THESIS. Andrew Jordan Clark, B.S. Graduate Program in Nuclear Engineering. The Ohio State University. Master s Examination Committee:

Transcription

1 Effectiveness of Surveillance Sampling Strategies for Detecting Steam Generator Tube Degradation THESIS Presented in Partial Fulfillment of the Requirement for the Degree Master of Science in the Graduate School of The Ohio State University By Andrew Jordan Clark, B.S. Graduate Program in Nuclear Engineering The Ohio State University 2015 Master s Examination Committee: Professor Tunc Aldemir, Advisor Professor Jinsuo Zhang

2 Copyright by Andrew Jordan Clark 2015

3 Abstract Nuclear power plants seeking to ext their operating license must first address the degradation of systems, structures, and components (SSCs) to ensure they can maintain a satisfactory level of reliability into the exted lifetime. Passive SSCs play an important role in determining the feasibility of life extension. Part of the feasibility analysis requires plants to demonstrate the viability and reliability of passive SSCs into the exted lifetime. The research carried out toward this thesis considers primary water stress corrosion cracking (PWSCC) of steam generator (SG) tubes as an example degradation mechanism. An empirical model for PWSCC crack growth is adopted to simulate crack growth over a 40-year operating lifetime. Surveillance and maintenance strategies similar to those performed by the industry are integrated with the PWSCC crack growth model to determine the effectiveness of surveillance strategies for detecting SG tube degradation. The results of this analysis were applied to a specific accident scenario in which steam generator tubes rupture following a depressurization of the secondary side due to the sudden rupture of a steam-line caused by flow-accelerated corrosion. Likelihood of a spontaneous steam generator tube rupture is also assessed. The analysis and application of the specific accident scenario indicates a maximum core damage frequency in the 16 th year. Sensitivity analyses into the ii

4 probability of detection (POD) and crack growth rates were also performed. As expected, the likelihood of the accident scenario occurring increased significantly as the maximum POD was decreased. When crack growth rates were slowed down, the overall likelihood of the accident scenario decreased and the expected occurrence of the accident scenario was delayed. iii

5 Dedication I dedicate this to my family. iv

6 Acknowledgements I want to thank Dr. Tunc Aldemir, Dr. Jinsuo Zhang and Dr. Richard Denning for being such valuable resources. Their continued support, guidance and wisdom have been invaluable. v

7 Vita June 2007 Adolfo Camarillo High School June 2013 B.S. Chemical Engineering, University of California, Irvine Fields of Study Major Field: Nuclear Engineering vi

8 Table of Contents Abstract... ii Dedication... iv Acknowledgements... v Vita... vi List of Figures... xi List of Tables... xiv List of Acronyms... xvi Chapter 1: Introduction Motivation and Research Objectives SG Tube Degradation General Overview SG Degradation Issues Current State of Steam Generators... 8 Chapter 2: Review of Previous Work Overview of Accident Scenarios Main Steam Isolation Valve vii

9 MSIV Modes of Failure Flow-Accelerated Corrosion (FAC) of Steam Line FAC Description Predicting Pipe Failure SCC in the Steam Generator Tubes Steam Generator Tube Material Stress Corrosion Cracking Susceptible Areas PWSCC Model Assumptions PWSCC Crack Initiation PWSCC Crack Growth Model Determination of Critical Crack Lengths Applying Lewandowski s SCC Model to Surveillance Scheme Used in this Work Chapter 3: Steam Generator Surveillance Sampling Strategies General Overview Steam Generator Program Elements Degradation Assessment Inspection viii

10 Integrity Assessment Tube Plugging and Repairs Additional Aspects of Steam Generator Program Surveillance Sampling Programs Original Licensing Surveillance Sampling Program Updated Surveillance Sampling Program Chapter 4: Non-destructive Examination (NDE) of SG Tubes Non-Destructive Examination Techniques Eddy Current Inspection Eddy Current Technology Eddy-current Probe Manipulators Eddy-current Probe Technology Eddy Current Reliability Chapter 5: Steam Generator Surveillance Assessment and Results SG Surveillance Assessment SG Surveillance Subsequent Assessments Analysis of the Results SLB-induced SGTR Spontaneous SGTR ix

11 5.4. Sensitivity Study Core Damage Frequency Analysis Chapter 6: Summary and Concluding Remarks Model Review Model Limitations Future Research Conclusion REFERENCES APPENDIX A: MATLAB Code for Simulation x

12 List of Figures Figure 1 Westinghouse-designed two-loop PWR plant [4] Figure 2 Schematic diagram of a U-tube SG [6] Figure 3 Degradation mechanisms contributing to SG tube degradation [10]: Stress corrosion cracking (SCC); intergranular attack (IGA); inner diameter (ID); outer diameter (OD) Figure 4 Main steam system in a PWR [1] Figure 5 Failure of an elbow downstream of a tee [1] Figure 6 Failure of a heat drain pump discharge piping [1] Figure 7 Factors influencing stress corrosion cracking [19] Figure 8 Roll transition zone [20] Figure 9 Semielliptical axial crack with depth c and length a [1] Figure 10 Temperature depent IGSCC initiation time. Initiation times are plotted vs. 1000/T and the figure was produced for 19mm outer diameter tubing of Alloy 600MA [11] Figure 11 Faction of tubes for which crack initiation occurs at the beginning of each oneyear operating cycle Figure 12 Fraction of tubes for which crack initiation occurs at the beginning of each two-year operating cycle xi

13 Figure 13 Observed crack lengths in Ringhals SG after 11 years of operation [22] Figure 14 Observed crack lengths in a simulation after 11 years of operation (adapted from [1]) Figure 15 Fraction of cracks belonging to each crack growth rate group using data from [1] Figure 16 Zetec SM 23 (left) and 23A (right) are two examples man-way-mounted probe manipulators [28] Figure 17 Typical bobbin probes used for SG tube inspection [28] Figure 18 EC patters generated by bobbin (left) and rotating probe (right) [28] Figure 19 Mitsubishi Intelligent Probe [28] Figure 20 Axial (L) and circumferential (C) IDSCC in tube sheet. "95 OSL" refers to one-sided 95% confidence limit [2] Figure 21 Surveillance assessment flowchart Figure 22 Surveillance assessment flow chart for defective tubes Figure 23 Probability of at least one tube failing in the event of a SLB per SG (2-year cycles) Figure 24 Probability of at least one tube failing in the event of a SLB per SG (1-year cycles) Figure 25 Probability of at least one tube failing in the event of a SLB. POD values set to the same values used by Lewandowski [1] Figure 26 Probability of spontaneous SGTR for 2-year cycles Figure 27 Probability of spontaneous SGTR for 1-year cycles xii

14 Figure 28 Probability of at least one tube failing due to sponatneous SGTR. POD values set to the same values used by Lewandowski [1] Figure 29 Effect of relaxing crack growth rates on the probability of SLB-induced SGTR Figure 30 Probability of at least one tube failing in the event of a SLB per SG (2-year cycles). Max POD = Figure 31 Probability of at least one tube failing in the event of a SLB per SG (2-year cycles). Max POD = Figure 32 Probability of at least one tube failing in the event of a SLB per SG (2-year cycles). Max POD = Figure 33 Probability of at least one tube failing in the event of a SLB per SG (2-year cycles). Max POD = Figure 34 Probability of at least one tube failing in the event of a SLB per SG (2-year cycles). Max POD = Figure 35 Core damage frequency for SLB initiating event and two year cycles Figure 36 Core damage frequency for SLB initiating event and one year cycles xiii

15 List of Tables Table 1 MSIV failure mode distributions (presented in [15] and adapted by [1]) Table 2 Assumptions made about the PWSCC crack growth model [1] Table 3 Crack growth rate amplitudes for 20 growth groups determined from Ringhals Unit 4 data [1] Table 4 Number of tubes (or cracks) belonging to a unique cohort [i, j] for two-year cycles based on Figure 12, Figure 15, and 3,592 tubes per SG Table 5 Number of tubes (or cracks) belonging to a unique cohort [i, j] for one-year cycles based on Figure 11, Figure 15, and 3,592 tubes Table 6 Criteria for inspection result categories (adopted from [13]) Table 7 Required actions for inspection results (adapted from [13]) Table 8 Required actions for inspection results (adapted from [13]) Table 9 Assessment of Cohort [1,9] during Cycle 6 for two-year cycles. Results are obtained from Eq. (7) with N = 9 and p = Table 10 Assessment of Cohort [1,9] during Cycle 6 for two-year cycles. Results are obtained from Eq. (7) with N = 9 and p = Table 11 Tube number criteria for each category of results given a SG containing 3,592 tubes xiv

16 Table 12 Assessment of Cohort [1,9] during Cycle 6 for two-year cycles. Results are obtained from Eq. (7) with N = 47 and p = Table 13 Conditional probabilities for each sub-cohort produced in Table 12 for Cohort [1,9] during Cycle Table 14 Results of applying Eq. (8) to probabilities in Table Table 15 Final results of the surveillance assessment during Cycle 7 for Cohort [1,9]. Also represents the sub-cohorts used in the next cycle assessment for this cohort xv

17 List of Acronyms ANL CDF CSV EPRI FAC IGSCC MA MSIV NDE NEI NRC ODSCC POD PORV PRA PWR PWSCC RTZ SCC SG SGTR SLB SSC TSTF Argonne National Laboratory Core Damage Frequency Code Safety Valves Electric Power Research Institute Flow Accelerated Corrosion Intergranular Stress Corrosion Cracking Mill Annealed Main Steam Isolation Valve Non-destructive Examination Nuclear Energy Institute Nuclear Regulatory Commission Outer Diameter Stress Corrosion Cracking Probability Of Detection Power Operated Relief Valve Probabilistic Risk Assessment Pressurized Water Reactor Primary Water Stress Corrosion Cracking Roll Transition Zone Stress Corrosion Cracking Steam Generator Steam Generator Tube Rupture Steam Line Break Systems, Structures, and Components Technical Specifications Task Force xvi

18 TT TW Thermally Treated Through-wall xvii

19 Chapter 1: Introduction The research described in this thesis supports the development of a methodology for making a nuclear power plant s probabilistic risk assessment (PRA) more representative of the actual condition of the plant as monitored by the plant s surveillance program and involves extensions made to earlier work by Radoslaw Lewandowski [1]. The specific degradation mechanism used as an example in this work is primary water stress corrosion cracking of steam generator (SG) tubes as monitored by eddy current testing. Chapter 1 of this thesis starts by explaining the motivation and research objectives for this study (Section 1.1), describing historically encountered mechanisms of steam generator tube degradation (Section 1.2), and identifying modifications that have been made in steam generators to better control steam generator degradation (Section 1.3). Chapter 2 reviews the work performed by Lewandowski and discusses the components of his efforts that are applied in this work. Chapter 3 discusses the approaches and sampling strategies used by the industry for carrying out SG tube inspections. Chapter 4 focuses on eddy-current examination and SG reliability. Chapter 5 describes the activities performed in the current research effort and presents the results of this effort. Chapter 6 provides conclusions as they pertain to the work carried out here. 1

20 1.1. Motivation and Research Objectives In a program sponsored by the U.S. Department of Energy entitled Methodology Development for Passive Component Reliability Modeling in a Multi-Physics Simulation Environment, The Ohio State University developed a concept for a condition-depent approach to probabilistic risk assessment (PRA) in which the results of surveillance activities of the plant could be incorporated into periodic updates of risk prediction for the plant [1]. The approach is based on mechanistic modeling of plant degradation mechanisms. The potential benefits of a condition-depent PRA include a more plantspecific characterization of accident risk, as well as risk-informing the plant s surveillance and proactive maintenance programs. In order to assess the feasibility of a condition-depent PRA, a MATLAB program was developed by Lewandowski [1] to assess the time-depent progression of SG tube degradation and the associated conditions leading to tube failure. The code was then used to assess the risk of spontaneous SG tube ruptures and a containment bypass scenario associated with a steam-line break leading to SG tube rupture. The assessment addresses the formation and growth of axial cracks in SGs made of Alloy 600, for which crack growth models have been previously developed. The results obtained with the code over the operating lifetime of a pressurized water reactor (PWR) were found to be consistent with SG tube degradation observed historically in plants operating with Alloy 600 SGs once the mechanistic models for predicting crack initiation and growth were tuned to reflect operational experience. 2

21 The U.S. Nuclear Regulatory Commission (NRC) study for the reliability of eddy current testing, which provides a basis for assessing probability of detection (POD) as a function of crack depth in a SG tube [2], had been adopted in the Lewandowski study [1]. Unfortunately, the NRC study only addresses surveillance reliability up to the point of penetration through the thickness of the tube wall. The critical crack lengths at which a SG rupture could occur are much longer than the likely point at which the crack would penetrate the wall [1]. Thus, Lewandowski only used a constant value of POD associated with through-wall penetration. In addition, lacking data on the effectiveness of subsequent surveillances of SG tubes for which surveillances has failed to detect a crack, Lewandowski assumed a 50:50 probability of success in each subsequent surveillance under the assumption that there could be some common cause reason why a particular crack has failed to be detected. Lewandowski also assumed 100% inspection of SG tubes rather than accounting for the complexity of limited sampling of SG tubes for inspection in outages. The objective of the current study is to ext the work performed by Lewandowski in two key areas: to determine the importance of probability of detection (POD) of surveillance instruments and to evaluate the effectiveness of sampling strategies in assuring that, by chance, degradation mechanisms could be undetected in a plant. 3

22 1.2. SG Tube Degradation General Overview Steam generators are effectively the boilers of PWRs. Figure 1 provides a general illustration of a PWR. The SG serves several functions in a PWR. From a power generation perspective, its main function is to generate the steam that turns a turbine and produces electric power. Heat is carried away from the reactor core by the primary loop. Heat transferred from the primary to secondary loop generates steam on the outside of the SG tubes. Steam produced in the SG is sent to the turbine. The SG tubes make up over 50% of primary side pressure boundary area [3]. In a PWR, boiling in the primary loop is prevented by keeping the primary coolant loop highly pressurized to pressures of approximately 15.5 MPa. Steam generator tubes provide a barrier that prevents the release to the environment of fission products that are present in the primary loop. If a leak or a rupture occurs in a SG tube, this radioactive material can spill into the secondary loop, bypass containment and then possibly be released to the environment. 4

23 Figure 1 Westinghouse-designed two-loop PWR plant [4]. Steam generators are large pieces of equipment that require significant amounts of effort to maneuver. A typical SG can measure as high as 70 feet, weigh close to 800 tons, and includes anywhere from 3,000 to 16,000 steam generator tubes [5]. The SG being considered in this thesis is a Westinghouse-designed U-tube recirculating steam generator, seen in Figure 2. Originally, steam generator tubes were manufactured from Alloy 600, a nickel alloy. Significant degradation issues were observed and the nuclear industry responded by substituting SG tubes with Alloy 690 [3]. 5

24 Figure 2 Schematic diagram of a U-tube SG [6] SG Degradation Issues Many of the SG degradation issues encountered over the last 40+ years were not fully anticipated. Degradation issues encountered at nuclear power plants have been met with remedial measures designed to counter degradation mechanisms. The degradation mechanisms that have contributed to SG tube degradation over the years are represented in Figure 3. In the early-to-mid 1970 s, the primary degradation mechanism that was observed was due to wastage of SG tubes. Wastage degradation is the process of wall thinning that occurs in the secondary loop because of the water chemistry. The water chemistry was 6

25 changed in response to wastage and this degradation mechanism has essentially been eliminated [7]. Steam generator tube denting became the next prevalent form of degradation. Tube denting is due to the presence of corrosion products in the secondary loop that are introduced because the chloride chemistry of the secondary water causes corrosion of the SG tubesheet and in the crevices between tube and tube support plates [8]. Proper chemistry control of the chloride levels proved to be very effective at eliminating degradation due to denting. Fretting degradation is due to flow induced vibrations that create continuous rubbing between SG tubes and other surfaces [9]. Pitting corrosion is described by small diameter wall penetrations that are a result of localized corrosion, specifically thought to be due to chloride or sulfate acids [9]. Figure 3 Degradation mechanisms contributing to SG tube degradation [10]: Stress corrosion cracking (SCC); intergranular attack (IGA); inner diameter (ID); outer diameter (OD). 7

26 Stress corrosion cracking (SCC) is generally described by a slow crack initiation phase followed by an accelerated crack propagation phase [11]. Thus, the extensive amount of SG tube degradation by SCC was not apparent till many years (approximately 10 years as seen in Figure 3) after PWRs began operating. The other forms of degradation, such as wastage and fretting, are largely chemical or mechanical in nature. Stress corrosion cracking is also affected by chemical and mechanical phenomenon, but SCC in SG tubes is largely attributed to the choice of SG tube material. For nuclear power plants built in the early-to-mid 1970 s, nearly all of the PWR SG tubes (except for one) were manufactured from mill-annealed Alloy 600 (Alloy 600MA) [5]. Unfortunately, Alloy 600MA would prove to be a poor choice of material because of its susceptibility to SCC. Thermally treated Alloy 690 (Alloy 690TT), on the other hand, has shown no instances of SCC [5] due to its much larger chromium composition; Alloy 600 chemical composition contains 14-17% chromium and Alloy 690 chemical composition contains 28-31% chromium [3] Current State of Steam Generators Remedial measures and structure replacements have been implemented to reduce instances of SG tube degradation as much as possible. In addition to changing the water chemistry, stainless steel tube support plates have replaced carbon steel support plates in an effort to minimize the amount of denting [5]. Alloy 690TT looks to be a promising material when it comes to preventing the large amount of tube degradation experienced with Alloy 600MA. 8

27 Although these measures have helped reduce the amount of degradation, it has not eliminated degradation completely. The industry employs several practices in order to detect instances of degradation. Surveillance strategies, such as though addressed in NEI [12], are used during refueling cycles in an attempt to identify degraded tubes before they lead to leak or rupture. The guidelines for performing examinations, PWR Steam Generator Examination Guidelines [13], have evolved in response to the degradation observed in SGs. Detecting degraded tubes is depent on the effectiveness of non-destructive examination (NDE) techniques. Eddy-current examination is the primary method of examination when it comes to examining SG tubes, as will be described in Chapter 4. Eddy-current technology continues to improve to make inspections quicker and more reliable. Eddy-current reliability experiments have been conducted in order to assess its reliability, most notably in NUREG/CR-6791 Eddy Current Reliability Results from the Steam Generator Mock-up Analysis Round-Robin [2]. 9

28 Chapter 2: Review of Previous Work Previous work conducted by Lewandowski [1] (Incorporation of Corrosion Mechanisms into a State-depent Probabilistic Risk Assessment) provides the framework for current study. This chapter describes the development of his work along with how it is applied in the current work. Section 2.1 reviews the accident scenario that was considered by Lewandowski. Sections 2.2 through 2.4 discuss the components that influence the accident scenario. Section 2.5 determines critical crack lengths associated with the accident scenario. Section 2.6 describes how the research conducted by Lewandowski is applied in this work Overview of Accident Scenarios Two accident scenarios are considered leading to steam generator tube rupture (SGTR). Historically, SGTRs have occurred in PWRs as the result of a variety of degradation mechanisms. In the Lewandowski analysis, spontaneous failure of SG tubes was assumed to occur when an axial crack length exceeds the critical value at which crack growth becomes unstable. In addition, a scenario was analyzed in which failure of a steam line leading to depressurization of the secondary side of a SG would lead to SGTR at a smaller crack length than for the spontaneous SGTR. This type of failure could be of particular interest because it could lead to a core damage event involving bypass of the 10

29 containment and release of radioactive material to the environment. A complex combination of failures is required for this type of scenario. Rupture of a steam line must occur with the indepent failure of a main steam isolation valve (MSIV) to close. The passive system failure mechanisms in this accident scenario are ones that have been addressed in NUREG/CR-6923: Proactive Materials Degradation Assessment [14]. Failure mechanisms for rupture of a steam-line are attributed to flow accelerated corrosion (Section 2.2) and SCC is the mechanism that leads to failure of SG tubes (Section 2.4) [14]. Failure of the MSIV to close involves failure of an active component. The initiating event in this scenario is the steam line break (SLB). In PWRs, this failure can occur either upstream or downstream of the MSIV. Steam lines penetrate the containment wall with the MSIV controlling the flow through the containment wall. If a failure occurs upstream of the MSIV, the material is released into containment. This is an event analyzed in a plant s safety analysis report as a design basis accident. If a SLB occurs downstream of the MSIV, then there exists the potential for material to spill outside of containment and into the environment. Thus, MSIV should close when a SLB event occurs to prevent the secondary loop from depressurizing and prevent material from spilling to the environment. In this accident scenario, it is assumed that the SLB occurs downstream of the MSIV and that the MSIV fails to close. With this series of events, the result is a depressurization of the secondary loop. Typically the SLB occurs some years after the initial startup of the plant [1] allowing SG tube degradation to progress as well. If SG tube cracks have grown large enough when the SLB/MSIV event occurs, then the 11

30 depressurization causes a SGTR. This progression of events, SLB/MSIV/SGTR, during operation will cause radioactive material initially confined to the primary loop to spill into the secondary loop and then the environment. In addition to the potential release of radioactive material in the primary coolant, there is a probability that core meltdown with the containment bypassed can occur [1] Main Steam Isolation Valve Estimation of the MSIV conditional probability of failure to close can be based on the NUREG/CR-6246 report, Effects on Aging and Service Wear on Main Steam Isolation Valves and Valve Operators [15]. The MSIVs are housed in a valve vault, which contains the SG power-operated relief valves (PORVs), the code safety valves (CSVs), the MSIVs, and the MSIV bypass valves. Steam lines pass through this valve vault before penetrating containment to deliver steam to the turbines. In the event that a SLB occurs, Engineered Safety Feature Actuation System automatically closes the MSIV. If SLB occurs inside containment and the MSIV does not close within 5 seconds, then reverse steam flow from intact SGs will cause the pressure in containment to increase [15]. A schematic representation of the SGs, valve vault and turbine building is displayed in Figure 4. 12

31 Figure 4 Main steam system in a PWR [1] MSIV Modes of Failure Three basic types of MSIVs are used in PWRs: globe, check and ball valves. Several different failure modes are possible for MSIVs: failure to open, failure to close, spurious valve opening, spurious valve closure, valve stem or shaft leakage, body-tobonnet leakage, and valve-seat leakage. The distributions of valve failures for each of the 3 types of valves are given in Table 1. Although this report provides considerable information on the numbers of valve failures that have occurred, it does not indicate the associated number of challenges. Based on assumptions made regarding the number of associated demands by Lewandowski, the failure rate per demand would be assessed as 13

32 1.9E-2. In comparison with NUREG-1150 [16] values for failure to close, this failure probability appears to be conservative. Table 1 MSIV failure mode distributions (presented in [15] and adapted by [1]). 2.3 Flow-Accelerated Corrosion (FAC) of Steam Line FAC Description Environments most susceptible to FAC are those that contain a flowing two phase flow (water and steam mixture). FAC typically occurs in areas of high turbulence, such as those encountered downstream of elbows and tees. Water chemistry, susceptible material, fluid dynamics and flow geometry contribute to the material loss and wall thinning phenomena characteristic of FAC [17]. The process is mainly due to dissolution of the oxide layer that forms at the metal surface. Thus, the process may never reach equilibrium; oxide layer dissolution consequently leads to the production of another oxide layer, which then also experiences dissolution and the process continues to repeat in this fashion. 14

33 Predicting Pipe Failure The FAC corrosion model adopted by Lewandowski is the KWU-KR model [18]. Operating conditions and parameters for two industry examples were applied by Lewandowski to the KWU-KR model to predict the time of failure of steam-lines subjected to FAC [1]. The first example is failure of an elbow downstream of a tee and the second example is the failure of a heat drain pump discharge piping. The two examples show the capacity of the piping material as a function of time and the load placed on the pipe which was assumed to be constant over time [1]. The capacity factor of the pipes as a function of time for the two examples is shown in Figure 5 and Figure 6. As seen in the two figures, the predicted times of failures were 9.4 and 13.5 years for example one and two, respectively. Figure 5 Failure of an elbow downstream of a tee [1]. 15

34 Figure 6 Failure of a heat drain pump discharge piping [1] SCC in the Steam Generator Tubes Steam Generator Tube Material Initially, most steam generators around the world used mill-annealed Alloy 600 (600MA). Alloy 600MA itself was largely responsible for the significant amount of SCC observed in steam generators. The objective of the mill-annealing process is to dissolve carbide, obtain a relatively large grain size, and cover grain boundaries with carbides with slow cooling [9]. This mill-annealing process would prove to be inadequate when many instances of SCC were showing up in SG tubes. To remedy further degradation, 600MA were also thermally treated (600TT) to relieve fabrication stresses and to further improve the microstructure [9]. The significance of the thermal treatment process is to increase the amount of carbides at the grain boundary and enhance diffusion of chromium to the grain boundaries which helps to prevent primary water stress corrosion cracking 16

35 (PWSCC) and outer diameter stress corrosion cracking (ODSCC), respectively. It was later found that SCC is highly depent on the chromium content [9]. Although Alloy 600TT has significantly reduced instances of SCC, almost all new SGs being built utilize Alloy 690TT because of its much larger chromium content Stress Corrosion Cracking Certain materials are prone to SCC when exposed to tensile stresses and a corrosive environment. Stress corrosion cracking is characterized by an initiation phase that leads to slow crack growth, followed by a propagation phase which usually entails an accelerated crack growth rate. The initiation phase is defined by an incubation period and a slow growth phase. Once the crack has reached a critical length, cracks enter the propagation phase in which crack growth increases significantly. It is this sudden and accelerated propagation phase that can lead to sudden and unexpected tube ruptures. The time required for initiation to begin and the mechanisms leading to SCC remain largely unknown. Mechanical loading, environmental conditions, and susceptible material conditions are required for SCC, as seen in Figure 7. It is difficult to quantify the amount to which each condition affects SCC. However, it is generally agreed that reduction in any of these conditions will effectively reduce instances of stress corrosion cracking. 17

36 Figure 7 Factors influencing stress corrosion cracking [19] Susceptible Areas Many areas in the steam generators, both primary and secondary sides, are susceptible to stress corrosion cracking. PWSCC can occur in roll transition zone (RTZ), U-bs, tube denting locations and on plugs and sleeves [10]. ODSCC has been known to occur in the tube-support-plate crevice, tube sheet region and free-span regions [10]. RTZ regions typically experience axial cracks, but circumferential cracks have been known to occur. The area of the RTZ that experiences the most serious degradation from PWSCC is the expanded zone above the tube sheet, as seen in Figure 8. Steam generator designers have aimed to circumvent this issue in several different ways, but fabrication errors still leave this region susceptible to PWSCC [10]. 18

37 Figure 8 Roll transition zone [20] PWSCC Model Assumptions Simulation of Lewandowski s crack growth model requires assumptions to be made about the SG, SG tube material, operating conditions, and crack morphology. Westinghouse-designed recirculating steam generators are modeled here and further assumptions are drawn from common knowledge of PWRs. These assumptions are summarized in Table 2. In this model, PWSCC is assumed to occur in the roll-transition zone where axial cracks are known to be the dominate crack morphology [10]. Thus, the crack morphology and region being considered is an axial growth of a semi-elliptical crack in the roll transition zone. Figure 9 gives a visualization of the semi-elliptical crack in a SG tube. The crack depth-to-length ratio is assumed to be 1:3 and remains constant during crack growth. 19

38 Table 2 Assumptions made about the PWSCC crack growth model [1]. 20

39 Tube Thickness Figure 9 Semielliptical axial crack with depth c and length a [1] PWSCC Crack Initiation Crack initiation is usually treated statistically because the mechanisms contributing to crack initiation remain largely unknown. Lognormal or Weibull distributions are usually used to describe crack initiation. Lewandowski s model applies the work done by Staehle [11], in which a lognormal distribution was used to describe crack initiation. Instead of describing the formation and evolution of cracks in the initiation phase, the lognormal distribution applied here indicates the instantaneous time at which a crack is introduced to a tube at a length of 0.1 mm. Figure 10 shows the temperature-depent initiation time for intergranular stress corrosion cracking (IGSCC) cracks in Alloy 600MA. Industry records indicate that SCC occurs predominately in the hot leg of SGs. Thus, this study looks at an average hot leg 21

40 temperature of 330 C which is indicated by the red lines in Figure 10. At this temperature, Figure 10 shows that a mean time of 9.3 years and a standard deviation years is obtained as indicated by red lines. Thus, it can be determined how many cracks are introduced at the beginning of each operating cycle using these parameters. In this thesis, two different cycle lengths are considered: Case 1 for one-year long operating cycles and Case 2 for two-year long operating cycles. Applying these parameters to the lognormal distribution produces Figure 11 and Figure 12 for one and two year cycles, respectively. In both cases, the majority of cracks are introduced in the first few cycles. Under these assumptions, about 67% of tubes would have a crack introduced over a 40-year lifetime. Figure 10 Temperature depent IGSCC initiation time. Initiation times are plotted vs. 1000/T and the figure was produced for 19mm outer diameter tubing of Alloy 600MA [11]. 22

41 Fraction of Tubes Affected Fraction of Tubes Affected f t = t 2π e (ln t 9.3) 2 2(3.162) Cycle (t) Figure 11 Fraction of tubes (f(t) integrated over the cycle length) for which crack initiation occurs at the beginning of each one-year operating cycle f t = t 2π e (ln (t) 9.3) 2 2(3.162) Cycle (t) Figure 12 Fraction of tubes (f(t) integrated over the cycle length) for which crack initiation occurs at the beginning of each two-year operating cycle. 23

42 PWSCC Crack Growth Model Crack initiation is followed by a phase of crack growth that proceeds in a nonlinear and accelerated fashion. Lewandowski adopted an empirical model for SCC [21] where crack growth rates for SCC are depent on the stress intensity factor around the crack tip, as seen in Eq. (1), with the crack tip intensity factor given by Eq. (2). da dt = α(k K th) β K = Fσ πa Eq.(1) Eq.(2) where α = crack growth amplitude K = crack tip stress intensity factor K th = crack tip stress intensity factor threshold β = exponent F = geometric factor σ = stress at crack tip a = crack dimension in the direction of crack growth Model parameters are determined empirically, such as was done by Peter M. Scott for primary water at 330 C using data collected for Alloy 600 by Smialowska et al. [21]. The crack tip stress intensity factor threshold was found to be 9 MPa m and the growth exponent as 1.16 [21]. The crack growth amplitude was determined for cold worked Alloy 600 and non-cold worked Alloy 600. Application of these parameters to Eq. (1) is shown in Eq. (3) and (4), respectively. 24

43 da dt = (K 9) 1.16 da dt = (K 9) 1.16 Eq.(3) Eq.(4) Equations (3) and (4) represent a single crack growth rate. Realistically, SG tubes experience a variety of crack growth rates leading to a distribution of crack lengths. Crack growth rates differ as a result of residual stresses left over in tubes (e.g. residual stresses vary as a result of the forming process in the U-b region) and environmental conditions, such as temperature. To better reflect the various crack growth rates, Scott s equation was applied [1] to industry data obtained from Ringhals Unit 4 after 11 years of operation, as reported by Wu [22] and shown in Figure 13. Wu reported a gamma distribution of crack lengths in SG tubes. As seen in Figure 13, there about 20 different crack lengths observed after 11 years of operation. 25

44 Figure 13 Observed crack lengths in Ringhals SG after 11 years of operation [22]. Using Figure 13 to reflect the various crack growth rates observed in SG tubes, 20 unique crack growth amplitudes were substituted into Scott s equation in [1]. Initially, amplitudes on the order of Scott s original amplitude were chosen. Cracks were then simulated over 11 years and compared to the data in Figure 13 and a new set of amplitudes were estimated based on a statistical fit. The resulting crack growth amplitudes and affected tube distribution are shown in Table 3 and Figure 14, respectively. Compared to Scott s original crack growth amplitude of , these amplitudes are about 1-2 orders of magnitude smaller. Figure 13 also provides information for α and β of the gamma distribution, where α = and β = The gamma distribution with these parameters is used to 26

45 determine the fraction of cracks that belong to each crack growth rate group, which is represented in Figure

46 Number of Tubes Table 3 Crack growth rate amplitudes for 20 growth groups determined from Ringhals Unit 4 data [1] Crack Length (mm) Figure 14 Observed crack lengths in a simulation after 11 years of operation (adapted from [1]). 28

47 Fraction of cracks in each group g x Crack Growth Rate Groups (x) = x e 1.395x Γ(3.393) Figure 15 Fraction of cracks g(x) belonging to each crack growth rate group using data from [1]. 2.5 Determination of Critical Crack Lengths Surveillance programs are applied to SG tubes during refueling cycles in order to identify tubes that may lead to leaks or catastrophic failure. However, if cracks go undetected and continue to grow, they will eventually reach lengths that make them susceptible to SGTR. Ruptures are determined by the pressure differential across the tube wall, which is consequently depent on the through-wall (TW) thickness. Formulation of the critical pressures and crack sizes that lead to the SGTR are given by Eq. (5a-e) and are discussed in NUREG/CR-6664, which was developed by Hahn and modified by Erdogan [23]. p cr = σ Rh mr = p b m Eq.(5a) 29

48 σ F = k(s Y + S U ) (with k = ) m = λ e 1.25λ λ = [12(1 v 2 )] 0.25 p b = σ h R c Rh = 1.82c Rh Eq.(5b) Eq.(5c) Eq.(5d) Eq.(5e) where p cr = critical pressure required for tube rupture σ F = flow stress h = wall thickness of tube R = mean tube radius p b = burst pressure of an unflawed virgin tube S Y = yield strength S U = ultimate tensile strength v = Poisson ratio 2c = axial crack length The accident scenario under consideration in [1] leads to a rapid depressurization of the secondary side and a change in pressure differential across the SG tube wall. Thus, the critical depth for the SLB accident scenario will be depent on the pressure differential during the accident scenario. It is also of interest to determine the critical depth corresponding to a spontaneous SGTR event during normal operation. Lewandowski applied Alloy 600MA and information about pressure differential for the 30

49 two cases and found that the critical lengths are 33.0 mm for the SLB event and 62.5 mm for a spontaneous SGTR event [1] Applying Lewandowski s SCC Model to Surveillance Scheme Used in this Work Following crack initiation (Section 2.4.5), cracks grow according to the modifications of Scotts crack growth model (Section 2.4.6) in [1]. During each cycle i, a fraction of the total SG tubes have cracks introduced in them, where the number of cracks introduced is depent on the lognormal distribution. As indicated in Section 2.4.6, Ringhals data was adopted in [1] to determine the distribution of introduced cracks among the 20 crack growth rate groups during cycle i which follows a gamma distribution (see Figure 15). Thus, during each cycle i, the initiated cracks are assigned to a crack growth rate group j by the gamma distribution. For clarity, the indices [i, j] is termed a cohort which is defined as the unique combination of initiation cycle i and crack growth rate group j. This leads to a matrix of cohorts associated with each of these unique combinations with dimensions [i j]. Applying the lognormal distribution (Figure 11 for one year cycles and Figure 12 for two year cycles) and gamma distribution (Figure 15) to the number of tubes in a SG (3,592 tubes per SG) at the beginning of each cycle, the number of tubes 1 belonging to each cohort, N i,j, is determined. Cycles lasting one and two years are investigated in this thesis which leads to slightly different cohort sizes as seen in Table 4 and Table 5 for cycles lasting two and one years, respectively. 1 Technically cracks belong to cohorts, not tubes. However, it is assumed that a tube contains only one dominant crack so that cracks in a cohort equates to tubes in a cohort. Therefore, cracks and tubes may be used interchangeably in this analysis. 31

50 For example, consider Cohort [1,9] in Table 4 for two-year cycles. The lognormal distribution in Figure 12 shows that the fraction of SG tubes that have cracks introduced in them during the first cycle is The gamma distribution in Figure 15 shows that of the 0.31 affected tubes, about of them have cracks that belong to the ninth crack growth rate group. Then for 3,592 tubes in the SG, , tubes would belong to Cohort [1,9] as seen in Table 4. To further clarify a cohort, note that a cohort is not necessarily defined by a crack length; it is the unique combination of cycle introduction and crack growth rate that defines a cohort so that the crack length at any point in time for all cracks in the cohort is determined by Lewandowski s crack growth model. In this model, only one crack is initiated and assumed to grow in a SG tube. Thus, all cracks in a cohort have the identical length at any point in time. Once a cohort is established it is associated with a specific number of SG tubes (or cracks). Additional tubes with cracks cannot be born into a cohort and no cracks may transfer from one cohort to another; the cohort remains its own entity. However, cracks may be removed from a cohort through the physical process of plugging a tube, which will be discussed further in Chapter 3. 32

51 Cycle Table 4 Number of tubes (or cracks) belonging to a unique cohort [i, j] for two-year cycles based on Figure 12, Figure 15, and 3,592 tubes per SG. Crack Growth Rate Groups [i,j]

52 Cycle Table 5 Number of tubes (or cracks) belonging to a unique cohort [i, j] for one-year cycles based on Figure 11, Figure 15, and 3,592 tubes. Crack Growth Rate Groups [i, j]

53 The crack growth model developed by Lewandowski was applied to the induced SGTR scenario (Section 2.1) and spontaneous SGTR to determine the probability that one or more cracks would grow to the critical lengths associated with these events while avoiding detection. In practice, tubes are selected from a SG and inspected using NDE techniques. Thus, the probability that a crack avoids detection is determined by first it being sampled from the SG and second by the probability of failure of the NDE techniques to detect cracks that have progressed to limits that require it to be taken out of service. Lewandowski s approach to selection and detection probabilities was only approximately based on industry practice and observations (Section 1.1). In this thesis, the crack growth model for PWSCC in SG tubes developed by Lewandowski is adopted, but the tube sampling approach and detection probabilities are different. The sampling approach used in this thesis more closely resembles the sampling strategies that the NRC requires all licensees to follow. These strategies are described in NEI [12] and supporting documents and are discussed in more detail in Chapter 3. In addition, the NDE reliability information used in this thesis better reflects the probability of detection as a function of crack depth. NDE reliability data is adopted from NUREG/CR-6791 [2] and the details of this data is described are Chapter 4. The final results produced from Lewandowski and in this thesis logically lead to differing results when assessing the two SGTR events. These results are presented in Chapter 5. 35

54 Chapter 3: Steam Generator Surveillance Sampling Strategies The purpose of Chapter 3 is to describe the technical specification requirements for steam generator surveillance and the details of the sampling strategy for periodic eddy current testing of steam generator tubes. Section 3.1 discusses some background and the documents that contribute to the plant s Steam Generator Program, which provides details of the steam generator surveillance program. Section 3.2 describes the mandatory assessments that licensee must include in their Steam Generator Program. Section 3.3 provides the details of the original and updated surveillance sampling strategies General Overview SG tubes provide both a pressure and heat transfer boundary between primary and secondary sides and they confine radioactive material to the primary side. Any leak or rupture of SG tubes will cause radioactive material to spill into the secondary side which can then leak to the environment and potentially affect the health of the public. Nuclear power plants are allowed to operate with a limited amount of leakage from fuel pins to the primary system. In a SGTR event, the secondary side relief valve can open and lead to significant release of radioactive material, in particular I-131, to the environment. A SGTR is in essence a small break Loss of Coolant Accident and represents a demand on the emergency core cooling system. Thus, a SGTR is a potential initiator of a severe 36

55 accident. Therefore, monitoring degradation mechanisms and SG tube crack propagation is important to maintain safe and reliable operation of PWRs. There are a number of elements to the Steam Generator Program. The framework for the Steam Generator Program is outlined in Nuclear Energy Institute (NEI) 97-06, Steam Generator Program Outlines [12]; Electric Power Research Institute (EPRI) PWR Steam Generator Examination Guidelines [13] and Technical Specifications Task Force (TSTF)-449 Steam Generator Tube Integrity [24] provide guidelines for the industry. These documents focus on issues related to management and repair of steam generator tubes. The requirements identified in NEI have been accepted by the NRC as requirements that all licensees must follow. The major goal of the SG program is to provide a balance of prevention, inspection, evaluation and repair, and leakage monitoring measures [12]. The following sections introduce some of aspects that licensee must address in their SG program Steam Generator Program Elements The simulation developed in this work only considers a few aspects of the Steam Generator Program. Each aspect plays an important role in the SG program, but the aspects regarding Degradation Assessment, Inspection, Integrity Assessment, and Tube Plugging and Repairs are applicable to work being carried out here. The following sections are briefly discussed here and will be covered in more detail in following sections and/or chapters: Degradation Assessment will be addressed in Section 3.3; Inspection and monitoring in Chapter 4; Integrity Assessment in Chapter 5; and Tube Plugging and Repairs is discussed throughout the document. 37

56 Degradation Assessment Before performing any type of inspection or integrity assessment, it is necessary to determine the scope of the inspection during the upcoming outage. The Degradation Assessment is outlined by EPRI Steam Generator Integrity Assessment Guidelines [25] which provides a licensee with the acceptable methods for this assessment. This assessment encompasses structures and components making up and aiding the support of the pressure boundary within the steam generator, such as SG tubes, plugs, and tube support plates. Proper assessment must consider operating experience, known degradation mechanisms in the SG, and regions known to be susceptible to degradation mechanisms. When complete, the Degradation Assessment will include the following features [12]: Identifying existing and potential degradation mechanisms Choosing techniques to test for degradation based on the probability of detection and sizing capability Establishing the number of tubes to be inspected Establishing the tube integrity limits for condition monitoring and operational assessment Inspection Eddy-current inspections are typically used for SG tube examinations, as discussed in Chapter 4. Guidance for carrying out inspections is outlined in EPRI PWR Steam Generator Examination Guidelines [13] and the scope of the inspection is 38

57 determined by the Degradation Assessment. Information obtained from an inspection is used for future Degradation Assessments, condition monitoring and operational assessments Integrity Assessment Tube integrity is assessed following the inspection results which contribute to the Integrity Assessment. Condition monitoring assesses the current state of the SG tubes and compares the integrity to the assessment carried out during the previous outage. This may consist of re-inspecting degraded tubes from the previous outage that did not require plugging or comparing the results to previous outages in order to determine any abnormal degradation in tubes or areas of the SG. Operational assessment uses the current integrity state to demonstrate that tube integrity will be maintained until the next outage or inspection. Evaluation methods and uncertainty considerations are assessed by following the guidance in EPRI Steam Generator Integrity Assessment Guidelines [25] Tube Plugging and Repairs Tube plugging and repairs are implemented to ensure that leaks or ruptures do not occur during operation. Repairs are implemented to restore and revitalize tube integrity so as to not remove it from service; repairs are usually done by welding sleeves to the outside of the tubes. If degradation has progressed considerably then a tube will be plugged rather than repaired. Plugging differs from a repair by removing the tube from service. Repairs seek to maintain the pressure boundary in a SG tube and plugging takes the tube out of service. Plugging prevents a SG tube from leaking or possible ruptures, 39

58 but it decreases the number of tubes in operation which effectively leads to a decrease in heat transfer. Plugging and repair guidelines are provided by EPRI PWR Steam Generator Examination Guidelines [13] Additional Aspects of Steam Generator Program In addition to the aspects discussed above, NEI addresses several other aspects in the Steam Generator Program. Primary-to-secondary leakage monitoring provides information for operators that signals that SG tube integrity is not maintained. For steam generators that indicate leakage, the operational primary-to-secondary leakage cannot exceed more than 150 gallons per day [12]. Maintenance of steam generator secondary-side integrity is important because corrosion of SG tubesheet can lead to SG tube denting that effects SG tube degradation. Secondary-side inspection can be carried out visually. Primary and secondary water chemistry, foreign material exclusion, contractor oversight, self-assessment, and reporting make up the other elements of the Steam Generator Program Surveillance Sampling Programs Inspection of SG tubes measures the depth of cracks in the walls of SG tubes. Tubes that have less than 20% through wall (TW) depth, or no degradation at all, are acceptable and do not require further attention. Degraded tubes are tubes that have TW depths between 20% and 40%. Degraded tubes that are detected must be inspected again after the next cycle. Defective tubes have TW depths greater than 40% [24]. Defective tubes that are detected are taken out of service by plugging the tubes. Tubes that reach 40

59 100% TW depths will logically contribute to leak rates. The TW depth criterion for acceptable, degraded and defective tubes applies to both the original and updated surveillance sampling programs. At the of each cycle, one SG is scheduled for inspection; the other SGs not inspected are labeled as unscheduled SGs. SG tubes are randomly selected from the scheduled SG and the number of tubes sampled deps on the sampling programs described in Section and After all tubes have been inspected and have been classified according to their TW depth, the results of the inspection are classified into one of three categories that are summarized in Table 6. All percentages in Table 6 are a percentage of the total number of tubes being inspected. The categories in Table 6 apply to both the original and updated surveillance sampling programs. Table 6 Criteria for inspection result categories (adopted from [13]). Category Degraded Criteria Defective Criteria C-1 Less than 5% 0 tubes C-2 Between 5% and 10% One or more tubes but less than 1% C-3 Greater than 10% Greater than 1% *Percentages are depent on the total number of tubes sampled during the inspection Original Licensing Surveillance Sampling Program It is well documented that the amount of SCC observed in SG tubes was not anticipated when SGs first began operation [10]. It wasn t till nearly a decade later when 41

60 the effects of SCC became apparent. This lack of foresight led to a surveillance sampling program that potentially provided inadequate coverage of SG tubes. As a minimum, the original surveillance strategy required sampling of number tubes that is 3% times the number of SGs in the plant. For instance, the Westinghouse designed PWR considered in this thesis contains 3 SGs so that 9% of the total tubes in the scheduled SG would be sampled. The original scope of inspection covered only the hot leg entry point to the first support plate on the cold leg side of the U-b. Although SCC and other degradation mechanisms are a function of temperature, the cold-leg side is still at a high enough temperature to warrant investigation [24]. Table 7 outlines the necessary course of action that must be carried out deping on the category of inspection results (see Table 6). In Table 7, the first column refers to the initial inspection of the scheduled SG. If the required action leads to unscheduled SGs being inspected, then the first column in this table applies to the first inspection in the unscheduled SGs. Following the initial inspection in either scheduled or unscheduled SGs, if the action requires that additional tubes be inspected, then the First Expansion column applies to the additional sampled tubes. Similarly, if action requires that additional tubes be inspected following the First Expansion, then the Second Expansion column applies to the additional sampled tubes. 42

61 Table 7 Required actions for inspection results (adapted from [13]). Expansion of a Sample Plan or Expansion into Unscheduled SG First Expansion Second Expansion Results Action Required Results Action Required Results Action Required C-1 None C-2 Inspect an additional 2S of all remaining tubes in scheduled SG. C1 None C-2 Inspect an additional 4S of all remaining tubes in the scheduled SG. C-3 Inspect all remaining tubes in scheduled SG and 2S of tubes in each unscheduled SG. S = 3% times the number of SGs in the plant C-3 Inspect all remaining tubes in scheduled SG and 2S of tubes in each unscheduled SG. C-1 None C-2 None C-3 Inspect all remaining tubes in scheduled SG and 2S of tubes in each unscheduled SG Updated Surveillance Sampling Program Over time, improvements have been made to the SG surveillance program to be more effective at detecting SG tube degradation. These improvements replace outdated 43

62 technical specifications with ones that consider information obtained from plant data and studies performed with SGs. Compared to the outdated surveillance program the updated surveillance program is better equipped to identify the various degradation mechanisms and manage them more effectively [24]. For instance, inspections now cover the entire length of the SG tube, instead of the length from the entry point to first support plate on the cold-leg side. Steam generator surveillance programs have also been updated to account for changes in SG tube material. Initially, most steam generators around the world used Alloy 600MA which turned out to be largely responsible for the significant amount of SCC observed in steam generators. To remedy further degradation, 600MA were also thermally treated, 600TT, to relieve fabrication stresses and to further improve the microstructure [9]. Although Alloy 600TT has significantly reduced instances of SCC, almost all SGs being built utilize Alloy 690TT because of its much larger chromium content. Stress corrosion cracking is highly depent on the chromium content and no instances of cracking have been observed in Alloy 690TT [9]. This information has led to inspection intervals based on the SG tube material. Specifically, 100% of the SG tubes must be inspected within an interval of 60 months for SG tubes of Alloy 600MA; Alloy 600TT intervals are 120 months, followed by 90 months, and then every 60 months afterwards; Alloy 690TT inspection intervals are 144 months, 108 months, 72 months, and then 60 months thereafter [24]. Another major change is the amount of tubes that are inspected during each outage. Before, only 3% times the number of SGs in the plant were sampled. The updated 44

63 sampling [12] has exted the amount of tubes sampled to be 20% of the total number of tubes in the SG. Table 6 still applies but the course of action differs as seen in Table 8. The differences between the updated and the original sampling programs are highlighted in Table 8. The details of the different expansions discussed for Table 7 also apply to Table 8. 45

64 Table 8 Required actions for inspection results (adapted from [13]). Expansion of a Sample Plan or Expansion into Unscheduled SG First Expansion Second Expansion Results Action Required Results Action Required Results Action Required C-1 None C-2 Inspect an additional 20% of all remaining tubes or tube section in the affected sample in this SG C-3 Inspect all remaining tubes in scheduled SG and 20% of tubes in each unscheduled SG. C1 None C-2 Inspect an additional 20% of all remaining tubes in the scheduled SG. C-3 Inspect all remaining tubes in scheduled SG and 20% of tubes in each unscheduled SG. C-1 None C-2 Inspect all remaining tubes in scheduled SG. C-3 Inspect all remaining tubes in scheduled SG and 20% of tubes in each unscheduled SG. 46

65 Chapter 4: Non-destructive Examination (NDE) of SG Tubes Chapter 4 describes NDE technology, equipment and reliability. Section 4.1 discusses NDE examination methods, specifically eddy-current testing as it is the primary means for examining SG tubes. Section 4.2 covers the eddy current examination equipment and its advancements. Section 4.3 introduces eddy current reliability and the formulation of the POD curve used in the analysis of this work Non-Destructive Examination Techniques Steam generator inspections should be quick and efficient to reduce the amount of outage time not only to reduce the direct cost of performing the inspections, but to also reduce economic losses from station downtime. NRC regulatory requirements for frequency of inspections and number of SG tubes that must be sampled impinge on the plant s availability. Industry demands have led to significant improvements in NDE technologies that have made inspection times shorter and more effective. NDE techniques are used to identify and determine flaw sizes in systems, structures and components (SSC). Non-destructive examination is advantageous because the integrity of SSCs is not impaired when performing these examinations. Although both ultrasonic and eddy-current examinations can be used for detecting flaws in SG tubes, eddy-current inspection is the primary means for in-service inspections [2]. 47

66 Eddy Current Inspection Eddy-current (EC) inspections utilize the principles of electromagnetic induction to detect cracks in SG tubes. An alternating current is applied to a probe made of conducting material which creates a magnetic field around the probe. The probe moves through the inside of the SG tubes which are electrically conductive. Magnetic fields generated from the probe induce eddy currents in the SG tubes; the induced eddy currents also generate magnetic fields. Thus, eddy-current inspections occur by observing how the generated magnetic field in the probe is opposed by the induced eddy current magnetic field in the SG tubes [26]. When a flaw is present, the eddy currents in the SG tube are distorted and the distortion is displayed on a screen for the operator to see. Some advantages of EC inspection are [27]: Inspections are portable and give instant results via display screens Able to detect surface and near surface cracks or defects Small cracks and other defects are readily detected 4.2. Eddy Current Technology Eddy-current Probe Manipulators Although eddy-current inspection probes do not require direct contact with the material, close proximity to the material is necessary. Eddy-current inspection probes are small and effective enough so that they can inspect tubes from inside the SG tubes. Before the advancement in technology and robotics, eddy-current operators had to crawl through manholes in the SG to setup the equipment in order to carry out the 48

67 inspection. As expected, this caused operators to acquire substantial amounts of radiation exposure from this procedure, especially if the procedure was not carried out quickly. Fortunately, probe manipulator technology has evolved so that human interaction is significantly reduced through the use of robotics [28]. Probe manipulators fit right over the manholes and video cameras located on the arms allow the operator to control the arm to guide the tubes into place. Zetec SM 23 and 23A, seen in Figure 16, are two such manway-mounted probe manipulators designed to perform EC inspections. Figure 16 Zetec SM 23 (left) and 23A (right) are two examples man-way-mounted probe manipulators [28] Eddy-current Probe Technology Probe technology has evolved from simple bobbin probes to more advanced rotating probes. Bobbin probes, Figure 17, are effective tools for detecting axial cracks, but it was found that the induced eddy-current makes it difficult to detect circumferential cracks [28], as seen in Figure 18. Rotating probes were developed in order to detect 49

68 circumferential cracks. As the name implies, rotating probes physically rotate in the device to induce circular eddy currents capable of detecting circumferential cracks, as seen in Figure 18. In Figure 18, pancake coils refer to a special type of rotating probes, where the detector is flat and rotates in place. Rotating probes are effective probes for detecting SCC, but it takes longer to perform inspections with these probes relative to the individual bobbin probes. The Mitsubishi Intelligent Probe shown in Figure 19 represents the advanced eddy-current probes used today which integrate bobbin and rotating probes onto the same device. Figure 17 Typical bobbin probes used for SG tube inspection [28]. 50

69 Figure 18 EC patters generated by bobbin (left) and rotating probe (right) [28]. Figure 19 Mitsubishi Intelligent Probe [28] Eddy Current Reliability Condition monitoring assesses the current state of SG tubes and operational assessment is used to determine if the SG tube structural integrity has been maintained over the operating cycle. Eddy current inspections and reliability data provide the necessary information to develop these assessments. Detection reliability is crack size depent as cracks remain largely undetectable (POD < 40%) until they reach 40-50% TW depth [10]. Thus, it is beneficial to establish quantitative information regarding POD as a function of TW depth. 51