimum analysis technique to draw the detection and categorization of bowing(BOW) signals; abnormal tube-to-tube proximity in the SG upper bundle free span area. The locations in which BOW signals are detected likely have latent degradation of ODSCC(Outer Diameter Stress Corrosion Cracking). For the sake of timely and correct detection of BOW signals and diagnosis of ODSCC, we carried out the experimental demonstrations using a reduced mock-up. And we validated the MRPC(Motorized Rotating Pancake Coil) analysis technique is better than the bobbin. Hence, it comes to conclusion that the optimum analysis technique can be a good alternative for the reliable SG tube examination.

• Nuclear Engineering and Technology
• /
• 제54권3호
• /
• pp.917-925
• /
• 2022

Steam generator (SG) tubes in a nuclear power plant can undergo rapid changes in pressure and temperature during an accident; thus, an accurate model to predict short-term creep damage is essential. The theta (𝜃) projection method has been widely used for modeling creep-strain behavior under constant stress. However, many creep test data are obtained under constant load, so creep rupture behavior under a constant load cannot be accurately simulated due to the different stress conditions. This paper proposes a novel methodology to obtain the creep curve under constant stress using a modified 𝜃 projection method that considers the increase in true stress during creep deformation in a constant-load creep test. The methodology is validated using finite element analysis, and the limitations of the methodology are also discussed. The paper also proposes a creep-strain model for alloy 690 as an SG material and a novel creep hardening rule we call the damage-fraction hardening rule. The creep hardening rule is applied to evaluate the creep rupture behavior of SG tubes. The results of this study show its great potential to evaluate the rupture behavior of an SG tube governed by creep deformation.

• Nuclear Engineering and Technology
• /
• 제41권1호
• /
• pp.125-134
• /
• 2009

This paper presents details on a tube inspection robotic system and a positioning method of the robot for a steam generator (SG) in nuclear power plants (NPPs). The robotic system is separated into three parts for easy handling, which reduces the radiation exposure during installation. The system has a supporting leg to increase the rigidity of the robot base. Since there are several thousands of tubes to be inspected inside a SG, it is very important to position the tool of the robot at the right tubes even if the robot base is positioned inaccurately during the installation. In order to obtain absolute accuracy of a position, the robot kinematics was mathematically modeled with the modified DH(Denavit-Hartenberg) model and calibrated on site using tube holes as calibration points. To tune the PID gains of a commercial motor driver systematically, the time delay control (TDC) based gain tuning method was adopted. To verify the performance of the robotic system, experiments on a Framatomes 51B Model type SG mockup were undertaken.

• 전자공학회논문지S
• /
• 제35S권2호
• /
• pp.137-145
• /
• 1998

In this paper, we propose a tube-hole center detection vision algorithm verifying the position of a tele-controlled robot and providing visual information for increasing reliability and efficiency in the diagnosis of steam generator (SG) tubes in nuclear power plant. A tele-controlled robot plays a role in carrying the probe used in inspecting the integrity of SG tubes. Thus accurately locating a tele-controlled robot on the desired tube-hole center is important issue for reliability of inspection. To do this work, we have to find the tube-hole center locations from the input image. At first, we apply the three-class segmentation method modified for this application. WE extract minimum bounding rectangles (MBRs) in the theresholded binary image. Second, for discriminating between MBR by tube and MBR by noise, we introduce the MBR rejection rules as knowledge-based rule set. MBRs are divided into the very dark region MBRs and the very bright region MBRs. In order to describe the region of complete tube-hole, the MBRs need a process of pairing each other. We then can find the tube-hole center from the paired MBR. For more accurately finding the tube-hole center in several sequential images, the centers of some frames need to be averaged. We tested the performance of our method using hundreds of real images.

• 비파괴검사학회지
• /
• 제29권4호
• /
• pp.374-382
• /
• 2009

원자력발전소의 1차측 및 2차측 냉각계의 장벽 역할을 하는 핵심 설비중 하나인 증기발생기(steam generator, SG) 전열관은 공공의 사회적 안전성과 효율적인 발전 용량을 유지하기 위해 구조적 건전성을 유지하여야 한다. 또한 결함을 함유하고 있는 전열관은 해당결함을 조기에 검출, 정량적으로 결함을 평가하여 필요한 경우에는 보수조치를 수행하여야 한다. 이러한 결함의 검출 및 정량화를 위해서 검사관련 고시 및 강화된 SG 관리프로그램(SGMP)에 근거하여 와전류탐상검사법(eddy current testing, ECT)을 적용, 검사를 수행하고 있다. SG 전열관에서 검출되고 있는 결함중 응력부식균열(stress corrosion cracking, SCC)은 미세한 경우 결함의 검출이 어려울 뿐 아니라 생성된 결함의 성장속도가 빠르기 때문에 SG 전열관의 건전성을 위협하는 주요결함 기구중 하나로 분류하고 있다. 본 논문에서는 다양한 결함 깊이 및 길이별로 방전가공(electric discharge machining, EDM)된 축방향 ODSCC에 대해 pancake, +point 및 shielded pancake 코일 등이 탑재된 3 coil형태의 +PT MRPC(motorized rotating pancake coils)를 적용하여 결함의 검출가능 여부 및 크기 측정을 위한 검사를 수행하였으며 본 실험결과를 통해 SG 전열관의 건전성 및 원전 운전의 안전성을 진단하는 공학적 평가 자료로써의 활용 가능성 뿐 아니라 와전류탐상검사의 신뢰도 향상을 도모하고자 하였다.

• 비파괴검사학회지
• /
• 제30권1호
• /
• pp.13-19
• /
• 2010

본 논문에서는 원전SG세관 결함 크기 추정을 위한 새로운 구조의 추정시스템에 대한 연구를 수행한다. 기존의 연구에서는 결함 크기를 추정하기 위하여 각각의 결함 형태별로 결함크기추정시스템을 설계하였다. 이와 같은 경우, 추정시스템의 구조가 복잡해지고 결함 크기 추정 이전에 수행하는 결함형태분류기의 정확성이 떨어질 경우 결함 크기 추정 성능도 결과적으로 악화될 수밖에 없다. 이에 본 논문에서는 결함 형태 분류 과정을 필요로 하지 않는 결함크기추정시스템의 성능을 분석하고 이를 향상시키기 위한 방안을 연구하였다. 기존의 추정시스템은 각각의 결함 형태별로 특화된 추정기를 사용하기 때문에 추정 성능이 훨씬 뛰어날 것으로 예상되었지만, 실험 결과 두 추정시스템의 성능 차이는 그리 크지 않다는 것을 알 수 있었다. 따라서 결함형태분류기의 정확성이 완벽하지 않을 경우, 본 논문에서 제안한 구조의 추정기가 효과적으로 사용될 수 있을 것으로 기대된다.

• 비파괴검사학회지
• /
• 제30권4호
• /
• pp.302-310
• /
• 2010

본 논문에서는 원자력 발전소 증기발생기 세관에 발생할 수 있는 결함의 크기측정에 사용되는 Bagging 신경회로망에 대한 연구를 수행하였다. Bagging은 부트스트랩(bootstrap) 샘플링에 기반을 둔 추정기 앙상블을 생성하는 방법이다. 증기발생기 세관의 결함 크기측정을 위하여 다양한 폭과 깊이를 갖는 4가지 결함패턴의 eddy current testing 신호를 생성하였다. 그 다음, 단일 신경회로망(single neural network; SNN)과 Bagging 신경회로망(Bagging neural network; BNN)을 구성하여 각 결함의 폭과 깊이를 추정하였다. SNN과 BNN 추정성능은 최대오차를 이용해서 측정하였다. 실험결과, 결함 깊이 추정시의 SNN과 BNN 최대오차는 0.117mm와 0.089mm 이었다. 또한, 결함 폭 추정 시에는 SNN과 BNN 최대오차는 0.494mm와 0.306mm 이었다. 이러한 실험결과는 BNN 추정성능이 SNN 추정성능보다 우수하다는 것을 보여준다.

• 한국소음진동공학회논문집
• /
• 제13권5호
• /
• pp.400-408
• /
• 2003

This paper investigates the vibration characteristics of steam generator (SG) U-tubes with defect. The operating SG shell-side flow field conditions for determining the fluidelastic instability parameters such as added mass are obtained from three-dimensional SG flow calculation. Modal analyses are performed for the U-tubes either with axial or circumferential flaw with different sizes. Special emphases are on the effects of flaw orientation and size on the modal and instability characteristics of tubes, which are expressed in terms of the natural frequency, corresponding mode shape and stability ratio. Also, addressed is the effect of the internal pressure on the vibration characteristics of the tube.

• 대한기계학회:학술대회논문집
• /
• 대한기계학회 2003년도 춘계학술대회
• /
• pp.610-616
• /
• 2003

Recently, the common concern of nuclear power industry in the development of technology mitigating and preventing the aging of steam generator tubes prevails, because the trends of steam generator flaws at Uljin unit #1,2 and KSNP(Korea Standard Nuclear Power Plant) impose a burden on the operation of nuclear power plant. While the regulatory agency is demanding the establishment of the advanced general performance maintenance system, the steam generator management program adapting advanced technology is being developed which may comply with EPRI PWR SG Guidelines based on NEI 97-06 ‘ General Guidelines including all the maintenance aspects consist of the tube integrity assessment criteria, repair limit, allowable leakage level, water chemistry will be composed in order to obtain the approval of regulatory agency and be applied to Nuclear power plant early 2005. This presentation is to introduce maintenance state including SG tube degradation and main contents of advanced SG management program being developed, and futhermore update present and future plan, and estimate the alternation after the completion.