Acceptance criterion for probabilistic structural integrity assessment: Prediction of the failure pressure of steam generator tubing with fretting flaws

2015 ◽  
Vol 281 ◽  
pp. 154-162 ◽  
Author(s):  
Xinjian Duan ◽  
Min Wang ◽  
Michael Kozluk
2008 ◽  
Vol 130 (4) ◽  
Author(s):  
Xinjian Duan ◽  
Michael J. Kozluk ◽  
Sandra Pagan ◽  
Brian Mills

Aging steam generator tubes have been experiencing a variety of degradations such as pitting, fretting wear, erosion-corrosion, thinning, cracking, and denting. To assist with steam generator life cycle management, some defect-specific flaw models have been developed from burst pressure testing results. In this work, an alternative approach; heterogeneous finite element model (HFEM), is explored. The HFEM is first validated by comparing the predicted failure modes and failure pressure with experimental measurements of several tubes. Several issues related to the finite element analyses such as temporal convergence, mesh size effect, and the determination of critical failure parameters are detailed. The HFEM is then applied to predict the failure pressure for use in a fitness-for-service condition monitoring assessment of one removed steam generator tube. HFEM not only calculates the correct failure pressure for a variety of defects, but also predicts the correct change of failure mode. The Taguchi experimental design method is also applied to prioritize the flaw dimensions that affect the integrity of degraded steam generator tubes such as the defect length, depth, and width. It has been shown that the defect depth is the dominant parameter controlling the failure pressure. The failure pressure varies almost linearly with defect depth when the defect length is greater than two times the tube diameter. An axial slot specific flaw model is finally developed.


Author(s):  
Carlos Alexandre de Jesus Miranda ◽  
Miguel Mattar Neto

A fundamental step in tube plugging management of a Steam Generator (SG), in a Nuclear Power Plant (NPP), is the tube structural integrity evaluation. The degradation of SG tubes may be considered one of the most serious problems found in PWRs operation, mainly when the tube material is the Inconel 600. The first repair criterion was based on the degradation mode where a uniform tube wall thickness corrosion thinning occurred. Thus, a requirement of a maximum depth of 40% of the tube wall thickness was imposed for any type of tube damage. A new approach considers different defects arising from different degradation modes, which comes from the in-service inspections (NDE) and how to consider the involved uncertainties. It is based on experimental results, using statistics to consider the involved uncertainties, to assess structural limits of PWR SG tubes. In any case, the obtained results, critical defect dimensions, are within the regulatory limits. In this paper this new approach will be discussed and it will be applied to two cases (two defects) using typical data of SG tubes of one Westinghouse NPP. The obtained results are compared with ‘historical’ approaches and some comments are addressed from the results and their comparison.


Author(s):  
Warren Bamford ◽  
John Hall

Service induced cracking in Alloy 600 has been known for a long time, having been first observed in the 1980’s in steam generator tubing and small bore piping, and later, in 1991, in reactor vessel control rod drive mechanism (CRDM) head penetrations. Other than steam generator tubing, which cracked within a few years of operation, the first Alloy 600 cracking was in base metal of Combustion Engineering small bore piping, followed closely by CE pressurizer heater sleeves. The first reactor vessel CRDM penetrations (base metal) to crack were in France, US plants found CRDM cracking several years later. Three plants have discovered weld metal cracking at the outlet nozzle to pipe weld region. This was the first known weld metal cracking. This paper will chronicle the development of service-induced cracking in these components, and compare the behavior of welds as opposed to base metal, from the standpoint of time to crack initiation, growth rate of cracks, and their impact on structural integrity. In addition, a discussion of potential future trends will be provided.


Author(s):  
Yong-Seok Kang ◽  
Hong-Deok Kim ◽  
Kuk-Hee Lee ◽  
Jai-Hak Park

Degraded steam generator tubing can affect its safety functions. Therefore, its integrity should be maintained for each degradation form and all detected degradation must be assessed to verify that if adequate integrity is retained. Determination of tube integrity limits includes identifying acceptable structural parameters such as flaw length, depth, and amplitude of signals. If we consider just single-cracked tubes, short and deep flaws are not likely to threaten structural integrity of tubes. But if it has multiple-cracks, we have to consider interaction effects of multiple adjacent cracks on its burst pressure. Because adjacent multiple cracks can be merged due to the crack growth then it can challenge against the structural performance limit. There are some studies on the interaction effects of adjacent cracks. However, existing works on the interaction effect consider only through-wall cracks. No study has been carried out on the interaction effects of part-through cracks. Most cracks existing in real steam generator tubing are not through-wall cracks but part-through cracks. Hence, integrity of part-through cracks is more practical issue than that of through-wall cracks. This paper presents experimental burst test results with steam generator tubing for evaluation of interaction effects with axial oriented two collinear and parallel part-through cracks. The interaction effect between two adjacent cracks disappeared when the distance exceeds about 2 mm.


Author(s):  
Russell C. Cipolla ◽  
James A. Begley ◽  
Robert F. Keating

General Design Criteria (GDC) 1, 2, 4, 14, 30, 31 and 32 of 10 CFR Part 50, Appendix A, define requirements for the reactor coolant pressure boundary (RCPB) with respect to structural and leakage integrity [1]. Steam generator tubing and tube repairs constitute a major fraction of the RCPB surface area. Steam generator tubing and associated repair techniques and components, such as sleeves, must be able to maintain reactor coolant inventory and pressure. The Structural Integrity Performance Criterion (SIPC) from Nuclear Energy Institute (NEI) 97-06 [2] was developed to provide reasonable assurance that a steam generator tube will not burst during normal or postulated accident conditions. This paper presents the SIPC and its technical basis.


Author(s):  
Kyu Jung Yeom ◽  
Yong Kwang Lee ◽  
Kyu Hwan Oh ◽  
Cheol Man Kim ◽  
Woo Sik Kim

Gas pipelines with mechanical damages could affect the structural integrity and causes local stress and strain concentration. Failures in gas pipeline as leakages that could affect the supply of gas, loss of production, and environmental pollution. It is important to determine if pipelines are fitness-for-service. ASME B31G code is still widely used criterion although the assessment method is the conservative method. Further examinations are needed on the effects of material grade and pipeline shape on the burst pressure of damaged pipelines. The goal of this paper is to predict the failure pressure of mechanical damaged made of API X65 and X70 pipelines, by conducting full scale burst tests and finite element analysis (FEA). Different pipeline grades, effects of gouges, and dent depths were considered for an integrity assessment. The full scale burst tests were performed for pipelines with artificial mechanical damage. The gouge defect was made in a V-notch shape and the dented pipeline was generated using a ball shaped indenter that was pressed into the pipe. A three dimensional FEA was performed to obtain the burst pressure of a pipe with gouge and dent defects as a function of defect depth and length. A FEA was used to simulate the and externally damaged pipes under internal pressure. Failure pressure was predicted with stress based and strain based assessments by the finite element method (FEM).


Author(s):  
Seung-Hyun Park ◽  
Jae-Boong Choi ◽  
Nam-Su Huh

Nowadays, nuclear power plant (NPP) has become one of the most important energy sources to generate electricity in the world. Steam generator (SG) is a heat exchanger included in primary system of NPP. Alloy 600 MA is widely used for SG tube material and this is well-known as weakness of stress corrosion cracking. In recent year, according to increase the number of long-term operation NPP, many axial surface flaws have been found on SG tube during an in-service inspection. Therefore, many researches have been carried out to maintain structural integrity of SG tube. Commonly, flaw shape needs to be idealized to calculate a burst pressure because detected flaw shape is complicated. In this paper, validation of EPRI’s weakest sub-crack model, one of the well-known flaw idealization rule, is conducted through finite element (FE) analysis. For this, three actual flaws are assumed and these are idealized by using four flaw shape idealization methods; semi-elliptical crack model, rectangular crack model, maximum length with effective depth crack model and weakest sub-crack model. Burst pressure of each model is calculated and compared with burst pressure of actual shape crack model. As a result, if actual flaw is idealized by weakest sub-crack model, it is expected that conservative and efficient structural integrity assessment will be possible.


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