• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.2
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• pp.35-46
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• 2020

In this paper, a fracture mechanics evaluation system which can be used to assess the leak-before-break (LBB) of nuclear piping is developed. Existing solutions for calculating the fracture mechanics parameters (J-integral and crack opening displacement) required for LBB evaluation were firstly presented. Then a module for calculating J-integral and COD was developed, with an additional module for predicting the critical load based on the crack driving force diagram to finally develop a fracture mechanics evaluation system. To confirm the validity of the proposed evaluation system, finite element (FE) analysis was performed, and the FE J-integral and COD results were compared with prediction results using the J-integral and COD estimations program. Furthermore, the critical load assessment module was verified by comparing the actual pipe test results (Battelle test data) with prediction results using the proposed program.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.24 no.11
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• pp.2796-2803
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• 2000

For traditional Leak-Before-Break(LBB) analyses, symmetric conditions were assumed for a pipe-nozzle interface to simplify the analysis in calculating J-integral. However. this assumption could result in an overly conservative design criteria for a pipe-nozzle interface, Since the pipe-nozzle interface is asymmetric due to the difference of stiffness between pipe and nozzle, it is required to develop a new methodology considering the nozzle effect. The objective of this paper is to evaluate the effect of nozzle no the development of LBB design criteria for nuclear pipings. For this purpose, extensive finite element analysis were performed to evaluate the effect of nozzle on Crack Opening Area(COA), Detectable Leakage Crack(DLC) length and J-integral values. In conclusion, it was proven that the application of LBB concept could be extended for more nuclear piping system by considering the nozzle.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.18 no.1
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• pp.19-25
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• 2022

Nuclear Power Plants (NPP) should be designed to have sufficient safety margins and to ensure seismic safety against earthquake that may occur during the plant life time. After the 9.12 Gyeongju earthquake accident, the structural integrity of nuclear power plants due to the beyond design basis earthquake is one of key safety issues. Accordingly, it is necessary to conduct structural integrity evaluations for domestic NPPs under beyond design basis earthquake. In this study, the Level 3 LBB (Leak Before Break) evaluation was performed by considering the beyond design basis earthquake for the surge line of a OPR1000 plant of which design basis earthquake was set to be 0.2g. The beyond design basis earthquake corresponding to peak ground acceleration 0.4g at the maximum stress point of the surge line was considered. It was confirmed that the moment behaviors of the hot leg and pressurized surge nozzle were lower than the maximum allowable loading in moment-rotation curve. It was also confirmed that the LBB margin could be secured by comparing the LBB margin through the Level 2 method. It was judged that the margin was secured by reducing the load generated through the compliance of the pipe.

• Transactions of the Korean Society of Mechanical Engineers
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• v.19 no.3
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• pp.731-740
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• 1995

The objective of this paper is to develop defect assessment technology for integrity evaluation of CANDU pressure tubes. In fracture mechanics analysis, three-dimensional and two-dimensional (line-spring model) finite element analyses were performed to obtain the stress intensity factor for axial and circumferential surface cracks. In leak before break (LBB) analysis, heat transfer analyses for through-wall cracks were performed by considering the cooling effect and the LBB application time was computed. It was shown that the analytical results obtained in this study provide less-conservative but accurate solution for defect assessment of CANDU pressure tubes.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.1
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• pp.42-48
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• 2020

The effect of improving the tensile and J-R fracture toughness properties of SA508 Gr.1a on the LBB margin for the main steam pipe is investigated. The material properties and microstructure images of the existing main steam piping material SA106 Gr.C used in domestic nuclear power plants and the newly selected material SA508 Gr.1a were compared. For each material, LBB margins were calculated and compared through finite element analysis and crack instability evaluation. The LBB margin of the improved SA508 Gr.1a is found to be greatly improved compared to that of the existing SA106 Gr.C and SA508 Gr.1a. This is because of the increased material's strength and J-R fracture toughness compared to the previous materials. In order to analyze the effect of physical property change on the LBB margin, the sensitivity of each LBB margin according to the variation of tensile strength and J-R fracture toughness was analyzed. The effect of the change in tensile strength was found to be greater than that of the change in fracture toughness. Therefore, an increase in strength significantly influenced the improvement of the LBB margin of the improved SA508 Gr.1a.

• Journal of the KSME
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• v.51 no.12
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• pp.37-41
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• 2011

이 글에서는 원전배관의 안전설계 개념인 양단순간파단(DEGB: Double Ended Guillotine Break) 및 파단전누설(LBB: Leak Before Break)에 대해 설명하고, 파단전누설 설계를 위한 다양한 실배관 파과저항시험 방법 및 실배관 파괴저항시험의 필요성에 대해 소개하고자 한다.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.70-76
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• 2013

Several finite element models for the leak-before-break (LBB) assessment of overlay dissimilar metal weldment were constructed and analyzed to develop a simple finite element modeling method. The J-integral, crack opening displacement (COD) and J-integral distribution along the crack front in thickness direction due to the applied moment were obtained from the analysis results of the constructed finite element models, and studied compared to the previous literatures. It is concluded that the modeling with base material only is simple and produces a slightly conservative results compared to the complex modeling composed with weld metal and base metal in the calculation of J-integrals and COD values which are used for the calculation of fracture toughness and postulated leakage crack length respectively.

• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.332-339
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• 2015

The leak before break (LBB) concept is widely used in designing pipe lines in nuclear power plants. According to the concept, the amount of leaking liquid from a pipe should be more than the minimum detectable leak rate of a leak detection system before catastrophic failure occurs. Therefore, accurate estimation of the leak rate is important to evaluate the validity of the LBB concept in pipe line design. In this paper, a program was developed to estimate the leak rate through circumferential cracks in pipes in nuclear power plants using the Henry-Fauske flow model and modified Henry-Fauske flow model. By using the developed program, the leak rate was calculated for a circumferential crack in a sample pipe, and the effect of the flow model on the leak rate was examined. Treating the crack morphology parameters as random variables, the statistical behavior of the leak rate was also examined. As a result, it was found that the crack morphology parameters have a strong effect on the leak rate and the statistical behavior of the leak rate can be simulated using normally distributed crack morphology parameters.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.28 no.5
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• pp.610-616
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• 2004

The aim of this study was to investigate the hydride embrittlement when the LBB evaluation was carried out for the integrity of PHWR Pressure Tubes. The transverse tensile and CCT toughness tests were performed at three hydrogen concentrations while the test temperatures were changed (RT to 30$0^{\circ}C$). Both the transverse tensile and the fracture toughness tests showed the hydrogen embitterment clearly at RT but this phenomenon was disappeared while the test temperature arrived at 25$0^{\circ}C$. Using the DHC test results, the CCL and LBB time were calculated and compared. The hydride embrittlement at the LBB evaluation made the LBB time short definedly. If the operating temperature, DHCV and LBB deterministic parameters such as A and m were known, LBB time could be estimated without the calculation of CCL.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.126-133
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• 2016

Recently, the research and development of Sodium-cooled Fast Reactors (SFRs) have made progresses. However, liquid sodium, the coolant of an SFR, is chemically unstable and sodium fire can be occurred when liquid sodium leaks from sodium pipe. To reduce the damage by the sodium fire, many fire walls and fire extinguishers are needed for SFRs. LBB concept in SFR might reduce the scale of sodium fire and decrease or eliminate fire walls and fire extinguishers. Therefore, LBB concept can contribute to improve economic efficiency and to strengthen defense-in depth safety. The LBB assessment procedure has been well established, and has been used significantly in light water reactors (LWRs). However, an LBB assessment of an SFR is more complicated because SFRs are operated in elevated temperature regions. In such a region, because creep damage may occur in a material, thereby growing defects, an LBB assessment of an SFR should consider elevated temperature effects. The procedure and method for this purpose are provided in RCC-MRx A16, which is a French code. In this study, LBB assessment was performed for PGSFR IHTS hot-leg pipe according to RCC-MRx A16 and the applicability of the code was discussed.