Structural integrity assessment of nuclear steam generator

Author(s):  
S. Majumdar ◽  
S. Bakhtiari ◽  
Z. Zeng ◽  
C.B. Bahn
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.


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):  
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.


2017 ◽  
Vol 140 (1) ◽  
Author(s):  
Seung-Hyun Park ◽  
Jae-Boong Choi ◽  
Nam-Su Huh ◽  
Sang-Min Lee ◽  
Yong-Beum Kim

This work investigates the applicability of the flaw shape idealization methods to carry out the structural integrity assessment of steam generator (SG) tubes under internal pressure with complicated axial inner and outer surface flaws that were typically found during the in-service-inspection (ISI). In terms of flaw shape, three different shapes of flaws which can be detected during an actual ISI are considered, i.e., long symmetric flaw, asymmetric inclined flaw and narrow, symmetric deep flaw. As for flaw shape idealization methods for the predictions of burst pressures of these flaws, four different flaw shape idealization models, i.e., semi-elliptical, rectangular, maximum length with effective flaw depth and weakest subcrack model proposed by the Electric Power Research Institute (EPRI) are employed in this work. In order to validate the applicability of these idealization methods, the burst pressures of SG tubes with these flaws are investigated by using the finite element (FE) analyses. By comparing the predictions of the burst pressures based on the four different flaw shape idealization methods with those based on actual flaw shapes, it is found that the weakest subcrack model proposed by the EPRI and maximum length with effective flaw depth model provide the better agreement with actual complex flaw.


Sign in / Sign up

Export Citation Format

Share Document