• Nuclear Engineering and Technology
• /
• 제36권1호
• /
• pp.112-120
• /
• 2004

The purpose of this paper is to analyze the piping failure frequency for the main feedwater system in domestic nuclear power plants(NPPs) for the application to an in-service inspection(ISI), leak before break(LBB) concept, aging management program(AMP), and probabilistic safety analysis(PSA). First, a database was developed for piping failure events in domestic NPPs, and 23 domestic piping failure events were collected. Among the 23 events, 12 locations of wall thinning due to flow accelerated corrosion(FAC) were identified in the main feedwater system in 4 domestic WH 3-loop NPPs. Two types of the piping failure frequency such as the damage frequency and rupture frequency were considered in this study. The damage frequency was calculated from both the plant population data and damage(s) including crack, wall thinning, leak, and/or rupture, while the rupture frequency was estimated by using both the well-known Jeffreys method and a new method considering the degradation due to FAC. The results showed that the damage frequencies based on the number of the base metal piping susceptible to FAC ranged from $1.26{\times}10^{-3}/cr.yr\;to\;3.91{\times}10^{-3}/cr.yr$ for the main feedwater system of domestic WH 3-loop NPPs. The rupture frequencies obtained from the Jeffreys method for the main feedwater system were $1.01{\times}10^{-2}/cr.yr\;and\;4.54{\times}10^{-3}/cr.yr$ for the domestic WH 3-loop NPPs and all the other domestic PWR NPPs respectively, while those from the new method considering the degradation were higher than those from the Jeffreys method by about an order of one.

• Nuclear Engineering and Technology
• /
• 제55권10호
• /
• pp.3581-3590
• /
• 2023

High-temperature design evaluations were conducted on Type 316L stainless steel piping for a 700 ℃ large-capacity thermal energy storage verification test loop (TESET) under construction at KAERI. The hot leg piping with sodium coolant at 700 ℃ connects the main components of the loop heater, hot storage tank, and air-to-sodium heat exchanger. Currently, the design rules of ASME B31.1 and RCC-MRx provide design procedures for high-temperature piping in the creep range for Type 316L stainless steel. However, the design material properties around 700 ℃ are not available in those rules. Therefore, a number of material tests, including creep tests at various temperatures, were conducted to determine the insufficient material properties and relevant design coefficients so that high-temperature design on the 700 ℃ piping may be possible. It was shown that Type 316L stainless steel can be used in a 700 ℃ high-temperature piping system of Generation IV reactor systems or a renewable energy systems, such as thermal energy storage systems, for a limited operation time.

• 대한기계학회:학술대회논문집
• /
• 대한기계학회 2003년도 춘계학술대회
• /
• pp.617-621
• /
• 2003

The purpose of this paper is to analyze piping failure trend of safety pipings In domestic nuclear power plants. First, database for the piping failure was constructed with 105 data fields. The database includes plant population data, event data, and service history data. 7 kinds of piping failures in domestic NPPs were investigated. Among the 7 cases, detailed root causes were investigated for 3 cases. The first one is pipe wall thinning in main feedwater pipings of Westinghouse 3 loop type plants. The root cause of the wall thinning was flow accelerated corrosion near welding area. The next one is leak event in chemical and volume control system(CVCS) due to vibration. Some cracks occurred in socket welding area. The events showed that the integrity or socket weld is very vulnerable to vibration. The last one is also a leak event in primary sampling line in Korean standard reactor due to thermal fatigue. Although the structural integrity was not maintained by the events, there was no effect on nuclear safety in the above 3 piping failure eases.

• 한국압력기기공학회 논문집
• /
• 제13권2호
• /
• pp.67-74
• /
• 2017

Recently there have been many concerns on structural integrity of nuclear piping under seismic loadings. In terms of failure of nuclear piping due to seismic loadings, an important failure mechanism is low cycle fatigue with large cyclic displacements. To investigate the effects of seismic loading on low cycle fatigue behavior of nuclear piping, the cyclic behavior of materials and nuclear piping needs to be accurately estimated. In this paper, the non-linear finite element (FE) analyses have been carried out to evaluate the effects of three different cyclic hardening models on cyclic behavior of materials and nuclear piping, such as isotropic hardening, kinematic hardening and combined hardening.

• 비파괴검사학회지
• /
• 제35권5호
• /
• pp.321-331
• /
• 2015

냉각 위상잠금 적외선열화상 기법을 이용하여 원전 배관 루프시스템의 가열결함 검출의 선행연구를 통하여 냉각결함 검출조건의 적용에 관한 연구를 수행하였다. 배관의 결함가공은 감육 길이, 감육 깊이를 변화시켜 결함조건을 가공하여 루프시스템을 제작하였다. 사용된 장비는 적외선열화상 카메라와 냉각장치를 사용하였으며 냉각장치와 대상 루프시스템과의 거리는 2m로 고정시켜 실험을 수행하였다. 실험 결과의 분석을 위하여 냉각온도 분포, 위상데이터를 확보하고, 이를 분석하여 결함 길이를 측정하였다. 냉각결함 검출조건은 적외선열화상 데이터보다 위상잠금 적외선열화상 데이터가 측정 결과의 신뢰도가 높았다.

• 대한기계학회논문집A
• /
• 제25권7호
• /
• pp.1031-1038
• /
• 2001

In order to verify the analytical methods predicting failure behavior of cracked piping, full-scale pipe tests are crucial in nuclear power plant piping. For this reason, series of international test programs have been conducted. However, full-scale pipe tests require expensive testing equipment and long period of testing time. The objective of this paper is to develop a test system which can economically simulate the full-scale pipe test regarding the integrity evaluation. This system provides the failure behavior of cracked pipe by testing a wide-plate specimen. The system provides the failure behavior of cracked pipe by testing a wide-plate specimen. The system was developed for the integrity evaluation of nuclear piping based on the methodology of hardware-in-the-loop (HiL) simulation. Using this simulator, the piping integrity can be evaluated based on the elastic-plastic behavior of full-scale pipe, and the high cost full-scale pipe test may be replaced with this economical system.

• Nuclear Engineering and Technology
• /
• 제26권2호
• /
• pp.225-236
• /
• 1994

SEBIM 밸브 상부에 위치한 밀봉수의 급격한 방출은 밸브 후단의 방출배관계통에 큰 운동량과 관성력의 작용을 초래한다. 본 연구는 밸브개방시 방출배관계통의 후단에 발생하는 열수력학적 과도현상을 분석하기 위한 해석절차 및 해석결과를 다루고 있으며, 이 분석을 위해 RELAP5 /MOD3 를 사용하였다. RELAP5 /MOD3 분석을 위하여, 방출관 계통과 SEBIM 밸브의 개방특성 및 밀봉수 방출등의 적절한 모델방법이 제시되었다. 또한 접합부(junction)와 체적(volume)의 제어 플래그 (flag)에서 옵션(option)의 적절한 선택을 위하여 민감도분석도 수행되었다. 분석결과, SEBIM 밸브 방출배관계통의 밀봉수 방출에 따른 열수력학적 과도현상을 분석하는데 RELAP5 /MOD3가 적절히 사용될 수 있음을 알 수 있었다. 민감도 분석결과로부터, 밀봉수 방출해석을 위해서는 적절한 기하학적 압력분포를 가지는 완만한(smooth) 면적변화 및 비평형 옵션(option), 적절한 시간간격(time step)의 사용이 필수적인 것을 알 수 있었다.

• 한국압력기기공학회 논문집
• /
• 제6권2호
• /
• pp.54-60
• /
• 2010

Piping failures due to thermal fatigue have been widely reported in normally stagnant non-isolable reactor coolant branch lines. Since the thermal fatigue due to thermal stratification was not considered in the piping fatigue design in old NPPs, it is important to evaluate the effect of thermal stratification on the integrity of branch lines. In this study, geometrical screening criteria for Up-horizontal branch lines in MRP-132 were applied to SI(Safety Injection) lines of KSNP 2-loop and WH 3-loop. Some computational fluid dynamic(CFD) analyses on the Reactor Coolant System(RCS) branch lines were also performed to develop the regulatory guidelines for screening criteria. As a result of applying MRP-132 screening criteria, KSNP 2-loop and WH 3-loop SI lines are determined to need further detailed evaluation. Results of CFD analyses show that both valve isolation and amount of leakage through valve can be used as technical bases for the screening criteria on the thermal fatigue analysis.

• 한국압력기기공학회 논문집
• /
• 제8권1호
• /
• pp.33-38
• /
• 2012

A PHE (Process Heat Exchanger) in a nuclear hydrogen system is a key component required to transfer heat energy of $950^{\circ}C$ generated in a VHTR (Very High Temperature Reactor) to a chemical reaction that yields a large quantity of hydrogen. Korea Atomic Energy Research Institute has established a small-scale gas loop for the performance test on VHTR components and recently has manufactured a medium-scale PHE prototype made of Hastelloy-X. A performance test on the PHE prototype is scheduled in the gas loop. In this study, high-temperature structural analysis modeling, and macroscopic thermal and structural analysis of the medium-scale PHE prototype by imposing the established displacement boundary constraints in the previous research were carried out under the gas loop test condition. The results obtained in this study will be compared with performance test results.

• 한국압력기기공학회 논문집
• /
• 제6권1호
• /
• pp.57-64
• /
• 2010

PHE(Process Heat Exchanger) is a key component required to transfer heat energy of $950^{\circ}C$ generated in a VHTR(Very High Temperature Reactor) to the chemical reaction that yields a large quantity of hydrogen. Korea Atomic Energy Research Institute established the gas loop for the performance test of components, which are used in the VHTR, and they manufactured a PHE prototype to be tested in the loop. In this study, as part of the high-temperature structural-integrity evaluation of the PHE prototype, which is scheduled to be tested in the gas loop, we carried out high-temperature structural-analysis modeling, thermal analysis, and thermal expansion analysis of the PHE prototype. The results obtained in this study will be used to design the performance test setup for the PHE prototype.