SMART 원자로 제어봉 구동 장치의 동특성해석

이장원(Jang-Won Lee),조상순(Sang Soon Cho),김동옥(Dong-Ok Kim),박진석(Jin Seok Park),이원재(Won Jae Lee) 대한기계학회 2010 大韓機械學會論文集A Vol.34 No.8

한국원자력연구원은 전력생산과 해수담수화를 동시에 수행하고 친환경적인 SMART 원자로를 개발하였다. SMART 원자로의 여러 구조물 중에 제어봉 구동 장치(CRDM)는 제어봉의 삽입 량을 조절하여 원자로의 출력을 조정하고 비상시 제어봉을 긴급 삽입하여 원자로를 정지시키기 위한 기기이다. 본 연구의 목적은 제어봉 구동 장치의 구조적 건전성을 확보하기 위해서 동특성해석을 수행하는 것이다. 또한 향후 내진해석에 활용될 단순모델의 활용을 위해 상세모델과의 비교, 검증을 수행하였다. 해석은 유한요소 해석기법을 활용하였고 상용해석 프로그램인 ABAQUS 와 ANSYS V12 를 사용하였다. 유한요소 해석모델은 상세모델인 3-D Solid 모델과 단순모델인 Beam 모델을 작성하여 비교하였고 추가로 단순모델을 오일러 보인 Beam4 요소와 티모센코 보인 Beam188 요소로 작성하여 비교검토하였다. 향후 SMART 원자로집합체의 단순모델을 작성하여 내진해석 등 다양한 해석에 활용될 계획이므로 단순모델은 상세모델과의 오차를 줄이기 위해서 모델 보정(model updating)이 수행되었다. The Korea Atomic Energy Research Institutes has been developing the SMART (System integrated Modular Advanced ReacTor), an environment-friendly nuclear reactor for the generation of electricity and to perform desalination. SMART reactors can be exposed to various external and internal loads caused by seismic and coolant flows. The CRDM(control rod drive mechanism), one of structures of the SMART, is a component which is adjusting inserting amount of a control rod, controlling output of reactor power and in an emergency situation, inserting a control rod to stop the reactor. The purpose of this research is performing the analysis of dynamic characteristic to ensure safety and integrity of structure of CRDM. This paper presents two FE-models, 3-D solid model and simplified Beam model of the CRDM in the coolant, and then compared the results of the dynamic characteristic about the two FE-models using a commercial Finite Element tool, ABAQUS CAE V6.8 and ANSYS V12. Beam 4 and beam 188 of simplified-model were also compared each other. And simplified model is updated for accuracy compare to 3-D solid.

SMART 원자로 제어봉구동장치의 동특성해석

이장원(Jang-Won Lee),조상순(Sang Soon Cho),김동옥(Dong-Ok Kim),박진석(Jin-Seok Park),이원재(Won-Jae Lee) 대한기계학회 2010 대한기계학회 춘추학술대회 Vol.2010 No.3

The Korea Atomic Energy Research Institutes has been developing the SMART (System integrated Modular Advanced ReacTor), an environment-friendly nuclear reactor for the generation of electricity and to perform desalination. SMART reactors can be exposed to various external and internal loads caused by seismic and coolant flows. The CRDM(control rod drive mechanism), one of structures of the SMART, is a component which is adjusting inserting amount of a control rod, controlling output of reactor power and in an emergency situation, inserting a control rod to stop the reactor. The purpose of this research is performing the analysis of dynamic characteristic to ensure safety and integrity of structure of CRDM. This paper presents two FE-models, 3-D solid model and simplified Beam model of the CRDM in the coolant, and then compared the results of the dynamic characteristic about the two FE-models using a commercial Finite Element tool, ABAQUS CAE V6.8 and ANSYS V12. Beam 4 and beam 188 of simplified model were also compared each other in addition.

원자로 상부헤드 제어봉구동장치 관통노즐 형상이 J-Groove 용접잔류응력에 미치는 영향

김주희(Ju Hee Kim),김윤재(Yun Jae Kim),이성호(Sung Ho Lee),허남용(Nam Young Hur),배홍열(Hong Yeol Bae),오창영(Chang Young Oh),김지수(Ji Soo Kim),박흥배(Heung Bae Park),이승건(Seung Geon Lee),김종성(Jong Sung Kim),허남수(Nam Su Huh) 대한기계학회 2011 大韓機械學會論文集A Vol.35 No.10

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. In pressurized water reactors (PWRs), the reactor pressure vessel (RPV) upper head contains numerous control rod drive mechanism (CRDM) nozzles. In the last 10 years, the incidences of cracking in alloy 600 CRDM nozzles and their associated welds has increased significantly. Several axial and circumferential cracks have been found in CRDM nozzles in European PWRs and U.S. nuclear power plants. These cracks are caused by primary water stress corrosion cracking (PWSCC) and have been shown to be driven by welding residual stresses and operational stresses in the weld region. Therefore, detailed finite-element (FE) simulations for the Korea Nuclear Reactor Pressure Vessel have been conducted in order to predict the magnitudes of the weld residual stresses in the tube materials. In particular, the weld residual stress results are compared in terms for nozzle location, geometry factor r_o/t, geometry of fillet, and adjacent nozzle.

제어봉구동장치 관통노즐 용접부의 잔류응력 완화를 위한 연구

이승건(Seung-Gun Lee),김종성(Jong-Sung Kim),진태은(Tae-Eun Jin) 대한기계학회 2004 대한기계학회 춘추학술대회 Vol.2004 No.4

In this study, we proposed new method to mitigate tensile welding residual stress for preventing PWSCC in CRDM nozzle. Residual stress analysis using finite element method is performed to confirm benefit of the new method. In case of applying existing method, tensile axial residual stress decrease by about 28% and tensile hoop residual stress decrease by about 33%. In case of applying the new method, tensile axial residual stress decrease by about 32% and tensile hoop residual stress decrease by about 43%. Therefore, we conclude the new proposed method is more effective to prevent PWSCC than existing method.

CRDM Nozzle의 TOFD 진단 신호를 활용한 딥 러닝 기반 결함 분석 연구

이수민,박준필,김훈희,박재석,이재선 한국비파괴검사학회 2023 한국비파괴검사학회지 Vol.43 No.4

The probability of defects occurring increases over time with the prolonged use of a control rod drive mechanism (CRDM) nozzle. While workers perform regular inspections for safety purposes, there is a likelihood that defects are missed, which can lead to safety risks and economic losses. To prevent this, an automated method for analyzing defects is needed. This study utilizes the compliant image validation analyzer (CIVA) tool, which specializes in non-destructive testing simulations, to generate data resembling real field B-scan images. The objective is to investigate the accuracy of measuring defects on the surface or welds of the CRDM nozzle using the high-sensitivity TOFD ultrasonic testing technique for defect detection. Simulations were conducted with different defect locations and sizes, and the obtained data was compared with data obtained from the actual field. Finally, to make the CIVA simulation data similar to actual field data, noise and distortion were added to the simulated images. A convolutional neural network (CNN) was trained using this data to create a model for defect analysis, and model validation was performed using CRDM nozzle diagnostic B-scan test images. CRDM(Control Rod Drive Mechanism) nozzle이 장기간 사용되면서 결함 발생 확률이 높아져 안전을위해 작업자가 정기적인 진단을 실행하고 있지만 결함을 놓칠 수 있다. 그로 인해 문제가 생기면 안전성이저하되고 경제적 손실이 생긴다. 이를 방지하기 위해 결함을 자동으로 분석하는 방법이 필요하다. 본 연구에서는 결함 검출 감도가 높은 TOFD(Time of Flight Diffraction) 초음파 검사 기법으로 CRDM nozzle의 표면 또는 용접부에 결함이 있으면 그 결함이 잘 측정될 수 있는지 알아보기 위해 비파괴 검사 시뮬레이션 전문 도구인 CIVA를 사용하여 실제 현장 B-scan 이미지와 유사하게 데이터를 생성하였다. 각각 다른 결함 위치, 크기 조건으로 시뮬레이션을 진행 후 얻은 데이터가 실제 현장에서 얻은 데이터와 유사한지 확인하였다. 최종적으로 CIVA 시뮬레이션 데이터가 실제 현장 데이터와 유사하도록 시뮬레이션 이미지에 노이즈 및 왜곡을추가하였고, 그 데이터를 사용하여 결함 분석을 위한 모델을 만들기 위해 CNN(Convolutional Neural Networks) 신경망을 훈련하였으며 임의의 CRDM nozzle 진단 신호 B-scan 테스트 이미지를 사용하여 모델 검증을 진행하였다.

JRTR 제어봉구동장치의 구조건전성 평가를 위한 내진시험

최명환(Myoung-Hwan Choi),김경호(Gyeong-Ho Kim),선종오(Jong-Oh Sun),조영갑(Yeong-Garp Cho) 한국소음진동공학회 2016 한국소음진동공학회 학술대회논문집 Vol.2016 No.4

A control rod drive mechanism is a reactor regulating system, which inserts, withdraws or maintains a control rod containing a neutron absorbing material within a reactor core to control the reactivity of the core. The top-mounted CRDM for Jordan Research and Training Reactor has been designed and fabricated based on the HANARO’s experience. This paper describes the seismic test results to evaluate the structural integrity of CRDM drive assembly. The tests are carried out at two test rigs simulating the upper part of reactor pool. The seismic test results show that the CRDM maintains the function and structural integrity without changes of natural frequencies before and after 5 OBE and 1 SSE load excitations. Also, the stepping motor during the OBE load is driven having a good stepping performance without a malfunction.

JRTR 제어봉구동장치의 내진시험

최명환(Myoung-Hwan Choi),김경호(Gyeong-Ho Kim),선종오(Jong-Oh Sun),조영갑(Yeong-Garp Cho) 한국소음진동공학회 2016 한국소음진동공학회 논문집 Vol.26 No.5

A control rod drive mechanism(CRDM) is a reactor regulating system, which inserts, withdraws or maintains a control rod within a reactor core to control the reactivity of the core. The CRDM for Jordan Research and Training Reactor with 5㎿ power has been designed and fabricated based on the HANARO’s experience through KAERI and DAEWOO consortium. This paper describes the seismic test results to demonstrate the operability, the drop performance and the structural integrity of CRDM during or after seismic excitations. The seismic tests are carried out under 5 OBE and 1 SSE loads at three Test Rigs simulating the reactor structure and the pool top. From the tests, the CRDM is smoothly driven without a malfunction of stepping motor under OBE load. The pure drop time under OBE and SSE loads is measured as 1.169s and 1.855s to meet the design requirement. Also, it is found that the CRDM maintains the structural integrity without a change of the function and natural frequency before and after seismic loads.

원자로 CRDM 관통노즐 J-Groove 용접부 잔류응력 예측을 위한 유한요소 변수 민감도 해석

배홍열(Hong Yeol Bae),김주희(Ju Hee Kim),김윤재(Yun Jae Kim),오창영(Chang Young Oh),김지수(Ji Soo Kim),이성호(Sung Ho Lee),이경수(Kyoung Soo Lee) 대한기계학회 2012 大韓機械學會論文集A Vol.36 No.10

최근 원자로 압력용기 상부헤드 관통노즐 J-groove 용접부 주변에서 균열로 인한 냉각수 누출사고가 발행하고 있다. 이러한 사고의 원인은 용접에 의한 인장잔류응력, 농축된 붕산수 및 응력부식에 민감한 재료로 인한 일차수응력부식균열(PWSCC : primary water stress corrosion cracking)인 것으로 판명되었다. PWSCC 평가는 원자로 건전성 평가의 주요 관심사로서 용접에 의해 발생되는 잔류응력을 정확하게 예측함으로써 가능하다. 본 연구에서는 유한요소해석을 이용하여 국내 원자로의 일반적인 J-groove 용접부의 해석절차를 소개하고, 용접해석 관련 변수의 민감도 해석을 통해 잔류응력 예측기법을 제시하고자 한다. 이를 위해 2차원 및 3차원 요한요소해석 방법을 바탕으로 변수 민감도 해석을 수행하였으며, 기존 연구결과와 비교를 통해 해석절차 및 방법의 유용성을 검정하였다. In nuclear power plants, the reactor pressure vessel (RPV) upper head control rod drive mechanism (CRDM) penetration nozzles are fabricated using J-groove weld geometry. Recently, the incidences of cracking in Alloy 600 CRDM nozzles and their associated welds have increased significantly. The cracking mechanism has been attributed to primary water stress corrosion cracking (PWSCC), and it has been shown to be driven by welding residual stresses and operational stresses in the weld region. The weld-induced residual stress is the main factor contributing to crack growth. Therefore, an exact estimation of the residual stress is important for ensuring reliable operation. This study presents the residual stress computation performed for an RPV CRDM penetration nozzle in Korea. Based on two and three dimensional finite element analyses, the effect of welding variables on the residual stress variation is estimated for sensitivity analysis.

요르단 연구용원자로 제어봉구동장치의 성능검증시험

최명환(M. H. Choi),조영갑(Y. G. Cho),김정현(J. H. Kim),이관희(K. H. Lee) 한국소음진동공학회 2015 한국소음진동공학회 논문집 Vol.25 No.12

A control rod drive mechanism(CRDM) is a reactor regulating system, which inserts, withdraws or maintains a control rod containing a neutron absorbing material within a reactor core to control the reactivity of the core. The top-mounted CRDM for Jordan Research and Training Reactor(JRTR) with 5 MW power has been designed and fabricated based on the HANARO’s experience through KAERI and DAEWOO consortium project. This paper describes the performance qualification test results to demonstrate the operability of a prototype and four production CRDMs during the reactor lifetime. The driving performance, the drop performance and the endurance tests for CRDM are carried out at a test rig simulating the actual reactor conditions. A vibration of internal components due to the coolant flow is also measured using a laser vibrometer. As a result, the CRDMs are driven having a good driving performance without a malfunction between command and output signals for the stepping motor. Also, the pure drop time and the impact acceleration are within 0.72 s and 4.2 g to meet the design requirements, and the vibrational displacement of control rod is measured as maximum 5.2 μm.

요르단 연구용원자로 제어봉구동장치의 성능검증시험

최명환(Choi, M.H.),조영갑(Cho, Y.G.),김정현(Kim, J.H.),이관희(Lee, K.H.) 한국소음진동공학회 2015 한국소음진동공학회 학술대회논문집 Vol.2015 No.10