감시시험 결과를 이용한 국내원전 압력용기 재료의 Chemistry Factor 및 RT_PTS 평가여유도 분석

이호진,윤지현,최권재,이봉상,Lee, Ho-Jin,Yoon, Ji-Hyun,Choi, Kwon-Jae,Lee, Bong-Sang 한국압력기기공학회 2011 한국압력기기공학회 논문집 Vol.7 No.3

The chemistry factor and RTPTS margin for domestic reactor pressure vessel materials were analyzed by using the surveillance data which have been obtained from 8 nuclear power plants in Korea. The surveillance data have been used to assess the integrity of the pressure vessel under the pressurized thermal shock (PTS) event. The chemistry factor, which is determined by the Cu and Ni contents of vessel materials, is considered a proper tool to assess the $RT_{PTS}$. The chemistry factors, which were obtained from the surveillance data of domestic reactor pressure vessels, were investigated and compared with those of Regulatory Guide 1.99 in this study. Regressions for ${\Delta}RT_{NDT}$ were performed to expect the chemistry factor as a function of Cu and Ni, and to estimate $RT_{PTS}$ margin. The margin analysis was performed by comparing the regression graphs and standard deviations with those of Regulatory Guide 1.99. The standard deviations calculated by using the domestic surveillance data for base metal and welds are almost same as the standard deviations which are suggested on Regulatory Guide 1.99, Rev.2.

SG전열관 2차측 이물질 검출 및 특성분석을 위한 ETSS 개발

신기석,문용식,민경만,Shin, Ki Seok,Moon, Yong Sig,Min, Kyong Mahn 한국압력기기공학회 2011 한국압력기기공학회 논문집 Vol.7 No.3

The integrity of the SG(Steam Generator) tubes has been challenged by numerous factors such as flaws, operation, atmosphere, inherently degraded materials, loose parts and even human errors. Of the factors, loose parts(or foreign materials) on the secondary side of the tubes can bring about volumetric defects and even leakage from the primary to the secondary side in a short period of time. More serious concerns about the loose parts are their unknown influx path and rapid growth rate of the defects affected by the loose parts. Therefore it is imperative to detect and characterize the foreign materials and the defects. As a part of the measures for loose part detection, TTS(Top of Tubesheet) MRPC(Motorized Rotating Pancake Coils) ECT has been carried out especially to the restricted high probability area of the loose part. However, in the presence of loose parts in the other areas, wide range loose part detection techniques are required. In this study, loose part standard tube was presented as a way to accurately detect and characterize loose part signals. And the SG tube ECT bobbin coil and MRPC ISI(In-service Inspection) data of domestic OPR-1000 and Westinghouse Model F(W_F) were reviewed and consequently, comprehensive loose part detection technique is derived especially by applying bobbin coil signals

고온로 설계 적합성평가 프로그램 개발

조두호,서한범,최재붕,허남수,최영환,Cho, Doo Ho,Surh, Han Bum,Choi, Jae Boong,Huh, Nam Su,Choi, Young Hwan 한국압력기기공학회 2013 한국압력기기공학회 논문집 Vol.9 No.1

In this paper, W-DCAP-HTR(Web-based Design Compatibility Assessment Program for High Temperature Reactor) which will be used to check the design criteria for high temperature reactor is newly proposed. To do this, the assessment procedure of the ASME Sec.III Div.5 such as time-dependent primary stress limit, accumulated inelastic strain, and creep-fatigue damage evaluation were investigated. Furthermore, the trend of candidate materials for high temperature reactor was also reviewed. Then, all assessment procedures for high temperature reactor have been computerized to enhance the efficiency and to reduce the possibility of human error during calculating procedure by hand calculation. It can be directly conducted by adopting the actual thermal and structural analysis results. The validation of W-DCAP-HTR has been demonstrated by benchmark analysis.

CANDU형 원전 압력관에 존재하는 축방향 균열의 응력확대계수

이국희,오영진,박흥배,정한섭,정하주,김윤재,Lee, Kuk-Hee,Oh, Young-Jin,Park, Heung-Bae,Chung, Han-Sub,Chung, Ha-Joo,Kim, Yun-Jae 한국압력기기공학회 2011 한국압력기기공학회 논문집 Vol.7 No.1

CANDU reactor core is composed a few hundreds pressure tubes, which support and locate the nuclear fuels in the reactor. Each pressure tube provides pressure boundary and flow path of primary heat transport system in the core region. In order to guarantee the structural integrity of pressure tube flaws which can be found by in-service inspection, crack growth and fracture initiation assessment have to be performed. Stress intensity factors are important and basic information for structural integrity assessment of planar and laminar flaws (e. g. crack). This paper reviews and confirms the stress intensity factor of axial crack, proposed in CSA N285.8-05, which is an fitness-for-service evaluation code for pressure tubes in CANDU nuclear reactors. The stress intensity factors in CSA N285.8-05 were compared with stress intensity factors calculated by three methods (finite element results, API 579-1/ASME FFS-1 2007 Fitness-For-Service and ASME Boiler and Pressure Vessel Code Section XI). The effects of Poisson's ratio and anisotropic elastic modulus on stress intensity factors were also discussed.

이종금속용접부 예방정비 방법에 따른 잔류응력 분포 고찰

송태광,최영환,박정순,정해동,오창영,Song, Tae-Kwang,Choi, Young Hwan,Park, Jeong Soon,Chung, Hae-Dong,Oh, Chang-Young 한국압력기기공학회 2011 한국압력기기공학회 논문집 Vol.7 No.4

This paper presents the effects of preventive maintenance schemes on the residual stress distributions in dissimilar metal welds. Dissimilar metal weld is known susceptible to PWSCC and thus, effective maintenance schemes to prevent PWSCC are needed. Three preventive maintenances schemes, i.e. weld overlay, MSIP and inlay weld which are widely used in nuclear power plants, are selected and their effects on welding residual stresses are investigated via finite element analyses. As results, weld overlay and MSIP were proved effective method to mitigate residual stresses and inlay weld, on the other hand, produces strong tensile residual stresses in the inner surface. Although Alloy 690 known to be resistant to PWSCC are used in inlay weld, continuous careful observation are needed since tensile welding residual stresses are key parameter for PWSCC.

국부 감육이 배관 굽힘 컴플라이언스에 미치는 영향

서기완,김재민,김윤재 한국압력기기공학회 2021 한국압력기기공학회 논문집 Vol.17 No.2

The thickness of pipe can be locally reduced during operation due to wall thinning. Due to its significance on structural integrity, many non-destructive detecting techniques and assessment methods are available. In this study, the elastic bending compliance of local wall-thinned pipe is presented in terms of the wall thinning geometry: wall thinning depth, circumferential angle and longitudinal length. Elastic finite element (FE) analysis further shows that the presented equation can be used for any wall thinning shape. The proposed solution differs from FE results by less than 6% for all cases analyzed. The bending compliance increases linearly with increasing longitudinal thinning length and non-linearly with increasing thinning angle and depth.

증기발생기 전열관에 작용되는 정적 하중 평가

박범진,박재학,조영기,Park, Bumjin,Park, Jai Hak,Cho, Young Ki 한국압력기기공학회 2011 한국압력기기공학회 논문집 Vol.7 No.1

If a plugged tube in a steam generator is broken, it may damage nearby intact tubes. To prevent this damage, it is recommended that a stabilizer is installed into the plugged tube. However, the installation cost of a stabilizer is very high. So studies are required to determine the conditions on which the installation is necessary. For this purpose static loads and dynamic loads applied on a tube should be known to estimate the residual strength and remaining fatigue and wear life of a plugged tube. Two-dimensional and three-dimensional computational fluid dynamics (CFD) analyses are performed to obtain the drag coefficient for cross flow to a tube. Using the obtained drag coefficient, the static load can be estimated and the residual strength of a plugged tube can be calculated. An inclined flow problem is also analyzed and the vertical and horizontal forces are obtained and discussed.

간극효과를 고려한 증기발생기 전열관의 3차원 유한요소해석

조영기,박재학,Cho, Young Ki,Park, Jai Hak 한국압력기기공학회 2011 한국압력기기공학회 논문집 Vol.7 No.4

Steam generator is one of the main equipments that affect safety and long term operation in nuclear power plants. Fluid flows inside and outside of the steam generator tubes and induces vibration. To prevent the vibration the tubes are supported by AVB (anti vibration bar). When the steam generator tube contact to AVB, it is damaged by the accumulation of wear and corrosion. Therefore studies are required to determine the effects of the gap between the steam generator tube and AVB. In order to obtain the stress and the displacement distributions of the steam generator tube, three dimensional finite element analyses were performed by using the commercial program ANSYS. Using the calculated the stress and the displacement distributions, the static residual strength of the steam generator tube can be evaluated. The results show that the stress and displacement of the steam generator tube increase significantly compared with the results from a zero-gap model.

원전 단종 밸브의 DED 방식 금속 3D프린팅 제작 및 성능시험

장경남 한국압력기기공학회 2021 한국압력기기공학회 논문집 Vol.17 No.2

The 3D printing technology is one of the fourth industrial revolution technology that drives innovation in the manufacturing process, and should be applied to nuclear industry for various purposes according to the manufacturing trend change. In nuclear industry, it can be applied to manufacture obsolete items and new designed parts in advanced reactors or small modular reactors (SMRs), replacing the traditional manufacturing technologies. A gate valve body was manufactured, which was obsolete in nuclear power plant, using DED(Directed Energy Deposition) metal 3D printing technology after restoring design characteristics including 3D design drawing by reverse engineering. The 3D printed valve body was assembled with commercial parts such as seat-ring, disk, stem, and actuator for performance test. For the valve assembly, including 3D printed valve body, several tests were performed, including pressure test, end-loading test, and seismic test according to KEPIC MGG and KEPIC MFC. In the pressure test, hydraulic pressure of 391kgf/cm2 was applied to 3D printed valve body, and no leak was detected. Also the 3D printed valve assembly was performed well in end-loading and seismic tests.

초임계압 화력 과열기 구조의 고신뢰도 건전성 평가 방법

이형연,주용선,최현선,원민구,허남수 한국압력기기공학회 2020 한국압력기기공학회 논문집 Vol.16 No.1

Integrity evaluations on a platen superheater were conducted as per ASME Section VIII Division 2(hereafter ‘ASME VIII(2)’) which was originally used for design with implicit consideration of creep effects. A platen superheater subjected to severe loading conditions of high pressure and high temperature at creep regime in a supercritical thermal plant in Korea was chosen for present study. Additional evaluations were conducted as per nuclear-grade high-temperature design rule of RCC-MRx that takes creep effects into account explicitly. Comparisons of the two results from ASME VIII(2) and RCC-MRx were conducted to quantify the conservatism of ASME VIII(2). From present analyses, it was shown that the design evaluation results exceeded allowable limits of RCC-MRx for the plant design conditions although limits of ASME VIII(2) were satisfied regardless of operation time, which means that design as per ASME VIII(2) might be potentially non-conservative in case of operation in creep range. A high-temperature design evaluation program as per RCC-MRx, called ‘HITEP_RCC-MRx’ has been used and it was shown that pressure boundary components can be designed reliably with the program especially for the loading conditions of long-term creep conditions.