• Title/Summary/Keyword: Leak-before-break curve

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LEAK-BEFORE-BREAK ANALYSIS OF THERMALLY AGED NUCLEAR PIPE UNDER DIFFERENT BENDING MOMENTS

  • LV, XUMING;LI, SHILEI;ZHANG, HAILONG;WANG, YANLI;WANG, ZHAOXI;XUE, FEI;WANG, XITAO
    • Nuclear Engineering and Technology
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    • 제47권6호
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    • pp.712-718
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    • 2015
  • Cast duplex stainless steels are susceptible to thermal aging during long-term service at temperatures ranging from $280^{\circ}C$ to $450^{\circ}C$. To analyze the effect of thermal aging on leak-before-break (LBB) behavior, three-dimensional finite element analysis models were built for circumferentially cracked pipes. Based on the elasticeplastic fracture mechanics theory, the detectable leakage crack length calculation and J-integral stability assessment diagram approach were carried out under different bending moments. The LBB curves and LBB assessment diagrams for unaged and thermally aged pipes were constructed. The results show that the detectable leakage crack length for thermally aged pipes increases with increasing bending moments, whereas the critical crack length decreases. The ligament instability line and critical crack length line for thermally aged pipes move downward and to the left, respectively, and unsafe LBB assessment results will be produced if thermal aging is not considered. If the applied bending moment is increased, the degree of safety decreases in the LBB assessment.

원자력 배관재료의 파괴저항곡선 예측 (Prediction of Fracture Resistance Curves for Nuclear Piping Materials(II))

  • 장윤석;석창성;김영진
    • 대한기계학회논문집A
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    • 제21권11호
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    • pp.1786-1795
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    • 1997
  • In order to perform leak-before-break design of nuclear piping systems and integrity evaluation of reactor vessels, full stress-strain curves and fracture resistance (J-R) curves are required. However it is time-consuming and expensive to obtain J-R curves experimentally. The objective of this paper is to modify two J-R curve prediction methods previously proposed by the authors and to propose an additional J-R curve prediction method for nuclear piping materials. In the first method which is based on the elastic-plastic finite element analysis, a blunting region handling procedure is added to the existing method. In the second method which is based on the empirical equation, a revised general equation is proposed to apply to both carbon steel and stainless steel. Finally, in the third method, both full stress-strain curve and finite element analysis results are used for J-R curve prediction. A good agreement between the predicted results based on the proposed methods and the experimental ones is obtained.

흰 광폭평판 시험을 이용한 원자력 배관의 파괴거동예측 (Prediction of Failure Behavior for Nuclear Piping Using Curved Wide-Plate Test)

  • 허남수;김윤재;최재붕;김영진;임혁순;정대율
    • 대한기계학회논문집A
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    • 제28권4호
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    • pp.352-361
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    • 2004
  • One important element of the Leak-Before-Break analysis of nuclear piping is how to determine relevant fracture toughness (or the J-resistance curve) for nonlinear fracture mechanics analysis. The practice to use fracture toughness from a standard C(T) specimen is known to often give conservative estimates of toughness. To improve the accuracy, this paper proposes a new method to determine fracture toughness using a nonstandard testing specimen, curved wide-plate in tension. To show validity of the proposed curved wide-plate test, the J-resistance curve from the full-scale pipe test is compared with that from the curved wide-plate test and that from the C(T) specimen. It is shown that the J-resistance curve form the curved wide-plate tension test is similar to, but that from the C(T) specimen is lower than, the J-resistance curve from the full-scale pipe test. Further validation is performed by investigating crack-tip constraint conditions via detailed 3-D FE analyses, which shows that the crack-tip constraint condition in the curved wide-plate tension specimen is indeed similar to that in the full-scale pipe under bending.

원자력 배관재료의 파괴저항곡선 예측 (Prediction of Fracture Resistance Curves for Nuclear Piping Materials)

  • 장윤석;석창성;김영진
    • 대한기계학회논문집
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    • 제19권4호
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    • pp.1051-1061
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    • 1995
  • In order perform leak-before-break design of nuclear piping systems and integrity evaluation of reactor vessels, full stress-strain (.sigma. - .epsilon.) curves and fracture resistance (J-R) curves are required. However it is time-consuming and expensive to obtain J-R curves experimentally. The objective of this paper is to develop two methods for J-R curve prediction. In the first method, elastic-plastic finite element analyses for a series of crack length / specimen width ratio were performed. Accordingly the load versus load line displacement (P .delta.) curve corresponding to the fracture strain is obtained and the J-R curve based on the generalized locus method is obtained. In the second method, the correlation between .sigma.-.epsilon. curves and J-R curves was statistically analyzed and an empirical equation to predict the J-R curve from the .sigma.-.epsilon. test result is proposed. A good correlation between the predicted results based on the proposed methods and the experimental ones is obtained.

REVIEW OF DYNAMIC LOADING J-R TEST METHOD FOR LEAK BEFORE BREAK OF NUCLEAR PIPING

  • Oh, Young-Jin;Hwang, Il-Soon
    • Nuclear Engineering and Technology
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    • 제38권7호
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    • pp.639-656
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    • 2006
  • In order to apply the leak before break (LBB) concept to nuclear piping systems, the dynamic strain aging effect of low carbon steel materials has to be taken into account, in compliance with the requirements of the Korean Standard Review Guide (KSRG) 3.6.3-1. For this goal, J-R tests are needed for a range of various temperatures and loading rates, including dynamic loading conditions. In the dynamic loading J-R test, the unloading compliance method can not be applied to measure the crack growth and direct current potential drop (DCPD) method; this method also has a problem defining the crack initiation point. The normalization method is known as a very useful method to determine the J-R curve under dynamic loading because it does not need additional equipment or complicated loading sequences such as electric current or unloading. This method was accepted by the American Society for Testing and Materials (ASTM) as a standard test method E1820 A15 in 2001. However, it has not yet been clearly verified yet if the normalization method is sufficiently reliable to be applied to LBB. In this study, the basic background of the J-integral, LBB and dynamic loading J-R test are explained, and the current status for dynamic loading J-R test methods are reviewed from the view point of LBB for nuclear piping. In particular, the theoretical and historical background of the normalization method which has received attention recently, is summarized. Recent studies for this method are introduced and future works are suggested that may improve the reliability of LBB for nuclear piping.

Dynamic Strain Aging on the Leak-Before-Break Analysis in SA106 Gr.C Piping Steel

  • Kim, Jin-Weon;Kim, In-Sup
    • 한국원자력학회:학술대회논문집
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    • 한국원자력학회 1996년도 춘계학술발표회논문집(3)
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    • pp.193-198
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    • 1996
  • The effect of dynamic strain aging (DSA) on the leak-before-break (LBB) analysis was estimated through the evaluation of leakage-size-crack and flaw stability in SA106 Gr.C piping steel. Also. the results were represented as a form of "LBB allowable load window". In the DSA temperature region. the leakage-size-crack length was smaller than that at other temperatures and it increased with increasing tensile strain rate. In the results of flaw stability analysis. the lowest instability load appeared at the temperature corresponding to minimum J- R curve which was caused by DSA. The instability load near the plant operating temperature depended on the loading rate of J-R data. and decreased with increasing tensile strain rate. These are due to the strain hardening characteristic and strain rate sensitivity of DSA. In the "LBB allowable load window". LBB allowable region was the narrowest at the temperature and loading conditions where DSA occurs.

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설계초과지진시 CPE를 고려한 밀림관 파단전누설 평가 (Leak Before Break Evaluation of Surge Line by Considering CPE under Beyond Design Basis Earthquake)

  • 김승현;김연정;이한걸;강선예
    • 한국압력기기공학회 논문집
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    • 제18권1호
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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.

Type 316N 스테인리스강의 OPR1000 및 APR1400 가압기 밀림관 적용성에 대한 연구 (A Study on Applicability of Stainless Steel Type 316N to the PZR Surge-line of OPR1000 and APR1400)

  • 유완;정성훈;박성호;손갑헌;이봉상;김민철
    • 대한기계학회:학술대회논문집
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    • 대한기계학회 2008년도 추계학술대회A
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    • pp.287-292
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    • 2008
  • The applicability of stainless steel type 316N to the PZR surge-lines of OPR1000 and APR1400 is investigated. So far, strainless steel type 347 has been used for the OPR1000 surge-lines. The degree of improvement in the leak-before-break(LBB) and component design margin is evaluated when stainless steel type 347 is substituted by type 316N. For the study, the tensile and J-R tests on type 316N and type 347 stainless steels were performed at 316 and the microstructure of both types was examined. Stainless steel type 316N shows the higher values on the stress-strain curves, J-R curves and stress intensity, Sm, compared to those of type 347. Therefore, stainless steel type 316N ensures the higher LBB and component design margins. As a result, this study shows that stainless steel type 316N could substitute type 347 for the surge-lines of OPR1000 and APR1400.

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굽힘피로 하중을 받는 배관의 피로균열 발생수명 예측 (Crack Initiation Life Analysis in Notched Pipe Under Cyclic Bending Loads)

  • 곽상록;이준성;김영진;박윤원
    • 대한기계학회논문집A
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    • 제25권10호
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    • pp.1528-1534
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    • 2001
  • In order to improve Leak-Be(ore-Break methodology, more precisely the crack growth evaluation, a round robin analysis was proposed by the CEA Saclay. The aim of this analysis was to evaluate the crack initiation life, penetration life and shape of through wall crack under cyclic bending loads. The proposed round robin analysis is composed of three main topic; fatigue crack initiation, crack propagation and crack penetration. This paper deals with the first topic, crack initiation in a notched pipe under four point bending. Both elastic-plastic finite element analysis and Neuber's rule were used to estimate the crack initiation life and the finite element models were verified by mesh-refinement, stress distribution and global deflection. In elastic-plastic finite element analysis, crack initiation life was determined by strain amplitude at the notch tip and strain-life curve of the material. In the analytical method, Neuber's rule with the consideration of load history and mean stress effect, was used for the life estimation. The effect of notch tip radius, strain range, cyclic hardening rule were examined in this study. When these results were compared with the experimental ones, the global deformation was a good agreement but the crack initiation cycle was higher than the experimental result.

영광원자력 배관소재의 재료물성치 평가 (1)-정지냉각계통- (Evaluation of Material Properties for Yonggwang Nuclear Piping System(I)-Shutdown Cooling System-)

  • 석창성;최용식;장윤석;김종욱
    • 대한기계학회논문집
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    • 제18권5호
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    • pp.1106-1116
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    • 1994
  • Leak Before Break(LBB) design concept is applied to piping systems of newly-built Yonggwang 3, 4 nuclear generating stations as a design alternative to the provision of pipe whip restraints, in recognition of the questionable benefits of providing such restraints. The objective of this paper is to evaluate the material properties (tensile and fracture toughness) of SA312 TP316 stainless steel and their associated welds manufactured for shutdown cooling system of Yonggwang 3, 4 nuclear generating stations. Effect of various parameters such as specimen orientation, test temperature, welding on material properties were examined.