• Nuclear Engineering and Technology
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• 제32권1호
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• pp.67-76
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• 2000

Significant wall thinning due to flow accelerated corrosion(FAC）was recently reported in main feedwater pipes in 3 Korean pressurized water reactor(PWR) plants. The main feedwater pipes in one plant were repaired using overlay weld method at the outside of pipe, while those in 2 other plants were replaced with new pipes. In this study, the effect of the wall thinning in the main feedwater pipes on piping integrity was evaluated using finite element method. Especially, the effects of both the overlay weld repair and the stress concentration in notch-type thinned area on the piping integrity were investigated. The results are as follows : (1) The piping load carrying capacity may significantly decrease due to FAC. In special, the load carrying capacity of the main feedwater pipe was reduced by about 40% during about 140 months operation in Korean PWR plants. (2) By performing overlay weld repair at the outside of pipe, the piping load carrying capacity can increase and the stress concentration level in the thinned area can be reduced.

• 한국압력기기공학회 논문집
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• 제12권1호
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• pp.101-106
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• 2016

NSSS (Nuclear Steam Supply System) and BOP (Balance of Plant) design works for PGSFR (Prototype Gen-IV Sodium Fast Reactor) have been conducted in Korea. NSSS major components, e.g. reactor vessel, steam generator and secondary sodium main pipes, are designed according to the rule of ASME boiler and pressure vessel code division 5, in which DBA (Design by Analysis) methods are used in the stress assessments. However, there is little discussions about detail rules for BOP piping design. In this paper, the detail methodologies of BOP piping stress assessment are discussed including safety systems and non-safety system pipings. It is confirmed that KEPIC MGE(ASME B31.1) and ASME BPV code division 5 HCB-3600 can be used in stress assessments of non-safety pipes and class B pipes, respectively. However, class A pipe design according to ASME BPV code division 5 HBB-3200 has many difficulties applying to PGSFR BOP design. Finally, future development plan for class A pipe stress assessment method is proposed in this paper.

• 플랜트 저널
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• 제8권2호
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• pp.52-58
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• 2012

Pipe means the connection of the tube in order to transfer fluid from one device to another device. The piping stress analysis is to analyze the structural stability considering the location and the features of piping support after completing the piping design, The allowable stresses comply with the requirements of the relevant standards by examining whether the support of the function and location of pipe or re-operation is confirmed. Allowable stresses are to make sure that the maximum stress should not exceed the allowable stress presented in the ASME B31.1 POWER PIPING code. ASME B31.1 POWER PIPING code ensures a smooth stress analysis can be performed during the initial pipe stress analysis as provided in the case of straight pipe to the horizontal distance between the supports. However, because there is no criteria set in the case of curved pipe, the optimum pipe supporting points were studied in this paper. As mentioned about the curved pipe, loads applied to the support of the position of 17% and 83% of the position relative to the elbow part have results similar to the load acting on the support of straight pipe.

• Nuclear Engineering and Technology
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• 제29권4호
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• pp.310-319
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• 1997

A finite element analysis of a trunnion pipe anchor is presented. The structure is analyzed for the case of internal pressure and moment loadings. The stress results are categorized into the average (membrane) stress, the linearly varying (bending) stress and the peak stress through the thickness. The resulting stresses are interpreted per Section III of the ASME Boiler and Pressure Vessel Code from which the Primary(B$_1$), Secondary(C$_1$) and Peak(K$_1$) stress indices for pressure, the Primary (B$_2$), Secondary(C$_2$) and Peak(K$_2$) stress indices for moment are developed. Based on the comparison between stress value by stress indices derived in this paper and stress value represented by the ASME Code Case N-391-1, the empirical equations for stress indices are effectively used in the piping stress analysis. Therefore, the use of empirical equations can simplify the procedure of evaluating the local stress in the piping design stage.

• 한국기계가공학회지
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• 제13권1호
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• pp.23-29
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• 2014

The purpose of this paper is to assess the structural integrity of piping penetrations for nuclear power plants. A piping qualification analysis describes loads due to deadweight, pressure difference acts normal to the plate, thermal transients, and earthquakes, among other events, on piping penetrations that have been modeled as an anchor. Amodel was analyzed using a commercial finite element program. Apiping penetration analysis model was constructed with an assembly of pipe, head fittings and sleeves. Normally, the design load, thus obtained, will consist of three moments and three forces, referred to a Cartesian coordinate system. When comparing the stress analysis results from each required cutting position, the general membrane stress intensities and local membrane plus bending stress intensities during a structural evaluation cannot exceed the allowable amount of stress for the design loads. Therefore, the piping penetration design satisfies the code requirements.

• 한국압력기기공학회 논문집
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• 제14권2호
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• pp.59-68
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• 2018

This study compares true stress-true strain curves obtained by tensile tests of various piping materials with bi-linear stress-strain approximation suggested in the JSME Code Case(CC) Draft, a guideline for piping seismic inelastic response analysis. Based on the comparisons, the reliability of the bi-linear approximation is evaluated. It is found that bi-linear stress-strain curve of TP316 stainless steel is in good agreement with its true stress-true strain curve. However, Bi-linear stress-strain curves of TP304 stainless steel and carbon steels determined by the approximation cannot appropriately estimate their stress-strain behavior. Accordingly new bi-linear approximations for carbon steels and low-alloy steels are proposed. The proposed bi-linear approximations for carbon and low-alloy steels, which include the temperature effect on strength and hardening of material, estimate their stress-strain behavior reasonably well.

• 에너지공학
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• 제27권2호
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• pp.14-25
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• 2018

복합화력발전 플랜트는 천연가스와 같은 연료로 가스 터빈을 작동시킨 후 잔류 열로 증기를 생산하는 사이클을 가지고 있다. 연료가스는 압축기 및 열교환기를 통해 4~5 MPa, $200^{\circ}C$ 수준의 상태로 가스터빈에 공급된다. 본 연구에서는 가스 연료공급 배관계통의 안전 운영 및 건전성 확보를 위해 배관 시스템 응력을 고려한 위험도 평가 기법 연구를 수행하였다. 위험도 평가 기법으로 잘 알려진 API 580/581 RBI 코드에서는 위험도에 배관 응력의 영향을 반영이 제한적이다. 따라서 배관 해석을 이용하여 배관의 시스템적 응력을 위험도의 파손확률로써 인자화하는 접근법을 제시하였다. 해석은 가상 발전 플랜트의 가스연료 공급 배관의 설계 데이터에 근거하여 배관 시스템 응력 해석을 수행하였다. API 코드에 의해서 평가된 파손확률 등급과 배관해석을 이용한 응력비 평가 결과를 비교하였다.

• 한국압력기기공학회 논문집
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• 제11권1호
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• pp.69-73
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• 2015

The study investigated effects of residual stresses and Charpy impact energy on brittle fractures of the butt weld between the valve and the piping, and of the valve body in nuclear power plants via a linear elastic fracture mechanics approach in the ASME B&PV Code, Sec.XI and finite element analysis. Weld residual stress in a butt weld between close check valve and piping, and residual stress in the valve due to casting process were assumed to be proportional to yield strength of base metal. Operating stresses in the butt weld and the valve body were calculated using approximate engineering formulae and finite element analysis, respectively. Applied stress intensity factors were calculated by assuming postulated cracks with specific sizes and then by substituting the residual stresses and the operating stresses into engineering formulae presented in the ASME B&PV Code, Sec.III. Plane strain fracture toughness was derived by using a correlation between Charpy V-notch impact energy and fracture toughness. Structural integrity of the weld and the body against brittle fracture was assessed by using the applied stress intensity factors, plane strain fracture toughness and the linear elastic fracture mechanics approach. As a result, it was identified that the structural integrity was maintained with decreasing the residual stress levels and increasing the Charpy V-notch impact energy.

• 대한기계학회논문집A
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• 제33권10호
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• pp.1177-1186
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• 2009

High temperature steam pipes of thermal power plant are subject to a severe thermal range and usually operates well into the creep range. Cyclic operation of the plant subjects the piping system to mechanical and thermal fatigue damages. Also, poor or malfunctional supports can impose massive loads or stress onto the piping system. In order to prevent the serious damage and failure of the critical piping system, various inspection methods such as visual inspection, computational analysis and on-line piping displacement monitoring were developed. 3-dimensional piping displacement monitoring system was developed with using the aluminum alloy rod and rotary encoder sensors, this system was installed and operated on the high temperature steam piping of "Y" thermal power plant successfully. It is expected that this study will contribute to the safety of piping system, which could minimize stress and extend the actual life of critical piping.

• Nuclear Engineering and Technology
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• 제51권2호
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• pp.561-572
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• 2019

In the design criterion for the nuclear power plant piping system, the limit state of the piping against an earthquake is assumed to be plastic collapse. The failure of a common piping system, however, means the leakage caused by the cracks. Therefore, for the seismic fragility analysis of a nuclear power plant, a method capable of quantitatively expressing the failure of an actual piping system is required. In this study, it was conducted to propose a quantitative failure criterion for piping system, which is required for the seismic fragility analysis of nuclear power plants against critical accidents. The in-plane cyclic loading test was conducted to propose a quantitative failure criterion for steel pipe elbows in the nuclear power plant piping system. Nonlinear analysis was conducted using a finite element model, and the results were compared with the test results to verify the effectiveness of the finite element model. The collapse load point derived from the experiment and analysis results and the damage index based on the stress-strain relationship were defined as failure criteria, and seismic fragility analysis was conducted for the piping system of the BNL (Brookhaven National Laboratory) - NRC (Nuclear Regulatory Commission) benchmark model.