Jin Ho Lee

Determination of equivalent single crack based on coalescence criterion of collinear axial cracks

In a nuclear power plant the steam generator tubes cover a major portion of the primary pressure-retaining boundary. Thus, very conservative approaches have been taken in the light of steam generator tube integrity. According to the present criteria, tubes wall-thinned in excess of 40% should be plugged whatever the cause. However, many analytical and experimental results have shown that no safety problems exist even with thickness reductions greater than 40%. The present criterion was developed about 20 years ago when wear and pitting were dominant causes for steam generator tube degradation, and it is based on tubes with single cracks regardless of the fact that the appearance of multiple cracks is more common in general. The objective of this study is to review the conservatism of the present plugging criteria of steam generator tubes and to propose a new coalescence model for two adjacent through-wall cracks existing in steam generator tubes. Using the existing failure models and experimental results, we reviewed the conservatism of the present plugging criteria. In order to verify the usefulness of the proposed new coalescence model, we performed finite element analysis and some parametric studies. Then, we developed a coalescence evaluation diagram.

Investigation on the interaction effect of two parallel axial through-wall cracks existing in steam generator tube

It is commonly required that steam generator tubes wall-thinned in excess of 40% should be plugged. However, the plugging criterion is known to be too conservative for some locations and types of defects and its application is confined to a single crack. In the previous study, the conservatism of the present plugging criterion of steam generator tubes was reviewed and a crack coalescence model applicable to steam generator tubes with two collinear axial through-wall cracks was proposed. Since parallel axial cracks are more frequently detected during in-service inspections than collinear axial cracks, the studies on parallel axial cracks spaced in circumferential direction are necessary. The objective of this paper is to investigate the interaction effect between two parallel axial through-wall cracks existing in a steam generator tube. Finite element analyses were performed and a new failure model of the steam generator tube with these types of cracks is suggested. Interaction effects between two adjacent cracks were investigated to explain the deformation behavior of cracked tubes.

A COUPLED CFD-FEM ANALYSIS ON THE SAFETY INJECTION PIPING SUBJECTED TO THERMAL STRATIFICATION

Thermal stratification has continuously caused several piping failures in nuclear power plants since the early 1980s. However, this critical thermal effect was not considered when the old nuclear power plants were designed. Therefore, it is urgent to evaluate this unexpected thermal effect on the structural integrity of piping systems. In this paper, the thermal effects of stratified flow in two different safety injection piping systems were investigated by using a coupled CFD-FE method. Since stratified flow is generally generated by turbulent penetration and/or valve leakage, thermal stress analyses as well as CFD analyses were carried out considering these two primary causes. Numerical results show that the most critical factor governing thermal stratification is valve leakage and that temperature distribution significantly changes according to the leakage path. In particular, in-leakage has a high possibility of causing considerable structural problems in RCS piping.

Round robin analysis of pressurized thermal shock for reactor pressure vessel

A comparative assessment study is performed here for deterministic fracture mechanics analysis of a pressurized thermal shock (PTS). A round robin problem is proposed using the data available in Korea and all organizations interested in the PTS analysis are invited. The problems consisting of two transients and 10 cracks are solved and their results are compared to generate a reference solution that could serve as benchmarks for future qualification of analytical method. Nine participants from seven organizations responded to the problem and their results are compiled in this paper.

Stress Distribution in the Dissimilar Metal Butt Weld of Nuclear Reactor Piping due to the Simulation Technique for the Repair Welding

이 보고된 이후 전세계 원자력 산업계에서는 이에 의한 손상을 예측하거나 방지하기 위해 많은 연구를 수행한 바 있다. 특히 PWSCC에 의한 균열 발생의 대표적인 원인 가운데 하나가 DMW 및 동종금속용접(Similar Metal Weld, SMW)에 의한 인장 용접잔류응력이기에 용접과정 중 발생하는 용접잔류응력을 정확하게 예측하기 위한 연구가 다양한 DMW 부위에 대해 수행되고 있다. 원자력기기 용접 공정 중에는 용접 공정 상 결함이 발생할 수 있는데 만약 용접 공정 중 결함이 발견되면 이를 제거하고 국부적으로 보수 용접을 Key Words: Dissimilar Metal Butt Weld(이종금속 맞대기 용접), Finite Element Analysis(유한요소해석), PWSCC(일차수 응력부식균열), Repair Welding(보수용접), Welding Residual Stress(용접잔류 응력) 초록: 이종금속용접부에 대한 실제 용접 공정 중 용접부에서 결함이 발견되면 이를 제거하고 보수용접이 수행된다. 일반적으로 보수용접을 수행하면 용접부에서 인장 잔류응력이 크게 증가될 수 있는 것으로 알려져 있다. 따라서 Alloy 82/182를 사용하여 보수용접이 수행된 이종금속용접부의 일차수 응력부식균열 현상을 평가하기 위해서는 보수용접에 의한 용접부의 응력 변화를 정확하게 평가해야 한다. 본 논문에서는 비선형 유한요소해석을 수행하여 보수용접에 의한 원자력 이종금속 맞대기 용접부의 응력 분포를 평가하였다. 특히 보수용접 공정 모사를 위한 여러 유한요소 해석방법이 이종금속용접부의 응력 분포에 미치는 영향을 평가하였다. Abstract: During welding, the dissimilar metal butt welds of nuclear piping are typically subjected to repair welding in order to eliminate defects that are found during post-weld inspection. It has been found that the repair weld can significantly increase the tensile residual stress in the weldment, and therefore, accurate estimation of the weld residual stress due to repair weld, especially for dissimilar metal welds using Ni-based alloy 82/182 in nuclear components, is of great importance in order to assess susceptibility to primary water stress corrosion cracking. In the present study, the stress distributions of dissimilar metal butt welds in nuclear reactor piping subjected to repair weld were investigated based on detailed nonlinear finite element analyses. Particular emphasis was placed on the variation of the stress distribution in the dissimilar metal butt weld according to the finite element welding analysis sequence for the repair welding process.

Evaluation of Representative Piping Systems Designed by Implicit Fatigue Concept

NUREG-1801 provides generic aging lessons learned to manage aging effects that may occur during continued operation beyond the design life of nuclear power plant. According to this report, the metal fatigue, among several age-related degradation mechanisms, is identified as one of time-limited aging analysis item. The objective of this paper is to introduce fatigue life evaluation of representative surge line and residual heat removal system piping which was designed by implicit fatigue concept. For the back-fitting evaluation employing explicit fatigue concept, detailed parametric CFD as well as FE analyses results are used. The well-known ASME Section III NB-3600 procedure is adopted for the metal fatigue and NUREG/CR-5704 procedure is further investigated to deal with additional environmental water effects. With regard to the environmental effect evaluation, two types of fatigue life correction factors are considered, such as maximum Fen and individual Fen . As a result, it was proven that a thermal stratification phenomenon is the governing factor in metal fatigue life of the surge line and strain rate is the most important parameter affecting the environmental fatigue life of both piping. The evaluation results will be used as technical bases for continued operation of OPR 1000 plant.© 2008 ASME

Statistical Reliability Assessment of UT Round-Robin Test Data for Piping Welds

Ultrasonic NDE is one of important technologies in the life-time maintenance of nuclear power plant. Ultrasonic inspection system is consisted of the operator, equipment and procedure. The reliability of ultrasonic inspection system is affected by its ability. The performance demonstration round robin was conducted to quantify the capability of ultrasonic inspection for in-service. Several teams employed procedures that met or exceeded with ASME Sec. XI code requirements detected the piping of nuclear power plant with various cracks to evaluate the capability of detection and sizing. In this paper, the statistical reliability assessment of ultrasonic nondestructive inspection data using probability of detection (POD) is presented. The result of POD using logistic model was useful to the reliability assessment for the NDE hit or miss data.

Effects of leak rate on LOCA probability of pipes in nuclear power plants

In the leak before break (LBB) concept, catastrophic pipe failure can be prevented if leak rate through a crack is more than the minimum detectable leak rate of a leak detection system before a crack grows to the critical crack size. Therefore accurate estimation of leak rate is important to evaluate the validity of LBB concept in pipe line design. Usually LOCA probability analysis is performed with constant crack morphology parameters. Since leak rate is affected by the crack morphology parameters, the LOCA probability may be affected by the parameters also. In this paper the effect of crack morphology parameters on LOCA probability was examined assuming the crack morphology parameters to be normally distributed random variables. A developed probabilistic fracture mechanics program, called P-PIE, was used in the analysis.

Stress Classification and Fatigue Life Assessment of Modular Component with Asymmetric Perforated Parts

In the ASME Code Section III ‘design by analysis’ approach, stresses are determined by numerical method and compared with corresponding stress limits. This approach provides several stress criteria for fatigue life assessment and procedures for categorizing the representative stress components. Since the stress criteria were derived from two-dimensional basis, however, it may inappropriate to delineate structural components with complex geometry. In this paper, detailed transient analyses are performed for modular pressurizer with an asymmetric geometry, which includes perforated parts to mount various piping and equipments. Also, the applicability of an effective elastic modulus to consider the perforation and the appropriateness of stress linearization method using stress classification line are assessed. Then, the cumulative usage factor as well as stress intensities at critical locations of the pressurizer are calculated and compared with corresponding allowable design stress limits. The key findings of this work can be used to make regulatory guides for evaluation and confirmation of structural intensity of components with asymmetric perforated parts.

Numerical Analyses of Surge Line Piping to Assess Thermal Stratification Phenomenon

During the last two decades, thermal stratification has been issued as a critical problem in the nuclear power industry. Since the problem caused by this phenomenon also became important in Korea, it is necessary to quantify the thermal stratification effect to ensure the safety of the piping system. In this paper, detailed stress analyses of the surge line, considering the thermal stratification, are conducted. Parametric sensitivity analyses to find out an optimum model were carried out using pipe element models and full 3-D element models. For instance, in case of the pipe element model, the effect of starting location of thermal stratification and boundary condition were investigated. And, in case of the 3-D solid element model, the effect of boundary condition and thermal loading condition were assessed. The stress analysis results showed that the thermal stratification phenomenon significantly affected the integrity of the surge line piping. Also, establishment of insurge and outsurge conditions was derived as one of the further investigations.Copyright