• Nuclear Engineering and Technology
• /
• 제54권5호
• /
• pp.1588-1596
• /
• 2022

In the nuclear power plants, stainless steel is widely used for fabrication of various components such as piping and pipe fittings. These piping components are subjected to cyclic loading due to start up and shut down of the nuclear power plants. The application of cyclic loading may lead to initiation of crack at stress raiser locations such as nozzle to piping connection, crown of piping bends etc. of the piping system. Crack initiation can also take place from the flaws which have gone unnoticed during manufacturing. Therefore, prediction of crack initiation life would help in decision making with respect to plant operational life. The primary objective of the present study is to compile various analytical models to predict the crack initiation life of the pipes with notch. Here notch simulates the stress raisers in the piping system. As a part of the study, Coffin-Manson equations have been benchmarked to predict the crack initiation life of pipe with notch. Analytical models proposed by Zheng et al. [1], Singh et al. [2], Yang Dong et al. [25], Masayuki et al. [33] and Liu et al. [3] were compiled to predict the crack initiation life of SA312 Type 304LN stainless steel pipe with notch under fatigue loading. Tensile and low cycle fatigue properties were evaluated for the same lot of SA312 Type 304LN stainless steel as that of pipe test. The predicted crack initiation lives by different models were compared with the experimental results of three pipes under different frequencies and loading conditions. It was observed that the predicted crack initiation life is in very good agreement with experimental results with maximum difference of ±10.0%.

• 한국지진공학회논문집
• /
• 제26권1호
• /
• pp.13-22
• /
• 2022

This study proposes a low-cycle fatigue life derived from measurement points on pipe elbows, which are components that are vulnerable to seismic load in the interface piping systems of nuclear power plants that use seismic isolation systems. In order to quantitatively define limit states regarding leakage, i.e., actual failure caused by low-cycle fatigue, in-plane cyclic loading tests were performed using a sine wave of constant amplitude. The test specimens consisted of SCH40 6-inch carbon steel pipe elbows and straight pipes, and an image processing method was used to measure the nonlinear behavior of the test specimens. The leakage lines caused by low-cycle fatigue and the low-cycle fatigue curves were compared and analyzed using the relationship between the relative deformation angles, which were measured based on each of the measurement points on the straight pipe, and the moment, which was measured at the center of the pipe elbow. Damage indices based on the combination of ductility and dissipation energy at each measurement point were used to quantitatively express the time at which leakage occurs due to through-wall cracking in the pipe elbow.

• Nuclear Engineering and Technology
• /
• 제28권1호
• /
• pp.27-35
• /
• 1996

본 논문에서는, 가압경수로 발전소의 가압기 밀림관내 비정상상태의 열성층현상에 대한 계산 모델을 제안하여 배관내의 온도분포, 유동특성 및 열응력에 대해 연구하였다. 경계면이 시간에 따라 변화가 없거나 정상상태에서 개발된 다른 모델과는 달리 본 모델에서는 고온 및 저온유체 사이의 경계면을 시간의 함수로 가정하였다. 열성층현상에 대한 무차원지배방정식은 SIMPLE 알고리즘을 사용하여 해를 구하였다. 본 수치계산의 결과는 주어진 조건하에서 무차원시간이 약 27.0 일 때 배관의 고온부 및 저온부사이의 최대무차원온도차는 0.78정도이었고, 이때의 열성층 현상에 의한 최대 열응력은 276 MPa로 계산되었다.

• 한국압력기기공학회 논문집
• /
• 제7권3호
• /
• pp.40-47
• /
• 2011

KSNPs(Korea Standard Nuclear Power Plant) have been applied the break exclusion criteria to the high energy lines passing through containment penetration area to ensure that piping failures would not cause the loss of containment isolation function, and to reduce the resulting dynamic effects. Systems with the criteria are the Main Steam system, Feed Water system, Steam Generator Blowdown system, and Chemical & Volume Control system. In accordance with FSAR(Final Safety Analysis Report), a 100% volumetric examination by augmented in-service inspection of all pipe welds appled the break exclusion criteria is required for the break exclusion application piping. However, it is difficult to fully satisfy the requirements of inspection because 12", 8" and 6" weldolet weldments of Main Steam pipe line have complex structural shapes. To resolve the difficulty on the application of conventional UT(Ultrasonic Testing) technique, realistic mock-ups and UT calibration blocks were made. Simulations of conventional UT were performed utilizing CIVA, a commercial NDE(Nondestructive Examination) simulation software. Phased array UT experiments were performed through mock-up including artificial notch type flaws. A phased array UT technique is finally developed to improve the reliability of ultrasonic test at main steam line pipe to 12", 8" and 6" branch connection weld.

• 대한기계학회논문집B
• /
• 제35권10호
• /
• pp.1097-1103
• /
• 2011

액적충돌침식은 증기나 공기에 포함된 액적이 금속 소재에 고속으로 충돌할 때 모재가 손상되는 현상이다. 액적충돌침식 손상은 증기터빈이나 빗방울과 부딪치는 항공기에서 주로 발생되어 왔으나 최근에는 원전 배관에서도 발생하고 있다. 원전 배관 중에서도 특히 높은 압력강하가 발생하고 2상 증기가 흐르는 배관에서 주로 발생한다. 실제 2011년 초반 국내 한 원전에서는 2상 증기가 흐르는 배관에서 액적충돌침식 손상으로 인한 누설이 발생한 바 있다. 본 논문에서는 액적충돌침식 손상이 발생한 배관에 대하여 손상을 억제할 수 있는 설계변경 방안에 관한 연구를 수행하였다. 설계변경은 유체 유동측면에서 분석하였으며, 상용 수치해석 코드인 FLUENT를 이용하였다.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
• /
• pp.435-440
• /
• 1996

Large Loss-of-Coolant Accidents analyzed in Final Safety Analysis Reports are usually covered by Reactor Inlet Header. Reactor Outlet Header and Primary Pump Suction breaks as representative cases. In this study we analyze the total (guillotine) break of an Emergency Core Cooling System (ECCS) pipe located at the ECCS injection point into the Primary Heat Transport System (PHTS). It was expected that thermal-hydraulic behaviors in the PHT and ECC systems are different from those of a Reactor Inlet Header break, having an equivalent break size. The main purpose of this study is to get insights on the differences occurred between the two cases and to assess these differences from the phenomenon behavior point of view. It was also investigated whether the ECCS line break analysis results could be covered by header break analysis results. The study reveals that as the intact loop has almost the same behavior in both analyzed cases. broken loop behavior is different mostly regarding sheath temperature in the critical core pass and pressure decrease in the broken Reactor Inlet Header. Differences are also met in the ECCS behavior and in event sequences timings.

• Corrosion Science and Technology
• /
• 제3권5호
• /
• pp.216-221
• /
• 2004

The operating experience showed that the fatigue is one of the major piping failure mechanisms in nuclear power plants (NPPs). The pressure and/or temperature loading transients, the vibration, and the mechanical cyclic loading during the plant operation may induce the fatigue failure in the nuclear piping. Recently, many fatigue piping failure occurred at the socket weld area have been widely reported. Many failure cases showed that the gap requirement between the pipe and fitting in the socket weld was not satisfied though the ASME Code Sec. III requires 1/16 inch gap in the socket weld. The ASME Code OM also limits the vibration level of the piping system, but some failure cases showed the limitation was not satisfied during the plant operation. In this paper, the fatigue behavior of the socket weld in the nuclear piping was estimated by using the three dimensional finite element method. The results are as follows. (1) The socket weld is susceptible to the vibration if the vibration levels exceed the requirement in the ASME Code OM. (2) The effect of the pressure or temperature transient load on the socket weld in NPPs is not significant because of the very low frequency of the transient during the plant lifetime operation. (3) 'No gap' is very risky to the socket weld integrity for the specific systems having the vibration condition to exceed the requirement in the ASME OM Code and/or the transient loading condition. (4) The reduction of the weld leg size from $1.09*t_1$ to $0.75*t_1$ can affect severely on the socket weld integrity.

• 대한기계학회:학술대회논문집
• /
• 대한기계학회 2008년도 추계학술대회A
• /
• pp.88-93
• /
• 2008

Weld overlay is one of the residual stress mitigation method which arrest crack. An overlay weld sued in this manner is termed a preemptive weld overlay(PWOL). PWOL was good for distribution of residual stress of dissimilar metal weld(DMW) by previous research. Because range of overlay welding is wide relatively, residual stress distribution on PWR is affected by welding sequence. In order to examine the effect of welding sequence, PWOL was applied to a specific DMW of KORI nuclear power plant by finite element analysis method. As a result, the welding direction that from nozzle to pipe is better good for residual stress distribution on PWR.

• Nuclear Engineering and Technology
• /
• 제52권9호
• /
• pp.2119-2129
• /
• 2020

Flow-accelerated corrosion (FAC), liquid droplet impingement erosion (LDIE), cavitation and flashing can cause continuous wall-thinning in nuclear secondary pipes. In order to prevent pipe rupture events resulting from the wall-thinning, most NPPs (nuclear power plants) implement their management programs, which include periodic thickness inspection using UT (ultrasonic test). Meanwhile, it is well known in field experiences that the thickness measurement errors (or deviations) are often comparable with the amount of thickness reduction. Because of these errors, it is difficult to estimate wall-thinning exactly whether the significant thinning has occurred in the inspected components or not. In the previous study, the authors presented an approximate estimation procedure as the first step for thickness measurement deviations at each inspected component and the statistical & quantitative characteristics of the measurement deviations using plant experience data. In this study, statistical significance was quantified for the current methods used for wall-thinning determination. Also, the authors proposed new estimation procedures for determining local wall-thinning to overcome the weakness of the current methods, in which the proposed procedure is based on analysis of variance (ANOVA) method using subgrouping of measured thinning values at all measurement grids. The new procedures were also quantified for their statistical significance. As the results, it is confirmed that the new methods have better estimation confidence than the methods having used until now.

• 한국안전학회지
• /
• 제26권2호
• /
• pp.6-12
• /
• 2011

The failure probabilities of pipes in nuclear power plants due to stress corrosion are obtained using the P-PIE program, which is developed for evaluating failure probability of pipes based on the existing PRAISE program. Leak, big leak and LOCA(loss of coolant accident) probabilities are calculated as a function of operating time for several pipes in a domestic nuclear plant. The sensitivity analysis is also performed to find out the important parameters for the failure of pipes due to stress corrosion. The results show that the steady state oxygen concentration and steady state temperature are important parameters and failure probability is very low when the oxygen concentration is maintained according to the regulation.