ML13030A451

SCE Brief on Issues Referred by the Commission, Attachments 1 to 5, Including Affidavits of Richard Brabec, Vickram Nazareth, Confirmatory Action Letter CAL 4-12-001

ML13030A451
Person / Time
Site:	San Onofre
Issue date:	01/30/2013
From:	Brabec R, Nazareth V Southern California Edison Co
To:	Atomic Safety and Licensing Board Panel
SECY RAS
Shared Package
ML130310300	List: ML13030A441 ML13030A451 ML13030A456 ML13030A462 ML13030A467
References
RAS 24059, 50-361-CAL, 50-362-CAL, ASBLP 13-924-01-CAL-BD01
Download: ML13030A451 (93)

Text

SCE ATTACHMENT 1 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD

)

In the Matter of )

) Docket Nos. 50-361-CAL & 50-362-CAL SOUTHERN CALIFORNIA EDISON COMPANY )

)

(San Onofre Nuclear Generating Station, ) January 30, 2013 Units 2 and 3) )

)

AFFIDAVIT OF RICHARD BRABEC I. PERSONAL QUALIFICATIONS

1. My name is Richard Brabec. I am employed by Southern California Edison Company (SCE). I have been the Project Manager for the San Onofre Nuclear Generating Station (SONGS) Steam Generator Recovery Project since February 2012. In that role, I have been responsible for managing the identification of inspections to be performed on the SONGS steam generators during the current outage, assessment of the results of the inspections (including the development of operational assessments (OAs)), and development of compensatory and corrective actions for restart of SONGS Unit 2 (including designation of tubes to be plugged).

2. My resume is attached to this Affidavit. In summary, I have over 20 years experience in a variety of roles in the commercial nuclear power generation industry including exercising Control Room Command Function of a dual unit nuclear power facility as a U.S. Nuclear Regulatory Commission (NRC) licensed Senior Reactor Operator (SRO). I also filled management roles in Nuclear Chemistry, major projects such as New Plant Construction, Steam Generator and Reactor Vessel Head Replacements, and Outage and on-line Work Management.

As an SRO certified training specialist, I trained licensed and non-licensed plant operators. My DB1/ 72617040.3 1

commercial nuclear power generation experience was preceded by over 10 years in the U.S.

Nuclear Navy. My naval career included several years as an instructor training and qualifying reactor operators (ROs), management of a submarine Reactor Controls Division as Division Chief, and management of a Nuclear Repair Department aboard a submarine repair facility. I received a Bachelor of Science degree (magna cum laude) in Nuclear Engineering Technology from the University of North Texas in 1998.

II. PURPOSE OF THE AFFIDAVIT

3. The purpose of this Affidavit is to provide support for Southern California Edison Companys Brief on Issues Referred by the Commission (SCEs Brief). In particular, my Affidavit attests to the accuracy of the factual statements in SCEs Brief regarding the SONGS steam generators that are supported by the information provided below.

III. SCES RESTART REPORT FOR SONGS UNIT 2

4. On October 3, 2012, SCE submitted its Restart Report for SONGS Unit 2 to the NRC.

The Restart Report describes the results of extensive inspections and analyses SCE performed on the steam generators, including various operational assessments for Unit 2. Specifically, the Restart Report consists of non-proprietary versions of the following:

• A letter responding to NRCs Confirmatory Action Letter (CAL) of March 27, 2012 regarding the SONGS steam generators. (Letter from P. Dietrich, SCE, to E. Collins, NRC, Confirmatory Action Letter - Actions to Address Steam Generator Tube Degradation (Oct. 3, 2012)).

• Enclosure 1 - - List of Commitments

• Enclosure 2 - - Unit 2 Return to Service Report o Attachment 1 - - SONGS Unit 2 Relevant Technical Specifications o Attachment 2 - - AREVA Document 51-9182368-003, SONGS 2C17 Steam Generator Condition Monitoring Report DB1/ 72617040.3 2

o Attachment 3 - - AREVA Document 51-9180143-001, SONGS Unit 3 February 2012 Leaker Outage - Steam Generator Condition Monitoring Report o Attachment 4 - - MHI Document L5-04GA564, Tube Wear of Unit-3 RSG -

Technical Evaluation Report o Attachment 5 - - MHI Document L5-04GA571, Screening Criteria for Susceptibility to In-Plane Tube Motion o Attachment 6 - - SCE - - SONGS U2C17 Steam Generator Operational Assessment Appendix A - - SONGS U2C17 Outage - Steam Generator Operational Assessment (AREVA non-TTW OA)

Appendix B - - SONGS U2C17 Steam Generator Operational Assessment for Tube-to-Tube Wear (AREVA OA)

Appendix C - - Operational Assessment for SONGS Unit 2 Steam Generators for Upper Bundle Tube-to-Tube Wear Degradation at End of Cycle 16 (Intertek OA)

Appendix D - - Operational Assessment of Wear Indications in the U-bend Region of San Onofre Nuclear Generating Station Unit 2 Replacement Steam Generators (Westinghouse OA)

Proprietary versions of the above documents were submitted to the NRC by SCE via a letter dated November 28, 2012.

5. The NRC Staff has issued a number of requests for additional information (RAIs) to SCE related to the Restart Report. SCE has submitted responses to many of these RAIs throughout January 2013 (RAI Responses).

6. The Restart Report and the RAI Responses were prepared under my supervision and control. They are complete and accurate to the best of my knowledge, information, and belief.

IV. DESCRIPTION OF THE STEAM GENERATORS

7. SONGS is located near San Clemente, California. SONGS Unit 1 ceased operation in 1992 and has since been decommissioned. SONGS Units 2 and 3 are pressurized water reactors DB1/ 72617040.3 3

using a Combustion Engineering (CE) design with two steam generators per unit. SCE is the operator of SONGS Units 2 and 3.

8. SCE put replacement steam generators (RSGs) into service at SONGS Units 2 and 3 in 2010 and 2011, respectively. The RSGs were intended to resolve corrosion and other degradation issues present in the original steam generators for the units. The RSGs were designed and manufactured by Mitsubishi Heavy Industries (MHI). The RSGs are vertical cylindrical U-tube heat exchangers, in which primary coolant is circulated inside the tubes and heat from the primary-side coolant is transferred to the secondary-side coolant outside of the tubes. The heat transfer converts the secondary-side coolant into saturated steam that is used to drive a turbine-generator to create electricity.

9. Each RSG has 9727 tubes, organized in 142 rows and 177 columns, in a triangular pitch configuration. The term pitch means the distance between the centers of two adjacent tubes.

Tube configuration refers to the arrangement of the tubes when viewed along the axis of the tubes. Thus, in a triangular pitch configuration, each tube is equidistant from its adjacent tubes, forming an equilateral triangle arrangement. In contrast, in a square configuration, the tubes in columns and rows are fully aligned, forming a square pattern.

10. In the straight-leg portion of the tubes, the tubes are supported by a series of tube support plates (TSPs) through which the tubes penetrate. The RSGs do not have a stay cylinder. A stay cylinder is a central structural column in the channel head of some steam generators.

11. The U-bend region is located at the top of the tube bundle and is supported by an anti-vibration bar (AVB) structure consisting of six sets of V-shaped AVBs between each adjacent set of columns of tubes. Each AVB is welded to a retaining bar, which in turn is attached to a DB1/ 72617040.3 4

bridge structure over the top of the U-bend region. The retaining bars with AVBs attached are also supported by 24 retainer bars that are just inside the tube bundle.

12. The configuration of the RSGs is depicted in Figure 1 below.

Figure 1 - Configuration of Replacement Steam Generators TSP AVBs U-Bend Region Retainer Bridge Retaining Bar Bar Straight Leg Region Tubes DB1/ 72617040.3 5

13. On June 27, 2008, SCE requested a license amendment for certain issues related to the SONGS Units 2 and 3 steam generator replacement (e.g., changes to certain SONGS Technical Specifications related to steam generator tube integrity). The NRC approved the license amendment on June 25, 2009.

V. COMPARISON OF SONGS RSGS WITH STEAM GENERATORS FOR OTHER COMBUSTION ENGINEERING PLANTS

14. The SONGS RSGs are designed for a higher power level than the steam generators that are the subject of Figures 4-3 and 5-1 of the AREVA OA (which the Licensing Boards order of December 7, 2012 refers to as the Tube-to-Tube Report). As a result, the RSGs have more tubes than the steam generators of those plants. By limiting operation of SONGS Unit 2 to 70%

power, the power level of the SONGS RSGs will be less than the power of one of the other plants and somewhat higher than the remainder, as shown in the following table.

15. Other than the difference in design power level, there is not a significant difference in the design of SONGS RSGs and the steam generators of the other plants in terms of factors such as tube material, tube size, tube pitch, tube configuration, TSPs, and AVBs. The following table provides information comparing the design of the SONGS RSGs with the design of the steam generators for the comparison plants discussed in the AREVA OA:

Plant SONGS A B C D E Tube material 690TT 690TT 690TT 690TT 690TT 690TT Tube outside diameter (inches) 0.750 0.750 0.750 0.750 0.750 0.750 Tube wall thickness (inches) 0.043 0.043 0.043 0.043 0.043 0.043 Tube pitch (inches) 1.000 1.000 1.043 1.031 1.080 1.080 DB1/ 72617040.3 6

Tube configuration Triangular Triangular Square Triangular Triangular Triangular Tube support plates Broached Broached Broached Broached Broached Broached Max AVBs per Tube 12 8 10 6 6 8 Stay Cylinder (Yes/No) No No No No No No 1210 at 70%

MWt per SG power 1355 825 895 868 1000

16. As this table demonstrates, the design of the tubes and tube supports for the SONGS RSGs is very similar to the design of the steam generators of the comparison plants. For example, the table shows plants with similar broached tube supports, tube pitch, and no stay cylinder. In the case of Plant A, the original steam generators had stay cylinders, but the replacement steam generators did not. Additionally, several CE plants besides SONGS have installed replacement steam generators without stay cylinders. The most significant difference is the design of the AVBs. As indicated in the table, the SONGS RSGs have more AVBs per tube than the steam generators in the comparison plants.

17. The purpose of the comparison is to show that the thermal-hydraulic conditions that can lead to fluid elastic instability (FEI) will not exist in the RSGs at 70% power, and that the comparison shows that the thermal hydraulic conditions in the RSGs at 70% power will be similar to the conditions in other plants that have not experienced FEI.

18. All other things being equal, having more AVBs per tube is beneficial in preventing FEI, i.e., a stability ratio of 1.0, which was the cause of the tube leak in SONGS Unit 3 in January 2012. However, in its assessment of the potential for FEI of the RSG tubes at 70% power, the AREVA OA assumed that none of the AVBs is effective. Thus, the number of AVBs was not a factor in that assessment. This is discussed in more detail in Section VIII below. In this regard, DB1/ 72617040.3 7

effective tube supports are defined based upon tube-to-AVB contact forces. Even if an AVB is not effective in providing contact force on a tube, the AVB is still needed to provide out-of-plane tube-to-AVB support and damping contribution to the calculation of in-plane and out-of-plane stability ratios.

VI. DESCRIPTION OF THE STEAM GENERATOR TUBE WEAR

19. On January 31, 2012, SCE identified a leak in a tube in one of the SONGS Unit 3 RSGs.

This leak was well below allowable limits in the Technical Specifications, and presented no hazard to the public health and safety. Pursuant to established procedures, SCE shut down Unit

3. At the time, SONGS Unit 2 already was shutdown and undergoing a refueling outage.

Additionally, SCE identified unexpected tube wear adjacent to small diameter retainer bars in the RSGs for Units 2 and 3 (the retainer bars in the RSGs have two different diameters).

20. On March 23, 2012, SCE sent a letter to the NRC committing to take certain actions to determine and address the causes of the leak and identified instances of tube wear prior to restart of the SONGS units. The NRC memorialized its understanding of the actions planned by SCE in the March 27 CAL, which confirmed the commitments made by SCE for actions to be taken prior to restarting either unit.

21. SCE assembled a team of experts to address the tube leak and its causes, including experts in thermal-hydraulics and steam generator design, manufacture, operation, and maintenance. The team included personnel from Westinghouse, AREVA, and Intertek, with support from MHI. The team performed extensive investigations into the causes of the tube leak and the unexpected wear of tubes adjacent to the retainer bars, and developed compensatory and corrective actions.

22. The investigations included root cause evaluations to determine the mechanistic causes of the tube degradation and failure. The root cause evaluations included analysis of a number of DB1/ 72617040.3 8

factors to determine whether the factors might have caused the unexpected tube wear. Based upon those analyses, SCE eliminated a number of factors as potential causes. For example, the root cause evaluations assessed issues related to the tubesheet and determined that those issues had no bearing on the tube wear experienced by SONGS. The thickness of the tubesheet has no bearing on the tube wear experienced by SONGS.

23. SCE has determined that the Unit 3 RSG tube leak was the result of tube-to-tube wear (TTW) that was caused by in-plane FEI from the combination of localized high steam velocity, high steam void fraction, and insufficient contact forces between the tubes and the AVBs. FEI is a phenomenon in which the tubes vibrate with increasingly larger amplitudes due to the flow velocity exceeding the critical velocity for a tube, given its supporting conditions and thermal-hydraulic environment. FEI occurs when the amount of energy imparted on the tube by the fluid is greater than the amount of energy that the tube can dissipate back to the fluid and to the supports. During in-plane FEI, tubes within the same column are excited by the fluid and move within the plane of the column, resulting in tube-to-tube contact and wear of the tubes.

24. SCE also determined the cause of the retainer bar vibrationturbulent flow caused the relatively long and narrow retainer bars to vibrate and contact the tubes.

25. In Unit 3, more than 150 tubes of the 9727 tubes in each RSG experienced TTW, including more than 100 tubes in each RSG with wear equal to or greater than 35% of the width of the tube wall (which is the criterion in SONGS Technical Specification 5.5.2.11 for removal of the tube from service by plugging of the tube).

26. In contrast to the extensive TTW in Unit 3, TTW existed in only a single pair of tubes (one contact location) in one of the two Unit 2 RSGs. The TTW in Unit 2 was so minor that it was not detectable using normal eddy current testing methods. Westinghouse has concluded that DB1/ 72617040.3 9

the TTW in Unit 2 was not due to FEI, but instead to proximity of the tubes in question and random vibration of those tubes. However, some of SCEs analyses (such as the Intertek OA and the AREVA OA) conservatively assumed that FEI could occur in Unit 2 at 100% power.

27. TTW due to in-plane FEI had not been previously experienced in U-tube steam generators. The difference in TTW between Units 2 and 3 is attributed to manufacturing differences in the RSGs of Units 2 and 3. A claim that Unit 2 poses the same risk as Unit 3 is manifestly incorrect, given the substantially lower TTW in Unit 2 than Unit 3 and the manufacturing differences between the units.

28. In addition to the TTW, SCE also identified other types of tube wear in Units 2 and 3, such as wear caused by vibration of tubes against AVBs and TSPs. Those types of tube wear had been previously identified in the steam generators of other plants and were expected for the RSGs (although the extent was higher than the industry average).

VII. DESCRIPTION OF THE COMPENSATORY AND CORRECTIVE ACTIONS FOR THE STEAM GENERATORS

29. SCE has taken a number of compensatory and corrective actions to prevent loss of integrity due to TTW in the Unit 2 RSGs. Compensatory actions are defined as temporary actions pending implementation of final corrective actions.

30. Tube Plugging - - As discussed in Section 8.2 of the Return to Service Report for Unit 2, SCE has plugged each end of those tubes that had wear exceeding 35% of the tube wall (which is a criterion for tube plugging in the SONGS Technical Specifications), that are adjacent to the retainer bars, or that are in approximately the same region that experienced TTW in Unit 3.

About 3% of the total number of tubes in each of the RSGs in Unit 2 have been plugged (which is far less than the approximate 8% of the tubes assumed to be plugged in the accident analyses in the SONGS Updated Final Safety Analysis Report (UFSAR)). The purpose of the tube DB1/ 72617040.3 10

plugging is to remove tubes from service that have been or may be susceptible to TTW or wear from vibration of the retainer bars. By removing such tubes from service, the plugged tubes would not be a source for release of primary coolant to the secondary system even if the tubes were to develop a through-wall leak or to burst. Tube plugging is not intended to address the root cause of the vibration.

31. Stabilization of Plugged Tubes - - Stabilizing involves the insertion of cables into selected plugged tubes, which provides additional reinforcement against full severance of a tube.

Because only plugged tubes are stabilized and because stabilizers provide additional protection against tube severance, stabilizers do not adversely affect any design function.

32. Administrative Limits on Power - - As a temporary compensatory measure pending a mid-cycle outage for a steam generator tube inspection, the power level of Unit 2 will be administratively limited to 70% of the maximum power specified in the license and Technical Specifications. Reducing power to 70% eliminates the type of thermal-hydraulic conditions that caused FEI and the associated TTW in SONGS Unit 3. This administrative limit is temporary and may change based upon the results of inspections, further analyses and long-term corrective actions.

33. Mid-Cycle Outage for Inspection of Steam Generator Tubes - - SCE will shut down Unit 2 for a mid-cycle steam generator tube inspection outage within 150 cumulative days of operation at or above 15% power. The purpose of the inspections is to confirm the effectiveness of the corrective and compensatory actions taken to address TTW in the Unit 2 RSGs. SONGS Technical Specification 5.5.2.11 requires periodic steam generator tube inspections. The specific nature of the inspections is not specified in the Technical Specifications. There is nothing inconsistent with the Technical Specifications or the UFSAR with conducting more frequent DB1/ 72617040.3 11

inspections. Furthermore, conducting more frequent inspections does not adversely affect any design function. Section 8.3 of the Return to Service Report describes the steam generator inspections that SCE will be conducting for Unit 2 during the mid-cycle outage. In summary, SCE will be performing the following inspections of the steam generator tubes:

• Eddy Current Bobbin Coil Examinations (ECT) of the full length of all in-service tubes.

• Rotating Coil Examinations of the following areas of the tubes:

o Tubes in the U-bend region (the inspection scope will repeat the pattern used during the current outage (approximately 1300 tubes per steam generator will be inspected)); and o Tube areas with current wear indications of greater than 20% due to contact with the TSPs or AVBs.

• Visual inspections will be performed on all installed tube plugs.

• 12 tubes in each steam generator will be unplugged and the stabilizer(s) removed, and ECT will be performed to assess the effectiveness of the TTW compensatory and corrective actions. Following these ECT inspections, all tubes will be re-plugged and stabilizers installed. The tubes will be selected as follows:

o The 2 tubes with previous TTW indications; o 5 tubes adjacent to tubes with TTW wear; and o 5 tubes selected from representative locations that were preventively plugged as part of the compensatory and corrective actions for TTW.

• Visual inspection of small diameter retainer bars and welds.

DB1/ 72617040.3 12

As discussed in Sections 6.1 and 7.1 of the Return to Service Report, these are the same types of inspections that have been conducted during the current outages of SONGS Units 2 and 3. These types of inspections are also commonly used in the industry, as indicated by EPRI 1019038, Steam Generator Management Program: Steam Generator Integrity Assessment Guidelines, Revision 3 (Oct. 2008).

34. Defense in Depth Actions - - In addition to the actions described above, Section 9 of the SONGS Return to Service Report identifies defense-in-depth actions for Unit 2. These actions consist of injection of Argon into the Reactor Coolant System (RCS) to facilitate detection of steam generator tube leakage into the secondary system, installation of a Nitrogen (N-16) radiation detection system on the Main Steam Lines also to facilitate detection of tube leakage, reduction of the administrative limit for RCS activity level to reduce impacts of tube leakage or a tube burst, and enhanced operator response to early indication of tube leakage. The only procedural revision involves a reduction in the level of leakage that triggers a shutdown. Such a reduction is a conservative change, and therefore does not adversely affect any safety function.

Those actions provide additional protection in the event of a tube leak or tube burst, do not constitute actions to prevent loss of integrity due to the causes of tube wear in the Unit 2 steam generator tubes, and do not adversely affect any safety function.

35. The Tube-to-Tube Report (pages 113-114) considers additional defense-in-depth factors by evaluating TTW if FEI were to occur in Unit 2 during operation at 70% power. As stated therein, the measures go beyond the technical case for restart of Unit 2 at 70% power, and add to the assurance that structural and leakage integrity requirements will be maintained throughout the next inspection interval. The Tube-to-Tube Report relies on operation at 70% to limit FEI, but the defense-in-depth measures evaluate TTW assuming FEI occurs. The Tube-to-Tube DB1/ 72617040.3 13

Report estimates between 2.5 months and 11 months for the tube wear. This estimate is for worst case wear. The Tube-to-Tube Report estimates 3.5 months for instability zone expansion (i.e., time for an in-service pressurized tube to be driven to instability with consequent development of TTW). This 3.5 months estimate is conservative because it is half of the 7 months estimated for instability zone expansion in Unit 3. When the instability zone expansion time is added to the tube wear time, then the value is less than the proposed operation period of 150 days.

36. Additional Actions - - Section 11 of the SONGS Return to Service Report describes additional actions that SCE will be taking for Unit 2, including improving the sensitivity of vibration monitoring equipment and using an analytic tool that will aid in diagnosis of equipment conditions. Neither of these actions adversely affects any design function. Moreover, the additional actions are intended to help in analysis of historical data, and do not constitute actions to prevent loss of integrity due to the causes of tube wear in the Unit 2 steam generator tubes.

37. Long-Term Corrective Actions - - SCE has not yet identified long-term corrective actions for the steam generator tubes for Unit 2. Any long-term corrective actions for Unit 2 will be subject to a separate evaluation pursuant to 10 C.F.R. § 50.59 to determine whether such actions require a license amendment.

38. Corrective Actions for Unit 3 - - SCE has not yet submitted any response to the CAL with respect to SONGS Unit 3. SCE has not yet identified the long-term corrective actions for Unit 3.

39. The CAL states that NRC must provide its approval before SCE can restart either SONGS unit. The NRC is currently reviewing the Restart Report for SONGS Unit 2.

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VIII. LICENSING BOARD QUESTION REGARDING 0.95 PROBABILITY AND 50%

CONFIDENCE LEVELS FOR STABILITY RATIOS

40. Stability ratio is defined as the ratio of the effective flow velocity to which the tube is subjected to the critical velocity. Critical velocity is the velocity at which the tube, with specific geometry and support conditions, becomes unstable. Stability ratios less than 1.0 represent a stable condition where the actual velocity is less than the critical velocity; stability ratios greater than 1.0 represent an unstable condition where the actual velocity exceeds the critical velocity.

Thus, if the stability ratio is less than 1.0, the type of TTW experienced in the RSGs of SONGS Unit 3 will not exist.

41. The flow velocity varies throughout the steam generators, based upon local conditions such as void fraction. As a result, the stability ratio varies from tube to tube in the steam generators. The criterion of 0.95 probability at 50% confidence is applied to ensure that no tube in the steam generators exceeds the target stability ratio; i.e., this criterion is applied to the worst case tube. The 50% confidence level simply means that a point estimate for probability is used.

For example, for a simulation involving 10,000 trials, the 0.95/50 acceptance criterion simply requires there be at least 9500 successful trials out of 10,000. If the tube with the highest stability ratio does not exceed the target stability ratio, the other tubes will have a stability ratio that is equal to or less than the highest stability ratio, and those tubes will also satisfy the target (with added margin for the most part).

42. The evaluation of the tubes against the criterion of a maximum stability ratio of 0.75 with 0.95 probability at 50% confidence level was performed to determine the margin to instability, not to provide reasonable assurance. In any event, the criterion that steam generator tubes maintain structural integrity with a 0.95 probability at 50% confidence level is based upon DB1/ 72617040.3 15

provisions in guidelines (EPRI 1012987) developed by the Electric Power Research Institute (EPRI), entitled Steam Generator Integrity Assessment Guidelines, Rev. 2 (July 2006) at 8-1:

The fundamental objective of an OA is to ensure that structural and leakage performance criteria will be met over the length of the upcoming inspection interval. It shall be demonstrated that the degradation detection sensitivity and/or NDE [non-destructive examination] sizing uncertainty combined with degradation growth rates leads to the expectation that structural and leakage integrity criteria will be met at the end of the next inspection interval. In terms of structural integrity, the fundamental OA requirement is that the projected worst case degraded tube for each existing degradation mechanism shall meet the limiting structural performance parameter with a 0.95 probability at 50% confidence.

This provision is also included in the latest version of the guidelines (EPRI 1019038). This EPRI Guideline is referenced in NEI 97-06, which in turn is referenced in SONGS Technical Specification Bases B3.4.17. Therefore, the 0.95 probability and 50% confidence criterion is already part of the SONGS licensing basis and is an industry-accepted measure for steam generator tube integrity. Thus, the criterion is an acceptable measure of reasonable assurance.

43. The AREVA OA has applied that criterion in an extremely conservative manner. Some of the conservatisms include the following:

• The AREVA OA applies that criterion to a stability ratio of 0.75, which is significantly less than 1.0 and therefore is significantly below the level at which FEI would occur. As the AREVA OA indicates, the calculation performed at a stability ratio of 0.75 was performed to demonstrate margin, not to determine reasonable assurance.

• In demonstrating this criterion is met at a stability ratio of 0.75, it was assumed that only 4 of the 12 AVBs were effective. In contrast, wear indications in Unit 2 indicate that at least 7 AVBs per tube were effective.

DB1/ 72617040.3 16

• The same section of the AREVA OA demonstrates that the stability ratio is less than 1, assuming that all of the 12 AVBs are ineffective.

44. In summary, the AREVA OA shows that FEI and TTW will not occur (i.e., the stability ratios are less than 1.0) at 70% power, even assuming that all of the AVBs are ineffective.

Furthermore, the AREVA OA demonstrates that there is a 0.95 probability at 50% confidence that the stability ratio will not exceed 0.75, which is well below the level that results in FEI.

That result indicates that there will be substantial margin to the onset of TTW in Unit 2. Thus, there is reasonable assurance that TTW will not occur at 70% power.

IX. BASIS FOR SCES REASONABLE ASSURANCE DETERMINATION

45. There are a number of bases for SCEs conclusions that FEI will not occur in Unit 2 at 70% power and that there is reasonable assurance that Unit 2 can be safely operated at 70%

power. The bases are discussed below.

46. Comparison with other Plants - - FEI is attributable to high flow velocities, high void fractions, and insufficient contacts forces between the tubes and the AVBs. Figure 5-1 in the AREVA OA compares bulk velocity ratio and mean void fraction ratio for SONGS at 70%

power and several other plants. For each of these parameters, SONGS operating at 70% power is bounded by at least one other plant, thereby demonstrating that FEI should not occur at SONGS because it is not occurring at the other plants. Furthermore, with respect to the mean void fraction ratio, SONGS at 70% power is the lowest of the six plants.

47. Low Void Fractions - - Void fraction is a critical factor in the determination of whether FEI will exist. Void fraction is equal to the volume of steam in the fluid divided by the total volume of the steam/water mixture. A high void fraction indicates less water in the fluid, which in turn results in lower damping and thus less resistance to FEI. As shown on Figure 8-2 of the Return to Service Report, all of the tubes with TTW occurred in a region where the void fraction DB1/ 72617040.3 17

exceeded 0.993. In contrast, at 70% power, the maximum void fraction in the SONGS Unit 2 RSGs will be 0.9258, as discussed in detail in Section 8.1 of the Return to Service Report. Thus, by limiting power to 70%, SCE will be reducing the void fractions to levels well below those associated with TTW in the SONGS RSGs.

48. Differences in TTW between Units 2 and 3 - - As discussed above, the RSGs for SONGS Unit 3 experienced extensive TTW. More than 300 tubes in the two RSGs in Unit 3 experienced TTW, including TTW resulting in a leaking tube, after only 11 months of operation. In contrast, the Unit 2 RSGs had 21 months of operation and only one instance of TTW (involving two tubes), and that TTW was so minor that it was not detected using normal eddy current testing techniques. This difference in operating experience indicates that the experience of the Unit 3 RSGs is not applicable to Unit 2. Furthermore, since Unit 2 experienced only one instance of TTW in 21 months of full power operation, operation at 70% power for 150 days will incur even less likelihood of any TTW.

49. Manufacturing Differences between Units 2 and 3 - - The design and operation of the RSGs for both units was essentially identical. The difference in TTW operating experience between Units 2 and 3 is attributable to manufacturing differences between Units 2 and 3. For the Unit 2 RSGs, there was more of a dispersion in the placement of the AVBs relative to the tubes, resulting in some AVBs with relatively high contact forces against the tubes (which helps prevent tube vibration) and other AVBs with little or no contact force. Since a tube only needs a few AVBs with relatively high contact forces to prevent FEI, the greater dispersion of the AVBs for the Unit 2 RSGs in retrospect was beneficial in preventing FEI and TTW. Therefore, Unit 2 is significantly different than Unit 3 with respect to manufacturing differences that result is substantially higher AVB contact forces with the tubes in Unit 2.

DB1/ 72617040.3 18

50. Absence of FEI in Unit 2 - - Westinghouse has performed an Operational Assessment for Unit 2, which demonstrates that there was no FEI at 100% power in Unit 2 and that the TTW in Unit 2 was not caused by FEI but instead due to the close proximity of the two tubes and random vibration.

51. Traditional Operational Assessment - - In a traditional Operational Assessment for a steam generator, the historic tube wear is extrapolated for the next operating period to verify that the tubes, at the end of that period, will satisfy the performance criteria. For Unit 2, one of SCEs contractors, Intertek, has performed a traditional industry Operational Assessment, based upon the TTW wear rates experienced in Unit 3 (i.e., it assumes that FEI and the TTW experience is applicable to Unit 2, even though the information discussed above demonstrates that it is not). This Operational Assessment demonstrates that there is reasonable assurance that the performance criteria will be satisfied for 16 months at 70% power level.

52. Deterministic Evaluation of the Potential for FEI - - One of SCEs contractors, AREVA, has performed a deterministic evaluation of the potential for FEI in Unit 2 operating at 70%

power, assuming that none of the AVBs is effective. That evaluation demonstrates that FEI will not occur for the duration of this assessment (18 months).

53. Plugging of Tubes in the Affected Region - - The local thermal-hydraulic conditions in the secondary side of steam generators vary greatly, depending upon the location within the tube bundle. The TTW in Unit 3 occurred in the region of the tube bundle with the highest flow velocity and void fractions. As a preventive measure, SCE has plugged the tubes in Unit 2 in that same region. Therefore, the tubes in Unit 2 that might potentially be subject to TTW based upon the experience in Unit 3 have been plugged and removed from service.

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54. As discussed in the October 3 cover letter for SCEs CAL response, there is reasonable assurance that Unit 2 can operate safely at 70% power for 150 days, based upon the following:

SCE has evaluated the causes of TTW in the Unit 3 SGs and . . .

has completed corrective and compensatory actions in Unit 2 to prevent loss of tube integrity due to these causes. Tubes within regions of the Unit 2 SGs that might be susceptible to FEI have been plugged. In addition, . . . SCE has established operational limits that eliminate the thermal-hydraulic conditions associated with FEI from the SONGS Unit 2 SGs. Specifically, operation of Unit 2 will be administratively limited to 70% power. Within 150 cumulative days of operation at or above 15% power, Unit 2 will be shut down for inspection to confirm the condition of the SG tubes. The analyses and OAs performed by SCE and independent industry experts demonstrate that under these conditions, tube integrity will be maintained. On this basis, SCE concludes that Unit 2 will operate safely.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on January 30, 2013.

Executed in Accord with 10 C.F.R. § 2.304(d)

/s/ Richard Brabec Richard Brabec SCE Project Manager for Steam Generator Recovery Project Southern California Edison Company P.O. Box 128 San Clemente, CA 92672 Phone: (949) 368-7418 E-mail: john.brabec@sce.com DB1/ 72617040.3 20

Resume of Richard C. (John) Brabec Experience:

San Onofre Nuclear Generating Station (SONGS) Manager, Unit 2 and 3 Steam Generator Recovery Project (2/2012 to present)

• Manage Southern California Edison (SCE) and vendor engineering and technical staff in analysis of unexpected Steam Generator (SG) tube wear.

• Manage field work implementation of corrective actions to prevent recurrence.

• Managed development of multiple Operational Assessments as part of the Stations Return to Service Plan.

SONGS Refueling Outage Director/ Outage Optimization Project Manager (5/2011-2/2012)

• Managed SONGS key strategic business initiative to dramatically improve SONGS Refueling Outage planning and execution.

• Outage Director during the spring 2012 Unit 2 Refueling Outage.

SONGS Units 2 and 3 Steam Generator Replacement (SGR) Implementation Manager (8/2009-2/2011)

• Provided Management of SCEs SGR Team. Completed planning, preparation and execution of two successful Steam Generator Replacement Outages (SGROs) over an 18 month period.

Comanche Peak Nuclear Power Plant (CPNPP) Chemistry Department Manager (1/2009-8/2009)

• Managed department of Staff Chemists, Technicians and Operators in the operation of the dual unit nuclear power generation facility.

CPNPP New Plant Construction Project Operations and Training Manager (6/2007 to 1/2009)

• Developed and initiated execution of the projects global integrated project plan. This included planning and scheduling of all plant staff training, plant construction, pre-operational testing, fuel load, and startup testing through full power operation. Project scope included the licensing, construction and startup planning of two 1700 MW nuclear generation facilities.

CPNPP Steam Generator and Reactor Vessel Head Replacement Project Implementation Manager (5/2004 to 6/2007)

• Management responsibility for all project construction oversight, scheduling, containment building management, operations interface, startup program development and deployment, and outage planning and execution for this major capital project.

• Project Outage Director during execution phases (three outages). Lead the execution of a world record setting 55 day steam generator and reactor vessel head replacement in spring 2007.

DB1/ 72617040.3

CPNPP Operations Department Nuclear Steam Supply System Outage Manager (1/2003-5/2004)

• Managed the planning, scheduling and execution of all refueling outage activities relative to the plants Nuclear Steam Supply System and its peripheral support systems.

CPNPP Control Room Senior Reactor Operator (SRO) (6/2000-1/2003)

• As a Licensed SRO, managed day-to-day station operations, maintenance and testing activities of the dual-unit nuclear power generation facility.

• Qualified Shift Technical Advisor (STA) and Shift Manager.

CPNPP Operator Training (11/1997-6/2000)

• Provided Training for the Plant Operations Department Staff (licensed and non-licensed plant operators) in the classroom and Control Room Simulator.

• Operations Department Supervisor during refueling outages CPNPP Operations Department (11/1992-11/1997)

• Plant Field Operations

• On-line and Outage work planning and execution UNITED STATES NAVY (5/1981-10/1992)

• Nuclear Submarine Repair Department Manager- Managed a multi-disciplined maintenance team in the planning and execution of nuclear plant outages aboard attack submarines.

• Reactor Controls Division Chief- Managed a division of staff Reactor Operators (ROs) responsible for the operation of the nuclear propulsion plant and performance of all Instrumentation and Calibration maintenance activities. Qualified RO, SRO and Engineering Watch Supervisor.

• Nuclear Prototype Training Instructor- Trained and qualified RO candidates in the operation of the nuclear propulsion plant. Qualified RO, SRO and Engineering Watch Supervisor.

Education:

Bachelor of Science, Nuclear Engineering Technology (magna cum-laude), University of North Texas (1998)

US Naval Nuclear Power School (1983)

Licenses/Professional Certifications:

USNRC Senior Reactor Operator License (2000-2008)

Project Management Professional (PMP) Certification (2006)

DB1/ 72617040.3

SCE ATTACHMENT 2 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD

)

In the Matter of )

) Docket Nos. 50-361-CAL & 50-362-CAL SOUTHERN CALIFORNIA EDISON COMPANY )

)

(San Onofre Nuclear Generating Station, ) January 30, 2013 Units 2 and 3) )

)

AFFIDAVIT OF VICKRAM NAZARETH I. PERSONAL QUALIFICATIONS

1. My name is Vickram Nazareth. I am employed by Southern California Edison Company (SCE) as a Manager in the Nuclear Fuels group at the San Onofre Nuclear Generating Station (SONGS). My responsibilities include maintaining the accident analyses in the Updated Final Safety Analysis Report (UFSAR) for SONGS Units 2 and 3.

2. My resume is attached to this Affidavit. In summary, I obtained a Bachelor of Science degree in Nuclear Engineering from the University of Florida in 1978. I also have obtained a Management Practice for Engineering and Technical Professionals Certificate from the University of California at Irvine in 2001. I have worked for SCE at SONGS since 1995, and have been responsible for various technical areas related to core physics, spent fuel criticality, fuel performance monitoring and analysis, fuel mechanical design, fuel manufacturing oversight, accident radiation doses and containment, and core thermal-hydraulic and UFSAR Chapter 15 transients. Prior to my employment with SCE, I held engineering positions with SUN Technical Services and Combustion Engineering, Inc.

DB1/ 72620338.1 1

II. PURPOSE OF THE AFFIDAVIT

3. The purpose of this Affidavit is to provide support for Southern California Edison Companys Brief on Issues Referred by the Commission (SCEs Brief). In particular, my Affidavit attests to the accuracy of the factual statements in SCEs Brief as they pertain to the UFSAR analysis of a steam generator tube rupture (SGTR).

III. STEAM GENERATOR TUBE RUPTURE

4. The SONGS UFSAR analyzes a SGTR event. Specifically, the SONGS UFSAR provides the accident analysis for a SGTR in Section 15.10.6.3.2, utilizing the dose models described in Appendix 15.10B. The SGTR event assumes the break of one steam generator tube, which is standard practice per the Nuclear Regulatory Commission Standard Review Plan Section 15.6.3.

5. For purposes of the UFSAR, the integrated flow through a broken tube was calculated to be 70,563 lbm (0 to 30 minutes). UFSAR Table 15.10.6.3.2-3. This rate is conservative in that it assumes a 45% double ended guillotine break. UFSAR Table 15.10.6.3.2-1. This limiting break size was determined by performing a parametric study for break size up to and including a 100% double ended guillotine break. The 45% double ended guillotine break resulted in prolonging the period in which the Main Steam Safety Valves (MSSVs) remain open; thereby maximizing radioactivity releases to the atmosphere and thus the dose consequences. UFSAR Section 15.6.3.2.3.B.

6. The radiological consequence analysis for the SGTR event includes the following additional conservatisms:

• The event is assumed to occur when the primary coolant is at the maximum activity concentration limits identified in Technical Specification 3.4.16, including a case with a pre-existing iodine spike, and a case with an accident induced iodine spike of 500 times DB1/ 72620338.1 2

greater than the primary coolant system iodine release rate corresponding to the Technical Specifications limit. UFSAR Table 15.10.6.3.2-3.

• The event is assumed to initiate with the secondary coolant activity concentration at the maximum limits specified in Technical Specification 3.7.19. UFSAR Table 15.10.6.3.2-3.

• The event assumes very adverse (95th percentile) atmospheric dispersion factors.

UFSAR Table 15B-4.

• The analysis conservatively assumes that operator action to isolate the affected steam generator is delayed until 30 minutes after initiation of the event. In reality, operators should recognize fairly quickly that a SGTR is in progress and isolate the affected steam generator. The break would be detected almost immediately due to the increase in radiation in the secondary system (similar to what happened during the steam generator tube leak at SONGS Unit 3 on January 31, 2012).

• The event assumes a loss of normal AC power at the time of reactor trip. This results in the loss of condenser vacuum and lifting of the MSSVs to relieve secondary pressure, increasing the release to the atmosphere. UFSAR Section 15.10.6.3.2.

• The MSSVs are assumed to open at the lowest possible pressure setpoint (including uncertainties), increasing the release to the atmosphere. UFSAR Table 15.10.6.3.2-1.

DB1/ 72620338.1 3

I declare under penalty of perjury that the foregoing is true and correct.

Executed on January 30, 2013.

Executed in Accord with 10 C.F.R. § 2.304(d)

/s/ Vickram F. Nazareth Vickram F. Nazareth Manager, Nuclear Fuel Management Southern California Edison Company San Onofre Nuclear Generating Station P.O. Box 128 San Clemente, CA 92672 Phone: (949) 368-3071 E-mail: Vick.Nazareth@sce.com DB1/ 72620338.1 4

VICKRAM(Vick) F. NAZARETH__________________

7325 E. Wrangler Circle, Orange, CA 92869 (949) 368-3071, Vick.Nazareth@sce.com EXPERIENCE 9/95-Present Southern California Edison, San Onofre Nuclear Generating Station Nuclear Fuel Management [Division Manager; Supervisor of Nuclear Fuel Analysis, Core Performance Analysis and Nuclear Safety Analysis]

Responsible for Fuel Strategic administration and goal development; fuel inspections and reconstitution; PRA application strategy; short term and long-term fuel budgets and forecasts; AREVA and Westinghouse fuel strategy and licensing; Executive, INPO, NRC and Participant interface and negotiation on Fuel performance and strategies; Fuel contract administration and negotiation; Employment decisions, interviews, hiring and compensation; Staff goal setting, training and development; Technical areas in nuclear fuel including core physics, spent fuel criticality, fuel performance monitoring & analysis, fuel mechanical design, fuel manufacturing oversight, accident radiation dose & containment, COLSS/CPC setpoints, core thermal-hydraulic and UFSAR chapter 15 transients Accomplishments include strategic lead in developing fuel failure mitigation strategies; strategic lead in new fuel contract negotiations with fuel fabricators; developer and negotiator of AREVA fuel assembly contracts; team lead in core physics and fuel cycle analysis; team leader in improving fuel integrity response to INPO AFI; development of high quality reload calculations for Units 2 and 3; reload project manager and mentor for several cycle reloads; organization and technical support of vendor fuel manufacturing audits; technical support for SONGS Replacement Steam Generator issues; lead in successful reload technology transfer from Westinghouse; lead in accident and technical support of NRC submittals and RAIs; development and incisive review of engineering procedures; management and evaluation of technical training of engineers; organization, interview and selection of job candidates; SONGS rep in W-CE Owners Group analysis committee; SONGS representative in Utility fuel performance reliability group; Organization of human performance training and engineering rollout; Management of notification backlog and cause evaluations directed assessment of fuel vendor change performance at Omaha 1/91-9/95 SUN Technical Services (contracted to Southern California Edison)

Design Engineering and Nuclear Fuel Management -- Performed all aspects of plant monitoring and protection system technical and licensing support for SONGS; Performed safety significances for LERs, DCPs and EOIs; Performed plant anomaly investigations and operability evaluations Page 1 of 2

Resume of Vick Nazareth Page 2 of 2 8/78-8/88 Combustion Engineering, Inc. (now Westinghouse Electric) Windsor, CT Nuclear Plant Operations Support - Primary responsibility for designing startup test programs for Arkansas Unit 2, SONGS Units 2 & 3, Waterford Unit 3 and Palo Verde Units 1, 2 & 3; Supervised and designed Fast Power Ascension Program to minimize startup test time by streamlining startup test requirements and procedures and automating startup criteria verification process; Reviewed plant startup and operation procedures; Represented CE as setpoint expert on site during SONGS & APS startups Plant performance and Licensing Analyses - Involved in original testing and thermal hydraulic analyses for SONGS and Palo Verde Units; Performed safety analyses and NSSS design calculations for CE 3410 and System 80 designs; technical lead in core thermal hydraulic and setpoint design for CE Korean plants; key analyst in development of Thermal-Hydraulic and power ascension computer codes Licensing and Training - Represented CE and assisted utilities in defending safety analyses, operating procedures and Technical Specifications before the NRC; Involved in the drafting of FSARs for SONGS and Palo Verde plants; Conducted training program at Arizona for CE digital plant engineers EDUCATION 1999-2001 University of California at Irvine, Management Practice for Engineering and Technical Professionals Certificate 1980-1984 University of Connecticut, Juris Doctris (JD); Performed legal study on Environmental Impact Statements for Nuclear Plants 1976-1978 University of Florida, Bachelor of Science in Nuclear Engineering 1974-1976 Valencia Community College, Orlando, Florida - Associate in Arts PUBLICATIONS V.F. Nazareth, et. al., "Optimized PWR Power Ascension Reload Testing", ANS annual meeting, Dallas, Texas, June 1987 V.F. Nazareth, et. al., "Fast Power Ascension Program for C-E Digital Protection System Plants", Third International Topical Meeting on Nuclear Power Plant Thermal Hydraulics and Operations, Nov. 1988, Seoul, Korea PROFESSIONAL ACTIVITIES Professional Engineer in Nuclear Engineering in the State of California Member of the American Nuclear Society

SCE ATTACHMENT 3 UNITED STATES NUCLEAR REGULATORY COMMISSION REGION I V 1600 EAST LAMAR BLVD ARLINGTON, TEXAS 76011-4511 March 27, 2012 CAL 4-12-001 Mr. Peter Dietrich Senior Vice President and Chief Nuclear Officer Southern California Edison Company San Onofre Nuclear Generating Station P.O. Box 128 San Clemente, CA 92674-0128

SUBJECT:

CONFIRMATORY ACTION LETTER - SAN ONOFRE NUCLEAR GENERATING STATION, UNITS 2 AND 3, COMMITMENTS TO ADDRESS STEAM GENERATOR TUBE DEGRADATION

Dear Mr. Dietrich:

On January 31, 2012, your staff at San Onofre Nuclear Generating Station (SONGS) Unit 3 performed a rapid shutdown because of indications of a steam generator tube leak on the 3E88 steam generator. Following extensive testing of 100 percent of the steam generator tubes in both Unit 3 steam generators, your staff identified unexpected wear caused by steam generator tubes rubbing against each other, as well as against retainer bars. Additional in-situ pressure testing of 129 steam generator tubes was performed for the tubes that exhibited the most wear.

Your staff identified that eight steam generator tubes in the Unit 3 3E88 steam generator had failed the pressure test. Failure of the in-situ pressure test is an indication that, for certain design basis events, such as a main steam line break, these steam generator tubes may not be able to maintain design structural integrity. You are continuing to evaluate these results to develop corrective actions for the Unit 3 steam generators.

SONGS Unit 2 was shutdown at the time of this event for a regularly scheduled refueling outage, and planned testing of 100 percent of the steam generator tubes was already in progress. Testing results on Unit 2 showed unexpected wear at retainer bars similar to the Unit 3 results, but did not show any wear from tubes rubbing against each other. Based on these results, your staff identified 6 tubes requiring plugging, and 186 additional tubes that were plugged as a precautionary measure. Evaluation for additional plugging or other corrective actions is continuing for Unit 2, based on ongoing evaluations of Unit 3 testing results.

P. Dietrich For both Units 2 and 3, this was the first cycle of operation with new replacement steam generators. Unit 2 replaced its steam generators in January 2010, and Unit 3 in January 2011.

Each steam generator has 9,727 steam generator tubes.

On March 23, 2012, you sent NRC a letter describing the actions you were committing to take prior to returning Units 2 and 3 to power operation (Agencywide Documents Access and Management System (ADAMS) Accession Number ML12086A182). In a phone conversation on March 26, 2012, I confirmed with you the commitments as described in your letter. This Confirmatory Action Letter (CAL) confirms that SONGS Unit 2 will not enter Mode 2, and SONGS Unit 3 will not enter Mode 4 (as defined in the technical specifications), until the NRC has completed its review of your actions listed below. The permission to resume power operations will be formally communicated to you in written correspondence.

Actions for Unit 2

1. Southern California Edison Company (SCE) will determine the causes of the tube-to-tube interactions that resulted in steam generator tube wear in Unit 3, and will implement actions to prevent loss of integrity due to these causes in the Unit 2 steam generator tubes. SCE will establish a protocol of inspections and/or operational limits for Unit 2, including plans for a mid-cycle shutdown for further inspections.

2. Prior to entry of Unit 2 into Mode 2, SCE will submit to the NRC in writing the results of your assessment of Unit 2 steam generators, the protocol of inspections and/or operational limits, including schedule dates for a mid-cycle shutdown for further inspections, and the basis for SCEs conclusion that there is reasonable assurance, as required by NRC regulations, that the unit will operate safely.

Actions for Unit 3

3. SCE will complete in-situ pressure testing of tubes with potentially significant wear indications in accordance with the Electric Power Research Institute (EPRI) Steam Generator In-situ Pressure Test Guidelines and will plug tubes in accordance with those guidelines.

4. SCE will plug all tubes with wear indications in excess of your Steam Generator Program Requirements (SGPR) and EPRI guidelines as well as perform preventive plugging or take other corrective actions to address retainer bar-related tube wear in Unit 3.

5. SCE will determine the causes of tube-to-tube interaction and implement actions to prevent recurrence of loss of integrity in the Unit 3 steam generator tubes while operating.

6. SCE will establish a protocol of inspections and/or operational limits for Unit 3, including plans for a mid-cycle shutdown for inspections. The protocol is intended to minimize the progression of tube wear, and ensure that tube wear will not progress to the point of degradation that could cause tubes not to meet leakage and structural strength test criteria.

P. Dietrich 7. Prior to entry of Unit 3 into Mode 4, SCE will submit to the NRC in writing the results of your assessment of Unit 3 steam generators, the protocol of inspections and/or operational limits, including schedule dates for a mid-cycle shutdown for further inspections, and the basis for SCEs conclusion that there is a reasonable assurance, as required by NRC regulations, that the unit will operate safely.

This CAL will remain in effect until the NRC has (1) reviewed your response to the actions above, including responses to staffs questions and the results of your evaluations, and (2) the staff communicates to you in written correspondence that it has concluded that SONGS Units 2 and 3 can be operated without undue risk to public health and safety, and the environment.

Issuance of this CAL does not preclude the issuance of an order formalizing the above commitments or requiring other actions on the part of SCE; nor does it preclude the NRC from taking enforcement actions for violations of NRC requirements that may have prompted the issuance of this letter. Failure to take the actions as described in this CAL may also result in an order if the NRC determines that failure to meet that action would result in a loss of reasonable assurance of the protection of public health and safety, and the environment.

Pursuant to Section 182 of the Atomic Energy Act of 1954, as amended (42 U.S.C. 2232), you are required to:

(1) Notify me immediately if your understanding differs from that set forth above; (2) Notify me if for any reason you cannot complete the actions and your proposed alternatives; and (3) Notify me in writing when you have completed the actions addressed in this Confirmatory Action Letter.

In accordance with 10 CFR 2.390 of the NRC's regulations a copy of this letter, and any response will be made available electronically for public inspection in the NRC Public Document Room or from the ADAMS, accessible from the NRC Web site at http://www.nrc.gov/readingrm/adams.html. To the extent possible, your response should not include any personal privacy, proprietary, or safeguards information so that it can be made available to the Public without redaction. If personal privacy or proprietary information is necessary to provide an acceptable response, then please provide a bracketed copy of your response that identifies the information that should be protected and a redacted copy of your response that deletes such information. If you request withholding of such material, you must specifically identify the portions of your response that you seek to have withheld and provide in detail the bases for your claim of withholding (e.g., explain why the disclosure of information will create an unwarranted invasion of personal privacy or provide the information required by 10 CFR 2.390(b) to support a request for withholding confidential commercial or financial information). If safeguards information is necessary to provide an acceptable response, please provide the level of protection described in 10 CFR 73.21.

P. Dietrich Please contact Ryan Lantz at (817) 200-1173 if you have any questions concerning this letter.

Sincerely,

/RA/

Elmo E. Collins Regional Administrator Docket No.: 50-361, 50-362 License No.: NPF-10, NPF-15 cc: Electronic Distribution

P. Dietrich Electronic distribution by RIV:

Regional Administrator (Elmo.Collins@nrc.gov)

Deputy Regional Administrator (Art.Howell@nrc.gov)

DRP Director (Kriss.Kennedy@nrc.gov)

DRP Deputy Director (Troy.Pruett@nrc.gov)

DRS Director (Anton.Vegel@nrc.gov)

DRS Deputy Director (Tom.Blount@nrc.gov)

Senior Resident Inspector (Greg.Warnick@nrc.gov)

Resident Inspector (John.Reynoso@nrc.gov)

Branch Chief, DRP/D (Ryan.Lantz@nrc.gov)

SONGS Administrative Assistant (Heather.Hutchinson@nrc.gov)

Project Engineer, DRP/D (David.You@nrc.gov)

Project Engineer, DRP/D (Brian.Parks@nrc.gov)

Public Affairs Officer (Victor.Dricks@nrc.gov)

Public Affairs Officer (Lara.Uselding@nrc.gov)

NRR PLBIV (Michael.Markley@nrc.gov)

Project Manager (Randy.Hall@nrc.gov)

Acting Branch Chief, DRS/TSB (Ryan.Alexander@nrc.gov)

RITS Coordinator (Marisa.Herrera@nrc.gov)

Regional Counsel (Karla.Fuller@nrc.gov)

Regional State Liaison Officer (Bill.Maier@nrc.gov)

Congressional Affairs Officer (Jenny.Weil@nrc.gov)

ACES (R4-RA-ACES.nrc.gov)

Director, NRR (Eric.Leeds@nrc.gov)

Director, DORL, NRR (Michele.Evans@nrc.gov)

Deputy Director, DORL, NRR (Louise.Lund@nrc.gov)

Director, DE, NRR (Patrick.Hiland@nrc.gov)

Branch Chief, NRR, SGCE (Gloria.Kulesa@nrc.gov)

DEDO (Martin.Virgilio@nrc.gov)

Regional Administrator, Region I (Bill.Dean@nrc.gov)

Regional Administrator, Region II (Victor.McCree@nrc.gov)

Acting Regional Administrator, Region III (Cynthia.Pederson@nrc.gov)

OEMail Resource OEDO (Lydia.Chang@nrc.gov)

File located: S:\DRP\DRPDIR\_SONGS SUNSI Rev Compl. Yes No ADAMS Yes No Reviewer Initials RL Publicly Avail Yes No Sensitive Yes No Sens. Type Initials RL DRP: PBD/BC ACES DRS: D DRP: D NRR RLantz HGepford TVegel KKennedy ELeeds

/RA/ /RA via Email/ /RA via Email/ /RA/ /RA via Email/

3/26/12 3/26/12 3/26/12 3/26/12 3/26/12 RIV: RA ECollins

/RA/

3/27/12 OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax

SCE ATTACHMENT 4 SOUTHERN CALIFORNIA Peter T. Dietrich EDISON EJ Senior Vice President & Chief Nuclear Officer An EDISON INTERNATIONAL Company October 3, 2012 Elmo E. Collins, Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 1600 East Lamar Blvd.

Arlington, Texas 76011-4511

Subject:

Docket No. 50-361 Confirmatory Action Letter - Actions to Address Steam Generator Tube Degradation San Onofre Nuclear Generating Station, Unit 2

References:

1. Letter from Mr. Peter T. Dietrich (SCE) to Mr. Elmo E. Collins (USNRC), dated March 23, 2012, Steam Generator Return-to-Service Action Plan, San Onofre Nuclear Generating Station

2. Letter from Mr. Elmo E. Collins (USNRC) to Mr. Peter T. Dietrich (SCE), dated March 27, 2012, Confirmatory Action Letter 4-12-001, San Onofre Nuclear Generating Station, Units 2 and 3, Commitments to Address Steam Generator Tube Degradation

Dear Mr. Collins:

On March 23, 2012, Southern California Edison (SCE) submitted a letter (Reference 1) to the NRC describing actions it planned to take with respect to issues identified in the steam generator (SG) tubes of San Onofre Nuclear Generating Station (SONGS) Units 2 and 3. On March 27, 2012, the NRC responded by issuing a Confirmatory Action Letter (CAL)

(Reference 2), describing the actions that the NRC and SCE agreed would be completed to address those issues and ensure safe operations. The purpose of this letter is to report the completion of the Unit 2 CAL actions, which are to be completed prior to entry of. Unit 2 into Mode 2 (as defined in the SONGS technical specifications).

Completion of the Unit 2 CAL actions is summarized below. Detailed information demonstrating fulfillment of Actions 1 and 2 of the CAL is provided in SCE's Unit 2 Return to Service Report which is included as Enclosure 2 of this letter. Enclosure 1 provides a list of new commitments identified in this letter.

P.O. Box 128 San Clemente, CA 92672 (949).368-6255 PAX 86255 Fax: (949) 368-6183 Pete.Dietrich@sce.com

Elmo E. Collins Regional Administrator October 3, 2012 U.S. Nuclear Regulatory Commission CAL ACTION 1:

"Southern CaliforniaEdison Company (SCE) will determine the causes of the tube-to-tube interactionsthat resulted in steam generatortube wear in Unit 3, and will implement actions to prevent loss of integrity due to these causes in the Unit 2 steam generator tubes. SCE will establish a protocol of inspections and/or operationallimits for Unit 2, including plans for a mid-cycle shutdown for further inspections."

COMPLETION OF CAL ACTION 1:

SCE has determined the causes of tube-to-tube interactions that resulted in SG tube wear in Unit 3, as summarized below. In addition, SCE implemented actions to prevent loss of tube integrity due to these causes in the Unit 2 SGs and established a protocol of inspections and operational limits, including plans for a mid-cycle shutdown. These are summarized under CAL Action 2.

Causes of Tube-to-Tube Interactions in Unit 3 As noted in Reference 1, the SG tube wear that caused a Unit 3 SG tube to leak was the result of tube-to-tube interaction. This type of wear was confirmed to exist in a number of other tubes in the same region in both Unit 3 SGs. Subsequent inspections of the Unit 2 SGs found this type of wear also existed in a single pair of tubes (one contact location) in one of the two Unit 2 SGs (SG 2E-089).

To determine the cause of the tube-to-tube wear (TTW), SCE performed extensive inspections and analyses, and commissioned the assistance of experts in the fields of thermal-hydraulics and in SG design, manufacturing, operation, and maintenance. Based on the results of these inspections and analyses, SCE determined the cause of the TTW in the two Unit 3 SGs was fluid elastic instability (FEI), resulting from the combination of localized high steam velocity, high steam void fraction, and insufficient contact forces between the tubes and the anti-vibration bars (AVBs). The FEI caused vibration of SG tubes in the in-plane direction that resulted in TTW in a localized area of the SGs. Details of SCE's investigation and cause evaluation are provided in Section 6 of Enclosure 2.

Corrective and Compensatory Actions, Inspections, and Operational Limits To prevent loss of integrity due to FEI and TTW in Unit 2, SCE implemented corrective and compensatory actions and established a protocol of inspections and operational limits, including plans for a mid-cycle shutdown. These are described in CAL Action 2 below.

CAL ACTION 2:

"Priorto entry of Unit 2 into Mode 2, SCE will submit to the NRC in writing the results of your assessment of Unit 2 steam generators,the protocol of inspections and/or operationallimits, including schedule dates for a mid-cycle shutdown for further inspections, and the basis for SCE's conclusion that there is reasonable assurance,as required by NRC regulations, that the unit will operate safely."

Elmo E. Collins Regional Administrator October 3, 2012 U.S. Nuclear Regulatory Commission COMPLETION OF CAL ACTION 2:

Assessment of Unit 2 Steam Generators SCE evaluated the causes of TTW in the Unit 3 SGs and the applicability of those causes to Unit 2 and inspected the Unit 2 SGs for evidence of similar wear. SCE determined the TTW effects were much less pronounced in Unit 2 where two adjacent tubes were identified with TTW indications. The wear depth was less than 15% through-wall wear, which is below the threshold of 35% through-wall at which tube plugging is required. These two tubes are located in the same region of the SG as those with TTW in Unit 3. Given that the thermal hydraulic conditions are essentially the same in both units, the significantly lower level of TTW in Unit 2 has been attributed to manufacturing differences that resulted in greater contact between the tubes and AVBs in Unit 2, providing greater tube support. Details of SCE's investigation and cause evaluation are provided in Section 6 of Enclosure 2.

Actions to Prevent Loss of Integrity due to TTW in Unit 2 SG Tubes Including Protocol of Inspections and Operational Limits SCE has taken actions to prevent loss of Unit 2 SG tube integrity due to TTW including establishing a protocol of inspections and operational limits to provide assurance that Unit 2 will operate safely. These actions are summarized below, with details provided in Section 8 of . The operational assessments performed to confirm the adequacy of these operational limits are described in Section 10 of Enclosure 2.

1. SCE will administratively limit Unit 2 to 70% reactor power prior to a mid-cycle shutdown (Commitment 1). Limiting Unit 2 power to 70% eliminates the thermal hydraulic conditions that cause FEI from the SONGS Unit 2 SGs by reducing the steam velocity and void fraction. Further, at 70% power, the SONGS Unit 2 SGs will operate within an envelope of steam velocity and void fraction that has proven successful in the operation of other SGs of similar design. Thus, limiting power to 70% ensures that loss of tube integrity due to FEI will not occur.

2. SCE plugged the two tubes with TTW in Unit 2. As a preventive measure, additional tubes were plugged in the Unit 2 SGs. Tubes were selected for preventive plugging using correlations between wear characteristics in Unit 3 tubes and actual wear patterns found in Unit 2 tubes. Removing these tubes from service will prevent any further wear of these tubes from challenging tube integrity.

3. SCE will shut down Unit 2 for a mid-cycle SG inspection outage within 150 cumulative days of operation at or above 15% power (Commitment 2). This shortened inspection interval will ensure that any potential tube wear will not challenge the structural integrity of the in-service tubes. The protocol for mid-cycle inspections is provided in Section 8.3 of Enclosure 2.

To ensure that these actions are effective in preventing a loss of tube integrity due to FEI, SCE retained the experience and expertise of AREVA NP, Westinghouse Electric Company LLC, and lntertek/APTECH. These companies routinely perform operational assessments (OAs) of SGs for the U.S. nuclear industry. AREVA and Westinghouse also have extensive steam generator design experience. SCE retained these companies to develop independent OAs using different methodologies to evaluate whether, under the operational limits imposed by SCE, SG tube

Elmo E. Collins Regional Administrator October 3, 2012 U.S. Nuclear Regulatory Commission integrity will be maintained until the next SG inspection. Each of these independent OAs demonstrates that operating at 70% power will prevent loss of tube integrity beyond the 150 cumulative day inspection interval.

The actions to operate at reduced power and shut down for a mid-cycle inspection within 150 cumulative days of operation are interim compensatory actions. SCE will reevaluate these actions during the mid-cycle inspection based on the data obtained during the inspections. In addition, SCE has established a project team to develop and implement a long term plan for repairing the SGs.

Defense-in-depth measures were developed to provide increased safety margin in the unlikely event of tube-to-tube degradation in the Unit 2 SGs during operation at 70% power. These actions, identified in Section 9 of Enclosure 2, will facilitate early detection of a SG tube leak and ensure immediate and appropriate plant operator and management response.

Basis for Conclusion of Reasonable Assurance SCE has evaluated the causes of TTW in the Unit 3 SGs and, as described in response to CAL Action 2 above, has completed corrective and compensatory actions in Unit 2 to prevent loss of tube integrity due to these causes. Tubes within regions of the Unit 2 SGs that might be susceptible to FEI have been plugged. In addition, as described in response to CAL Action 2 above, SCE has established operational limits that eliminate the thermal-hydraulic conditions associated with FEI from the SONGS Unit 2 SGs. Specifically, operation of Unit 2 will be administratively limited to 70% power. Within 150 cumulative days of operation at or above 15% power, Unit 2 will be shut down for inspection to confirm the condition of the SG tubes.

The analyses and OAs performed by SCE and independent industry experts demonstrate that under these conditions, tube integrity will be maintained. On this basis, SCE concludes that Unit 2 will operate safely.

We understand that the NRC will conduct inspections at SONGS to confirm the bases for the above information.

Please call me or Mr. Richard St. Onge at (949) 368-6240 should require any further information.

Sincerely,

Enclosures:

1. List of Commitments

2. Unit 2 Return to Service Report cc: NRC Document Control Desk R. Hall, NRC Project Manager, San Onofre Units 2 and 3 G. G. Warnick, NRC Senior Resident Inspector, San Onofre Units 2 and 3 R. E. Lantz, Branch Chief, Division of Reactor Projects, Region IV

ENCLOSURE 1 List of Commitments

Enclosure 1 List of Commitments This table identifies actions discussed in this letter that Southern California Edison commits to perform. Any other actions discussed in this submittal are described for the NRC's information and are not commitments.

Description of Commitment Scheduled Completion Date Prior to a mid-cycle shutdown of Unit 2, SCE will mid-cycle shutdown of administratively limit operation of Unit 2 to 70% power (refer to Unit 2 cover letter, Completion of CAL Action 2).

2 SCE will shut down Unit 2 for a mid-cycle steam generator within 150 cumulative (SG) inspection outage. During this outage, inspections of days of operation at or Unit 2 SG tubes will be performed to confirm the effectiveness above 15% power of the corrective and compensatory actions taken to address tube-to-tube wear in the Unit 2 SGs. (refer to cover letter, Completion of CAL Action 2).

3 SCE will install a temporary N-16 radiation detection system prior to Unit 2 entry into (refer to Enclosure 2, Section 9.2). The temporary N-16 Mode 2 detectors will be located on the Unit 2 main steam lines and be capable of detecting an increase in steam line activity.

4 SCE Plant Operators will receive training on use of the new prior to Unit 2 entry into detection tools for early tube leak identification and on lessons Mode 2 learned from response to the January 31, 2012, Unit 3 shutdown due to a steam generator (SG) tube leak (refer to Enclosure 2, Section 9.4.2).

5 SCE will upgrade the Unit 2 Vibration and Loose Part Monitor prior to Unit 2 entry into System (refer to Enclosure 2, Section 11.1). The new system Mode 2 will provide additional monitoring capabilities for steam generator secondary side noise.

6 SCE will install analytic and diagnostic software (GE Smart prior to Unit 2 entry into Signal) utilizing existing instrumentation (refer to Enclosure 2, Mode 2 Section 11.2).

SCE ATTACHMENT 5

• SOUTHERN CALIFORNIA

• EDISON0 An EDISOQ INIRNI't 4TI ALZCompnny SONGS Unit 2 Return to Service Report ATTACHMENT 4 MHI Document L5-04GA564 Tube Wear of Unit-3 RSG Technical Evaluation Report

[Proprietary Information Redacted]

SJ Non-proprietary Version ) (P.4)

Document No.L5-04GA564(9)

At Table of Contents

1. In tro d u c tion .................................................................................................................... 10

2. Sum mary of RSG Design for SONGS ....................................................................... 10 2.1 Overall RSG Design ........................................... 10 2.2 Tube Bundle Design ........................................................................................... 11

3. Description of Events ................................................................................................ 13 3 .1 Un it-2 ...................................................................................................................... 13 3.1.1. Abstract ...................................................................................................... 13 3.1.2. Sequence of Events .................................................................................... 13 3.2 Unit-3 ..... ................................. ............... 14 3.2.1. Abstract ...................................................................................................... 14 3.2.2. Sequence of Events .................................................................................... 14

4. Investigation of Wear Condition ................................................................................ 15 4.1 ECT Inspection Results ....................................................................................... 15 4.1.1. Types of Tube W ear .................................................................................. 18 4.1.2. Tube Wear in Unit-2 (for reference only) ......................... 50 4.1.3. Tube Wear in Unit-3 .................................................................................... 52 4.2 Visual Inspection Results of the Tube Bundle ..................................................... 54 4.2.1. Observations Common to Unit-2 and Unit-3 ........................ 54 4.2.2. Observations in Unit-3 .................................................................................. 54 4.2.3. Observations in Unit-2 .................................................................................. 54

5. Mechanistic Cause Analysis .................................................................................... 57 5.1 Thermal Hydraulic Condition in the Secondary Side ......................................... 57 5.2 Evaluation of U-bend Supports Condition ......................................................... 64 5.2.1. Out-of-Plane Direction Support ................................................................... 64 5.2.2. In-Plane Direction Support ......................................................................... 64 5.2.3. Differences between Unit-2 and Unit-3 ....................................................... 65

6. Tube W ear Causes ....................................... I............................................................. 70 6.1 Type 1 W ear (TTW) .......................................................................................... 71 6.2 Type 2 W ear (AVB wear) .................................................................................... 71 6.3 Type 3 W ear (TSP wear) ..................................................................................... 73 6.4 Type 4 W ear (RB wear) ...................................................................................... 73

7. C o n c lu s io n s ................................................................................................................... 81

8. Countermeasures for Return to Service .................................................................. 82 8.1 Tube Plugging ..................................................................................................... 82 8.1.1. Type 1 Wear ................................................................................................ 82 8.1.2. Type 2 W ear and Type 3 W ear ..................................................................... 82 MITSUBISHI HEAVY INDUSTRIES, LTD.

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At 8 .1.3. Type 4 W ear ............................... :......... :.................................................... . . 8 2 8.2 Operating at a Lower Thermal Power ................................................................ 84

9. References ................. :................. ;...................................... ................ 86 Appendices Appendix-1 ECT Data Evaluation of tubes with wear around Retainer Bar ............. 1-1 Appendix-2 FEI Evaluation of Tube Straight Portion for Unit-2/3 ............................. 2-1 Appendix-3 FEI Evaluation of Tube U-bend Portion for Unit-2/3 .............................. 3-1 Appendix-4 Investigation of Unit-2/3 Manufacturing and Inspection Records .......... 4-1 Appendix-5 Analytical Simulation of Tube Bundle Rotation and Hydro Static Test ....... 5-1 Appendix-6 Investigation of ISI ECT Data for AVB Support Condition for Unit-2/3 ....... 6-1 Appendix-7 Visual Inspection Results for U-Bend Region for Unit-2/3 .................... 7-1 Appendix-8 SG Tube Flowering Analysis for Unit-2/3 .............................................. 8-1 Appendix-9 .Simulation of Manufacturing Dispersion for Unit-2/3 ............................ 9-1 Appendix-1 0 SG Tube Wear Analysis for Unit-2/3 ........................................................ 10-1 Appendix-1l (D e le te d ) ........................................ ....................................................... 11-1 Appendix-1 2 Thermal Hydraulic Evaluation of Area Plugging ...................................... 12-1 Appendix-1 3 (D e le te d ) ............................................................................................... 13 -1 Appendix-14 Analytical evaluation of the impact on the Tube Support Plate and Tube Bundle due to Tubesheet deflection during Divider Plate detachment .... 14-1 Appendix-15 (D e le te d ) ............................................................................................... 15 -1 Appendix-1 6 Fatigue Evaluation of the Tube due to In-Plane Vibration ....................... 16-1 MITSUBISHI HEAVY INDUSTRIES, LTD.

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Non-proprietary Version { ](P.10)

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Document No.L5-04GA564(9)

At

1. Introduction After approximately 11 months of power operation following the steam generator replacement, SONGS Unit-3 underwent an unplanned shut down on Jan. 31, 2012 as a result of leakage of primary coolant to the secondary side from a tube in the 3B (3E-088) steam generator (SG).

The maximum leakage rate was at approximately 82 gallon/day (-13 liters/hour). Subsequent investigation revealed that the direct cause of the leakage was tube-to-tube wear.

At the time of the Unit-3 leak, SONGS Unit-2 had already completed one cycle of power operation (-22 months) after the steam generator was replaced in the refueling outage since Jan. 9, 2012. Eddy-Current Testing (ECT) inspections were performed on all tubes in both Unit-3 SGs and wear indications on many of the tubes were found. This report presents the evaluation of the mechanistic cause of tube wear, and the countermeasures required for Unit-3 return to service.

2. Summary of RSG Design for SONGS 2.1 Overall RSG Design The SONGS RSGs were specified, designed and fabricated as replacements on a like-for-like basis for the original steam generators in terms of fit, form and function with limited exceptions, and were replaced under the 10CFR50.59 rule. The CDS for the design and fabrication of the RSGs (S023-617-01, Revision 3) specified the limiting design parameters and materials.

Thus, replacement steam generator design with 3/4" tube diameter arranged in 1" triangular pitch, which was the same as in the original steam generators, and the larger heat transfer area than in the original steam generators, was optimal. The, other parameters/materials not specified by CDS were established/ selected in the design process. The SONGS RSGs were designed and fabricated to achieve an "effective zero gap" as required by CDS Rev. 3 in order to minimize its potential for tube wear. The CDS also states that the tube support/tube bundle assembly shall be fabricated such as to ensure no damage to the tubes and subsequent operation of the RSG with minimal vibration.

MITSUBISHI HEAVY INDUSTRIES, LTD.

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) (P.11)

Document No.L5-04GA564(9)

AO 2.2 Tube Bundle Design The major concern with the large U-tube SGs is their propensity for tube wear in the tube bundle U-bend region. Consequently, minimizing tube wear was given the first priority in the SONGS RSG specification, design and fabrication, and the tube support design and fabrication was discussed by MHI and SCE in numerous design review meetings. As a result, the tube bundle U-bend support design and fabrication was as follows:

1) Six (6) V-shaped AVBs (three sets of two) were provided between each tube column.

2) The AVB thickness was set such as to provide an effective "zero" tube-to-AVB gap under operating (hot) conditions.

3) The AVB end-caps were welded to the retaining bars with the U-bend in the gravity neutral position to achieve uniformity of the gap size and AVB parallelism, using spacers between the AVBs sized based on a mockup test.

The tube bundle and AVB structure configuration and components (AVBs, retaining bars, bridges and retainer bars) are shown in Fig.2-1. MHI investigated field experience with U-bend tube degradation using INPO, NRC and NPE data bases, and concluded that tube wear in the operating U-tube SGs was mostly being caused by out-of-plane tube motion. Consistent with this and Reference 7, only out-of-plane vibration of the SG U-tubes was evaluated because tube U-bend natural frequency in the out-of-plane direction is lower than natural frequency in the in-plane direction and out-of-plane vibration is more likely to occur than in-plane vibration.

No SG problems stemming from in-plane tube motion were identified by MHI and thus MHI concluded that the design and fabrication processes described above were sufficient for minimizing tubewear in the SONGS RSGs.

MITSUBISHI HEAVY INDUSTRIES, LTD.

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Aw L~IJ Bridges E~nd Cap Retaine BarJ Retaining Ba I

LAI1L Fig.2-1 Tube Bundle and AVB Structure Configuration MITSUBISHI HEAVY INDUSTRIES, LTD.

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Location of Free Span Wear Indication (2tubes)

• [~illll *' ll i~ l] i~ ll il~ Il l l l i~

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MITSUBISHI HEAVY INDUSTRIES, LTD.

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MITSUBISHI HEAVY INDUSTRIES, LTD.

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MITSUBISHI HEAVY INDUSTRIES, LTD.

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MITSUBISHI HEAVY INDUSTRIES, LTD.

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• r-V !2 C2 V t2 V2 C 7 C . z 0 A ' A C '! 0 00 ROW-141 ROW-141 ROW-131 ROW-1 31 ROW-121 ROW-121 ROW-111 (V N' ROW-111 ROW-101 ROW-91i ". lt ROW-9i

/

ROW-81 ROW-81 ROW-71 / ROW-71 ROW-61 ROW-61 N

ROW-51 ROW-51 ROW-41 ROW-41 ROW-31 ROW-31 ROW-21 ROW-21 K

ROW-11 ROW-1 ROW-1 ROW-I ROW-l1 9 ROW-i ROW-11 ROW-21 ROW-21 ROW-31 I o 0 *ROW-31 ROW-41 IROW-41 ROW-51 ROW-51 ROW-61 ROW-61 o- / ROW-71 ROW-71 ROW-81 /" ROW-81 ROW-91 , ROW-Si ROW-101 ROW-101 ROW-111 ROW-111 ROW-121 IROW-121

>40-I ROW-131 ROW-131 ROW-141 ROW-141 0000 0 0 0 0 0 0 0 0 0 0 0 0 0

00 0 80 0 0 q5 y5 yyq y 0 05 05 C, C) 05 0 05 0 0 0 £5 V~~~~

C V CCV V V N~~~~~~~~

N N 0C6 N

9V 9C 99999 6 y y y> 99 0 00 C2 C2 C N N -CC C N N N . y Fig 4.1.1-6 (2/7) Tubes with wear indications at TSPs MITSUBISHI HEAVY INDUSTRIES, LTD.

Page 33 of 474 S023-617-1-M1538, REV. 0

INon-proprietary Version j P.4 Document No L5-04GA564(9) 3A-SG #3TSP L.J_

00 0~ 0L 3 0 0co ~ c 0 0 000000 0 Q

0 0 L) 0 0 0 0

0 0 0

0

) 0 00 Q2Q 9 x y9 90 0 0 0 03 0 y yy, y

6 v A 4 y 9 ROW-141 ROW-141 ROW-131 ROW-131 ROW-121 ROW-121 ROW-111 ROW-Ill ROW-101 ROW-101 ROW-91 ROW-91 ROW-81 ROW-81 ROW-71 ROW-71 ROW-61 ROW-61 ROW-51 ROW-51 ROW-41 ROW-41 ROW-31 /

ROW-21 ROW-21 ROW-11 ROW-11 ROW-1 ROW-1 ROW-1 El V timliw R(*N-1 ROW-11 ROW-11 0 4 0 4 ROW-21 ROW-21 ROW-31 ROW-41 ROW-41 I

ROW-51 ROW-S1 ROW-61 ROW-61 ROW-71 ROW-71 ROW-81 ROW-81 ROW-91 ROW-91 ROW-101 ROW-101 ROW-111 ROW-111 ROW-121 ROW-121 ROW-131 ROW-131 ROW-141 ROW-141 00 0 0 0 0 0 0 0 0 0 0 0 0 0~- ~- ~- ~-O ~-0 ~-0 ~- 0 ~- 0 ~- 0 0 6 04§ oo 9 o 0 0 0 Fig 4.1.1-6 (3/7) Tubes with wear indications at TSPs MITSUBISHI HEAVY INDUSTRIES, LTD.

Page 34 of 474 S023-617-1-M1538, REV. 0

I Non-proprietary

( Version I ) (P.35)

Document No.L5-04GA564(9)

Ar 3A-SG #4TSP 00 oo 0 0 yy y 0 ~9 0 0 9 y y9 o 0o0o0 0 00'o 0 0 0 0,-~,40 y y 0

• 90o 0 0 0 0 0

~- 0 0 0 o o0 o* o, o~-o o~-o o~-

0 0 0 0 0 0 0 0 00 09 y y d,

U** y* y y L U

YY 0, v, 4

-h4 . v d, d, 4 4 9 Y YY ROW-141 ROW-141 ROW-131 ROW-1 31 ROW-121 N ROW-121 N

ROW-111 ROW-111 ROW-101 ROW-91 ROW-81 ROW-71

/

I ROW-101 ROW-91 ROW-81 ROW-71 ROW-61 \ ROW-61

/ ROYW-51 ROW-51

(

ROW-41 ROW-41 ROW-31 \ROW-31 ROW-21 ROW-21 ROW-11 ROW-11 ROW-1 ROW-1

______I _____

1 _______ -____ 1 ROW-1 ROW-1 4 ROW-11 ROW-11 ROW-21 ROW-21 ROW-31 4 4 0 ROW-41 ROW-41 ROW-S1 ROW-Si ROW-61 ROW-61 ROW-71 ROW-71 ROW-81 ROW-81 ROW-91 ROW-91 ROW-101 ROW-101 ROW-111 ROW-111 ROW-121 ROW-121 ROW-131 N ROW-131 ROW-141 ROW-141 00 0 0 0 0 00 00 0 00 00 0 9999090099y3y 0 0 0 00 0 0 0 000 0 0 0 00 0 0 0 0 0 0 0 0 0ý 0 - ..

9 9 y y 000 0 0 00 44 4 4444 444 M4 6 4400 Fig 4.1.1-6 (4/7) Tubes with wear indications at TSPs MITSUBISHI HEAVY INDUSTRIES, LTD.

Page 35 of 474 S023-617-1-M1538, REV. 0

INon-proprietary Version ) R6 Document No L5-04GA564(9) 3-S:G #5TSP 00 0 0 0* 0*J0 0 00 0 0 0L) 0 00 00 0 O0 0' UJ0 0 0 00 00 000 00 0 0 0 0 0 0 0- --

9999999 9 y 9 0 Q 9 9 9 9 0 00 4 4 h 4 04 44 -T, '4N ROW-141 ROW-141 ROW-131 ROW-131 ROW-121 ROW-121 ROW-111 ROW-111 ROW-101 ROW-101 ROW-91 ROW-91 ROW-81 ROW-81 ROW-71 ROW-71 ROW-61 ROW-61 ROW-51 ROW-51 O¸ ROW-41 ROW-31 /

ROW-21 ROW-21 I

ROW-11 ROW-11 ROW-1 0~ ROW-1 Li*

ROW-1

____________________________ I ROW-1 ROW-11 ROW-11 ROW-21 ROW-21 ROW-31 O IROW-31 ROW-41 O ROW-41 ROW-51 ROW-51 ROW-61 ROW-61 0

ROW-71 ROW-71 ROW-81 ROW-81 ROW-91 ROW-91 ROW-101 ROW-101 ROW-111 ROW-111 ROW-121 ROW-121 ROW-131 ROW-1 31 ROW-141 ROW-141

. -. -4. -4 -4. .J .* - 4 -4 -4 .J 00 0 0 0 0 0 0 0 0 0 0 0 0 0 0-------------

00 0U 9y 9 U 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 ---

~

"~~~ ~ 9 00000-y9y O9 9 9 CCto t 9 9 9 C Ct t C~~tttrt Q Q Q C C CON N Q Q C

ýC 0 00 0 0 dS 4-Fig 4.1.1-6 (5/7) Tubes with wear indications at TSPs MITSUBISHI HEAVY INDUSTRIES, LTD.

Page 36 of 474 S023-617-1-M1538, REV. 0

Non-proprietary Version 1] (P.37)

Document No.L5-04GA564(9) 3A-SG #6TSP 00 0 0 0 0 0 0 0 0 0 9Q 0 0 q0 9 0 0 9 Q Q

0 0

0 0

y 0

0 0 q0 0 0

y0 00 y0 0

000000000 0 0 y

000 y

0 00 y y0 y 0 0 0 0 y

4 y 0

0 o 0 - ~-

000 NN N R00 64 4A4 6 4 6 4 64 v4 o 9 ROW-141 ROW-141 ROW-131 ROW-131 ROW-121 ROW-121 ROW-111 ROW-111

//

ROW-101 ROW-101 ROW-91 ROW-91 ROW-81 // ROW-81 ROW-71 ROW-71 ROW-61 ROW-61 ROW-51 \ ROW-si ROW-41 ROW-41 ROW-31 \ROW-31 ROW-21 ROW-21 ROW-11 ROW-11 ROW-i ROW-1 3 i ___* 7K_,ZJ R ROW-1 ROW-1 ROW-li ROW-11 ROW-21 ROW-21 ROW-31 !ROW-31 ROW-41 ROW-41 ROW-51 / ROW-51 ROW-61 ROW-61 ROW-71 ROW-71 ROW-81 ROW-81 ROW-91

'ii.

ROW-91

/ ROW-101 ROW-101 ROW-111 ROW-111 ROW-121 ROW-121 ROW-131 ROW-131 ROW-141 ROW-141 00 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 "

0, 9p 9 00 0 0 0 0 0 y 0 0 0 (J 0 0 L* 0 0 0 80 80 0J0{ 0 0 0 L? y,

• J0L) y y Y" L? ) 4 8J y y, y" y 0 y 0 0 00

.000 YY Fig 4.1.1-6 (6/7) Tubes with wear indications at TSPs MITSUBISHI HEAVY INDUSTRIES, LTD.

Page 37 of 474 S023-617-1-M1538, REV. 0

Non-proprietary Version I

( ) (P.38)

Document No.L5-04GA564(9)

AO 3A-SG #7TSP 00ý 0(4 0(D 0,-4 0_. 0-4 0_4 0.4 .4 04J 0... 0_4 .j 0 0.

U00 0 0 0 0 0 0 0 0 0 0 0 0 0- -

64 d9 4 4 h 4, d4 9 99 9 9 Q 9 9 9 9 9 9 9 0 00 t2t" t ROW-141 RCOW-141 ROW-131 RCOW-131 ROW-121 ROW-121 ROW-111 ROW-101

/

• AVB 3141%40%

1 4 I

7 tt lrl-i-i-ý

ýk L

-h 0 9 Fig 4.1.1-7 (2/4) Tubes with wear indication (at AVBs)

MITSUBISHI HEAVY INDUSTRIES, LTD.

Page 40 of 474 S023-617-1-M1538, REV. 0

I Non-proprietary

( Version I )](P.41)

Document No L5-04GA564(9) 3BR-S';G AV B71 x

I I 'I I L L L L I-T 7 T-i- T - r - r - r r r 7 . . . . . . .

o"l A L 1. -1 -- - - - - - -

7 F T- r -

T 3B-SG AVB8 7:,  :: 0:0%,

LEGEND 1, J :AVB ID%10 j J :AVBt1 20%

AVB21 30%

-- -- -- - :AV 31 40" 7- 7 L- - - -

3B-SG AVB91 - -- LEGEND

- *-0: 0%

-  :*:AVB I 10%

-:AVB1 20%

AVB21 - 30%

- *:AVB 31 - 40%

• AVB >41%

1 4- I~ I Fig 4.1.1-7 (3/4) Tubes with wear indication (at AVBs)

MITSUBISHI HEAVY INDUSTRIES, LTD.

Page 41 of 474 S023-617-1-M1538, REV. 0

( A I Non-proprietary Version I

). (P.42)

Document No L5-04GA564(9) 3B-SG AVB104 3-GAVR111 13RSflAV/R1 21 Fig 4.1.1-7 (4/4) Tubes with wear indication (at AVBs)

MITSUBISHI HEAVY INDUSTRIES, LTD.

Page 42 of 474 S023-617-1-M1538, REV. 0

I Non-DroDrietarv Version I

] (P.43)

Document No L5-04GA564(9) x 3B-SG #1TSP 00000 Y

99999999 0 0 0 0 0 0 0 0 C) - ~0 - ~0 0~-0 0~-0 NO~~ 0 9 U

~ 0 0 0 Y Ou0Q~00000 0

0 0

0 0000 0

0 0

0 0 0 0 000 ROW-141,-* ....--. ROW-141 ROW-131 -. ROW-131 ROW-121 ROW-121 ROW-111 " ROW-111 ROW-101 / ROW-101 ROW-91 // ROW-91 ROW-81 / ROW-81 ROW-71 ROW-7i ROW-61 ROW-61 ROW-51 / ROW-51 ROW-41 3ROW-41 RW4 ROW-31 / ROW-31 ROW-21 ROW-21 ROW-11 ROW-11 ROW-1 0 0 0 ROW-i ROW-1 ROW-1 ROW-li ROW-11 ROW-21 ROW-21 ROW-31 0 0 ROW-31 ROW-41 / ROW-41 ROW-si \ / ROW-51 ROW-61 - / ROW-61 ROW-7i 1 ROW-71 ROW-1 ' ROW-81 ROW-91 ROW-9i ROW-101 ROW-101 ROW-111 ",jROW-111 ROW-121 ROW-121 ROW-131 ROW-131 ROW-141 .. ---- ROW-141 00 0 0 0 0 0 0 0 0 0 0 0 0 00 99 99 9 y y 9 9y 9 y9 y99 0 0 0 00 0 0 0 0 0 0 0 0 0 0 0 00 SoNy ~....,-t

, *, *, *, o, o 0 o , y yM 4y o0000000000000 y y0 y o yo o o o o o 0 00 y.49 Fig 4.1.1-8 (1/7) Tubes with wear indications at TSPs MITSUBISHI HEAVY INDUSTRIES, LTD.

Page 43 of 474 S023-617-1-M1538, REV. 0

Non-proprietary Version 1 )( .4 Document No L5-04GA564 (9) 3-:SG #2TSP 00 oo 0 0 0 0U00U00U 009990 0 0 000 0 0 0 0 0 0 0 0 -

y y 00000 ~ 9 Y0 Ui Y9 q 9Y 9 9 9 9 y 00 0 00 0 9 9 y 9 00M 0 g 9

0' 0 00 0-?

ROW-141 ROW-141

• '*-*ROW-131 ROW-131

"* ROW-121 ROW-121 ROW-111 *'* ROW-111 A.

ROW-101 *\ROW-1 01 ROW-91 ROW-91

• W-81 RCY ROW-81 ROW-71 ROW-71

/ ROW-61 ROW-61 ROW-51 ROW-51

\*ROW-41 ROW-41 ROW-31 ROW-31 ROW-21 ROW-21 RCOW-11 ROW-11 ROW-1 ROW-1 ROW-9

...__iL ROW-11 ROW-1 0~

ROW-11 ROW-21 ROW-21 ROW-31 ROW-31 ROW-41 ROW-4i1 ROW-51 ROW-51 ROW-61 ROW-61 ROW-71 ROW-71 ROW-81 ROW-81 C/ ROW-91 ROW-91

/ ROW-101 ROW-101 ROW-111 ROW-111

'C ROW-121 ROW-121 YYL)L~yU QROyW0 0 ROW-131

/ yOW-9ROW-131 ROW-141 ROW-141 9.9.. 0 00 000 UUU 0 0 00 0 RO -i U 0 00 0 0 00 00, 0o UUUU 04 CO CO .4 C. 4- t o yyo y 00 Fig 4.1.1-8 (2/7) Tubes with wear indications at TSPs MITSUBISHI HEAVY INDUSTRIES, LTD.

Page 44 of 474 S023-617-1-M1538, REV. 0

I Non-proprietary Version I

] (P.45)

Document No.L5-04GA564(9)

Ak 3B-SG #3TSP 000 0 000 00 0 00 00 0 4.

4 CCO CC yC O 00 0C 0C 00 090 0 0 0 90 9 09 C00 0 0 0 O N fN . 8N yOO CO q~ Oy q~ y~ y CC y0C y Y 0 000O ROW-141 ROW-141 ROW-1 31 ROW-131 ROW-121 ROW-121 ROW-111 ROW-111 ROW-101 ROW-101 ROW-91 ROW-91 ROW-B1 ROW-81 ROW-71 ROW-71 ROW-61 ROW-61 ROW-51 \ ROW-51 ROW-41 ROW-41 ROW-31 ROW-31 ROW-21 ROW-21 ROW-11 ROW-11 ROW-1 ROW-1 F-I I---

ROW-1 ROW-i1 ROW-11 ROW-11 ROW-21 ROW-21 ROW-31 I ROW-31 ROW-41 \ / ROW-41

/ROW-51 ROW-51 ROW-61 C / ROW-61 ROW-71 ROW-71 ROW-81 / ROW-81 ROW-91 ROW-101 4 C./

/ ROW-91 ROW-101 ROW-111 ROW-111 ROW-121 ROW-121 ROW-131 ROW-131 ROW-141 ROW-141

.*................,888.... .........

00 0 0 0 0 000000000000000000000000) ' (

• a o'J ' e 0 0 0 0 00L 0 0 0 0 .4 4 .2 .4. .4 Z4 A4..

• 00 Fig 4.1.1-8 (3/7) Tubes with wear indications at TSPs MITSUBISHI HEAVY INDUSTRIES, LTD.

Page 45 of 474 S023-617-1-M1538, REV. 0

I Non-DroDrietarv Version I

.... j (R.46)

Document No L5-04GA564(9) 3B-SG #4TSP 00 0 0 0 0 0 0 00*, 0- 0, 0 y0 9- - 0 0 0 -- 0o 0 - 00 000

1O 6 ý m - m t I U y0 to tot t: :2t- :2 ROW-141 ROW-141 ROW-131 ROW-131 ROW-121 ROW-121 ROW-111 ROW-111 ROW-101 ROW-101 ROW-91 ROW-91 ROW-81 ROW-81 ROW-71 ROW-71 ROW-6i ROW-61 ROW-S1 ROW-51 ROW-41 ROW-41 ROW-31 / \ROW-31 ROW-21 ROW-li ROW-11 ROW-1 *ROW-1 LIZ ~ Z10 111 wow C ROW-1 ROW-il RaWV-11 ROW-11 ROW-21 ROW-21 ROW-31 o , C ROW-31 ROW-41 ROW-41 ROW-51 ROW-S1 ROW-61 ROW-61 ROW-71 ROW-71 ROW-81 ROW-91 ROW-91 ROW-101 ROW-101 ROW-111 ROW-111 ROW-121 ROW-121 ROW-131 ROW-131 ROW-141 ROW-141 00 yy

,-tv 0 0 0 0 0 _CJ oo y 0_J JQ0( _

ch d, 0 0 0 4v 0 0 0 o oo - -o --

t* 0 Co CO 0 CO NJ NJ -*

O'C 4mv 0

o

-J

_J 00 o CI C9Y9C- ~

0 0aaaaaaaaaaaaaaaaaa

? -J Q0000

-Jo t Co ICC Co C 999Q o C N C CO Q9 o

000

• 4 Fig 4.1.1-8 (4/7) Tubes with wear indications at TSPs MITSUBISHI HEAVY INDUSTRIES, LTD.

Page 46 of 474 S023-617-1-M1538, REV. 0

Non-proprietary

( Version )) (P.47)

Document No L5-04GA564(9) x 3B-SG #5TSP 00 0 0 0 0 0 0 0 0 0 0 0 0 00 --

00 9 0 0 q9y 0 0 0 . 0 00 0 0 0 0000.0..00 0 0 0 0 00

................. v 'C 9g Y Y NU C 944 Y U 4 tC fl 0Q C C'-yQy ROW-141 ROW-141 ROW-131 ROW-131 ROW-121 ROW-121 ROW-111 ROW-101 j

r-:4V

-:!ý ROW-111 ROW-101 ROW-91 ROW-91 ROW-81 Z

ROW-81 ROW-71 ROW-71 ROW-61 /

ROW-5i ROW-51 ROW-41 ROW-41 ROW-31 I ROW-31 ROW-21 ROW-21 ROW-11 ROW-11 ROW-1 ROW-1 ROW-1 ROW-1 ROW-11 ROW-11 ROW-21 ROW-21 ROW-31 ROW-31 ROW-41 C ROW-41 ROW-51 / ROW-51 ROW-61 ROW-61 ROW-71 ROW-71 ROW-81 ROW-81 ROW-91 ROW-91

/

ROW-101 ROW-101 ROW-111 ROW-111 ROW-121 ROW-121 N

ROW-131 ROW-131 ROW-141 ROW-141

00. 0 0 00 0 0 00000000'O0'0-00 y 0 yy 000 00000 0 0 0 0 0 0 0 0 0 0 0 0 -*-O 0 0 000 0 0 0 00 00 Fig 4.1.1-8 (5/7) Tubes with wear indications at TSPs MITSUBISHI HEAVY INDUSTRIES, LTD.

Page 47 of 474 S023-617-1-M1538, REV. 0

I Non-proprietary Version I

) (P.48)

Document No L5-04GA564(9) x 3B!-:SG #6TSP 00 0 0 0 0 0 0 0 0 0 0 0 0 0 0- -

(9 99y 0 9 9 y 0y y q y y y 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 -

Cz a, , 000 Y~ Y 9 9009Y0 00 NN C C C C C 4. 9ý99 ROW-141 ROW-141 ROW-131 ROW-131 ROW-121 ROW-121 ROW-111 ROW-111 ROW-101 ROW-101 ROW-91 ROW-91 ROW-81 ROW-81 a * ,¸ ROW-71 0 ROW-71 ROW-61 ROW-61 ROW-51 ROW-51 ROW-41 ROW-41 ROW-31 / a 0 ROW-21 ROW-21 ROW-11 ROW-11 ROW- 1 ROW-1 ROW-Li LTh71 ROW-1 ROW-11 Q ROW-11 ROW-21 O ROW-31 /ROW-31 ROW-41 ROW-41 ROW-51 ROW-51 ROW-61 ROW-61 ROW-71 O ROW-71 ROW-81 ROW-81 ROW-91 ROW-91 ROW-101 ROW-101 ROW-111 ROW-111 ROW-121 ROW-121 ROW-131 ROW-131 ROW-141 ROW-141 00 0 0 0 0000000 0 0 0 0 0 0 0 0 U Q 0 0 99 0' 0 0 -

0 oO

~- ~-

000 0 00

~- 00~- 0 000 00 0 00 0 0% 4 6 4 64 CC4 ,Y 9 y9 9 y9 99999Q99q9Q0 00 t2 CC CCt2 1 C V 0 C 1 CC2 CT C XC Cn 6

- - - - -C~~~~~~~~~~

C C ' '

Fig 4.1.1-8 (6/7) Tubes with wear indications at TSPs MITSUBISHI HEAVY INDUSTRIES, LTD.

Page 48 of 474 S023-617-1-M1538, REV. 0

I Non-proprietary Version I j (P.49)

Document No.L5-04GA564(9)

AO 3B-SG #7TSP 00 0 0 0 0 909 0 0 0 0 0 0 0 0 0 0-y0 Y0 0 YU 0 0E 0 00 0 0 0 0 0 flt 0 0 0 0 0 0 yN y y Ny -

dd 0 00o

- 7- I yy O O O O ROW-141 ROW-141 ROW-131 ROW-131 ROW-121 ROW-121 x ROW-111 ROW-111 ROW-101 ROW-101 ROW-91 0I#

0% ROW-91 ROW-81 / ROW-81 ROW-71 ROW-71 ROW-61 // ROW-61 ROW-51 / ROW-51 ROW-41 ROW-41 ROW-31 ROW-31 ROW-21 ROW-21 ROW-11 ROW-11 ROW-1 RC1W-1 L---J ROW-1 ROW-1 ROW-11 ROW-11 ROW-21 ROW-21 ROW-31 @

/ ROW-31 ROW-41 ROW-41 ROW-51 ROW-51 ROW-61 ROW-61 ROW-71 ROW-71 ROW-81 ROW-81 ROW-91 ROW-91 ROW-101 ROW-101 ROW-111 ROW-111 ROW-121 ROW-121 ROW-131 ROW-131 ROW-141 ROW-141 00 0 0 0 0 0 0 00 0 0 0 0 0 0 -

0 0 0 0 00 0 0 0 0 0000y u o *oo oo499 0N 00 0 0 oo-0 8 oo00 0 4- o00 0 0 - 00000 00,

.y y.y Fig 4.1.1-8 (7/7) Tubes with wear indications at TSPs MITSUBISHI HEAVY INDUSTRIES, LTD.

Page 49 of 474 S023-617-1-M1538, REV. 0

• Non-roprietar version

) (P.70)

Document No.L5-04GA564(9)

6. Tube Wear Causes In general, there are 3 types of tube bundle vibration phenomena occurring in fluid environment (see Fig.6-1):

(1) Vortex Shedding Vibration (vibration due to Karman vortex)

In a single-phase flow when fluid is flowing perpendicularly to a tube, a pair of vortices, known as Karman vortices, will form periodically on the right and left side, and downstream of the tube. When the vortices move away from the tube surface periodically, the reaction forces created by them will cause the tube to vibrate. This phenomenon is called vortex shedding vibration.

The fluid flow across the SG tubes in the region of interest (U-bend region) is a two-phase flow with high void fraction (>I 1). Therefore, no Karman vortices are expected to form periodically downstream of the tube and no vortex shedding induced tube vibration is expected to occur. Empirical data confirms that vortex shedding vibration typically does not occur in two-phase flow environments where the void fraction is greater than 15%

(see Reference 1).

(2) Random Vibration Random vibration is a phenomenon where the tubes vibrate due to forces created by turbulent flow as a result of fluid velocity and density fluctuations. Vibration amplitudes due to random vibration are generally small (smaller than those due to tube fluid-elastic instability).

(3) Fluid Elastic Instability (FEI)

FEI is a phenomenon where the tubes vibrate with increasingly larger amplitudes due to the fluid effective flow velocity exceeding its specific limit (critical velocity) for a given tube and its supporting conditions and a given thermal hydraulic environment. This occurs when the amount of energy imparted on the tube by the fluid is greater than the amount of energy that the tube can dissipate back to the fluid and to the supports.

In the case when vibration occurs in a two-phase flow such as in the SG tube U-bend region, there is a possibility that it is either due to random vibration or FEI. Based upon the abovementioned study of vibration phenomena, the mechanism of each tube wear type is evaluated next based on the fault tree evaluations shown in Fig.6-2 and Fig.6-3.

MITSUBISHI HEAVY INDUSTRIES, LTD.

Page 70 of 474 S023-617-1-M1538, REV. 0

I Non-provrietarv Version I

( ,) (P.71)

Document No.L5-04GA564(9) 6.1 Type 1 Wear (TTW)

Based on the results from the rotating pancake coil ECT inspections and visual inspections, MHI concluded that the Type 1 wear (TTW) occurred due to tube in-plane motion (vibration) with a displacement (amplitude) greater than the distance between the tubes in the adjacent rows, resulting in tube-to-tube contact. Tube in-plane motion might have been caused by tube random vibration or FEI. Because the amplitude of random vibration is generally very small, the mechanistic cause of this type of wear is typically tube FEI (refer to Fig.6-1 for the difference between random vibration and FEI).

U-tube out-of-plane direction is more susceptible to flow-induced excitation than the in-plane direction due to lower U-bend natural frequency in the out-of-plane direction. U-tube FEI in the in-plane direction has never been observed in the U-tube SGs before its occurrence in the SONGS SGs. However, recent academic studies (Reference 2 and 3) report that FEI may also occur in the in-plane direction, if tube motion in the in-plane direction is possible (no tube in-plane supports or low tube contact forces with the out-of-plane supports, as concluded by MHI).

As described in Section 5.1, the void fraction (steam quality) and the flow velocity are high in the SONGS SGs which means that their tubes are generally more susceptible to vibration.

Furthermore, the average tube-to-AVB contact force in the Unit-3 SGs is concluded to be smaller than in the Unit-2 SGs, as described in Section 5.2, which makes the Unit-3 tubes to be even more susceptible to vibration and likely to FEI. Therefore, MHI concludes that in-plane tube motion which caused the Type 1 wear was due to tube FEI. The wear at the AVBs and at TSPs on some of the tubes with the Type 1 wear is an additional effect of these tubes being unstable (refer to Fig.6.1-1).

6.2 Type 2 Wear (AVB wear)

Based on the visual inspections, MHI concluded that the Type 2 wear occurred due to tube vibration which caused the tubes to wear against the AVBs at the tube-to-AVB intersections.

Tube wear at AVB intersections might have been caused by tube random vibration or FEI.

However, because most likely there were no significant gaps between the tubes and AVBs during operation (see Section 5.2.1), the occurrence of tube motion due to FEI is very unlikely (refer to Appendix-3). As described in Section 5.1, the SONGS SG tubes are susceptible to vibration (high void fractions and high flow velocities). Therefore, MHI concludes that the tube MITSUBISHI HEAVY INDUSTRIES, LTD.

Page 71 of 474 S023-617-1-M1538, REV. 0

Non-proprietary Version I

( ) (P.72)

Document NoL5-04GA564(9) wear at AVB intersections, which caused the Type 2 wear, was due to the random vibration (refer to Fig.6.2-1).

The Type 1. and Type 2 wear are simulated in the tube wear analysis as shown in Appendix-1 0.

MITSUBISHI HEAVY INDUSTRIES, LTD.

Page 72 of 474 S023-617-1-M1538, REV. 0

Non-proprietary Version )

] (P.7) "

Document No.L5-04GA564(9)

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6.3 Type 3 Wear (TSP wear)

The tubes with the Type 3 wear are located mostly near the TSP flow slots and at the periphery of the tube bundle where the velocities of cross-flow are high (refer Fig.6.3-1). Tube vibration in cross-flow may be caused by tube random vibration or FEI. However, the size of the gap between the TSP tube hole land surface and the tube is limited (design size is! 1). Thus, the occurrence of tube FEI is unlikely. Therefore, MHI concludes that the Type 3 wear is caused by cross-flow induced random vibration in the region where secondary fluid cross-flow velocities are high (refer Fig.6.3-2). The results of the FEI and random vibration analysis are shown in Appendix-2.

6.4 Type 4 Wear (RB wear)

The tubes with the Type 4 wear have no indications of TTW or AVB wear, or TSP wear, which suggests that it is caused by only the retainer bars vibrating. SONGS SGs have two types of retainer bars, I Imm (1 1) in diameter andl Imm (1 1) in diameter. Tube wear was found on the tubes adjacent to the retainer bars, but only at the smaller diameter retainer bars.

The retainer bars with the smaller diameter have also a relatively long span as compared with the other SGs fabricated by MHI, which means that the natural frequency of these retainer bars is lower and thus they are more likely to vibrate. Therefore, MHI concludes that the Type 4 wear is caused by random vibration of the retainer bars induced by the secondary fluid exiting the tube bundle(see Reference 4 for details).

MITSUBISHI HEAVY INDUSTRIES, LTD.

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Ak Downstream

> In a single-phase flow when fluid is flowing perpendicularly to a tube, a pair of vortices, known as Karman vortices, will form periodically on the right and left side, and downstream of the tube. When the vortices Karman Vortex move away from the tube surface periodically, the reaction forces created by them will cause the tube to vibrate. This phenomenon is called vortex shedding vibration.

Upstream Random vibration is a phenomenon where the tubes vibrate Downstream due to forces created by turbulent flow as a result of fluid velocity and density fluctuations. Vibration amplitudes due Each tube to random vibration are generally small (smaller than those vibrates G due to tube fluid-elastic instability). Independently \J Fli V Force Upstream I

I v'Fluid El astic Instability (FEI) I I

>FEI is a phenomenon where the tubes vibrate with Tube vibration affects surrounding fluid and causes Downstream increasingly larger amplitudes due to the fluid effective flow velocity exceeding its specific limit fluctuation which propagates to other us tubes 11 Fluid Elastic Instability A Random n tustream Vibration motion" due to tube (ube motion Vibration affects surrounding fluidI Coupled Wen tube motion and fluid force l flctuate at the right time, tubes vibrateI flIv11gorously) I,/

Fig.6-1 Flow Induced Tube Bundle Vibration MITSUBISHI HEAVY INDUSTRIES, LTD.

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E Part Assumed cause Evaluation Concusion Wear of tubes for Unit-3, U-bend.region Tube-to-Tube Wear Fluid Elastic Instability Progress of wear in very short Yes I(FEI) .eiiod can be causeddb FEI.

RandomRandm vbraion vibrationtubeoube

• /vibration ibra-tion amplitudes due to radm are genierally small and not -

w No Karman vortex shedding

[ocur Karman vortex shedding does not in two-phase flow environments.

No at AVB FEI Because most likely there were no No significant tube-to-AVB gaps'dudng operation, the occurrence of tube out-of-plane FEI is viery unlikely.

Random vibration H SONGS SG tubes are susceptible to Yes random vibration due to high void U) fractionsand high flow velocities.

M t-U 4Karman vortex sheddinq (same as tube-to-furbe wear'  ! No I (D M at retainer bar vibration of tube e ubes with retainer bar wear No

-4

> -haveNo indications-0f TTW of AVR Lwear, or TSP wear, which suggests Zthat it is caused by only the retainer

-_Z *-be-bars vibrating.

viraio of)reaiebr Retainer bars have a relatively smalll Yes X-4 rn;

..I Idiiifieter and long span as corftpared with the other SGi Yes

_l fabricated by MHI, which means that _ _

the natural frequency of these - r z retainer bars is lower and thus they 0 F* ____ _ - 'lare more likelyto vibrate. 7______'_

at Tube support plate (TSP) FEI The size of the gap between the No 0 TSP, tube hole land surface and the '--

o tube is limited. Thus. the occurrence CD of tube FEI is unlikely. (D R vibration TSP wear can caused by the cross- Y <

-4 flow induced random vibration in the region where secondary fluid cross- U .

flow velocities are high. 0 C.71~~~~~~(T 01 vortex sheddin Kam~an (same as tube-to-tube wear) No)

,._anvrtxsh,,7t~

Fig.6-2 Fault Tree Evaluation for the Causes of Wear V m

FEI and Theomsl Irlncrease of polential of tube Higher void fraction t U-bend region T h drmal a)lyset SaruSuructs in awo-wfiase flow fiWl have lower resiae o Yes Random .

II _______

LIv on

_____R ifydraHic H vit retion when a void fraction or steanm quality is high Sesfctirn 5,1 for de"Itsl)

L lcur AVB ~ ", de* tili [Ro n handigtc. of RSGa durig on AVB iertin depth by bobbin I s i e that AVB inertion depth is not changed dtion manufcturing ECT signals fonm hti design condition for the representative colurns (S4. Section Apprnd-4 for diitel.)

Contact foroe of AVO to Manufacturing dimensional disperaon Tube diameter ovskity (G vakie) Tubes uLed for Unt-2 havo larger variation (standard Ys tube devilton) of tube 0 value tihn those for Unit-3.

ItIsaessumed that the crtstd force of AIS to tubo for LUk-2 Ib relatively large compared with Urit-3.

(See Section 52,3 and Apenpdx-, for details.)

AVB tier end thickness AVB blia *A thickneee of Unit-2 we Igr then those o ye$

Unk-3 because of tie diffsreio. of AVB prsesing loed.

(See Section 52.3 and Appendix-9 for details)

('3 ECT date (Ding signa1) [The tube4o-AV1B contact forces of the Unit-2 SGe are yes SLgreater than those in the Unit-3 Ga, especially t tihe AVS not.e locations, which Is evidenced by more ding signals In the Unlt-2 SGs rtuan in the Unit-3 SOs,

=r" l(See Section 5.2,3 aid A ilendxi-for detals.)

>Rotation. hancing. etc. of RSGs during Research of history end recorda of mwufacnuactng The Unit-3 SGs underwerl I[more rota*ons then tie No ILbundle chdange In the tube support condtion bevieen t*e Unit-3 z and Unit-2 SGs due to the difference In number of rotaions wes found to be negligitily malt gh tfee Section 5.2.3 and A4pe,.dix-5 for delalls.)

Tubeto-A gap Rotation. handg. atet.O RSGs uit, eserch of iTtohy ant records of /4 I l-3,Smansufing undenwentl Imore rotation$ then the No manufacturing at Kobe shop aid deformation sanlysee of tube lUnt-2 SGs due to the divider plate repair. However, tie undl chane In the tube support condition between "h Unit-3

" land Unit-2 SGs due to the difference in number of Z (SIee Section 5.2.3 and App*enx-. for detaIl) 7 Vll inspection of inside of U-band region The were no signifcant gapebetween tie AV" and No 0 hsbeswhioh might have contilited to Uvessive tube 0 vjib on becatuse, tieW Ssappears to be Mrally in contect with tubel D Dacbmsiaoni of U-bend tDynanmio pressure of secondary flid Desformation analyses oftube bundlasby taking The tube bundle deformation analysis Indlcatsee that the

]

No <

region during operation and dfference of thermal expansion Into count of. -contact force between the tubes end AVBS pmrduce the CD

> DFighc presure of secondary fuid rlM forcs which pevent id on of the A)!M II Di expansion thncoftermsl Istrcture assamm*yand tie dynarili pressure durIng I operaion do"j not Iloibetive tH o-AW gaps. 6 o M Fig.6-3 Fault Tree Evaluation for the Causes of FEI and Random Vibration ,.

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Document No.L5-04GA564(9)

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[:::::Characteristics of SONGS RSG

[Thermal Hydraulics]-I

/Design with High Steam Quality in U-Bend(maxl 1)

[AVB Structure]

V Tube between 2 flat AVBs

-AVB Design Assumes Out-of Plane Vibration Since out-of-plane FEI is more likely to happen compared to in-plane FEI. AV13sare Void Fraction placed at the sides of tube to preven Out-ot plane vibration Distribution in v/ 6 V-Shaped AVBs (12support points) 100% Output

-Number of AVB Support Points are designed by FE!

Evaluation based on ASME Sec.l11 (Out-of-Plane FEI Wil not occur even if one of supports is inactive as design basis)

/Designed and fabricated for "Zero" Gap between Tube and AVB in hot condition Tubes Fig.6.1 -1 Type 1 Wear (TTW) Mechanism MITSUBISHI HEAVY INDUSTRIES, LTD.

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[Visu Inspection Results]

=*Configurabon r Out-of-Plane Vibration Tube contacts AVB dueto is Confirmed Random Vibration IRNo signifiant gaps

[ Visualnpeto between Results ] tubes and AVB.

Low probability of Out-of-Plane FEI ICauses of Random Vibration I High Void Fracion Cauies High two Phase Flow Average Velocity *Increae In Vibraton Excitaton Force

" High Void Fraction Causes Low Damping j Characteristics of SONGS RSG

[Thermal Hydraulic] [Structure] I (Very Dry Steam Void Fraotin (Max 0.996 Steam Quality (Max.0.$)

ZeoGap b3etween Tube

• nAV8 in esig and I Manufacturing I I I

f 2-hseFow Average VloctyInrease II Vi1bration Resistance (Damiping Rato) Decrease I Low contact force between Tuibe ndAV$ during I Random Vibration Excitation Force Increase ,I Opraio Iccreneof Randm ibaj Fig.6.2-1 Type 2 Wear (AVB wear) Mechanism MITSUBISHI HEAVY INDUSTRIES, LTD.

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S

\jeu All Row.1 Tube 4, I 0

'I 2*

4, A-4, a

C Height from Tubesheet [ft]

Ou~o*

Out-of- Axbe

• Tube Plane Ax 04, In-Plane Ce Height from Tubesheet [ft]

Fig.6.3-1 Flow Velocity Distribution at TSPs (Actua Desgn] I Characteristics of SONGS RSG I

-Low probability of FEI due to support condition Tube contacts Land of Tube Straight Region (extremely email gap -- Region of BEC Hole between land region of BEC Hole and tube due to Random (Nornikial Gap: 0.31rmm))

Vibration Narrow Pitch Tube Arrangement Main cause of Random Vibration L High Local Flow Velocity in the Horizontal Direction due to Tube Arrangement (Vibration Excitation Force Increase)

! High Flow Velocities in Tube Straight Region Occurrence of Randomibrion Fig.6.3-2 Type 3 Wear (TSP wear) Mechanism MITSUBISHI HEAVY INDUSTRIES, LTD.

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0 0

The tube which has wear at retainer bar region tI-however does not have wear at AVB region It is cnlddthal (It is considered that Retainer Bar itself isVibrating)

E CL E C

C C.- .0 0

C. -SONGS RSGs have 2 types of retainer bars, C-z which are retainer bars with small diameter and Vibatonof Retainer Bai with large diameter. However, wear indications CU C6 are only found at contact regions between tubes WJ and retainer bars with small diameter.

Z-7, I Cause of Random Vibration 4- D)

V Retainer Bar Design with Low Natural Frequency._ a, Retainer Bar '/;i efnto Is to hold AVI "anMbfr to the tub. bundle drnin *eachstag. of instdton ad opacv SSONGS RSGahaveretatna bars at 24 1Won~m 112Locat~n. Each of Large Dtuseter# 010.. & Sinai Dkuatar R4.T e in)ng a Retaining Bar