Probabilistic structural integrity assessment for Inconel690 alloy steam generator tube with volume defect

2021 ◽  
Vol 371 ◽  
pp. 110949
Author(s):  
Song Huang ◽  
Hu Hui ◽  
Zhizhen Peng
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):  
Tae-Young Ryu ◽  
Han-Beom Seo ◽  
Jong-Min Kim ◽  
Jae-Boong Choi ◽  
Young-Jin Kim ◽  
...  

For integrity assessment of structure containing crack, evaluation method based on fracture mechanics such as linear-elastic and elastic-plastic fracture mechanics has been relatively common method and becoming more widespread. However, it can be used only if the crack opening or tearing is occurred. If the crack exists on the piping components subjected to internal pressure, net-section stress is occurred in the direction on crack opening no matter where crack locate. On the contrary to this, if the external pressure is applied to piping components, net-section stress is occurred opposite direction and it is expected crack opening not to be occur. The subject of this study is SMART steam generator tube which is designed as helical geometrical feature and to be pressurized outside. Three dimensional finite element analyses are carried out to investigate crack behavior under external pressure considered various crack geometries and locations. Furthermore, the possibility of failure of SMART steam generator tube under design pressure is investigated.


2008 ◽  
Vol 22 (11) ◽  
pp. 1057-1061 ◽  
Author(s):  
HEE JONG LEE ◽  
MIN WOO NAM ◽  
CHAN HEE CHO ◽  
DONG HYUN JEE

The bobbin probe is now widely accepted as a basic and important ECT technique among various ECT techniques for steam generator tube integrity assessment that is practiced during each plant outage. The bobbin probe is also one of the essential components which consist of the whole ECT examination system, and provides us with decisive data for the evaluation of tube integrity in accordance with acceptance criteria described in specific procedures. Accordingly, the design of ECT probe is especially important to examination results because the quality of acquired ECT data is determined by the optimized probe design characteristics, such as coil geometry, electrical properties, operation frequency, and so on. In this study, the relationship between electric characteristic changes and differential coil gap variation for the optimization of the ECT probe signal was investigated. As a result, for the examination of volumetric flaws fabricated on the O.D. side of ASME type calibration standard, we propose that the best Lissajous signal for the examination will be obtained when the coil gap of differential bobbin probe is approximately 1.2 ~ 1.6 mm .


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.


2012 ◽  
Vol 249 ◽  
pp. 297-303 ◽  
Author(s):  
A. Erhard ◽  
X. Schuler ◽  
F. Otremba

Author(s):  
Hyun Su Kim ◽  
Jong Sung Kim ◽  
Tae Eun Jin ◽  
Hong Deok Kim ◽  
Han Sub Chung

The steam generator in a nuclear power plant is a large heat exchanger that uses heat from reactor to generate steam to drive the turbine generators. Rupture of a steam generator tube can result in release of fission products to environment. Therefore, an accurate integrity assessment of the steam generator tubes with cracks is of great importance for maintaining safety of a nuclear power plant. The steam generator tubes are supported at periodic intervals by support plates and rotations of the tubes are constrained. Although it was reported that the limit load for a circumferential crack was significantly affected by boundary condition of the tube, existing limit load solutions do not include the constraining effect of tube supports. This paper provides detailed limit load solutions for circumferential cracks in steam generator tubes considering the actual boundary conditions to simulate the constraining effect of the tube supports. Such solutions are developed based on three dimensional (3D) finite element analyses. The resulting limit load solutions are given in a polynomial form, and thus can be simply used in practical integrity assessment of the steam generator tubes.


Author(s):  
Ki-Wahn Ryu ◽  
Bong-Ho Cho ◽  
Chi-Yong Park ◽  
Su-Ki Park

The characteristics of fluid-elastic instability for the KSNP steam generator tubes were investigated numerically. The information on the thermal-hydraulic data of the steam generator has been obtained by using the ATHOS3-MOD1 code and the fluid-elastic instability analysis has been conducted by using the PIAT (Program for Integrity Assessment of Steam Generator Tube) code. The KSNP steam generator has the concentrated plugging zone at the vicinity of the stay cylinder inside the steam generator. To investigate the cause of the concentrated plugging, the fluid-elastic instability analysis has been performed on various column and row number of the KSNP steam generator tubes. From the results of this study the stability ratio due to the fluid-elastic instability in the concentrated plugged zone tend to have larger values than those of the outer zone. Even though the further study will still be required, these results seem to be related with concentrated plugging inside the steam generator. And the stability ratio of plugged tube does not have any consistent advantages for all modes over the normal one. This seems to be caused by the decrease of mass, the increase of natural frequency, and the change of mode shape after plugging.


Sign in / Sign up

Export Citation Format

Share Document