• Nuclear Engineering and Technology
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• 제30권5호
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• pp.403-415
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• 1998

The CANDU 6 reactor assembly consists of an array of 380 pressure tubes, which are installed horizontally in a large cylindrical vessel, the Calandria, containing the low pressure heavy water moderator. The pressure tube is located inside the calandria tube and the annulus between these tubes, which forms a closed loop with $CO_2$ gas recirculating, is called the Annulus Gas System(AGS). It is designed to give an alarm to the operator even for a small pressure tube leak by a very sensitive dew point meter so that he can take a preventive action for the pressure tube rupture incident. To judge whether the operator action time is enough or not in the design of Wolsong 2,3 & 4, the Leak Before Break(LBB) assessment is required for the analysis of the pressure tube failure accident. In order to provide the required data for the LBB assessment of Wolsong Units 2, 3, 4, a series of leak detection capability tests was performed by injecting controlled rates of heavy water vapour. The data of increased dew point and rates of rise were measured to determine the alarm set point for the dew point rate of rise of Wolsong Unit 2. It was found that the response of the dew point depends on the moisture injection rate, $CO_2$ gas flow rate and the leak location. The test showed that CANDU 6 AGS can detect the very small leaks less than few g/hr and dew point rate of rise alarm can be the most reliable alarm signal to warn the operator. Considering the present results, the first response time of dew point to the AGS $CO_2$ flow rate is approximated.

• 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.

• 한국압력기기공학회 논문집
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• 제9권1호
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• pp.62-69
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• 2013

A probabilistic assessment code, PRO-LOCA ver. 3.7 which was developed in an international co-operative research program, PARTRIDGE was evaluated by conducting sensitivity analysis. The effect of some variables such as simulation methods (adaptive sampling, iteration numbers, weld residual stress model), crack features(Poisson's arrival rate, maximum numbers of cracks, initial flaw size, fabrication flaws), operating and loading conditions(temperature, primary bending stress, earthquake strength and frequency), and inspection model(inspection intervals, detectable leak rate) on the failure probabilities of a surge line nozzle was investigated. The results of sensitivity analysis shows the remaining problems of the PRO-LOCA code such as the instability of adaptive sampling and unexpected trend of failure probabilities at an early stage.

• Nuclear Engineering and Technology
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• 제30권5호
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• pp.435-443
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• 1998

Leak-before-break(LBB) approach has been shown to be both cost effective and risk reductive when applied to high energy Piping in nuclear Power Plants. For the Korean Next Generation Reactor (KNGR) development, LBB application is considered for the Main Steam Line(MSL) piping inside containment. Unlike the primary system leakages, the MSL leak detection systems must be based on principles other than radioactivity measurements. Among humidity, heat and acoustic noise currently being considered as indicators of leakage, we explored humidity as an effective one and developed ceramic-based humidity sensor which can be qualified for LBB applications. The ceramic material, sintered and annealed MgCr$_2$O$_4$-TiO$_2$, is shown to increase its electrical conductivity drastically upon water vapor adsorption over the entire temperature range of interest. With this ceramic sensor specimen, we suggested installation-inside-the-piping method by which we can detect leakage more rapidly and sensitively. In this paper, we describe the progress in the development and characterization of ceramic humidity sensor for the LBB application to the MSL of KNGR.

• 한국가스학회지
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• 제3권1호
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• pp.1-7
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• 1999

알루미늄합금(5083-0)의 표면균열시험편의 관통전후의 피로거동을 조사하였다. 관통전의 피로균열형상은 거의 반타원형이며, 측정된 형상비는 Newman-Raju식의 K를 이용하여 계산한 값보다 더 크다. 관통후의 뒷면에서의 피로균열성장거동은 거의 유사하였으며 3영역으로 나누어졌다. 표면균열길이가 긴 경우, 피로균열전파법칙을 사용함으로서 표면균열재의 판두께 관통후의 피로균열 전파특성을 정량적으로 평가될 수 있음을 알 수 있었다.

• 소음진동
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• 제6권6호
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• pp.827-834
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• 1996

본연구에서는 핵연료집합체의 검증계획의 일환으로 2차측 배관파단의 영향을 조사하였다. 원자로노심의 상세모델을 이용한 동적해석으로 배관파단에 의한 응답을 구하였다. 파단적 누설개념의 적용으로 10인치 이상의 고에너지 배관에 대하여 양단 파단이 설계에서 배제됨에 따라 본 연구에서는 주증기관과 급수관의 파단을 가정 하였다. 핵연료 집합체의 전단력, 굽힘모우멘트, 변위 및 지지격자체의 충격하중에 대하여 자세히 고찰하였고 이들 동적해석 결과를 이용하여 핵연료집합체의 구조적 건전성을 평가하였으며 사고조건에서 2차측 배관파단이 핵연료집합체의 구조적 건전성 에 미치는 영향을 검토하였다.

• 한국해양공학회지
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• 제12권1호
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• pp.50-57
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• 1998

The leak before break(LBB) concept is generalized on the design of LNG tanks, pressure vessels and nuclear reactor in that any leakage of containment, in whatever amount, will not result in catastropic failure. For this purpose it is necessary to determine the surface crack shape, the opening displacement and the risk of catastropic brittle fracture when it becomes a through crack. In this study the crack propagation behavior of surface flaws and the crack opening displacement of through cracks under combined membrane and bending stresses were investigated with fatigue tests and fracture toughness test of aluminium alloy A5083-O. And fracture mechanics analysis of the crack opening displacement of through cracks were made in order to develop a new model expressing the behaviors of COD under combined membrane and bending stresses.

• 한국안전학회지
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• 제14권2호
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• pp.32-41
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• 1999

If the Leak-Before-Break (LBB) concept is applicable to the nuclear piping design, it is not necessary to consider the dynamic effect due to pipe rupture. Therefore, the construction cost can be significantly reduced by eliminating unnecessary pipe whip restraints and jet impingement devices. The objective of this paper is to develop the Piping Evaluation Diagram (PED) for efficient application of LBB concept to piping system at an initial piping design stage. For this purpose, the 3-D finite element analyses were performed to evaluate the crack stability. And the stress-strain curve based on the pipe material tests were used to calculate the detectable leakage crack length. Finally, the present PED which was composed as a function of NOP load and allowable SSE load, was developed for an application of LBB concept to the safety injection and shutdown cooling line in Korean Next Generation Reactor (KNGR).

• 한국가스학회지
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• 제4권1호
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• pp.33-39
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• 2000

긴 표면균열을 갖는 알루미늄합금(A5083)의 평판을 사용하여, 균열관통후의 응력확대계수를 검토하였다. 짧은 표면균열의 관통후에 제안된 식에 의하여 긴 표면균열 관통시의 평가는 균열종횡비가 적을수록 오차가 크게 되므로, 진전거동을 정확하게 평가하기 위해서는 수정이 필요하였다. 따라서 긴 표면균열 관통시의 종횡비를 고려하여, 응력확대계수의 평가식을 수정함으로써 관통후 고정도의 해석이 가능하였다.

• 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.