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.

No abstract

Ontario Power Generation (OPG) has developed and implemented a systematic managed process for steam generators at all of its facilities. One of the key requirements of this managed process is to have in place long range Steam Generator Life Cycle Management (SG LCM) plans for each of its reactor units. The primary goal of these plans is to maximize the value of the nuclear facility through safe and reliable steam generator operation over the expected life of the units. These SG LCM plans integrate and schedule all steam generator actions such as inspection, operation, maintenance, repairs, modifications, assessments, performance monitoring, research and development, and feedback. This paper provides an overview of how structural and leak-rate testing, being conducted by OPG, is being used to support fitness-for-service assessments for fretting degradation in the U-bend region of the recirculating steam generators at the Darlington Nuclear Generating Station.

The first CANDU (CANadian Deuterium Uranium) pressurized heavy water reactor (PHWR) went into operation in July 1971. Today, there are several units in operation at the Pickering, Bruce, and Darlington sites in Ontario, Canada. The steam generator tubing materials were manufactured from Monel 400, Inconel 600, and Incoloy 800 for the Pickering, Bruce, and Darlington respectively and are subjected to different operating conditions. This paper presents a review of some of the various types of degradation mechanisms that have been observed on these tubing materials over the operating period of the respective plants. The results presented are based on the metallurgical examination of removed tubes. The mechanisms that have been observed include pitting, stress corrosion cracking, intergranular attack, fretting, and erosion corrosion. The nature of the flaws and causative factors (if known) are discussed.

Canadian nuclear standard CSA N285.4 requires the periodic metallurgical examination of removed ex-service steam generator tubes. This paper describes the practices used for the characterization and structural integrity tests of ex-service steam generator tubes at Ontario Power Generation (OPG). It shows that there is no degradation in mechanical properties of Monel 400 tubes after 7 to 18 Effective Full Power Years (EFPY) of operation and Incoloy 800 tubes after more than 10 EFPY of operation.