ML20141C241
| ML20141C241 | |
| Person / Time | |
|---|---|
| Issue date: | 04/30/1997 |
| From: | Karwoski K Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-1604, NUDOCS 9705160211 | |
| Download: ML20141C241 (171) | |
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NUREG-1604 Circumferential Cracking of Steam Generator Tubes U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation 1
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AVAILABILITY NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources:
1.
The NRC Public Document Room, 2120 L Street, NW., Lower Level, Washington, DC 20555-0001 2.
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Circumferential Cracking of Steam Generator Tubes t
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Manuscript Completed: April 1997 j
Date Published: April 1997 Prepared by K. J. Karwoski i
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Division of Engineering Office of Nuclear Reactor Regulation j
U.S. Nuclear Regulatory Commission
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i ABSTRACT On April 28, 1995, the U.S. Nuclear Regulatory This report is intended to be representative of Commission (NRC) issued Generic Letter (GL) significant operating experience pertaining to 95-03, 'Circumferential Cracking of Steam Generator circumferential cracking of steam generator tubes Tubes." GL 95-03 was issued to obtain information from April 1995 through December 1996. Operating needed to verify licensee compliance with existing experience prior to April 1995 is discussed regulatory requirements regarding the integrity of throughout the
- report, as necessary, for steam generator tubes in domestic pressurized-water completeness.
reactors (PWRs).
This report briefly describes the design and function of domestic steam generators and summarizes the staff's assessment of the responses to GL 95-03. The report concludes with several observations related to steam generator operating experience.
iii NUREG-1604
l CONTENTS j
i AB STRA CT........................................................... iii
- LIST OF FIG URES....................................................... vii LIST OF TABLES.......................................................ix EXECUTI VE SUM M ARY.................................................... xi 1 INTR ODUCTION..................................................... 1 - 1 l
2 STEAM GENERATOR DESIGNS........................................... 2-1 l
3 GL 95-03, "CIRCUMFERENTIAL CRACKING OF STEAM GENERATOR TUBES"
..........3-1 3.1 Content of G L 95-03............................................. 3-1 3.2 Generic Assessment of the Industry's Responses to GL 95-03..................... 3 - 1 4 INSPECTION, REPAIR, AND ASSESSMENT OF STEAM GENERATOR TUBES............ 4-1 4.1 Purpose of Steam Generator Tube Insnections............................... 4-1 4.2 Eddy Current Testina Tube Repai rs................................................... 4 2 4.3
.4-4 4.3.1 Pl uggin g............................................... 4 -4 4.3.2 Sleeving............................................... 4-4 4.3.3 Repl acement........................................
Sum =~v.....................................................4-4 4.4
..... 4-5 5 BABCOCK AND WILCOX ONCE-TliROUGH STEAM GENERATORS.................. 5-1 5.1 Desien.......................................................5-1 5.1.1 Generic Featu res.......................................... 5 1 5.1.2 Plant-Specific Features...................................... 5-1 5.2 Locations Suscentible to Circumferential Crackine............................ 5 3 5.2.1 Uppermost Span of Unsleeved Tubes in the ime/ Wedge Region............. 5-3 5.2.2 Expansion Transitions and Crevice Region.......................... 5-3 5.2.3 Dented / Dinged Regions...................................... 5-5 5.2.4 S leeve Joints............................................. 5 -6 5.3 Justification for Continued Operation.................................... 5-6 5.4. Tube lnspections
................................................ 5-7 6 COMBUSTION ENGINEERING STEAM GENERATORS............................ 6-1 6.1 Desi gn...................................................... 6-1 6.1.1 Generic Features..........................................
6-1 6.1.2 Plant-Specific Features
..................................... 6-2
6.2 Locatiom
Susceptible to Circumferential Crackine
............................ 6-3 6.2.1 Top of Tubesheet Region..................................... 6-3 6.2.2 Dented Locations Including Dented Tube Support Areas.................. 6-4 6.2.3 U-Bend Region........................................... 6-5 6.2.4 S leeve Joints............................................. 6 -5 y
i 6.2.5 Circumferential Cracking Experiences in the Pre-Replacement Palisades and Millstone Steam G enerators.......................................... 6-5 6.2.5.1 Pre-Replacement Palisades Steam Generators................. 6-5 6.2.5.2 Pre-Replacement Millstone 2 Steam Generators
...............6-7 i
6.2.5.3 Summary of Operating Experience from the Pre-Replacement Palisades j
and Millstone 2 Steam Generators
.......................6-9 j
6.3 Justification for Continued Ooeration.................................... 6-9 i
6.4 Tube lnsoections.............................................. 6-10 1
I 7 WESTINGHOUSE STEAM GENERATORS.................................... 7-1 7.1 Ds81gn....................................................... 7-1 7.1.1 Generic Features................................. '.........
7-1 l
7.1.2 Plant-Specific Features...................................... 7-3 7.2 locatinai Suecaatible to Circumferaaei=1 Crackia-............................ 7-3 7.2.1 Expansion Transition and/or Top of Tubesheet Region................... 7-4 4
I 7.2.1.1 Partial-Depth Hardroll Steam Generators
...................7-4 7.2.1.2 Full-Depth Hardroll Steam Generators..................... 7 5 7.2.1.3 WEXTEX Steam Generators........................... 7-8 7.2.1.4 Hydraulic Steam Generators........................... 7-9
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7.2.2 Small-Radius U-Bends......................................
7-11 i
7.2.3 Dented locations.........................................
7-14 7.2.3.1 Fatigue Cracks at North Anna 1 and Indian Point 3............ 716 7.2.3.2 Non-Fatigue Cracks............................... 7-17 7.2.4 Sleeve Joints............................................ 7-21 7.2.5 Preheater Expansion Transitions................................ 7-22 7.2.6 S ummary.............................................. 7-23 j
7.2.6.1 Partial-Depth Hardroll Steam Generators.................. 7-23 7.2.6.2 ~
Full-Depth Hardroll Steam Generators................... 7-23 1
7.2.6.3 WEXTEX Steam Generators.......................... 7-24 7.2.6.4 Hydraulic Steam Genentors.......................... 7-24 7.3 Justification for Continued Ooeration................................... 7 24 j
7.4 Tube lnsoections............................................... 7-26 8 STEAM GENERATORS WITH ALLOY 690 TUBES............................... 8-1 8.1 Demian and Imcations Susceptible to Circumferential Crackine
....................8-1 8.2 Justification for Continued Operation.................................... 8-2 8.3 Tube inaaactions................................................ 8-2 9 CON CLUS I ON S...................................................... 9 - 1
- APPENDIX A: STEAM GENERATOR TUBE SLEEVES
.............................A-1 APPENDIX B: PLANT LI STING S............................................ B 1 APPENDIX C: A CR O NY M S............................................... C-1 i
APPENDIX D: REFERENCES..
...........................................D-1
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i NUREG 1604 vi l
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l LIST OF FIGURES 2-1 PWR recirculating steam generator........................................ 2-2 2-2 B&W once-through steam generator........................................ 2-3 2-3 CE steam generator.................................................. 2-4 4-1 Bobbin coil probe.................................................. 4-7 4-2 Rotating pancake coil probe............................................. 4-8 4-3 Plus-point coil on a rotating probe head
.....................................4-9 4-4 Cecco probe.....................................................4-10 5-1 Typical B&W OTSG tube pattern......................................... 5-9 5-2 B&W OTSO tube support plate design..................................... 5-10 6-1 Typical CE recirculating steam generator (RSG) tube pattern
.......................612 6-2 Partial-and full-depth expansion transitions.................................. 6-13 6-3 Typical tube support configurations....................................... 6-14 7-1 Cutaway view of s typical RSG
.........................................7-28 7-2 Typical Westinghouse Model 51 RSG tube pattern.............................. 719 7-3 Typical Westinghouse steam generator with a preheater (Model D-2).................. 7-3 0 7-4 U-bend portion of a small-radius tube..,....................,...........,. 7-31 A1 Westinghouse hybrid expansion joint configuration............................... A-5 A-2 Westinghouse hybrid expansion joint sleeve................................... A-6 A-3 B&W kinetically welded tube support plate sleeve............................... A-7 A-4 B&W kinetically welded tubesheet sleeve
....................................A-8 A-5 B&W mechanical sleeve............................................... A-9 A-6 CE TIG-welded expansion transition zone sleeve............................... A-10 A-7 CE TIG-welded tubesheet sleeve......................................... A-11 A-8 CE TIG-welded tube support plate sleeve
................................... A-12 A-9 Westinghouse laser-welded full-length tubesheet sleeve
........................... A 13 A-10 Westinghouse laser-welded elevated tubesheet sleeve............................. A 14 A-11 Westinghouse laser-welded tube support plate sleeve............................. A-15 l
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f vii NUREG-1604
LIST OF TABLES 51 Inspections at the expansion transition region in B&W plan s 5-11 5-2 Inspections at the lane / wedge region in B&W plants 5-12 5-3 Inspections at dented locations in B&W plants 5-13 5-4 Inspections at sleeve joints in B&W plants 5-14 6-1 Millstone 2 steam generator tube inspections (1987 to replacement) 6-15 6-2 Inspections at the expansion transition region in CE plants 6-16 6-3 Inspections at dented locations in CE plants 6-18 6-4 Inspections in the U-bend region of small-radii tubes in CE plants 6-20 6-5 Inspections at sleeve joints in CE plants 6-22 7-1 Inspections at the expansion transition region in Westinghouse partial-depth hardroll plants 7 32 7-2 Inspections at the top-of-tubesheet dented locations in Westinghouse partial-depth hardroll plants 7-33 7-3 Inspections in the U-bend region of small-radii tubes in Westinghouse partial-depth hardroll plants 7-34 7-4 Inspections at dented locations in Westinghouse partial-depth hardroll plants 7-35 7-5 Inspections at sleeve joints in Westinghouse partial-depth hardroll plants 7 36 7-6 Inspections at the expansion transition region in Westinghouse full-depth hardroll plants 7-37 7-7 Inspections in the U-bend region of small-radii tubes in Westinghouse full-depth hardroll plants 7-38 7-8 Inspections at dented locations in Westinghouse full-depth hardroll plants 7-39 7-9 Inspections at sleeve joints in Westinghouse full-depth hardroll plants 7-40 7-10 Inspections at the expansion transition region in Westinghouse WEXTEX plants 7-41 7-11 Inspections in the U-bend region of small-radii tubes in Westinghouse WEXTEX plants 7-42 7-12 Inspections at dented locations in Westinghouse WEXTEX plants 7-43 7-13 Inspections at sleeve joints in Westinghouse WEXTEX plants 7-44 7-14 Inspections at the expansion transition region in Westinghouse hydraulic plants 7-45 7-15 Inspections in the U-bend region of small-radii tubes in Westinghouse hydraulic plants 7-47 7-16 Inspections at dented locations in Westinghouse hydraulic plants 7-49 7 17 Inspections at sleeve joints in Westinghouse hydraulic plants 7-51 8-1 Inspections at the expansion transition region in plants with alloy 690 steam generator tubes 8-4 8-2 Inspections in the U-bend region of small-radii tubes in plants with alloy 690 steam generator tubes 8-5 8-3 Inspections at dented locations in plants with alloy 690 steam generator tubes 8-6 B-1 Plant Listing by Name B-2 B-2 Plant Listing by Tube Material B-4 B-3 Plant Listing by Vendor B-5 B-4 Westinghouse Plant Listing by Tube Expansion Type and Material B-6 B-5 Westinghouse Plant Listing by Steam Generator Model B-7 B-6 Plants with Replacement Steam Generators (December 19%)
B-8 ix NUREG-1604
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EXECUTIVE
SUMMARY
Steam generator tubes in pressurized-water reactors a few severe indications were detected which (PWRs) were manufactured using materials warranted inspections at more frequent intervals to susceptible to various forms of degradation. As provide reasonable assurance that a safety-signincant steam generators in operating PWRs age, the number indication will not develop before the next inspection.
of tubes susceptible to degradation is expected to increase. Consequently, the U.S. Nuclear Regulatory Because of the number of tubes identified with Commission (NRC) has monitored steam generator circumferential indications and the extent (i.e., size) operating experience for many years and has issued of some of these indications, the NRC staff held many generic communications related to steam several meetings with the PWR owners groups, the generator tube integrity (Refs. I through 33),
industry (Electric Power Research Institute), and various licensees. These meetings focused on efforts The types of degradation affecting steam generator being undertaken to improve the ability to detect and tubes include pitting, wear, thinning, wastage, and size circumferential indications. During the period intergranular stress corrosion cracking (IGSCC). The when these meetings were being conducted, the orientation of these types of degradation is either licensee for Maine Yankee Atomic Power Station axial, circumferential, or volumetric. Historically, began to identify a number of circumferential these forms of degradation have been detected and indications that were larger than anticipated for the characterized with a bobbin coil probe, a type of 6-to-7-month operating time between inspections.
eddy current inspection probe. The habbin coil probe These results, in part, led the licensee to perform is primarily sensitive to degradation with an axial additional inspections using a more sensitive component (i.e., axial and volumetric flaws).
technique. These inspections identified many more Circumferentially oriented tube degradation is tubes with circumferential indications than had particularly noteworthy since detection of this form of previously been identified. Penetrant testing and degradation requires the use of specialized eddy destructive examination of some tubes that had been current inspection probes.
removed from the steam generators confirmed that several indications identified with this more sensitive Circumferentially oriented steam generator tube technique were, in fact, circumferential cracks.
degradation has been identified at various locations as a result of inspections and primary-to-secondary In consideration of the events at Maine Yankee, the leaks. Circumferential indications have been detected NRC staff issued Generic Letter (GL) 95-03, in the expansion-transition region, dented locations "Circumferential Cracking of Steam Generator (including dented tube-to-tube support locations),
Tubes," on April 28,1995 (Ref. 2). GL 95-03 tube-to-sleeve joints, the U-bend region of requested, in part, that all holders of PWR operating small-radius tubes, and the 15th tube support plate licenses or construction permits submit a safety and upper tubesheet secondary face (UTSF) in the assessment justifying continued operation, as well as lane / wedge region of once-through steam generators a summary of their inspection plans for the next (OTSGs).
The circumferential indications are scheduled steam generator tube inspection outage as attributed primarily to lGSCC and, to a lesser extent, they pertained to the detection of circumferential to high cycle fatigue.
cracking.
Large circumferential indications (in circumfenential By late June 1995, the NRC staff received the extent and depth of degradation) were identified at a majority of the PWR licensee submittals in response few plants. In addition, the licensees for a few plants to GL 95-03. An initial review of these submittals identified a significant number of circumferential and other supporting documentation (e.g., steam indications. Although a number of tubes exhibited generator inspection reports) provided by the various circumferential indications, the majority of these licensees and vendors did not identify any significant indications were of little or no safety significance, safety concern that would warrant immediate NRC given the current practice of repairing all action. However, the NRC review did identify circumferential indications upon detection. However, additionalinformation that was needed for the staff to xi NUREG-1604
complete its evaluation. As a result, a number of (5) the operating time until the next steam plant-specific requests for additional information generator tube inspections were to be conducted (RAls) were prepared and sent to the licensees. In or repairs (i.e., steam generator replacement) many instances, these RAls asked licensees to were to be implemented by the individual PWR explicitly address in their responses all areas licensees potentially susceptible to circumferential degradation.
In most cases, these RAls were provided to the (6) the risk and potential consequences of a range licensees before the next scheduled steam generator of steam generator tube rupture events as tube inspection outages to ensure that the licensees discussed in NUREG-0844, "NRC Integrated evaluated all appropriate areas before (or during)
Program for the Resolution of Unresolved their next scheduled outages. Responses to the RAls Safety Issues A-3, A-4, and A-5 Regarding were received from September 1995 through Steam Generator Tube Integrity," which was December 1996.
issued in September 1988 (Ref. 27)
In evaluating the responses to GL 95-03, the staff In general, reasonable assurance of tube integrity was took into consideration that most circumferential provided through a combination of inservice indications are attributed to IGSCC, a time-dependent inspection, tube repair, primary-to-secondary leak material degradation process which depends on many rate monitoring, preventive measures (such as plant-specific factors. Rese factors include tube chemistry control), and analyses (e.g., condition material (e.g., alloy 600, alloy 690), operating monitoring and operational assessments) to verify that characteristics (e.g.,
cumulative operating time, safety objectives were being met. Hese methods temperature), tube microstructure (e.g., grain size were used to provide defense-in-depth, and all plants and carbide distribution), and water chemistry relied on these methods to provide reasonable practices. The staff also considered that IGSCC is assurance of steam generator tube integrity and plant exacerbated by the presence of crevices, residual
- safety, stresses, material sensitization, cold work, and corrosive environments.
The staff has assessed the results of several inspections performed since GL 95-03 was issued.
After reviewing various licensee submittals, the staff These assessments confirmed the various licensees concluded that, in all cases, PWR licensees provided justifications for operating until the next steam sufficient justification to operate their facilities as a generator tube inspection outage. In some cases result of circumferential steam generator tube (e.g., Byron 1), many circumferential indications degradation until inspections could be conducted or were detected; however, the structural and leakage repairs (e.g., steam generator replacement) could be integrity of the tubes was not compromised by these implemented. The staff based its conclusions on the indications. It is important to note that although a following factors:
number of tubes exhibited circumferential indications, the size of the circumferential indications is of greater (1) the scope and results of the prior inspection safety significance than the number ofindications.
(2) the operating experience at similarly designed and operated units (3) the operating conditions at the plant (e.g.,
hot-leg temperature, water chemistry practices, leakage monitoring, cumulative steam generator operating time)
(4) mitigating actions taken by the licensee (e.g.,
shotpeening, rotopeening, in situ stress relief)
NUREG-1604 xii
1 INTRODUCTION Steam generator tubes in pressurized-water reactors Rotating probes and Cecco probes are currently used (PWRs) are susceptible to v:.rious forms of to inspect locations susceptible to circumferentially degradation, including pitting, wear, thinning, oriented tube degradation. Rotating probes frequently wastage, and intergranular stress corrosion cracking have three eddy current coils per probe head, and the (IGSCC). Historically, these forms of degradation coils typically used in rotating probes for the have been detected and characterized with a bobbin detection of circumferential degradation include the coil probe, a type of eddy current inspection probe.
pancake coil and the plus-point coil. Cecco probes ne bobbin coil probe is primarily sensitive to are available in several different designs.
The degradation with an axial component (i.e.. axial and designs most commonly used in the United States are volumetric flaws). As a result, forms of degradation the Cecco-3 and Cecco-5 probes. Both of these which are circumferentially oriented are particularly designs have multiple transmit / receive coil pairs.
noteworthy since detection of these forms of degradation requires the use of specialized eddy In the early 1990s, a few plants identified current inspection probes.
circumferential indications which were large in circumferential extent and depth of degradation. In Inspections with the bobbin coil probe proved addition, a significant number of circumferential
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ineffective at detecting circumferentially oriented tube indications were identified at a few plants, including I
degradation that began to emerge in the late 1970s.
Arkansas Nuclear One, Unit 2 (ANO-2) and Maine As a result, the nuclear industry developed new Yankee Atomic Power Station.
l probes to supplement inspections performed with the bobbin coil probe. These new probes were more Because of the n imber of tubes identified with effective at detecting both circumferentially and circumferential ind. cations and the size of the axially oriented tube degradation. Inspections with indications, the NRC staff held several meetings with these probes and primary-to-secondary tube leaks the PWR owners groups, the industry (Electric Power resulted in the identification of circumferentially Research Institute), and various licensees.
On oriented tube degradation at various locations in the September 26, September 27, and November 15, steam generator, including the expansion-transition 1994, the NRC staff discussed circumferential steam
- region, dented locations (including dented generator tube cracking and the importance of tube-to-tube support locations), tube-to-sleevejoints, detecting this degradation with the Westinghouse, the U-bend region of small-radius tubes, and the 15th Combustion Engineering (CE), and Babcock &
tube support plate and upper tubesheet secondary face Wilcox (B&W) owners groups, respectively. In (UTSF) in the lane / wedge region of once-through addition, on January 12 and February 22,1995, the steam generators (OTSGs).
Identification of NRC staff and industry representatives from the circumferentially oriented tube degradation resulted in Steam Generator Stu.Qic Management Program the publication of several U.S. Nuclear Regulatory (SGMP) met to discuss ongoing efforts to improve Commission (NRC) information notices (Refs. 8,11, the methods for detecting and sizing circumferential 14, and 17).
indications. Representatives from the SGMP stated that qualified techniques for detecting circumferential Circumferentially oriented steam generator tube indications were available and that a qualified i
i degradation is generally attributed to IGSCC. Axially technique for sizing circumferential indications was l
and circumferentially oriented IGSCC has been under development. Subsequent to these meetings, observed for a number of years and was identified in the SGMP shared with the industry points to consider 1990 by the NRC as a source of significant in circumferential crack detection and length sizing degradation to tubes in PWR steam generators (Ref.
that reflected the results of plant inspections and 17). In addition to IGSCC, circumferential cracks laboratory testing up to that time (Ref. 34).
have also been attributed to high cycle fatigue.
Several NRC generic communications were issued in On December 23,1994, the NRC issued Information response to the detectionof high cycle fatigue cracks Notice (IN) 94-88, " Inservice Inspection Deficiencies I
in steam generator tubes (Refs.16 and 23).
Result in Severely Degraded Steam Generator l
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11 NUREG-1604
Tubes," to all holders of operating licenses for PWRs licensees before the next scheduled steam generator (Ref. 8). This notice provided information related to tube inspection outages to ensure that the licensees inspection findings at Maine Yankee. In particular, evaluated all appropriate areas before (or during)
Reference 8 highlighted weaknesses in the licensee's their next scheduled outages. Responses to these analysis of pancake coil data. (A pancake coil is one RAls were received between September 1995 and type of coil that is sensitive to circumferential December 1996, degradation.) After Reference 8 was issued, the licensee performed additional pancake coil inspections Since the issuance of GL 95-03, many PWR licensees during a refueling outage. These inspections led the have inspected their steam generator tubes to identify licensee to identify a number of circumferential indications of circumferentially c-iented degradation.
indications that were larger than anticipated for the Rese inspections have revealed the presence of this 6-to-7-month operating time between the inspections, form of degradation; however, in all cases evaluated These results, in part, led the licensee to perform by the staff, the severity of the indications was additional inspections using a rotating probe with a limited (i.e., the tubes were capable of performing plus-point coil.
He plus-point coil inspections their intended safety function with additional identified many more tubes with circumferential regulatory margins).
indications than had previously been identified with the pancake coil screening criteria. Penetrant testing This report provides background information and the and destructive examination of some tubes that had current status of circumferential cracking in PWR been removed from the steam generators confirmed steam generator tubes. Section 2 describes PWR that several indications identified with the plus-point steam generator designs. Section 3 discusses GL coil were circumferential cracks.
95-03 and presents a generic summary of the staff's assessment of the responses.
Section 4 briefly In consideration of the events at Maine Yankee, the summarizes the nondestructive examination NRC staff issued Generic Letter (GL) 95-03, techniques typically used to inspect steam generator "Circumferential Cracking of Steam Generator tubes for degradation.
Sections 5, 6, and 7 Tubes," on April 28,1995 (Ref. 2). GL 95-03 respectively discuss the plant-specific reviews of the j
requested, in part, that all holders of operating GL 95-03 responses for Babcock & Wilcox (B&W),
licenses or construction permits for PWRs submit a Combustion Engineering (CE), and Westinghouse safety assessment justifying continued operation until plants with steam generators containing alloy 600 the next planned steam generator tube inspections. In tubes.
Similarly, Section 8 discusses the addition, GL 95-03 requested a summary of the plant-specific review for plants with steam generators inspections to be performed during the next steam containing alloy 690 tubes. Section 9 gives the generator tube inspection outage as they pertained to staff's conclusions regarding the issue of the detection of circumferential cracking.
circumferential cracking of steam generator tubes.
By late June 1995, the NRC staff received the This report also includes a number of appendices.
majority of the PWR licensee submittals in response Appendix A discusses the various types of sleeve to GL 95-03. An initial review of these submittals designs and the associated operating experience with and other supporting documentation (e.g., steam respect to circumferentially oriented degradation, generator inspection reports) provided by the various Appendix B lists plants by name, vendor, steam licensees and vendors did not identify any significant generator tube material, and steam generator tube safety concern that would warrant immediate NRC expansion technique. Appendix C lists acronyms and action. However, the NRC review did identify abbreviations used in this report. Appendix D lists additional information that was needed for the staff to the references.
complete its evaluation. As a result, a number of plant-specific requests for additional information This report is intended to be representative of (RAls) were prepared and sent to the licensees. In significant operating experience pertaining to many instances, these RAls asked licensees to circumferential cracking of steam generator tubes explicitly address in their responses all areas from April 1995 through December 1996. Steam potentially susceptible to circumferential degradation.
generator operating experience prior to April 1995 is in most cases, these RAls were provided to the included in this report, as appropriate, for NUREG-1604 12
completeness. The figures provided in this report are l
for illustrative purposes only ar.d are not to scale.
This decwt only summarizes some of the more pertinent information supplied by licensees and their vendors and is not intended to replace the original material. For additional information pertaining to plant 4pecific actions and responses, the reader should refer to the appropriate reference material.
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1 1-3 NUREG-1604
2 STEAM GENERATOR DESIGNS
'Ibe steam generators in PWRs are shell-and-tube end located near the bottom of the recirculating steam heat exchangers that use the heat from the primary generator vessel. The tubes are expanded into the reactor coolant to make steam to drive turbine tubesheet either for the entire thickness of the generators. A typical PWR plant has two to four tubesheet (full-depth expansion) or for a portion of steam generators per reactor, each with several the tubesheet (partial-depth expansion). The tubes thousand tubes (up to 15,531). The primary reactor are supported above the tubesheet with plates or coolant passes through the steam generator tubes and eggerate-type dividers at a number of fixed axial boils water on the outside of the tubes (secondary locations along the tube bundle and with various side) to make dry saturated steam for turbine shaped bars and small plates in the U-bend region of generators and auxiliary systems.
The steam the tube bundle (Ref. 24). All of the steam generator generators also serve as a beat sink for the reactor tubes in the Westinghouse-designed steam generators coolant system under normal, abnormal, and are U-shaped, while the larger-radius tubes in the CE emergency conditions.
steam generators have two 90-degree bends, as shawn in Figure 2-3.
The steam generators are designed to confine radioactivity from neutron activation or fission The once-through steam generators (OTSGs) products to the primary coolant during normal manufactured by B&W use straight heat exchanger operation. However, the prirnary reactor coolant is tubes with a tubesheet at both the top and bottom of at a higher pressure than the secondary coolant, so the tube bundle, as shown in Figure 2-2. The B&W any leakage through defects in the tubes (or PWR plants have two steam generators per plant, each with tubesheets) is from the primary (radioactive) to the approximately 15,500 tubes arranged in a triangular secondary (nonradioactive) side. Consequently, a pattern. The tubes are installed in the tubesheet in defect penetrating through a steam generator tube can the same way as the tubes in the recirculating steam release radioactivity outside the reactor containment generators (i.e., partially expanded and welded). The through the pressure relief valves, the condenser OTSG design has an "untubed" lane, which provides off-gas, or other paths in the secondary system (Ref, access for secondary-side inspections (Ref. 24).
24).
To minimize the potential for release of radionuclides, the steam generator tubes must have adequate integrity.
Steam generators currently operating in the United States are of two major types: the recirculating U-tube type, manufactured primarily by Westinghouse and CE (Figure 2-1), and the l
once-through type manufactured by B&W (Figure 2-2). The following paragraphs briefly describe each of these steam generator types with an emphasis on those aspects that may affect steam generator tube degradation and, in particular, circumferential cracking.
A more detailed description of the different designs is presented in Sections 5.1, 6.1, and 7.1 for B&W, CE, and Westinghouse steam generators, respectively.
Figure 2-1 presents a cross-section of a simplified PWR recirculating steam generator (U-tube type). In this design, the tube bundle consists of many thousands of individual tubes, each welded to a thick plate (called a tubesheet) with a hole for each tube 2-1 NUREG-1604
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Co7Le Figure 2-1 PWR recirculating steam generator NUREG-1604 2-2
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Tube expansion, typical at upper and lower Uppermost span,
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Figure 2-2 B&W once-through steam generator NUREG-1604 2-3
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Figure 2-3 CE stearn generator NUREG-1604 2-4
l 3 GL 95-03, "CIRCUMFERENTIAL CRACKING OF STEAM GENERATOR TUBES" 3.1 Content of GL 95-03 (3) Develop plans for the next steam generator tube inspections as they pertain to the detection of On April 28, 1995, the NRC issued GL 95-03, circumferential cracking. The inspection plans
'Circumferential Crackirg of Steam Generator should address, but not be limited to, scope Tubes," (Ref. 2) to all holders of operating licenses (including sample expansion criteria, if or construction permits for PWRs. The NRC issued applicable), methods, equipment, and criteria GL 95-03 for three principal reasons:
(including personnel training and qualification),
(1) Notify addressees about the safety significance To document the outcome of these actions, the NRC of the recent steam generator tube inspection staff requested that addressees prepare and submit the findings at Maine Yankee Atomic Power following:
Station.
(1) a safety assessment justifying continued (2) Request that all addressees implement the operation, predicated on the evaluation actions described within the generic letter.
performed in accordance with requested actions I and 2 (above)
(3) Require that all addressees submit to the NRC a written response regarding implementation of (2) a summary of the inspection plans developed in the requested actions.
accordance with requested action 3 (above) and a schedule for the next planned inspection In addition, GL 95-03 alerted addressees to the importance of performing comprehensive 3.2 Generic Assessment of the industry's examinations of steam generator tubes using Remonses to GL 95-03 techniques and equipment capable of reliably detecting the types of degradation to which the steam The NRC's review of the plant-specific responses to generator tubes may be susceptible. The staff also GL 95-03 specifically addressed the two areas listed noted that the performance of steam generator tube above.
examinations is controlled, in part, by Appendix B to Title 10, Part 50, of the Code offederal Regulations la evaluating thejustification for continued operation, (10 CFR Part 50).
the staff considered the following factors:
In GL 95-03, the NRC staff also requested that (1) past inspection scope and results licensees take the following actions:
(2) operating parameters (such as the hot-leg (1) Evaluate recent operating experience with operating temperature (i. e., T-hot) and j
respect to the detection and sizing of cumulative steam generator operating time) circumferential indications to determine the applicability to their plants.
(3) tube material and properties (e.g.,
heat treatment and stress relieving actions, tube (2) On the basis of the evaluation in item (1) loads, heat of material, tube location) above, as well as past inspection scope and results, susceptibility to circumferential (4) past steam generator operating experience cracking, threshold of detection, expected or inferred crack growth rates, and other relevant (5) schedule for the next steam generator tube factors, develop a safety assessment justifying inspection continued operation until the next scheduled steam generator tube inspections.
3-1 NUREG 1604
j
\\
r (6) defense in-depth. measures (such as steam generators. To the knowledge of the NRC, no primary-to-secondary leakage monitoring and circumferential cracking has been observed at these emergency operating procedures) locations. Tubes have also been expanded at the hot-leg tube support plate elevations at Byron I and (7) risk and potential consequences of a range of Braidwood I for implementation of a voltage-based steam generator tube rupture events tube repair criterion that applies to axially oriented tube degradation. Dese tubes are not in service To facilitate its reviews, the NRC divided the (i.e., they are plugged), although they are relied upon j
industry's steam generators into several groups.
to limit the motion of the tube support plate under j
First, plants with alloy 600 steam generator tubes postulated accident conditions.
Similar tube
{
were distinguished from plants with alloy 690 steam expansions have also been performed in several CE generator tubes. The plants with alloy 690 steam units at drilled tube support plate locations. These generator tubes were treated as a single distinct tubes are not in service.
j group. The plants with alloy 600 steam generator tubes were further subdivided on the basis of steam For OTSGs, locations with axial stresses include generator vendor (B&W, CE, or. Westinghouse).
expansion transition locations, dented regions, and he Westinghouse alloy 600 steam generators were sleeve joints. For these steam generators, only a then subdivided on the basis of the type of expension limited number of service-induced circumferential transition (i.e., partial-depth roll expansion, full-depth indications have been observed at these locations.
roll expansion, explosive expansion, and hydraulic However, at TM11 a significant number of expansion),. since most circumferential indications circumferential indications near the expansion
. bave been observed at this location. The staff elected transition in the UTS region were observed as a l
to group the plants on the basis of expansion result of a significant chemistry excursion when the transition type, since plants with similar expansion unit was shutdown, as discussed in Section 5.1.2.
l transitions typically are of the same vintage. All CE Most of the circumferential cracks in B&W OTSGs plants with alloy 600 steam generator tubes have occur in the lane / wedge region as a result of high tubes which were explosively expanded the full length cycle fatigue.
(i.e., depth) of the tubesheet. All B&W OTSGs have tubes with partial-depth roll expansions, with the In order to evaluate the acceptability of each PWR exception that Three Mile Island I (TMI-1) has licensee's justification for operation to the next steam kinetic expansions in the upper tubesheet (UTS) as a generator tube inspection, the NRC reviewed the i
result of a significant chemistry excursion which is licensee's initial GL 95-03 response. This initial l
discussed in Section 5.1.2.
review focused on determining whether the licensee had adequately monitored all locations that are To evaluate the GL 95-03 responses, the staff first potentially susceptible to circumferentially oriented determined which locations were most likely to degradation (with particular emphasis on the exhibit circumferential cracking. (De staff based expansion transition, since the majority of this determination primarily on operating experience.)
dreumferentially oriented degradation occurs at this Circumferential cracking of steam generator tubes beation).
normally occurs at the tube locations where significant axial stresses are present.
In instances where locations were susceptible to circumferential cracking and monitored with For recirculating steam generators, these locations techniques capable of reliably detecting such include expansion transition locations, dented degradation, the NRC reviewed the licensee's locations, the U-bend region of tubes (particularly inspection results. In all of these instances, either the
)
small-radius U-bend tubes), and sleeve joints.
results were acceptable, or the licensees had already Arnong the entire PWR steam generator population, implemented the appropriate action (e.g., limited the circumferential cracking has been observed at all of amount of time between tube inspections).
these locati<ms. Other portions of tubes where axial stresses occur include tubes expanded into the tube In instances where locations susceptible to support plate. %ese tubes are normally located on circumferential cracking were either not monitored os the cold-leg side of Westinghouse Model D4 and D5 were not inspected with techniques capable of.eliably NUREG-1604 32 1
i detecting such degradation, the staff evaluated various the staff held discussions with the licensees regarding parameters to determine whether the justification for their inspection scope and results before, during, and continued operation was acceptable.
These after the steam generator tube inspection outages to evaluations considered such parameters as the provide additional confidence that the licensees had operating ties of the steam generators at the plant, performed comprehensive stea m generator tube the operating experience of the particular steam mapections including a detailed evaluation of the generators and similarly designed steam generators, inspection results. The results of the NRC evaluation the time interval until the next steam generator tube are contained in Sections 5 through 8. These sections mspections, and the leakage monitoring program for primarily contain operating experience related to the plant.
circumferentially oriented tube degradation; however, these sections also contain information pertaining to On the basis of these reviews, the NRC concluded the scope of the next inspectian.
that, in all cases, sufficient technical justification existed tojustify operation of the units until the next steam generator tube inspection or until replacement of the steam generator could be implemented.
s The staff then used a similar approach to evaluate each licensee's mspection plans for the next scheduled steam generator tube inspection. That is, the staff reviewed the operating experience of the plant and similarly designed plants, as well as the scope of the licensee's inspection program, to ensure that all locations susceptible to circumferentially oriented degradation were being monitored or that there was adequate justification not to monitor those locations at this time. (For example, a newly replaced steam generator with limited operating time would not necessarily require a full examination of all tube locations susceptible to circumferential cracking with a technique capable of detecting such degradation.)
As discussed above, the staff reviewed the information provided by the various licensees and vendors and did not identify any significant safety concern that would warrant immMiate NRC action.
However, the review did ' identify additional information the staff needed to complete its review.
As a result, a number of plant-specific requests for additional information (RAls) were prepared and sent to the licensees In many instances, these RAIs focused on having the licensee explicitly address all areas potentially susceptible to circumferential degradation.
In most cases, these RAIs were provided to the licensees before the next scheduled steam generator tube inspection outages to ensure that the licensees evaluated all appropriate areas before (or during) their next scheduled outages.
Responses to the RAIs were received from September 1995 through December 1996. In many instances, 3-3 NUREG-1604
i
(
4 INSPECTION, REPAIR, AND ASSESSMENT OF STEAM GENERATOR TUBES 4.1 Purnose of S**== Gaaa ator Tube fa==artions previous inspections have revealed minor degradation. These intervals are reduced or extended De requirements for the mapection of steam on the basis of the categorization of inspection generator tubes are intended to ensure that this results, as defined in the plant's technical portion of the reactor coolant system maintains its specifications.
structural and leakage integrity. Structural integrity refers to maintaining adequate margins against gross Although the technical specifications include a general failure, rupture, and collapse of the steam generator provision to extend surveillances by 25 percent of the tubes.
Leakage integrity refers to limiting specified interval, this provision is not considered primary-to-secondary leakage during normal applicable to steam generator tube inspections; the operation and postulated accidents to within above criteria indicate the only conditions under acceptable limits.
which the surveillance interval for steam generator tube inspections may be changed. This position was The structural criteria that the tubes are intended to delineated in NRC GL 91-04, " Changes in Technical meet are specified in Regulatory Guide 1.121, " Bases Specification Surveillance Intervals to Accommodate for Plugging Degraded PWR Steam Generator Tubes" a 24-Month Fuel Cycle,' issued on April 2,1991 (Ref. 33).
Adequate leakage integrity during (Ref. 3). As a practical matter, however, utilities transients and postulated accidents is demonstrated by generally perform inspections at each refueling showing that the resulting leakage from the tubes will outage, which typically occurs every 12 to 18 not exceed a rate that would violate offsite or control months.
room dose criteria. Dese criteria are specified, in part, in 10 CFR Part 100 and in General Design Since the purpose of the steam generator tube Criteria 19 of Appendix A to 10 CFR Part 50.
mspections is, in part, to ensure adequate structural and leakage integrity of the tube bundle, more To provide assurance of adequate structural and frequent inservice inspections may be required, leakage integrity, licensees perform inservice depending on the severity of the indications detected, inspections of the steam generator tubes. These To ensure that the frequency was adequate for the mspections are intended to detect mechanical or prior cycle, licensees for PWRs should assess the corrosive damage to the tubes, which may result from inspection results following every outage to ensure manufacturing and/or inservice conditions.
In that the tubes, individually and collectively, retained addition, the inservice inspections of the steam adequate structural and leakage integrity. This type generator tubes provide a means of characterizing the of assessment is typically referred to as ' condition nature and cause of any tube degradation so that monitoring. " In addition, licensees should project the corrective measures can be taken. Tubes that show condition of the tubes from the current inspection to an indication of degradation that exceeds the tube the next inspection to ensure that the tubes will retain repair limits specified in a plant's technical adequate integrity for the next operating interval, specifications are removed from service (by plugging)
This type of assessment is typically referred to as an or are repaired (by sleeving), as discussed in Section
" operational assessment.' These assessments should 4.3.
be performed since the inspection frequencies and tube repair criteria specified in the technical Re frequency of the inservice inspections of the specifications were established on the basis of specific tubes in at least one steam generator is generally assumptions concerning various parameters such as every 12 to 24 calendar months, as specified in a the forms of degradation to which the tubes may be plant's technical specifications.
The specified susceptible (if any), limitations of nondestructive maximum interval may need to be reduced to every examination techniques, and the rate of steam 20 months in cases where previous inspections have generator tube degradation.
If any of these shown extensive degradation, and may be incremaad parameters exceed what was assumed during the to as much as every 40 months in cases where development of the inspection intervals, the basis for 4-1 NUREG-1604
)
i the inspection frequency and tube repair criteria e a general inability to permit characterization of would no longer he considered valid.
identified degradation (e.g., axial, circumferential, or volumetric; single or multiple axial indications; In summary, the inse.rvice mspection of steam etc.)
generator tubes should be conducted at appropriate intervals, such that the structural and leakage
- relative insensitivity to detecting circumferentially integrity of the steam generator tubes is maintained oriented tube degradation with appropriate margins. These inspections should be adequate to detect degradation (such as e limited capability to detect degradation in regions circumferential cracking) at a sufficiently early stage with geometric discontinuities (e.g., expansion to preclude the progression of the degradation to the transitions, U-bends, and dents) and deposits point that the regulatory criteria regarding steam generator tube structural and leakage integrity can no In the late 1970s and early 1980s, other forms of tube longer be met during the interval between degradation began to emerge. These new forms of inspections.
tube degradation, in combication with the limitations of the bobbin coil discussed above, led to the 4.2 Eddy Current Testina development and refinement of additional eddy current inspection techniques.
Some of these Eddy current testing (EC) is the primary means for improvements were also a result of advancements in mapecting steam generator tubes.
This method computer technology, involves inserting a test coil inside the tube and pushing and pulling the coil so that it ernernes the Among these emerging improvements in inspection tube length. De test coil is then " excited" by technology, multifrequency techniques enabled test alternating current, thereby creating a magnetic field coils (such as the bobbin coil) to be " excited" at that induces eddy currents in the tube wall.
multiple frequencies rather than a single frequency.
Disturbances of the oddy currents caused by flaws in Multifrequency techniques were an improvement in the tube wall produce corresponding changes in the eddy current technology since they could more electrical impedance as seen at the test coil terminals.
readily detect very small-volume flaws (such as Instruments are used to translate these changes in test intergranular attack, stress corrosion cracks, fatigue
]
coil impedance into an output that can be monitored cracks, and small pits) that were traditionally hard to by the data analyst. The depth of certain types of detect with the single-frequency ECT methods. This flaws can be determined by the observed phase angle improvement in detection capabilities resulted, in response of this output signal. The test equipment is part, from the ability to isolate the signal produced by calibrated using tube specimens containing artificially a defect by suppressing the changes produced by induced flaws of known depth.
Geometric other unwanted parameters such as support plates and discontinuities (such as the expansion transition and denting. Because the responses from the defect dents) and support structures (such as the tubesheet signal and those produced by the unwanted and tube support plates) also produce eddy current parameters are frequency-dependent, the analyst can I
signals, making it very difficult to discriminate defect essentially subtract the unwanted signal from the signals at these locations. Reference 24 contains a signal consisting of the defect and the unwanted discussion regarding some of the basic principles of signal. The signals used in this signal mixing process EC.
are from two different frequencies.
In the 1970s, single-frequency bobbin coil probes During the late 1970s and early 1980s, both (illustrated in Figure 4-1) were typically the only circumferential and axial cracking mechanisms were probes used to inspect steam generator tubes. This beginning to emerge as the dominant degradation probe was adequate for detecting and characterizing mechanism. At the same time, the industry was wastage, which was the dominant form of developing probes that could more readily detect degradation otnerved at that time, ne bobbin coil these forms of degradation. For example, in the probe permits a rapid screening of the tube for 1982 time-frame, a 4x4 differentially linked surface degradation; however, it has several limitations:
riding pancake probe was used at Palisades to detect circumferentially oriented tube degradation.
In NUREG 1604 42
subsequent outages, more advanced probes such as axially oriented degradation). In late 1994 and early the 4C4F probe were used. The 4C4F probe was 1995, coils such as the plus-point coil and similar to the 4x4 probe except that it could traverse high-frequency pancake coil were used in the field.
the bend in the tubes. With the pancake coil used in The plus-point coil reduced volumetric influences and these probes, flaws of any orientation (axial, was sensitive to both axially and circumferentially circumferential, or branched) can be detected since oriented degradation. It was originally developed for the eddy currents induced into the tube are impeded surface examination of reactor vessel welds and was by flaws of any orientation, designed to reduce geometry and permeability effects.
(A plus-point coil on a rotating probe is depicted in In the mid-1980s, additional inspection probes were Figure 4-3.) This probe configuration differentially developed and implemented in the, field. These pairs the axial and circumferential coils in one probes included the 8x1 pancake array probe and the gimbal-mounted surface riding coil shoe, as depicted rotating probe with a pancake coil (typically referred in Figure 4-3, to reduce the effects of geometry to as a rotating pancake coil probe), illustrated in variations in the tube (Ref. 36).
Figure 4-2.
The licensee for Farley I and 2 was among the first to perform extensive examinations of Currently, the types of coils included on the rotating the expansion transition region of the steam generator probe head depend on the user's needs.
For tubes. After conducting these examinations using the example, the user may request that the rotating probe 8xl probe; however, the licensee indicated that the head include a mid-range pancake coil, a signal-to-noise characteristics of this probe were of high-frequency pancake coil, and a mid-range such an order that little improvement over bobbin phis-point coil, in addition, the licensee may specify probes was realized (Ref 35). As rotating pancake whether the coil is magnetically biased or not.
coil probes evolved during the 1980s, they were Magnetically biased coils can reduce interfering initially used on a limited basis to characterize signals (i.e., permeability variations). The coils used various bobbin coil indications such as from the in the rotating probe head at a specific plant depend tubesheet region. When it became practical to use on many factors including optimizing the coils for rotating pancake coil probes for large inspection detecting the forms of degradation to which a tube programs, some utilities began inspecting 100 percent may potentially be susceptible, of the tubes at the expansion transition region.
Each of the above mentioned test coils can be Since the late 1980s and early 1990s, additional designed and driven at specific frequencies to ensure improvements have been made in eddy current an. optimal inspection of the tubing. In general, inspection probes. Currently, inspections of steam lower frequencies are better for detecting degradation generator tubes generally employ both a bobbin coil initiating from the outside diameter of the tube, while probe and an additional probe, such as a rotating higher frequencies are better for detecting degradation probe, a Cecco probe, or both. The bobbin coil initiating from the inside diameter of the tube. The probe permits rapid screening of the tube for advantages of the rotating probes are that they are degradation and can be pulled through a tube at sensitive to circumferentially oriented degradation speeds exceeding 40 inches per second, while the (which the bobbin probe is nos), can better rotating probes and/or Cecco probes circumvent some characterize the defect, and are less sensitive to of the limitations of the bobbin coil (discussed geometric discontinuities. The major disadvantage of above).
the rotating probes is their slow inspection speed (typically less than 1 inch per second). Because of Early versions of the rotating probes generally this slow inspection speed, rotating probes are only contained one to three specialized test coils, which used at specific locations (e.g., U-bends, sleeves, usually included at least one pancake coil that was expansion transitions, dents, locations where there is sensitive to both axially and circumferential!y a bobbin coil indication, and locations where a more oriented degradation. Other test coils used on these sensitive inspection ia needed),
early rotating probe heads (early 1990s), if any, included an axially wound coil (which was sensitive Cecco probes (Cecco 3 and Cecco 5) operate to circumferentially oriented degradation) and/or a differently from rotating probes. For example, a I
circumferentially wound coil (which was sensitive to Cecco probe (depicted in Figure 4-4) contains 43 NUREG-1604
multiple transmit and receive coils (rather than just a failure provided by such support structures as the combined coil) and the Cecco probe is not rotated as tubesheet and tube support plates.
it is pulled through the tube. Like the rotating probes, the Cecco probes are sensitive to Because of its conservative basis, the depth-based circumferentially oriented degradation; however, limit tends to be inappropriate for highly localized characterization of degradation with these probes is flaws (such as stress corrosion cracks) and flaws currently limited. A major advantage of the Cecco within the tubesheet. As a result, the industry has probes is that they are capable of a much higher developed, and the NRC has approved, various inspection speed (i.e.,12 to 15 inches per second) alternative forms of repair criteria for specific forms than the rotating probes.
of steam generator tube degradation. (For example, the GL 95-05 (Ref.1) voltage-based limits are used 4.3 Tube Repairs for predominantly axially oriented outside diameter stress corrosion cracking (ODSCC) at tube support ne plant technical specifications set plugging and plate elevations, and the F-star limits are used for repair limits for the maximum allowable wall degradation confined within the tubesheet below the degradation beyond which the tubes must be removed expansion-transition.)
from service by plugging or repaired by sleeving.
Tube degradation is typically discovered during 4.3.1 Plugging scheduled inservice examinations of steam generator tubes, and tube repair (plugging or sleeving) is The plugging technique involves installing plugs at required for all tubes with indications of tube the tube inlet and outlet. After plugging, the tube no degradation exceeding the tube repair limits. All longer functions as the boundary between the primary plants have a depth-based repair limit that is and secondary coolant systems.
applicable to all forms of steam gercrator tube degradation. Alternatives to this deg.n-based limit 4.3.2 Sleeving have been approved; however, no alternative tube repair limits have been approved for circumferential To prolong the life of severely degraded steam indications (in part because no qualified method exists generator tubes, some utilities, with prior NRC for characterizing circumferentiel indications in a approval, have repaired defective tubes by sleeving, manner that can be reliably related to the integrity of as discussed in Appendix A.
After sleeving, the those indications that may be left in service). The repaired tube may remain in service, depth-based repair limit varies from plant to plant, but is typically 40 percent of the tube wall thickness.
4.3.3 Replacement (That is, tubes with indications of degradation greater Aan or equal to 40 percent must be plugged or To avoid the need for derating the plant and the repaired.)
excessive downtime associated with inspecting and repairing steam generator tubes, some utilities are The plugging and repair limits are established on the coneidering or have elected to replace severely basis of the minimum tub wall thickness necessary degraded steam generators. The decision to replace to provide adequate structural margins (in accordance a steam generator is largely economic rather than with Reference 33) during normal operating and technical. Several utilities have decided to replace postulated accident conditions. These limits allow for their original steam generators, others have chosen to eddy current error and incremental wall degradation operate the plant with the original steam generators that may occur before the next inservice inspection of until it is no longer economically viable to operate the the tube. nese plugging and repair limits are plant. The duration of an outage for steam generator conservatively established according to an assumed replacement varies; however, at the present time, mode of degradation in which the walls are uniformly steam generators can be replaced in 2 to 3 months.
thinned over a significant axial length of tubing.
These limits do not consider additional structural To minimize the potential for the modes of tube margins associated with defects such as small-volume degradation that have been identified to date, the thinning and pitting, and they do not consider the replacement steam generators currently being installed external structural constraints against gross tube contain the following improvements:
NUREG-1604 44
1 1
(1) stainless steel lattice grid tube supports -
with the pancake coil at Maine Yankee contributed to Stainless steel is more corrosion resistant than the issuance of GL 95-03.
However, it is not carbon steel resulting in a lower potential for necessarily the number of indications which is the tube support corrosion which can lead to tube prime concern, but the severity of those indications.
denting and stress corrosion cracking. A lattice In the case of Maine Yankee, evaluation of the grid tube support configuration can result in a inspection findings was completed after GL 95-03 reduced amount of contact area between the was issued. This evaluation indicated that the tubes tube and the tube support when compared to had adequate integrity (structural and leakage drilled hole tube support plates. This lower integrity). These results indicate that even though the contact area can result in greater secondary ECT techniques at Maine Yankee had limitations, coolant flow around the tubes which can lower they enabled the licensee to detect the flaws at an the potential for accumulating corrosive early enough stage in the prior outage such that they deposits in the tube-to-tube support crevice.
did not become significant (i.e., from a stmetural and leakage integrity standpoint) following the operating (2) thermally treated alloy 690 tubes - These interval between the inspectims. However, it should features (i.e., thermal treatment and alloy 690 be noted that the inspection interval at Maine Yankee tubes) reduce the potential for SCC throughout was only approximately 6 months.
the tube bundle. Thermal treatment involves subjecting the mill-annealed tubes to a final In summary, tube integrity can be ensured given a heat treatment, thereby relieving fabrication knowledge of the limitations of the techniques stresses and further improving the tube's employed at a specific plant (including the procedures microstructure and its corrosion resistance.
and analysts) and the implementation of appropriate restrictions on operating parameters (e.g., bot-leg (3) full-depth hydraulically expanded tubes -
temperature, water chemistry, and operating interval Hydraulically expanding the tubes the full duration). Nonetheless, it is important to note that length (i.e., depth) of the tubesheet eliminates the techniques used at one plant may not ensure tube the crevice between the tube and the tubesheet integrity at another plant.
This is a result of and reduces the stresses at the expansion plant-specific circumstances, including the transition (when compared to other expansion susceptibility of the tubes to degradation, the growth techniques such as mechanically rolling the tube rate of the degradation, the nature of any signals that into the tubesheet),
may interfere with the eddy current data analysis, and the frequency of inspections.
Appendix B lists plants that have replaced steam generators.
As a result of these plant-specific circumstances affecting the ability to ensure tube integrity, it is 4.4 Summary important to implement the following precautions:
Inspections of steam generator tubes at operating
- Optimize examination methods to minimize noise plants have demonstrated the capability to reliably and interfering signals in order to maximize flaw detect certain forms of degradation (such as tube detection, wear and wastage). However, the threshold for reliebte detection of other forms of degradation (e.g.,
- Evaluate the influence of interfering signals on SCC) continues to be an issue. In addition, the flaw detection, reliable sizing of indications, both above and below the tube repair limit (s), remains an issue since very
- Evaluate the examination and analysis procedures few degradation mechanisms can reliably be sized.
to maximize the discrimination of flaws from unavoidable noise and interfering signals.
Limitations in the apparent ability of eddy current techniques to detect degradation led, in part, to the
- Evaluate unique plant-specific circumstances issuance of GL 95-03. Specifically, the identification necessitating special examination techniques or of many more indications of circumferential!y analysis procedures.
oriented tube degradation with the plus-pointcoil than 4-5 NUREG-1604
Furthermore, it is important to have (1) site-specific Assessment of tube integrity is necessary to ensure steam generator eddy current data analysis guidelines that the operstmg interval between inspections is that afford the data analyst valuable information on appropriate. This===== ment should be performed the unique conditions present at the plant; (2) regardless of whether a tube, or a population of site-specific training and performance demonstration tubes, is repaired. Since pressure testing of tubes programs for all steam generator data analysts; (3) and tube pulls can not readily be performed for all qualified inspection techniques to detect known or tubes, aampling of the worst indications is necessary.
postulated tube degradation mechanisms and to However, given the inability to reliably size address the conditions that could exist in the steam indications, questions arise regarding the ability to generator tube bundle; (4) a protocol for ensuring determine which indications to test. Fortunately, adequate inspections (e.g., Lah;=ahat two-party ECT data analysts are capable of qualitatively review of the data; briefing analysts during the course assessing the relative severity of indications of tube of the outage when their evaluation differs from the degradation.
That is, a limited population of final resolution of the indication) to reduce the indications can be identified as being qualitatively probability of flaw signals being missed and not more severe than another set of indications although reported; (5) a tube removal program to validate absolute determination of the neverity of the nondestructive examination measurements, to assess indications in this limited population can not be tube structural conditions, and to determine the cause made.
of the degradation mechanism; and (6) an assessment of the significance of the nondestructive exanunation Tubes removed and analyzed from Byron 1 in the results with regard to safe, reliable plant operation.
1994-1996 time-frame confirmed that it is possible to qualitatively assess the severity of indications. This As mentioned above, limitations exist in the ability to permits licensees to select several (e.g.,10) of the size certain forms of degradation. The inability to most oevere tubes for further evaluation (e.g., tube reliably size degradation with current ECT methods pulls, pressure testing) in order to assess whether results in the following costly and undesirable tube integrity margins were met. However, caution actions:
must be used when comparing results from outage to outage since differences in acquisition techniques, the
- Tubes may be repaired upon detection of a effects of deposits, and other interfering signals may l
degradation mechanism that can not be reliably change from outage to outage. Consequently, it may sized, since it is unknown whether the indication be necessary to perform additional pressure tests exceeds the tube repair limits, and NRC and/or tube pulls in order to confirm the inspection regulations require the use of qualified methods results and the degradation morphology, for performing nondestructive examination.
- Other techniques (such as ultrasonic testing and penetrant testing) may be required to further characterize the degradation and/or to ensure that tube integrity has not been compromised.
l e In situ pressure testing of the tubes may be l.
necessary to demonstrate that an indication has not j
compromised tube integrity.
I e Tube pulls (removal of tube specimens) may be necessary to confirm the degradation mechanism, identify the root cause of the degradation, assess tube integrity, and/or assess inspection reliability (to detect and size degradation).
i NUREG-1604 4-6
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Probe Body Bobbin Coits
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5 BABCOCK AND WILCOX ONCE-TIIROUGII STEAM GENERATORS 5.1 Design partial-depth expansion of the tubes resulted in a tube to tubesheet crevice of approximately 58 cm (23 The designs of the currently operating B&W OTSGs inches). The entire steam generator was heat treated are very similar; however, there are some differences after tube installation to reduce residual stresses from between the various OTSGs as a result of design tube fabrication and installation and to increase considerations or the operation of the steam resistance to primary water stress corrosion cracking generator.
Generic features of the OTSGs are (PWSCC) by developing more carbides at the grain discussed in Section 5.1.1, while features unique to boundaries. Ilowever, this process also resulted in a given plant (or set of plants) are discussed in sensitization (chromium depletion at grain boundaries), making the tubing susceptible to other Section 5.1.2.
forms of corrosion (including SCC in oxidizing acidic 5.1.1 Generic Features conditions).
j The B&W OTSGs have straight heat exchanger tubes The sensitized, mill-annealed alloy 600 tubes are with an upper and lower tubesheet (UTS and LTS),
supported and aligned between the tubesheets by 15 as shown in Figure 2-2. Primary coolant is pumped carbon steel tube support plates. The first 14 tube through the tubes from top to bottom, while the support plates have broached holes to provide secondary coolant moves around the outside of the three-point support (i.e., broached trefoil design) and tubes from bottem to top in a counter-flow direction.
a space for axial flow around each tube, as depicted The secondary-system water enters a feed annulus in Figure 5-2. The highest (fifteenth) tube support above the ninth tube support plate, where it mixes plate has approximately 1600 drilled holes in the with steam aspirated from the tube bundle area and is outer periphery of the plate, with the remaining holes preheated to saturation. The saturated water flows being of the broached trefoil design, down the annulus, across the LTS, and up into the tube bundle where it becomes steam. Boiling occurs The B&W OTSG design has an "untubed" lane, about two-thirds of the way up the tube bundle. The which can be used to provide access for secondary area above the boiling region is the steam region in side inspections.
This untubed inspection lane which the tubes are exposed to steam. The steam is extends from the periphery to the center of the tube superheated in this region. After passing through the bundle along Row ~76 of the tube layout, as superheated region near the UTS, the steam flow conceptually depicted in Figure 5-1.
shifts from vertically upward to radially outward near the top of the tube bundle and then down the annulus 5.1.2 Plant-Specific Features to the steam outlet connection.
Most of the secondary coolant is completely evaporated in a There are unique features that affect steam generator single pass through the steam generator (Refs. 24 and tube integrity in several of the B&W OTSGs. These features primarily involve differences in, or near, the 31).
expansion transition region (i.e., the transitional There are currently seven plants operating with B&W region of the tube where the tube changes from fully OTSGs. These have two steam generators per plant, expanded against the tubesheet to non-expanded).
each with approximately 15,500 tubes arranged in a The following paragraphs describe some of the more triangular pattern as shown in Figure 5-1. The tubes pertinent features that may affect a tube's were made from mill-annealed alloy 600 and have a susceptibility to circumferential cracking or its nominal outside diameter of approximately 15.875 inspectability, millimeters (mm) (0.625 inch) and a wall thickness of 0.889 mm (0.035 inch). During manufacture, each in many of the B&W OTSGs, there is a limited tube was mechanically roller expanded approximately population of tubes that were not stress relieved at the 25 mm (1 inch) into the primary face of both the expansion transition. During the manufacturing UTS and LTS. Since the tubesheet has a thickness of process, several tubes were re-rolled following the approximately 61 centimeters (cm) (24 inches), this full furnace stress relief to temporarily seal the tube NUREG-1604
)
5-1 l
I during the shop hydrostatic tests (i e known non-stress
.., less than 200 relieved tubes exist).
carly 1980s.
expansions did not have a post-roll stres These In which makes them more November OTSGs during a reactor coolan s relief,
- 1981, cracking.
likely candidates for n both test. Subsequentdetailedexaminations rostatic Crystal River 3) have tubesheet bore sizeFour of the operat defective tubes.
revealed many portions of tubes confirmed that the tub ncluding
+0.15/-0.05 mm (0.635 +0.006/-0 002 i h) were initiated from the primary side (i s of 16.13 ures diameter) of the tubes in the form of c are smaller than those of the other B&W unit nc which
.e., inside licensee for Crystal River 3 expects thes intergranular stress corrosion cracks
- s. He m erential hole sizes resulted in a lower strain to me h chemical irnpurity causing the corrosion was s The active e smaller roll the 15.875 mm (0.625-inch) outside di reduced forms, which had been c anically u urin OTSG tube into ameter introduced into the reactor coolant inadvertently manufacturing process.the tubesheet sodium thiosulfate that had been used in during the system from being equal, the licensee believes that dWith all other parameters containment spray additive for iodine remov l past as a such as PWSCC would be expected to occ 7.6 cm (2 to 3 inches) of the primary egradation
- a. The areas of higher residual stresses (Ref 37) ur first in
. to transition), As a result, some tubes wc e of the 61 River 3 are somewhat unique. Several tubehe expansion tra g nal roll (explosively) expanded part-length tha ere plugged, rystal the UTS of the "B" OTSG were dam ends in netically of the breakup of a burnable poison rodaged as a result (Refs. 29 and 38).to create a new load carry (i.e., a loose part) in 1978 and 1979 (R f These m tingjoint assembly 37).
transitions were not stress relievedkinetic expansion tubes (i.e., the weld betweerThe resulting damage to the seal weld of th e s. 31 and e
from these tubes. primary face of the tubesheet) caused some l k. the tube and the involves the "untubed" laneAnother unique tubes to leak, a tube end repair wasTo reduce the potential for these TSGs ea age and LTS of the TMI-l"A"OTSG werSpecifically, the UTS the course of several outages His pperformed over before implementation of the open lane d i e dnsfed diameter which would more readilyof drilling out the flattened order to accommodate the untubed lane rocess consisted es gn. In The LTS holes were plugged by ins n inside
, the UTS passage of an eddy current probe. Tube end r permit the ee plugs.
thick-walled tube section that ext were followed by a field re-rollingperformed during refueling out epairs g a short tubesheet. He primary side of the tubp ends from the
)
stress relief wasperformed following the fieldperformed over the
, which was ote the tubesheet and seal welded using the e was rolled to No In 1996, the repair process was revised su h h protruding above the tubesheet, w same process re-roll.
general, the seal weld wu left intact eliminati c tat,in e u e, need for the field re roll.
welded end cap (Ref. 39).
ng the repair process was adopted because ofHis modification to the a
In addition to the unique plant-specific f susceptible to circumferential crackingregarding the creation of concerns been explosively expanded full leng entially eatures refueling outage field re-rolling. These nine re-rolls were10 (1996), only nine tube During ave UTS and LTS.
s required explosive expansion transition may differ froS ot the es o result of ov performed repair process, er-drilling of the tube end during the in a typical roll-expanded transition m those susceptibility of these locations to degradation, the relative The TMI-! OTSG expansion transitions in th U different.
may be are also significantly different than th e TS roll-expanded transitions in the other B&W OTS e typical as a result of a significant chemical excursi Gs, on in the NUREG-1604 5-2
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1 5.2 Locations Susceptible to circumferential in the tube bundle, contributing to local degradation.
Craciang De fatigue is vibration induced as a result of steam I
exiting the tube bundle (Ref. 28). He steam flow in At the time of the original respnoses to GL 95-03, the uppermost tube span is primarily across the tubes the B&W owners and licensees identified four regions as it turns to exit the bundle (Ref. 38).
as being potentially susceptible to circumferential e
cracking. Specifically, these regions included the Circumferential cracks in the lane / wedge region have uppermost span of tubes in the lane / wedge region; resulted in a number of outages as a result of tube j
tube-to-tubesheet expansion transitions, including leaks; however, no tube ruptures have occurred as a approximately 189 tubes that have not been result of this degradation mechanism. Defective stress-relieved and are therefore expected to be more tubes with circumferential indications are plugged and susceptible to cracking; tube kinetic expansion frequently stabilized from the UTS to prevent damage i
transitions that have not been stress relieved (TMI-1 to an adjacent tube should the degradation in the UTS only); and tube / sleeve expansions at sleeve roll plugged tube continue to the point that the tube severs transitions (Ref. 38).
(Ref. 28).
(
5.2.1 Uppermost Span of Unsleeved Tubes in the ne uppermost span of some tubes in the lane / wedge Lane / Wedge Region region were sleeved. Sleeving has been performed in all operating B&W OTSG plants either to repair In June 1995, when the original GL 95-03 responses defective tubes or to stiffen the tubes to limit the j
were submitted, the B&W Owners Group reported potential for tube failure as a result of high cycle that distinct service-induced circumferential cracking fatigue. As discussed in Appendix A, these sleeves in B&W units had only been observed in the are mechanical sleeves which are approximately uptwrmost span of unsleeved tubes in the lane / wedge 203 cm (80-inches)long. The sleeves are installed in region. He uppermost span includes the portion of the uppermost span of the tube, which is the location a ses between the uppermost (fifteenth) tube support operating experience has indicated is susceptible to plate and the upper tubesheet secondary face (UTSF).
high cycle fatigue. He discussion above pertains to unsleeved tubes in the lane / wedge region.
The lane / wedge region has been defined as a region of tubes abutting the outer half of the inspection lane 5.2.2 Expansion Transitions and Crevice Region (i.e., the "untubed" or open lane) and then flaring out into a wedge shape near the periphery of the tube Circumferential indications have been detected in bundle. The lane / wedge region, graphically depicted B&W OTSGs at the expansion transition and in the in Figure 5-1, includes about three rows of tubes on tubesheet crevice region. For the circumferential either side of the inspection lane and a few additional indications detected at the expansion transition, the rows at the periphery of the tube bundle. The indications were not service-induced. However, for lane / wedge region, however, has not been defined in the circumferential indications detected in the crevice the technical specifications for all plants, and the region, both service-induced and non-service-induced tubes generally referred to as lane / wedge tubes in a indications were observed, although the j
plant's technical specifications may vary from plant service-induced indications were associated with to plant.
intergranular attack. (Intergranular attack is typically volumetric in nature and more readily detectable with Many of the circumferential cracks that occur in this the bobbin coil than distinct circumferential cracks.)
region are corrosion initiated. These initiations sites ne following paragraphs provide additional details propagate through-wall and continue circumferentially related to these events.
in both directions around the tube as a result of a high-frequency, low-stress fatigue mechanism.
Non-service-induced circumferential cracks were Corrosion is initiated by concentrated chemical observed in the TMI-l steam generators as a result of species ca -ied by small water droplets in the a chemical excursion in the early 1980s (as discussed inspection ine and then deposited on the tubes. %e above). Dese circumferential cracks occurred while open lane is a cooler area of the generator, which in the unit was shut down (i.e., they were not turn allows wet steam to deposit contaminants higher service-induced) and were the reason for repairing the 5-3 NUREG-1604 1
i
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tubes using the kinetic expansion method mentioned likelihood of circumferential cracking in the above.
%ese circumferentially oriented cracks transitions is reduced (Ref. 38).
initiated from the inside diameter and the vast j
majority occurred within 5.1 to 7.6 cm (2 to 3 inches) of the primary face of the 61 cm (24-inch)
Although the tube roll transition area is potentially susceptible to circumferential cracking, no thick UTS (i.e., near the original roll transition).
service-induced circumferential indications have been i
%e repair process for these tubes involved kinetically reported to occur at this location. - In addition, no (explosively) expanding the tubes part length through appreciable, if any, axial cracking had been observed the UTS.
Rese kinetic expansions were not at this location before June 1995.
- However, subjected to a stress relief process; therefore, they subsequent to the original submission of responses to are considered a potential site for circumferential GL 95-03 (June 1995), the expansion transition cracking.
region was identified and confirmed to be susceptible to service-induced cracking (axial rather than Service-induced circumferential indications were circumferential cracks were observed).' Specifically, identified in the UTS crevice region in at least one axial indications associated with the expansion i
plant (ANO-1). At ANO-1, one tube was removed transitions have been noted in at least three B&W for destructive examination to verify the type of units (Crystal River 3, Davis-Besse, and ANO-1).
j indication and determine the damage extent and The inspection findings at these plants are discussed 1
l mechanism.
As a result of the destructive below.
l examinations, the indications were attributed to intergranular attack initiating from the outside At Crystal River 3, a single axial indication was diameter of the tube. Some of the indications were detected in the roll transition in a tube that was circumferentially oriented (Ref. 40). The licensee re-rolled following the full bundle stress relief and a concluded that the mechanism leading to the multiple axial indication was detected in the tube end, j
development of these indications most likely involved above the shop re-roll in the same tube. These reduced sulfur species in an acidic environment indications were located in the roll transition in the i
(Ref. 40). The circumferential indications on the UTS (i.e., the hot leg). The eddy current data l
pulled tube were located at several axial elevations provided clear indications that the' tube had been 1
along one axis of the tube. Since the pulled tube was rolled multiple times. He licensee believes the from the lane region and the indications were axially indication is attributable to PWSCC (Ref. 37).
aligned, the licensee believed that axial stress may have played some role in their formation.
In At Davis-Besse, an axially oriented indication was addition, these circumferential indications in the detected during the examination of what was believed pulled tube were not in the uppermost span of the to be a non-stress-relieved roll transition. Similar to tube as defined above in Section 5.2.1 (i.e., they Crystal River 3, this indication was in the roll were above the UTSF).
transition in the UTS (i.e., the hot leg). Subsequent review of shop records indicated that the espansion The tube roll transition area (i.e., the expansion transition had not been re-rolled and was, therefore, transition)is potentially susceptible to circumferential stress relieved (Ref. 4!). The licensee waoved the cracking. As mentioned previously, a small number roll transition portion of this tube for oestructive 1
of tubes were not stress relieved following re-rolling.
examination. De destructive examination indicated
{
The B&W Owners Group concluded that these tubes that the indication was attributable to PWSCC. To j
were more Mely candidates for cracking at the roll ascertain whether the tube had been stress relieved, transition ".han tubes that were stress relieved. This the licensee performed additional analyses and is becarse stress relieving reduces the residual testing; however, the results were unavailable at the stresses and enhances the resistance of alloy 600 to time this document was prepared (Ref. 37).
I caustic corrosion. In addition, the B&W Owners l
Group concluded that since the roll transitions in At ANO-1, several axially oriented and volumetric B&W OTSG units are normally in compression indications were detected in stress-relieved roll during operation, unlike a recirculating steam transitions in the UTS (i.e., the hot leg). The axial generator in which the tubes are under tension, the indications were attributed by the licensee to inside diameter initiated degradation (i.e., PWSCC). The NUREG 1604 5-4
_ ~ ~
I volumetric indications were outside diameter initiated abandoned and secured auxiliary feedwater header.
which may indicate that the indications are nis header is located in the span between the attributable to intergranular attack or closely spaced fifteenth tube support plate and the UTS secondary cracks. To further characterize the nature and cause face. The dents were caused by loose parts (i.e.,
j for the upper roll transition indications (and other dowel pins) contacting the tubes and by contact with j
I indications), the licensee for ANO-1 removed several the header itself. Damaged tubes in this area have tube sections for destructive examination. He results generally been plugged or sleeved. The auxiliary from this examination were not available at the time feedwater ring header is a circular pipe attached to this document was prepared (Ref. 42).
the inner wall of the steam generator which distributes water into the steam generator from the These events indicate that roll transitions in B&W auxiliary feedwater system. The ring header moved OTSGs are susceptible to cracking. Furthermore, because of the collapse of steam inside the header these events indicate that both stress-relieved and during occasional injection of relatively cold auxiliary
(
non-stress relieved transitions are susceptible.
feedwater (Ref. 29). Denting in the tubesheet, in at l
l least one plant (e.g., TMI-1), was caused by the 5.2.3 Dented / Dinged Regions distortion of the tubesheet ligament as a result of the installation of explosive plugs in adjacent tubes. The Denting is the plastic deformation (constriction or dents are generally midway between the primary and mechanical deformation) of the steam generator secondary faces of the LTS (Ref. 39).
tubes, which has resulted from both corrosion and mechanical processes. Sections 6.1.2, 6.2.2, and Dented locations were not identified as being 7.2.3 contain further details on tube denting, with potentially susceptible to circumferential cracking in Section 7.2.3 providing additional details regarding the original GL 95-03 response made by the B&W the nature and consequences of tube denting.
Owners Group.
Subsequent to the original submission of responses to GL 95-03 (June 1995),
Tube diameter reductions have been referred to as dented (dinged) regions were identified as being dings in most OTSGs by the B&W owners to potentially susceptible to cracking in B&W OTSGs.
distinguish them from the more severe classical In fact, degradation has been observed in these denting observed in the recirculating steam
" minor" dents (dings) in once-through steam generators.
The classical denting observed in generators. Dented locations are a possible area of recirculating steam generators was to such an extent susceptibility to circumferential cracking because of that it frequently caused licensees to use smaller than the associated increase in stress levels within the nominally sized eddy current test probes to ensure tubing (Ref.44).
Circumferentially oriented that the probe could pass through the tube. The indications associated with dented locations have been majority of the denting currently observed in observed in at least two plants with B&W OTSGs recirculating steam generators, however, is minor (ANO-1 and Oconee 1). The inspection findings at compared to the " classical" denting observed in the these plants related to these indications are discussed late 1970s and early 1980s. That is, the majority of
- below, the dents currently being observed at most plants with recirculating steam generators are of such a size that At ANO-1 in 1993, two volumetric indications with a nominally sized inspection probe can pass through cin:umferentially oriented crack-like indications the dented region of the tube. As a result, although associated with a dent were identified using a their causes may be different, dents and dings can be motorized rotating pancake coil probe at the UTS l
viewed as similar in terms of potential sites for secondary face (Ref. 44).
This plant had also degradation.
observed circumferentially oriented intergranular attack indications at non-dented locations in the same For the B&W OTSGs, denting is most prevalent at region; therefore, one licensee reported that it could the secondary faces of the UTS and LTS; however, not determine whether the indications were denting has also been observed at the tube support attributable to the dents or to general intergranular I
plates, within the tubesheets, and in free span attack coincidental to the dent ('Ref. 37). In the 1996 f
locations (Refs. 39 and 43). Free-span denting has steam generator tube inspection outage, the licensee occurred in periphery tubes in the region of the for ANO-1 identified an axially oriented crack-like NUREG-1604 5-5
i indication associated with a free-span dent (Ref. 42) been used on a very limited basis. (Oconee I has indicating that cracks can potentially occur at these two tuke with a total of four sleeves currently in dented locations in B&W OTSGs. His indication service.) De B&W Owners Group also identified leaked under main steam line break differential these joints as potential sites for circumferential i
pressure conditions during an in sits pressure test cracking.
(Ref. 42).
5.3 Justification for Odaw Onermdna At Oconee 1 in 1995, indications were also detected j
at dented locations at a tube support plate within the The staff evaluated each of the B&W owner's i
upper boiling region of the OTSGs.
D ese responses to GL 95-03 to confirm that each plant indications were associated with dents in the drilled could safely operate until the next scheduled steam l
hole portion of the fiAeenth tube support plate. One generator tube inspection outage. The operating of the indications was circumferential in nature, the experience to date has shown that widespread l
other was volumetric (Refs. 37 and 45),
service-induced circumferential cracking of safety l
significance is not occurring in B&W OTSGs. In i.
At Crystal River 3 in 1996, an eddy current fact, before June 1995, circumferential cracks indication was detected at a dented location. His (excluding the TMI l circumferential cracks which indication was detected with the bobbin coil and were initiated following a chemical excursion while confirmed with a pancake and plus-point coil to be the plant was shut down) had, for the most part, only volumetric in nature (Ref. 37).
Volumetric been detected in OTSG tubes at the top (fifteenth) indications can be a result of closely spaced cracks, tube support plate and at the bottom of the UTSF in although it is not known whether this is the cause of the lane / wedge region. As stated above, these cracks this indication.
were primarily a result of high cycle fatigue. As a preventive measure, the OTSG plant licensees have In summary, circumferential indications can occur as installed sleeves in the top span of some of the tubes a result of tube denting. These indications can occur in the lane / wedge region. These sleeves stiffen the at mechanically or corrosion induced dents regardless tubes and provide an inner pressure boundary (Ref, of the location in the tube bundle.
38).
5.2.4 Sleeve Joints ne staff concluded that all of the B&W units could operate until their next scheduled steam generator Sleeve joints were also identified (in June 1995) as tube inspection. The following factors contributed to 3
being potentially susceptible to circumfematial this conclusion:
cracking. As discussed in Appendix A, various sleeve designs exist and some have exhibited (1) the minimal number and severity of the circumferentially oriented degradation.
B&W circumferential indications detected in the B&W mechanical sleeves have been installed in all OTSGs operating B&W OTSG plants in order to mitigate tube leaks caused by high cycle fatigue and to repair (2) the preventive measures taken by the licensees tubes with other indications of degradation. De (e.g., sleeving of tubes in the lane / wedge i
mechanical sleeves have roller-expanded joints at region) each end to seal them into the parent tube. Dese tube /sleevejoints have not undergone any stress relief (3) the operating time until the next scheduled pmcess. As a result, the B&W Owners Group steam generator tube inspections identified this region as a potential site for circumferential cracking. De material composition (4) the requirement to monitor of the sleeves is either alloy 600 or alloy 690.
primary to-secondary leakage and to shut down I
the plant when leak rate limits are exceeded Hydraulically expanded sleeves (i.e., sleeves that have been hydraulically expanded into the tube rather (5) the use of procedures (including emergency than mechanically (i.e., roll) expanded) have also operating procedures) to diagnose and address steam generator tube leaks and ruptures l
NUREG 1604 5-6
i 1
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(6) the risk and potential consequences of a range degradation wem recorded as 'None" in Tables 5-1 of steam generator tube rupture events as through 5-4. In instances where the results of the discussed in NUREG-0844, "NRC Integrated tube mapections were readily available, the results Program for the Resolution of Unresolved were included in the tables. For example, at Crystal Safety Issues A-3, A-4, and A-5 Regarding River 3, the RAI was submitted subsequent to the Steam Generator Tube Integrity," which was completion of the 'next" steam generator tube issued in SW.R 1988 (Ref. 27) - In inspection. As a result, the licensee provided their Reference 27, the staff estimated the risk actual scope and some of the results from the contribution due to the potential for single and ex==ination. These results were included in the multiple steam generator tube ruptures. In tables, as appropriate. Acronyms and abbreviations addition, this study exanuned the expected used in the tables are explained in Appendix C.
consequences of steam generator tube rupture scenarios, including beyond design basis As can be seen from evaluating the data in Tables 5-1 situations, such as the potential for release due through 5-4, there are plant-specific differences in the to contamment bypass via failed tubes inspection piens (e.g.,
probe type, scope of concurrent with a breach of secondary system examination). nese differences in the inspection integrity.
plans were considered along with other plant-specific circumstances (e.g., preventive measures taken) in 5.4 ' Iube Insnections evaluating the acceptability of a licensee's response as discusaad in Section 5.3. For example, even though
,GL 95-03 requested, in part, a safety assessment a licensee may have implemented a smaller initial justifying continued operation on the basis of past mapection scope than another licecsee, this may have inspection results and a summary ofinspection plans been considered acceptable if the cumulative for the next scheduled steam generator tube operating time for the plant was less than that of the inspection outage as they pertain to the detection of other plant (all other parameters being equal).
circumferential cracking. The inspection plans were to consist of both an initial scope and sample Many of the licensees stated that their mspections expansion criteria. For the B&W units, the staff would be expanded on the basis of an engineering summarized in several tables at the end of this section evaluation of the results. He staff agrees that this is some of the information provided by the licensees a prudent measure; however, the staff notes the with respect to the previous and next inspection for following considerations:
each of the areas identified as being potentially t
susceptible to circumferential cracking.
He (1) ne expansion criteria defined in the technical designation of " previous" refers to mspections specification must be followed for all performed before issuing or responding to GL 95-03.
examinations performed, regardless of probe i
ne designation of "next" (and/or " future") refers to type.
inspections performed after issuing or responding to GL 95-03. The phrase, "if detect", is used to (2) Because of time constraints in an outage, it describe the inspection expansion criteria when a would be beneficial to have a well thought out circumferential indication is detected.
In many tube mspection expansion plan, which instances, the next (and/or future) inspections have anticipates possible results before commencing already been completed as a result of the time taken the steam generator tube inspections.
to prepare this document for publishing.
(3) De root cause of any "new form" of i
Table 5-1 summarizes the scope of the past and degradation will not likely be known during the future inspections at the expansion transition for each outage (since tube pulls are necessary for a
)
of the seven operating B&W units, along with some conclusive root cause evaluation and for pertinent notes. Tables 5-2, 5-3, and 5-4 provide confirming that the degradation at one plant is similar information for the lane / wedge region, dented similar to that at another). Consequently, the locations, and sleeve joints, respectively. Tube inspection expansion criteria should include a inspections performed using a technique not capable sampling of tubes not believed to be susceptible j
1 of reliably detecting circumferentially oriented to this "new form" of degradation to confirm 5-7 NUREG-1604 l
that the inspections performed bounded the affected areas.
(4) Any tube inspection expansion criteria should also be contingent on providing reasonable assurance that all tubes in the steam generator would retain adequate structural and leakage integrity, consistent with other regulatory guidance, until the next scheduled mapection for those tubes. Such an evaluation would
- require, in
- part, technically justified assumptions regarding (a) the size of flaws in tubes not inspected (most likely similar si:zes to those found in the tubes inspected), (b) the probability of detection, (c) the uncertainty in estimating the size of the flaws, and (d) the growth rate of the flaws.
The staff has reviewed the submissions provided by the licensees with B&W OTSGs and has concluded that they contain the information requested in GL 95-03.
General conclusions regarding the responses are discussed in Section 9.
NUREG-1604 5-8
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d i
Table 5-1 Inspections at the expansion transition region in B&W plants l
Plant Expsedom Transitlom EW= Transtalon Future Notes Past laspection laspection Arkansas Nuclear One - 1 None 100% of non-stress-relieved joints wnh 13 uansidons rerolled following stress relief.
ndia H cire. cracking technulue.
References 44 and 46.
If t, will evaluate expanding sample into stress-relieved transitions.
i Bellefonte Will provide response to GL 95-03 prior to inidal fuelloading. Reference 47.
Crystal River 3 Nonc 100% RPC of non-stress-reheved -
10 transindns rerolled following strem rehef.
joints. If detect, will evaluate 0.635' bore size in tubesheet (smaller than i
expanding sample into stress-relieved several other B&W SGs). Detected SA) and MAI i
transitions. [ Based on finding an axial in ITTS in one tube during 1996 anspecdon. MAI indication, expanded sample to include above shop re-roll. Indications beleved to be tubes rerolled for tube end repairs.)
PWSCC, Damaged tube ends in SG B as a result of a loose part - repaired by drilling tube end and l
rerolling during 8R,9R, and 10R (retolling not l
performed for all repairs in 10R). References 37 and 48.
Davis-Besse None 100% of non-stress-relieved joints with 8 transidons rerolled following stress relief.
Appendix H circ. cracking technique.
Rese 8 transitions are believed to be stress If detect, will evaluate expanding relieved (i.e., they were not rerolled) based on sample into stress-relieved transitions.
subsequent evaluadons. References 41,43, and 49.
Oconeei 100% RPC of 100% RPC of non-suess-relieved 132 transitions rerolled following stress relief.
non-stress-relieved HL joints. If detect,20% of stress-relieved References 50,51, and 52.
(upper tubenheet) joints.
transitions Oconec 2 100% RPC of 100% RPC of non-stress-relieved 4 transitions rerolled following stress relief.
non-stress-relieved joints. If detect,20% of stress-relievnt References 50,51, and 52.
transitions joints.
100% RPC o(non-stress-reheved 4 transitions rerolled following stress relief. 3 Oconee 3 100% RPC of non-stress-relieved joints. If detect,20% of stress-relieved tubes explosively expanded into the upper and transitions joints. 3 explosively expanded tubes lower tubesheet. References 50,51, and $2.
will be inspected or removed from
- service, nree Mile Island 1 Roll transidons (C1/LTS):
100% RPC of non-stress-rcheved joints 11 transidons rerolled following stress relief-i None. Kinetic Expansions (C11LTS). If detect, will evaluate records are not clear on exact locations. As a
)
(HIllfTS): 8x1 array coil expanding sample into stress-relieved result of a sulfur intrusion which resulted in ID j
for tubes in lane / wedge transitions. For kinede expansions, all initiated circ. cracking, the tubes in the UTS in region (217 tubes per SO unsleeved tubes in the lane / wedge both SGs were repaired by kinetically expanding 1
were inspected) region will be examined [589 tubes the tubes into the tubesheet. Here is at least a 2' were examined with RPC). If detect, defect-free unexpanded section associated with a expand based on erigineering 17' or 22' expansion length. Most expansions evaluation. [No circ. cracks observed ) (-14,000) are 17*, LTS in 50 A has short thick-walled tube secdon in the holes drilled in the open lane region. References 39,53, and 54.
5-11 NUREG-1604
1,.
i I
Table 5-2 Inspecak. sat the inne/wedse res en in saw pients s
M l
tusnes)
Arkansas Nuclear One. I Nonc 1 the bonier around sleeved region i tube was confirmed via a ate in the early 1
with Appendia H cire, crackins 1980s to have circumferential W
technique at 15th TSP and LTTSF. If with IGA in the LTTS crevice region. References detect, espand until bounded.
44 and 46.
i t
}
I Bellefame Will provide response to GL 9543 prior to initial fuelleading. Reference 47.
Crystal River 3 None: Preventive sleeving i ahe border around sleeved region References 37 and 48.
'l with RPC (185 hbes at 15th TSP and j
(TTS ancondary face). [No indications observed.]
j J
1 Davis-Besse None I tube border around sleeved regon References 41,43, and 49.
with Appendix H cire, cracking technique at 15th TSP and LTTSF. If detect, expand uruil bounded.
Oconee i 1005 RPC at 15di TSP 2 tube border around sleeved region References 50,51, and 52, and LTTS with RPC probe at 15th TSP and LTTSF. If detect, expand until bounded.
nc-2 1005 RPC at 15th TSP 2 tube border around sleeved region References 50,51, and 52.
and (TIS with RPC probe at I$th TSP and ITTSF. If detect, expand until
- bounded, j
Oconee 3 100% RPC at 15th TSP 2%be border around sleeved region References 50,51, and $2.
and (TIS with RPC probe at 15th TSP and LTTSF. If detect, expand urrdt bounded.
Three Mile Island i 217 tubes w/8xl in each 1 tube border around sleeved region A lealtage event associated with a TMI-!
SG at 15th 'ISP and LTTS with RPC at 15th TSP and ITTSF. If lane / wedge tube is discussed in NRC Informasion secondary face detect, expand until bounded. [No.
Notice 9143 (Ref.16). References 39,53, and defects observed.)
54, i
i I
N' UREG-1604 5-12
Table 5-3 Inspections at dented locations in B&W plants Plant Other Dested lecations Other Deuted : y.*'-a (e.g., TTS)
Notes (e.g., T!3) Past Future laspection L_,-_ Pan Arkansas Nuclear One - 1 27 dere w/RPC 10% of all dents at UTS saxi 15th TSP. 2 volumetric indicadons with circumferennally riented crack-like indradons were detected 10% of all dents at LTS. Inspect with o Appendix H circ. cracking technique. during IRll at the UTS secondary face. Deze if detect. 20% of remaining dents indstadons were associated with denu. All denu within the affected area.
inspected between 15th TSP and UTS with RPC I
along with a sample of dena at LTS in 1R11.
1R12 inspection included 27 dents. Dents selected will be largest by bobbin coil signal amplitude. A voltage threshold is not used in determining the threshold for examining denu.
l No further expansions specified. References 44 and 46.
l Will provide response to GL 9543 prior to imual I
Bellefonte fuel loading. Reference 47.
Crystal River 3 None 31 denu w/RPC. Il volumeuic Not a commitment to GL 95-03 (inspecdon was indication associated with a dent was completed before licensee received RAI). Record detected with a bobbin coil and dents at 5 volts. Calibration: 6 volts on 4-100%
confirmed with RPC and PP.]
through-wall holes at 600 kIh and 600/200 mix frequency. LTS dents believed to be deposits based on tube pulls. No distinct crack-like indications detected (small volumetric indication detected). UTS dents selected for examinadon were biased towards tubes which border tubes plugged via explosive charge. References 37 and 4g.
Davis-Besse
- Linuted* RPC sample 100% at 14th TSP arx! above with Dents examined will be those that have been Appendix H circ, cracking techruque.
previously detected. If a new dent is detected, 20% of remaining dents in SGs. If the need to inspect these dents will be evaluated.
detect. capand based on engineering Denu rwar secured internal auxiliary feedwater header. References 41. 43, and 49.
evaluation.
Oconee1 larger voltages w/RPC Rautinely monitored w/RPC.
Dents located at 10th TSP, upper and lower tubesheet secondary face. Not currendy believed to be susceptible to circumferential cracking.
References 50,51, and 52.
Oconec2 larger voltages w/RPC Routinely monitored w/RPC.
Denis located at 10th TSP, upper and lower tubesheet secondary face. Not currently believed to be susceptible to circumferential cracking.
References 50,51, and 52.
Oconee 3 12rger voltages w/RPC Roudnely monitored w/RPC.
Dents located at 10th TSP, upper and lower tubesheet secondary face. Not currendy believed to be suscepdble to circumferential cracking.
References 50,51, and 52.
nree Mile Island 1 Not addressed Dents above LTS inspected with RPC; Support structure diameter reductions are derus; Dents at LTS above 40 volts inspected freespan diameter reductions are dings.
with RPC - some samplir4 below 40 Calibration: 400 kHz differential set to 10 volts volts. [No defects observed.)
on 4-20% holes. Dent / ding recorded if greater than or equal to 10 volts. Denting occurs in tubesheet as a result of explosive plugs being installed in adjacent tubes (i.e., ligarnent distortioa). References 39,53, and 54.
NUREG-1604 5-13
Table 5-4 Inspections at sleeve joints in B&W plants Plant Sleeve Past losee Sleeve Future !=v 1--
Notes Ark =a Nuclear One - 1 222 sleeves with PP 20% with Appendix 11 circ. cr.cking References 44 and 46.
technique. If detect,100% cxam or until bounded.
Bellefonte Will provide response to GL 954)3 pror to iniual fuel loading. Reference 47.
Crystal River 3 Installation inspections 20% PP. If detect,100% cxam or until References 37 and 48.
bounded. [No crack like indrations observed.]
Davis-Besse None (crosswound) 20% PP. !! detect,100% cxam or until References 41,43, and 49.
bounded.
Ce=+e 1 100% RPC at upper rolls 20% PP, If detect,100% etam or until 2 tubes contain hydraulic sleeves at 15th TSP and bounded. The 2 tubes w/ hydraulic sleeves and the 32 sleeves at the 14th UTS. 32 sleeves at 14th TSP. Siceves installed TSP will be inspected or removed from in 10/87. References 50,51, and 32.
service.
Oconec 2 100% RPC at upper rolls 20% PP. If detect,100% exam or until lane / wedge sleeves installed in 9/90 and 10/94.
bounded References 50,51, and $2.
Oconce 3 100% RPC at upper rolls 20% PP. If detect,100% cxam or until lane / wedge sleeves installed in 8/88 a bounded.
References 50,51, arrt 52.
Three Mile Island 1 None (crosswound) 20% PP. If detect,100% exam or until References 39,53, and 54.
bounded. [No defects observed.]
NUREG-1604 5-14
- -.~. _ - - -
. ~ -. - -
l l
1 i
6 COMBUSTION ENGINEERING STEAM GENERATORS 6.1 Desian Only the Millstone 2 replacement steam generators have alloy 690 tubes; these are discussed in Section CE plants have steam generators with either alloy 600 8.
He remaining steam generators (including the or alloy 690 steam generator tubes. Sections 6.1.1 pre-replacement Millstone 2 steam generators and the and 6.1.2 describe the generic and plant-specific pre-replacement and replacement Palisades steam features of CE plants that have steam generators with generators) have alloy 600 tubes and are discussed in alloy 600 steam generator tubes. CE plants with this section.
l alloy 690 steam generator tubes are discussed in Section 8.-
With the exception of Maine Yankee, which has three steam generators, all CE plants have two steam 6.1.1 Generic Features generators.
The number of tubes in a steam generator varies from approximately 5,000 to 11,000 The CE recirculating sta generators (RSGs) are and these tubes are arranged in a triangular pattern as vertical, shell, and U-tube heat exchangers with shown in Figure 6-1. There is an open lane down integral moisture-separation equipment. A large the center of the steam generator between the legs of cylindrical vessel encloses an inverted U-shaped tube the innermost U-tubes.
l l
bundle consisting of many thousands of individual tubes, each welded to a thick plate (called a A number of different CE steam generator designs tubesheet) with a hole for each tube end located near exist in the United States. _ (An example of a CE the bottom of the steam generator vessel. The steam generator is illustrated in Figure 2-3.) These reactor coolant enters the hemispherical bottom head designs have been categorized by one CE licensee as through an inlet nozzle, flows through the U-tubes early models (i.e., before series 67), series 67,3410 l
and exits the lower plenum through an outlet nozzle.
series, and system 80 (Ref. 55). Differences exist l
A divider plate in the lower plenum below the between the four categories, as well as within tubesheet separates the inlet and outlet primary individual categories. Design differences that are coolant and directs the flow through the tubes.
pertinent to steam generator tube integrity (and particularly circumferential cracking) are discussed in For most of the CE steam generators, the secondary the following paragraphs.
system water (feedwater) is fed through a feedwater nozzle, to a feedring into the downcomer, where it ne steam generater tube bundles in CE plants mixes with recirculating water draining from the generally consist of both U-shaped tubes and tubes moisture separators. Tia d vncomer water flows to with two 90-degree bends. The U-shaped tubes the bottom of the steam gcnerator, across the top of (sometimes referred to as tubes with 180-degree the tubesheet, and then up through the tube bundle bends) are located in the lower row (smaller bend where steam is generated. Some of the CE steam radii) tubes. By contrast, the tubes with two-90 generators (e.g., Palo Verde 1, 2, and 3), however, degree bends are located in the higher row (larger have a separate economizer section (preheater) in the bend radii) tubes. Most of the tubes in a typical CE steam generator near the cold-leg outlet.
The steam generator tube bundle are of the design with feedwater flows into the preheater through nozzles two 90-degree bends.
located in the lower part of the vessel. Heat from the
{
primary fluid leaving the steam generator is used to The tubes in CE steam generators were explosively preheat the feedwater nearly to saturation temperature expaniled through the full length (i.e., thickness) of before it enters the evaporator section (Refs. 24 and the tubesheet, thereby eliminating the j
55).
tube-to-tubesheet crevice that is present in other steam generator designs. As a result, the expansion There are currently 15 operating CE units. Of these transition region where the tube changes from fully units,13 are currently operating with their original expanded to non-expanded (i.e., nominal tube stma generators, while 2 (i.e., Millstone 2 and diameter), is nominally located at the top of the Palisades) have replaced their steam generators.
tubesheet (i.e., the secondary face of the tubesheet);
6-1 NUREG-1604 l
__. _ _. _ _ _ _ _ _ _ _ _ _._ - _ _ _ - -. _ ~ - - -
however, there are a minimal number of tubes that seven full eggerate tube supports, two partial eggerate have not been fully expanded. Partial-and full-depth tube supports, and two partially drilled tube support tube expansions are illustrated in Figure 6-2. b plates (Ref. 46).
tubes are expanded into the tuw6t in both the primary coolant inlet region (hot leg) and the outlet 6.1.2 Plant-Specific Features region (cold leg).
As a result, the expansion transitions are frequently referred to as the hot-leg
%ere are unique features that affect steam generator and cold-leg expansion transitions, as appropriate.
tube integrity in several of the CE RSGs. The b tubes are seal welded to the tubesheet cladding in following pa, graphs describe some of the more both the hot and cold legs of the steam generator.
pertinent fea nres that may affect a tube's susceptibility to circumferential crecking or its b tubes in most CE steam generators are supported inspectability, with plates or eggerate-type dividers at a number of fixed axial locations along the tube bundle, as well as As mentioned above, the tube support structures can variously shaped bars and small plates in the U-bend vary from plant to plant. In general, stainless steel region of the tube bundle. (Examples of tube support tube support structures have shown to be relatively configurations used in steam generators are illustrated corrosion resistant compared to carbon steel tube in Figure 6-3.)
nese tube supports are either support structures. His corrosion resistance is horizontally, vertically, or diagonally oriented. Most important in reducing the likelihood of tube denting, of the supports in the CE steam generators are which has, in some instances, led to the development formed from a lattice arrangement of bars (i.e.,
of circumferential cracks. Denting is the plastic eggerate tube supports); however, drilled plates are deformation (constriction or mechanical deformation) also used. The tube rupport material is generally of the steam generator tubes; it typically occurs when either carbon steel (in earlier models) or stainless tube support structures (e.g., carbon steel tube steel (in later models such as Palo Verde).
support plates) corrode. Such corrosion results in the buildup of corrosion products (typically magnetite) in h horizontally oriented tube supports are primarily the crevice between the tube and the tube support eggerate suppo ts; however, drilled plates are also plate. This buildup of magnetite (iron oxide) leads used. He drilled tube support plates are primarily directly to the mechanical deformation of tubes where located in the upper portion of the tube bundle near t'.ey meet the tube support plate intersections; when the U-bend (or 90-degree bend) region and are this buildup is extensive, denting can lead to the frequently partial plates; that is, they do not support deformation and cracking of the tube support plates all of the tubes. However, some of the drilled plates themselves. In general, many of the newer steam are located in the lower portion of 6e bundle. For generators (including Palo Verde 1,2, and 3 and the -
example, Palo Verde 1,2, and 3 have a drilled plate, Palisades replacement steam generators) have tube which serves as a flow distribution baffle, located support structures made with stainless steel. Most of directly above the tubesheet (Ref. 55). Similarly, the the remaining CE units have carbon steel tube lower six tube supports in the original Palisades support structures.
steam generators were drilled plates.
The steam generators at Fort Calhoun contain 19.05 h betwing supports, referred to as anti-vibration mm (0.75 inch) thick orifice plates installed in the bars (AVBs) in Westinghouse steam generators, are primary channel heads on the hot-leg tubesheets (i.e.,
used in the U-bend region to stiffen the tubes and/or primary face of the tubesheet). These plates limit the prevent tube-to-tube contact. These supports are primary coolant flow to the tubes in rows 1 through diagonally oriented in the tube bundle. Vertical
- 18. He orifice plates were installed before initial straps are also used in many CE steam generators to plant startup to eliminate the potential for phosphate provide vertical support for the horizontal section of corrosion caused by steam blanketing in the U-bend the tube between the 90-degree bends.
area of the low row tubes. Dese plates were not removed when the licensee decided to operate with an b number of supports vary from plant to plant.
all-volatile treatment secondary water chemistry I
For example, ANO-2 has a total of eleven rather than phosphates.
However, the licensee horizontally oriented tube supports. These include believes that the orifice plates have probably played NUREG-1604 6-2
i i
4 a role in reducing the susceptibility of the steam o dented regions (particularly dented tube support generator tubes to both steam blanketing assisted regions) (Refs. 36,46,56, and 58) 1 4
intergranular stress corrosion cracking (IGSCC) in il the U-bend region and circumferential cracking at the o expansion transitions (Refs. 36, 46, 56, and 58) expansion transitions. This is because the orifice 1
plates reduce the primary coolant flow and
- sleeve joints (Refs. 36 and 46) subsequently lower tube wall temperatures in the tubes with orifices. De orifice holes in this plate are The operating experience for the pre-replacement too small to pass an inspection probe through the tube Millstone 2 and Palisades steam generators is from the hot-leg side. As a result, the orifice plates discussed separately (in Section 6.2.5) for clarity, render approximately 22-percent of the tubes inaccessible for inspection from the hot-leg side of 6.2.1 Top of Tubesheet Region j
the steam generators.
{
As discussed above, the expansion transition location Subsequent to their original GL 95-03 response, the in CE steam generator tubes exists at the top of the l.
licensee for Fort Calhoun decided to remove the tuh=haar and is susceptible to circumferential i
existing orifice plates from both steam generators.
cracking. As of June 1995, circumferential cracking I
4 This decision arose from the uncertainty regarding in this region was observed in 10 of the 14 CE plants j
the actual ternperature differences that may be with alloy 600 steam generator tubes, excluding the i
encountered behind the orifice plates compared to pre-seplacement Millstone 2 and Palisades steam other locations in the steam generator, as well as the generators. Depending on the plant, this cracking potential susceptibility of these tubes to initiated from either the inside diameter, outside circumferential cracking. After tube inspection and diameter, or both. As of June 1995, all of the repair, the licensee will install new orifice plates to circumferential indications were observed at the maintain reactor coolant system total flow equivalent hot-leg expansion transition, with the possible to current conditions (Ref. 56),
exception of one cold-leg expansion transition indication at Maine Yankee. (The licensee for Maine 6.2 Locations - Susceotible to Circumferential Yankee has stated that this cold-leg indication, which Crackine was detected and plugged in 1990, may have been mistakenly diagnosed as circumferential cracking on At the time of the original responses to GL 95-03, the basis of the techniques available at the time (Ref.
i the CE Owners Group identified three regions as 59).)
haviag exhibited circumferential cracking.
Specifically tiese regions included the top of the Several utilities have removed the expansion j
tubesheet, the U-bend (or the 90-degree bend), and transition portion of tubes from their plants for the tube-to-tube support intersections. However, destructive examination.
These utilities include j
circumferential cracking of steam generator tubes in ANO-2 (Ref. 46), Maine Yankee (Ref. 59), Millstone service as of June 1995 had only been observed at the 2 (Ref 57), and Calvert Cliffs 2 (Ref. 60). The top of the tubesheet. The circumferential cracks in results from these tube pulls generally confirm that j
the U-bend (or 90-degree bend) region and at, or the circumferential cracks at the hot-leg expansion j
near, the tube-to-tube support intersections were transition initiated from either the inside or outside l
observed in the pre-replacement Palisades steam diameter of the tube and are attributed to IGSCC.
generators; circumferential cracks at, or near, the De tube pulls from several plants indicate that both j
tube-to-tube support intersections were also detected inside and outside diameter initiated circumferential j
in the pre-replacement Millstone 2 steam generators cracks generally consist of a series of short j
(Ref. 57).
circumferential cracks connected by thin ligaments of non-corroded material (Refs. 46 and 61). The axial In general, most CE licensees consider the following elevation of these cracks varies within the crack locations currently susceptible to circumferential network (i.e., they are typically not coplanar).
j cracking:
I Subsequent to the original GL 95-03 responses, the licensees for a few plants detected circumferential a
6-3 NUREG-1604
- _ ~
- - - = -
i
)
indications at the cold-leg expansion transition. At These dents are generally a result of steam generator l
ANO-2, 36 circumferential indications were detected fabrication and loose parts in the steam generator.
at the cold-leg expension transition during an l
inspection in the fall of 1995 (Ref. 62), in addition, Dented regions are of concern because they can lead 3
1 10 cold-leg expansion transition circumferential to axially and circumferentially oriented mbe indications were detected at St. Lucie 1 during an degradation as a result of increasing the stress levels inspection in the spring of 1996 (Ref 63). Deee within the tubing.
In CE steam generators, inspection results indicate that hot-and cold-leg circumferential indications have occurred at dented 7
expansion transitions are susceptible to cracking.
regions (generally at corrosion-related dents). In d
addition, circumferential indications have been l
Virtually all CE units (including several of the newer observed at locations where there is ao dent, but units such as' Waterford 3) have exhibited where denting of tubes is believed to have contributed circumferential cracking at the expansion transition to the development of the circumferential indications.
(Ref. 64).
Circumferential cracks near tube support locations 6.2.2 Dented Locations including Dented Tube (e.g., eggerates, drilled plates, and betwings) were 3
Support Areas detected in the pre-replacement Palisades and Millstone 2 steam generators (as discussed in Sections As discussed in Section 6.1.2, denting is the plastic 6.2.5.1 and 6.2.5.2, respectively).
- Recently, j
deformation (constriction or mechanical deformation) circumferential indications were identified at St.
l of the steam generator tubes, which has resulted from Lucie 1 at both dented and non-dented tube support both corrosion and mechr.nical processes. (Section locations. These circumferential indications were 7.2.3 contains further dewis regarding the nature and identified on both the hot-and cold leg sides of the i
consequences of tube denting.)
Corrosion and steam generator.
4 mechanically induced stents have been observed at a number of locations in CE steam generators.
De top two supports (ninth and tenth) in the St.
a Lucie Unit I steam generators are partial drilled Corrosion induced dents aenerally occur at tube plates. The ninth tube support plate extends from locations next to support st uctures fabricated from tube row 91 through 140. The tenth drilled tube carbon steel (e.g., eggerate tube support, drilled hole support plate extends from tube row 117 through j
tube support plates, tie rods). Dese dents result 140. He inner edge of the drilled plates (row 91 from corrosion of the tube support structure and have and 117) is typically referred to as the chord or i
been observed in some CE steam generator tubes scallop bar, and surrounds only half of the tube 1
near horizontally oriented tube support structures and circumference for tubes in these rows. To relieve I
at the top of the tubesheet next to the tie rods. Tie denting stresses at these tube support plate levels, the l
rods are carbon steel rods (similar in diameter to support lug attachments for these two plates were cut j
tubes) that are used at various locations in the tube and the edges of the plates were trimmed to allow for bundle to maintain the support structure alignment movement of the plate (in the horizontal direction).
before completion of tube installation.
It is Selected tubes were expanded and staked to limit the postulated that corrosion of the tie rods at the top of vertical movement of the plate.
I the tubesheet, similar to the corrosion of tube support structures (as discussed in Sections 6,1,2 and 7.2.3),
During the 1996 inspection at St. Lucie 1, four results in tie rod expansion. It is further postulated circumferential indications were identified in the that tie rod expansion is transmitted to adjacent tubes, drilled hole tube support plates (i.e., the ninth and i
resulting in the potential for denting of the adjacent tenth supports). Three of the four indications were i
tubes (Ref. 65),
located at the ninth support and are along the chord edge in tube row 91 or 92. Two of these three Mechanically-induced dents generally occur indications were in the hot leg and one was in the throughout the steam generator tube bundle and are cold leg. De fourth indication was located at the j
frequently located in the tube free span (i.e., the 10th tube support, nine rows from the chord edge in region of the tube between tube support structures).
tube row 126. Two of the four indications were associated with minor dents (i.e., one of the ninth NUREG-1604 6-4 1
I m
---r-
tube support hot-leg circumferential indications and 6.2.4 Sleeve Joints the 10th tube support hot-leg circumferential indication). Of the three tubes with circumferential As discussed in Appendix A, various types of sleeve indications at the ninth tube support plate, two were designs exist, and some have exhibited determined to face toward the divider plate while one circumferentially oriented degradation. The steam faced away from the divider plate. The licensee generators at CE plants use several sleeve designs, postulated that accumulated stresses due to denting, including the B&W double kinetically welded sleeves, lockup, movement, and bending of tubes could have CE tungsten inert gas (TIG) welded sleeves, and contributed to the development of the circumferential Westinghouse laser-welded sleeves. At the time the indications (Ref. 66).
original GL 95-03 responses were submitted (June 1995), the only CE plants that had installed sleeves In summary, circumferential indications can occur as (or were in the process of installing them) were a result of tube denting. These indications can occur ANO-2 and Maine Yankee. During the summer of at mechanically or corrosion induced dents regardless 1995, Maine Yankee was in the process of installing of the location in the tube bundle. In addition, Westinghouse laser-welded sleeves at the hot-leg circumferential indications have been postulated to expansion transition region in all tubes that were to occur at non-dented regions as a result of remain in service (Ref. 59); ANO-2 had already accumulated stresses related to the denting installed B&W double kinetically welded sleeves in phenomena.
1992 (Ref. 44) and was in the process of installing CE TIG-welded sleeves in the fall of 1995 (Ref. 46).
6.2.3 U-Bend Region 6.2.5 Circumferential Cracking Experiences in the Circumferential indications have been detected in the Pre-Replacement Palisades and Millstone U-bend portion of tubes with small bend radii (i.e.,
Steam Generators rows 1 and 2) in certain Westinghouse steam generators (as discussed in Section 7.2.2); however, Since the pre-replacement Palisades and Millstone 2 no such indications have been detected in the steam generators are similar, in some respects, to the tight-radius U-bends (i.e., sows 1 and 2) of any currently operating CE steam generators, the steam generator supplied by CE (Refs. 60 and 67).
operating experience for these steam generators are summarized in the following sections.
Circumferential indications have, nonetheless, occurred in the upper portion of CE steam generator 6.2.5.1 Pre-Replacement Palisades Steam Generators tube bundles. These indications were detected in the pre-replacement Palisades and Millstone 2 steam The retired Palisades steam generators were unique generators.
('Ihe circumstances surrounding the with respect to the design of the tube supports, which development of these indications are discussed in consisted of solid tube support plates with flow holes, Sections 6.2.5.1 and 6.2.5.2, respectively.)
plates without flow holes but with flow slots, full eggerates, and partial drilled plates (Ref. 68).
Although no circumferential indications have been detected in the U-bend region of tubes with small Circumferential cracks were identified at several bend radii in CE steam generators, axial indications locations in the pre-replacement Palisades steam have been detected in this region in at least one CE generators.
In March 1982, the first two plant (e.g., Palo Verde 2). These indications are circumferential indications were observed as a result believed to have initiated from the outside diameter of a primary-to-secondary leakage event. These of the tube rather than the inside diameter. The indications included one in the tube at row 20, line indications observed in the U-bend region of small 83, which was slightly (approximately 2.5 mm (0.1 radii tubes in Westinghouse plants are predominantly inch)) above the ninth tube support plate and inside diameter initiated, approximately 10 cm (4 inches) below the U-bend and one in the horizontal section of the tube at row 127. line 22, at the first batwing support past the hot leg U-bend (actually a 90-degree bend). Neither of these indications was associated with a dent greater 6-5 NUREG-1604
l i
than 0.025 mm (0.001-inch)(i.e., denting levels were various horizontally oriented tube support elevations less than I mil in both tubes); however, these and were typically located within the bounds and near indications were associated with dents (Refs. 69 and the top edge of the affected tube support. In 70).
Since these tubes were not removed for addition, as mentioned above, circumferential destructive examination, the results are predicated on indications had also been detected at other regions in l
the ECT capabilities at that time.
the upper portion of the tube bundle (i.e., at the betwing, AVBs, diagonal straps, bend region, and/or
'Ibe licensee removed more than 50 tubes for vertical supports) (Ref. 73).
destructive exammation during their 1983/1984 steam
)
generator tube inspection outage, in part, to The licensee identified two possible causes for the characterize the circumferential crack-like indications cracking (specifically the cold-leg cracking):
that were being detected.
The destructive 4
I exammation revealed that the defects cont =iaM (1) in-plane loads caused by severe denting i
relatively large amounts of sulfur, and the tubes containing the defects were highly sensitized. It is (2) out-of-plane loads caused by thermal expansion believed that the corrosion was caused by a reduced form of sulfur on sensitized, alloy 600 tubing.
On the basis of operating experience, dented tubes j
(Sulfur, in the form of sodium sulfite, was used in are known to be susceptible to circumferential the 1973 to 1974 time-frame for controlling the cracking. The dented tubes in the original Palisades a
oxygen content in the secondary water at Palisades.)
steam generators would be considered susceptible to Furthermore, most defects were located adjacent to circumferential cracking, since the denting in these dents at the upper or lower edge of the tube support steam generators would be considered severe (i.e., an plate.
All three of the circumferential cracks average of 0.508 mm to 0.762 nun (0.020- to confirmed during the 1983/1984 tube pulls were 0.030-inch) of denting at one tube support plate)-
located within the confines of the ninth tube support compared to the denting currently (1996) being plate and were associated with a deep patch of observed in most steam generators that have exhibited intergranular attack. In summary, the results of the circumferential indications at dented tube support tube pulls indicated, in part, that the type of locations.
corrosion observed in the steam generator tubes was primarily sulfur-induced intergranular attack and, in In addition to the loads' caused by denting, the some cases, was accompanied by circumferential Palisades licensee postulated that higher-than-normal 1
IGSCC (Ref. 71).
thermal evansion loads could lead to circumferential cracking.
These higher-than-normal loads were During a 1985 mspection, no circumferential postulated Io occur in tubes adjacent to areas of crack-like indications wr,re detected (Ref. 72);
plugged tubs The licensee believed that this was however, in December 1987, a tube leak occurred as particularly true in the case of small-radius tubes, a result of a circumferential indication associated with which would have the ' hottest" primary water at the a small dent at the lower edge (both inside and cold-leg tubesheet. Since most of the tubes in rows outside) of the thirteer.th tube support plate. During 1 through 12 were plugged (as a result of wastage),
the December 1987 inspection, a total of 12 the unplugged tubes closest to the divider plate should circumferential indications were detected at the tube see the highest out-of-plane loads in the steam support plates, the tertical straps (sometimes referred generators because of thermal expansion stresses. In to as the betwing and AVBs 3,2, 3,4, and 5 by the addition to these thermal expansion loads, the licensee), and the cold-leg bends.' (These cold-leg licensee postulated that another probable area of high bend indications were frequently associated with the stress would be at tubes located in the lug (i.e.,
diagonal straps, which were sometimes referred to as wedge) regions. The restraining force of the lug, in batwings and AVBs by the licensee).
addition to thermal expansion stresses between the
" cold" lug and a ' hot
- tube, were postulated to make From 1987 until eteam generator replacement in this a high stress area that could lead to 1991, add;tional circumforential indications were circumferential cracking of the tubes (Ref. 73).
detected in both the hot and cold legs of the Palisades steam grierators. These indications occurred at NUREG-1604 6-6
. - ~ -. -. -.- -
4 l
I In summary, the retired Palisades steam generators
- Existing microcracks propagated as additional had not experienced circumferential cracking at the microcracks initiated, i
top of the tubesheet, but had experienced l
circumferential cracking in the upper portion of the
- As the microcracks propagated, the inside steam generator tube bundle. The postulated failure diameter of the tube was breached when a crack j
mechanism for the Palisades circumferential cracks reached approximately 14 to 21 degrees around the l
was ODSCC. The tubes were considered to be circumference.
I locked into the tube support plates as a result of denting. During operation, the plates and tubes
- The through-wall portion of the crack enlarged moved relative to one another, thereby imposing until it began to leak.
bending stresses on the tubes. b bending stresses j
then initiated circumferential cracks (Ref. 68).
After operating through cycle 9, the licensee detected several hundred circumferential cracks at the top of 6.2.5.2 Pre-Replacement Millstone 2 Steam the tubesheet in February 1989 (i.e., EOC 9).
j-Generators Cracks were observed on both the hot and cold legs of the steam generator near the top of the tubesheet.
He pre-replacement Millstone-2 steam generators De licensee concluded that this location had a were CE series 67 units that contained seven full corrosive environment and high tube stress. The eggerate tube support structures, two partial corrosive environment was a result of the sludge pile i
eggerates, two partial drilled tube support plates, two located on top of the tubesheet. The high tube stress diagonal (batwing) supports, and tluee vertical was a result of tubesheet denting and tube support j
supports. All of these were fabricated from carbon corrosion, which imparted stresses on the tube as a i
steel.
result of shifting of the tube bundle and bowing of
^
the tubes. Tube bowing was confirmed by the Circumferential SCC indications were first detected licensee through the inability to successfully pass long l
in Millstone 2 as a result of a leaking tube in January rigid objects (such as sleeves and stabilizers) through i
1987.
The leaking tube had a circumferential the top of the tubesheet region of some tubes. A indication at the top of the tubesheet. His tube was midcycle inspection was performed in October 1989 pulled for destructive exammation during the (outage 10A) to identify stress corrosion cracks that subsequent refueling outage in January 1988 (i.e.,
had developed to a detectable size since the last end-of-cycle (EOC) 8). Approximately 25 additional inspection and to ensure that the corrosion process hot-leg circumferential indications were detected as a responsible for the cracking was under control.
result of the EOC 8 examination. The tube pull During the inspection, approximately 100 indicated that the tube had a circumferentially top-of-tubesheet cracks were detected. Three tubes oriented outside diameter initiated stress corrosion were removed during the October 1989 outage for crack, which was greater than 50 percent burst testing and destructive examination.
through-wall over 260 degrees (190 degrees of which was 100 percent through-wall). The remauung 100 On the basis of the three previous inspections, the degrees of the tube circumference contained several cracks in the steam generator tuima were j
microcracks less than 50 percent through-wall. The characterized by the licensee as follows:
licensee evaluated the indicated crack sizes and concluded that the tube should not have been able to
- The cracks are located at the top of the tubesheet, withstand operating loads; therefore, the cracks must within approximately 5.1 mm (0.2 inches) from have grown during the time the tube was plugged where the expansion transition meetr. the nom %al (i.e., from the Jawry 1987 to January 1988 outage),
tube diameter, and are circumferentially oriented.
On the basis of careful examination of the fracture surface, the licensee postulatal:
- The macrocrack, as defined by rotating probe eddy current tests, is made up of several e caustic species were concentrating in the sludge discontinuous microcracks separated by ligaments area and initiated circumferentially oriented of sound material.
microcracks on the outside diameter of the tube.
6-7 NUREG-1604
s
- The n&rocracks e located at different axial During the May 1991 inspection, the licensee planes (i.e., they are not coplanar), typically identified approximately 35 circumferential within a 2.5 mm (0.1-inch) band (Ref. 74).
indications in the hot and cold legs near the top of the tubesheet region (i.e., near the expansion transition) j Additional planned and unplanned steam generator (Ref. 77).
)
tube inspections were performed by the licensee from the October 1989 outage until steam generator In January 1992, the licensee performed another k
replacement in late 1992 and early 1993. In May midcycle inspection in response to increasing j
1990, the licensee performed a second midcycle primary-to-secondaryleakage. Duringtheinspection, inspection (outage 10B) as a result of increasing two leaking tubes were identified. One leaking tube i
primary-to-secondary ler.kage (Ref. 75). His time, had a circumferential indication about 5.1 cm (2 the licensee attributed the leakage to leaking plugs inches) above the. top of the tubesheet in the sludge and sleeves; however, 23 tubes were identified that pile region. This defect was located in a region 5
had circumferential indications at the top of the where other similar indications had been identified.
tubesheet (Refs 76 and 77). Indications were found The other leaking tube, in row 87 (row 87 is in the j
in both the hot and cold legs o the steam generator, chord region for tube support 10), had a r
circumferential indication in the 90-degree bend area Following the May 1990 midcycle ' outage, the adjacent to, and oriented with, a betwing support.
j licensee performed inspections of the steam generator The licensee concluded that the cause of both defects tubes during a scheduled. refueling outage in was SCC. Both defects were found in regions
[
' September 1990 (i.e., EOC 10).
During this containing high copper deposits. He scope of the j-inspection, the licensee : found seven additional tube examinations during this outage were limited to i
circumferential indications at both-the hot-and the areas inunediately surrounding the leaking tubes cold-leg expansion transition areas. In April 1991, (Ref. 78).
the licensee performed a limited inspection of the steam ' generator tubes as a result of "he replacement of the original Millstone 2 steam primary-to-secondary leakage. This leakage was generators began in approximately June 1992. A 4
4 attributed to a circumferential crack in the outermost summary of the mspection findings is provided in tube of row 37 and 7.6 mm (0.3-inch) below the top Table 6-1.
of eggerate 8, which is an eggerate support just below the 90-degree bend of this tube.
The In summary, circumferential indications were inspections performed during this outage were identified in the pre-replacement Millstone 2 steam concentrated in row 37 and two adjacent rows each gen.erators.
Circumferential indications were side of row 37 in the area where the indication was identified in both the hot and cold legs of the steam observed (i.e., tube support 8 near the 90-degree generators near the top of the tubesheet. Many of bend). Noadditionalcircumferentialindicationswere these indications were located in a region of the tube detected, although one axial indication was detected.
bundle where sludge accumulates (i.e., the sludge The leaking tube was located in the chord region pile). Circumferential indications were r. Iso identified (i.e., the region of the tube bundle near the edge of at or near the horizontally oriented h,ee supports and a partial tube support).
betwing supports (near the bend area).
In late May 1991, the licensee performed another The licensee believed that the upper bundle cracking midcycle outage (outage 11B) as a result of another was probably caused by caustic SCC, similar to the primary-to-secondary leak. During this inspection, SCC at the top of the tubesheet. In addition, the the licensee examined the top of the tubesheet region licensee believed that the upper bundle indicatio_ns and in and around the chord region for eggerate were caused by bending stresses at the eggerate and supports 8 and 9 (rows 32 through 38 and 62 through 90-defree bend producing local circumferential 68, respectively). The leaking tube, in row 64, was cracks. The bending stresses were probably induced j
identified to have a circumferential crack at the as a result of extensive tube denting and bowing, as
)
90-degree hot-leg bend above the ninth eggerate tube discussed below.
support. The crack was at the lower edge of the first i
batwing strap, and the tube was in the chord region.
NUREG-1604 6-8 n.
~
-~
I l.
Degradation of the eggerates acd tube support plates ne staff evaluated each of the GL 95-03 responses occurred in the original Millstone 2 steam generators, submitted by CE plant owners with alloy 600 steam i
as a result of acidic chemistry conditions early in the generator tubes to confirm that the plants could safely '
operating life of the steam generators. D e acidic operate until the next scheduled steam generator tube i
conditions caused the carbon steel tube support inspection outage. The staff concluded that all of material to corrode and form magnetite. He build these CE units could operate until their next i
up of the magnetite corrosion product in turn caused scheduled steam generator tube inspection. De staff i
denting of the tubes and shifting of the supports. The based this conclusion on the following factors:
8 progression of the denting was reported by the
}
licensee to be essentially zero since cycle 7. A visual (1) scope and results of the prior inspection - In l
inspection of the first eggerate tube' support during steam generators that do not exhibit signals that j
the EOC 9 refueling outage confirmed that tube interfere with eddy current inspections (e.g.,
l bowing was occurring (Ref. 74). One licensee copper, denting), inspections may have been indicated that denting-related expansion and, shifting more effective in identifying tube degradation.
of the eggerates, and locking of the tubes at the
]
betwing and vertical supports, produced the undesired (2) preventive measures taken (e.g., sleeving of all j
stresses in the Millstone 2 steam generators and led, tubes at the hot-leg expansion transition) in part, to the circumferential cracks (Ref. 68).
(3) removal of tubes for destructive examination -
6.2.5.3 Summary of Operating Experience from the Data from steam generator tubes removed for j
Pre-Replacement Palisades and Millstone 2 destructive examination provide useful Steam Generators information concerning the causal effects and i
morphology of the degradation.
In many He tight U-bend circumferential cracking that has instances, tube pull data provide information 3
occurred in some Westinghouse steam generators has that tubes with circumferential indications are not been observed in CE steam generators; however, able to withstand the pressure loadings specified j
circumferential cracking did occur in the upper in Reference 33. Because of the inability to j
bundle region of the pre-replacement Palisades and reliably size and characterize the degradation, i
i Millstone 2 steam generators in tubes with double tube pulls are sometimes necessary to confirm 90-degree bends. Hese cracks occurred either in the tube integrity and degradation morphology.
90-degree bends at or adjacent to diagonal supports, or in the horizontal run between the bends at or (4) in situ pressure test data indicating that the adjacent to vertical supports (Refs. 60 and 67).
circumferential indications are capable of 1
Circumferential cracking also occurred at or near withstanding the pressure loadings specified in eggerate and drilled plates in these steam generators.
Reference 33 - When all circumferential j
Dentmg of the tubes and/or shifting of the tube indications are removed from service before an supports with respect to the tubes is believed to have operating interval (i.e., the time between tube contributed to the circumferential cracking.
inspections) and the most severe indications i
identified at the end of this operating interval 6.3 Justification for Continued OperatioD have adequate integrity based on in situ testing, it would be reasonable to conclude that the J
For the CE plants with original steam generators, the steam genrrator could be safely operated for a majority of circumferential indications have been similar operating interval, observed at the expansion transition region. De severity of these indications has led to midcycle (5) operatin; conditions at the plant (e.g., hot-leg inspections at one plant (i.e., ANO-2) and to the opeutmg temperature, water chemistry sleeving of all hot-leg expansion transitions at another practices) - Stringent water chemistry control plant (i.e., Maine Yankee).' In addition, a limited in accordance with industry standards designed number of circumferential indications, typically of to prevent uncontrolled tube degradation (i.e.,
less severity, have been observed at dented locations mitigate the initiation and propagation of SCC) and in sleeve joints, provides confidence that tube integrity will be 4
6-9 NUREG-1604
l l
l maintained consistent with previous (9) the risk and potential consequences of a range j
observations. Sludge lancing and cheaucal of steam generator tube rupture events as cleaning to reduce sludge accumulation discussed in Reference 27 - In Reference 27, provides confidence that impurities that may the staff estimated the risk contribution due to affect the rate of crack initiation and growth the potential for single and multiple steam and the ability to reliably inspect the tube are generator tube ruptures. In addition, this study controlled.
exanuned the expected consequences of steam generator tube rupture scenarios, including Since SCC is a thermally activated process, beyond design basis situations, such as the reducing the hot-leg operating _ temperature potential for release due to contammant bypass takes advantage of the temperature dependence via failed tubes concurrent with a breach of
=
of SCC growth and initiation rates (i.e., the secondary system integrity.
growth and initiation would be slower, with all other parameters being equal).
6.4 Tube Inspections (6) operating time until the next steam generator GL 95 03 requested, in part, a safety assessment tube inspections were to be conducted justifying continued operation on the basis of past inspection results and a summary of inspection plans (7) ' requirement to numitor primary-to-secondary for the next scheduled steam generator tube leakage and to shut down the plant when leak inspection outage as they pertain to the detection of rate limits are exceeded - Iankage monitoring circumferential cracking. The inspection plans were is a defense-in-depth operating practice that can to consist of both an initial scope and sample provide operators with a timely indication of a expansion criteria. For the CE units with alloy 600 steam generator tube leak or tube rupture, steam generator tubes, the staff summarized some of Nitrogen 16 monitors can permit faster the information provided by the licensees with respect i
identification and isolation of a steam generator to the previous and next aspection for each of the that has a tube that has degraded to the point of areas identified as being potentially susceptible to leakage.
Improved integrated leak rate circumferential cracking.
The designation of monitoring programs provide added confidence
' previous
- refers to inspections perfonned before tht mbe integrity will be maintained These issuing or responding to GL 95-03. The designation programs inture administrative limits on of "next" (and/or " future") refers to an inspection primary-to-secondary leakage. 'Ibey also use performed aAer issuing or responding to GL 95-03.
equipment and procedural upgrades to enable The phrase, "if detect", is used to describe the plant operators o detect and respond to changes inspection expansion criteria when a circumferential in steam generator primary-to-secondary indication is detected. In many instances, the next leakage, and to shut down the unit before a (and/or future) mspections have already. been significant leak or steam generator tube rupture completed as a result of the time taken to prepare this if tube degradation should exceed expected document for publishing.
values.
Table 6-2 summarizes the scope of the past and j
(8) - use ' of procedures (including emergency future inspections at the expansion transition along j
operating procedures) to diagnose and address with some pertinent notes. Tables 6-3,6-4, and 6-5 t
steam generator tube leaks and ruptures -
provide similar information for the dented locations, j
Procedures, equipment, and training programs U-bend region, and sleeve joints, respectively. Tube i
that are in place to identify and mitigate the inspections performed using a technique not capable consequences of failed tubes provide confidence of reliably detecting circumferentially oriented that, in the event of a loss of tube integrity, the degradation were recorded as 'None" in Tables 6-2 plant can be safely operated. These programs through 6-5.
In instances where the results of the include simulator training on steam generator tube inspections were readily available, the results tube leaks and ruptures for control room were included in the tables, as appropriate. For operators.
example, at St. Lucie 2, the results of the inspection were readily available, so the results were NUREG-1604 6-10
incorporated into the tables, as appropriate.
Acronyms and abbreviations used in the tables are explained in Appendix C.
As can be seen from evaluating the data in Tables 6-2 through 6-5, there are plant-specific differences in the mapection plans (e.g.,
probe type, scope of examination). 'these differences in the inspection plans were considered along with other plant-specific circumstances (e.g., preventive measures taken) in 4
evaluating the acceptability of a licensee's response as discussed in Section 6.3. For example, even though a licensee may tave implemented a smaller initial inspection scope than another licensee, this may have been considered acceptable if the cumulative operating time for the plant was less than that of the other plant (all other parameters being equal).
l The staff has reviewed the submissions provided by the licensees that have CE steam generators with alloy 600 tubes and has concluded that they contain the information requested in GL 95-03. General conclusions regarding the responses are discussed in Section 9.
iI f
e4 n
ll j
.. !i}
n NUREG-1604 6-11
1 00o 6
E Is E
I 5
i A
e
~d 5
s O
O 3
NUREG-1604 6-12 1
~
i
+
1
}
Secondary Face of Tubesheet (i.e., top \\
c of tubesheet) s y
y <
- Tube s
s s
Tube-to-Tubesheet
' Crevice
/
y Tubesheet Prima Face of Expansion Transition Tubes set j
i Partial-Depth Expansion Transition i
g Tube Secondary Face of Expansion ubesheet ( e., top l
s f Transition
-Tubesheet Full-Depth Expansion Transition Figure 6-2 Partial and full-depth expansion transitions 6-13 NUREG-1604
i Annulus Support Support Tube Flow Holes O
v O
o f
Broach-Quatrefoil (w
Iw les)
Support Tube Tube
/ Q4 r
Broach-Trefoil Eggerate Figure 6-3 Typical tube support configurations l
NUREG-1604 6-14 i
_ _ _ _ _ _ _ _ _ _ _ _. _ _. _... _., - _. _. _ _. _. _ _ _.~_.._. _.._..
i i
4 i
f i
Table 6-1 Millstone 2 steens generator tube ' -. Am (1987 to replacessent) 4 i
i j
Outage Date/ Designation Comments t
January 1987 (SA)
Outage as a mauk of a leaking circumferential crack near the top of tubeabeet on the hot-leg side.
January 1988 (EOC 8)
Hot-leg circumferentialindications detected near the top of tubeahest. 1987 leaker was pulled.
February 1989 (EOC 9)
Circumferential indications wm detected in both the hot and cold legs near the top of the tubesheet.
October 1989 (10A)
Scheduled midcyclo inspection (in part, because of the large number of cracks detected in EOC 9).
Circumferemialindications were detected in both the hot and cold legs near the top of the tubeabeet May 1990 (108)
Increasing primary-to-secondary leakage led to tube inspection cutage. Leakage attributed to leaking plugs and sleevw. Circumferentialindications were detected in both the hot and cold legs near the top of the tubesheet.
Septoneer 1990 (EOC 10)
Circumferentialindications were detected in both the hot and cold legs near the top of the tubesheet.
April 1991 (1I A)
Increasing primary-to-secondary leakage led to tube inspection outage 14akage attributed to a circumferential crack in chord region (near eggcrate support 8 and just below the 90-degra bend).
May 1991 (llB) lacreasing primary-to secondary leakage led to tube inspection outage. Iaakage attributed to a circumferential crack in chord region (at the 90-degree hot-leg bend above the ninth eggerate suppon). Circumferential indications were also detected in both the hot and cold legs near the top of the tubenheet.
January 1992 lacrening primary-to-secondary leakage led to tube inspection outage. Leakage attributed to a circumferential crack near the top of the tubeahat and a circumferential crack near the 90-degne hot-leg bend.
NUREG-1604 6-15
Table 6-2 Inspections at the expansion transition region in CE plants (Part 1)
Plant F r===Aa= Treasition F-.
- Transition Future Notes b
Past laspection laspection Arkansas Nuclear One 2 100% RPC in HL sludge 100% in HL with Appendix H circ.
Performing mid-cycle outages. Studge pile pile; 20% RPCinCL cracking technique.
constitutes -65% of tubes. References 44 and sludge pile in 2R10 46.
+
4 Calvert Chffs i 100% RPC in HL; 100% PP in HL Only 3 circumferential indicadons detected.
Sample distorted bobbin in References 58 and 60.
CL Caivert Cliffs 2 100% PP in HL; Sample 100% PP in HL distorted bobbin in CL Assessment of PP and RPC irxiicate reladvely the same detection threshold for condidons at the plant. 115 circumferential indications.
j i
References 58 and 60.
Fort Calhou.m 40% RPC in HL i
100% in HL with A eracking technique.ppendix H circ.Orifce plate prevents RPC insertion in HL for R1 will expand to CL through R18. Orifice plate irstalled to climinate l
based on EPRI uiteria, potenual for phosphate corrosion due to steam blanketing in the U-bend area. Orifice plate affects 22% of tubes. Will remove orifme plate in 1996 for inspections and install a new plate after inspecuans are complete. References 56 and 79, 1
Maine Yankee Sleeving HL; 100% RPC 20% C5 or PP in CL. If detect,100%
Penetrant testing on several tubes. Sleeving l
in CL with some PP and in CL entire HL Referein:es 59 and 80.
high frequency RPC in CL Palisades 200 tubes w/2 coil RPC in 20% RPC in HL(320 PP exams from References 68 and 81.
HL 20% sampic). If detect.100% in HL If widespread crackmg,20% RPC in CL i
Palo Verde 1 100% PPinHL; 20% PP 100% PP in HL References 36 and 82.
i in CL Palo Verde 2 100% PP in HL and CL in 100% PP in HL 500 PP exams in CL References 36 and 82, SO 21; 37% PPin HLin if detect in CL,100% PP in CL If SO 22; 100% PP in CL detect more than 50 indicathms in HL, in SO 22 20% PPin CL.
Palo Verde 3 100% RPC in sludge pilc 100% PP in HL 10% PP in CL References 36 and 82.
(i.e.,3% in SO 31 and 6% in 50 32) 1 San Onofre 2 800% RPC in HL 100% PP in HL.
75 % of cracks are ID iniunted based on phase
{
angle analysis. 12 circ. cracks idendf.ed in 1993 and 27 in 1995. References 65 and 83.
San Onofre 3 100% RPC in HL 100% PP in HL References 65 arx183.
NUREG 1604 6-16 I
l
Table 6-2 Inspections at the expansion transition region in CE plants (Part 2)
Notes Plant Expansion Transition Expandon Trandtion Future Past Inspection Inspection St. Imae i 100% RPC inllL; 3%
100% RPC in llL. 3% RPC in CL References 67 and 84.
RPC in CL St. Imcie 2 100% RPC inIIL: 3%
100% RPC in HL. 3% RPC in CL References 67 and 84.
RPC in CL
[ Performed 20% in CL sludge pile region with no iwlications detected.)
Waterford 3 65% RPC in IIL includmg 20% in ill plus 500 additional tubes References 85 and 86.
100% of sludge pile with Appendix H circ. cracking technique focused in sludge pile area, if detect,100% of both SGs.
NUREG-1604 6-17
Table 6-3 Inspections at dented locations in CE plants (Part 1) i Dented TSP Past l
Plant laspectica Dented TSP Future lagdlon Notes Arkansas Nuclear One -2
-20% RPC of 10th TSP dents None Will reassess every outage.
Dentmg observed in partial drilled TSPs.
locations =ith greater than 0.001 inch radial deformation were recorded as dents. Calibrano 0.001-inch radial dent corresponds to IV on 40 n:
0 kHz absolute. References 44 and 46.
Calvert Cuffs i RPC sampling of largest solid TSP dents 20% PP of dents > SV. If detect, expand based on EPRIguidelines, Calibranon. 4-20% holes set to 4V on all Calvert Cliffs 2 differernial channels. References 58 and 60.
20 PP exams oflargest solid TSP dents 20% PP of dents > SV. If detect, Cahbranon:
expand based on EPRIguidelines.
4-20% holes set to 4V on all differential channels. References 58 and 60.
Fort Calhoun 44 RPC exams 20% of dents > $V at 1(1 with Appendix H circ. cracking technique.
Cahbranon: 0 001 inch radial dent corresponds to IV on 400 kHz absolute. References 56 and 20% of dents at Hi < SV. If detect.
79.
100% at Hl 20% at H2 and so on until none detected. 20% sample of HL derus equal to or exceeding 0.010-inch deformation will be inspected with probe quahfied for circ. crack detecuen.
Maine Yankee 25 RPC exams 20% of dents > SV at drilled tube Cahbrauon: 0.010 radial dent corresponds to supports. If detect,100% of dents >
10V on 400 kHz absolute. Reported dents at IV.
SV and 20% of dents between IV and References 59 and 80.
SV. If detect in I to SV range,100%
of dents between IV and SV and 20%
of nondented intersections. If detect in nondented intersections,100% of all intersecnons and expand to cold leg drilied supports.
Pansades N/A N/A No dennng observed. Confirmed by visual examinanon in upper tube bundle region in 1 SG.
Not suscepuble (stainless steel supports and rigid secnndary water chemistry). References 68 and Palo Verde i 81.
N/A N/A Nocorrosiondenting. Cahbranon:
50 %
AVB/BW wear set to 5 volts (Vmax) on the 500/100 absolute mix New dents or changes in dent signals ( > -20 volts) require an RPC exam.
Extensive inspections at verucal straps and upper two tube supports. References 36 and 82.
Palo Verde 2 N/A N/A No corrosion deramg. Calibration:
50 %
AVB/BW wear set to 5 volts (Vrnax) on the 500/100 absolute mix. New dents or changes in i
dent signals (> -20 volts) require an RPC exam.
Extensive inspecnons at vertral straps and upper two tube supports. References 36 and 82.
hio Verde 3 N/A N/A Nocorrosiondentmg. Calibration: 50 %
VB/BW wear set to 5 volts (Vmax) on the A
500/100 absolute mix. New dents or changes in dent signals (> =20 volts) require an RPC exam.
Extensive inspections at verucal straps and upper two tube supports, References 36 and 82.
, San Onofre 2 N/A i
N/A No denung at support locations, Denting at 1TS-adjacent to tie rods. References 65 and 83.
San Onofre 3 N/A N/A No dennng at support locauona. Dentmg at 1TS-adjacent to tie rods. References 65 and 83.
NUREG-16G4 6-18
f i
i Table 6-3 Inspections at dented locations in CE plants (Part 2)
Plant Deuted TSP Past Dested TSP Future inspection Notes laspect&om St.1.acie i RPC sarnpling 20% of HL dents in at least i 50.
Cahbrauon: 4-20% ASME holes set to SV on 400/100 kHz mix. Larger dents given priority in sampling scheme. References 67 and 84.
St. Imcie 2 RPC sampimg 20% of HL dents in at least i SG.
Calibration: 4-20% ASME holes set to SV on
[irspected HL and CL dents with RPC 400/100 kHz mix. Larger dents given priority in with no indications detected.]
sampling scheme. References 67 and 84.
Waterford 3 RPC inspect any location RPC inspect any location with a bobbm Very little denting. Tubes adjacent to stay rods at with a bobbin signal signal greater than 10 volts and top of tubesheet and eggcrate supports are greater than 10 volts and significant growth (historic practice).
consulered mnat susceptible to denting.
significant growth References 85 and 86.
l l
t I
8 I
l i
(
1 i
e I
i l
I i
I.
2 t
6-19 NUREG-1604 1,
.c
i Table 64 Inspections in the U-bend region of small-radii tubes in CE plants (Part 1)
N U-bend Past inspectica U-bend Future Imp-tion Notes Arkansas Nuclear One. 2 None 20% PP in Rt. If detect,100% PP in References 44 and 46.
Rt. If detect in expansion, capand into R2.
Calvert Cliffs 1 None 20% PP in RI and R2. If detect.
References 58 and 60.
expand based on EPRI guidelines.
Calvert Cliffs 2 None 20% PP in R1 and R2. If detect, References 58 and 60.
expand based on EPRI guidelines.
Fort Calhoun None 20% in Rt through R4 with Appendix References 56 and 79.
H circ. cracking technique.
Maine Yankee 100% RPC in R1 through 100% RPC in R1 through R12 References 59 and 80.
R12
\\
Palisades None None.
References 68 and 81.
Palo Verde 1 100% RPC in R1 and R2 None.
First 18 rows are U-bend tubes. Remaining rows have double 90-degree bend tubes. Srnall radius U bends not susceptible to cire. crackmg based on evious exammation results. References 36 Palo Verde 2 100% RPC in R1 and R2 100% RPC in R1 and R2.
First 18 rows are U-bend tubes. Remaining rows have double 90 degree bend tubes. Small radius U-bends txt susceptible to circ. cracking based on previous examination results. References 36 and 82.
Palo Verde 3 100% RPC in R1 and R2 None.
First 18 rows are U-bend tubes. Remaining rows have double 9(kiegree bend tubes. Small radius U-bends not susceptible to circ. cracking based on previous examination results. References 36 and 82.
San Onofre 2 None None.
References 65 and 83.
San Onofre 3 None None.
References 65 and 83.
NUREG-1604 6-20
-.. -. - ~. _ -... - -.- ~.
~
_.. ~ - - -.
'I 4
1 l
h h
f j
Table 64 Inspections la the U-bend region of small-radil tubes in CE plants (Part 2) 1 Pinet U-bend Past leapact6es 14eed Fotore inspection Notes 1
St. Lucie i None None.
References 67 and 84.
4 St. Lucie 2 None None.
References 67 and 84.
1 i
i
]
Waterford 3 300 tubes in upper bundle None.
Steam blanket region not present in SGs.
region suscepnble to steam References 85 and 86.
drying i.
j i'
i i
1 f
v I
i ii i -
(
3 t
t i
i j
j i
i i
e i
i d
e i
a i
6-21 NUREG-1604
Table 6-5 Inspections at sleeve joints in CE plants (Part 1)
Plant Sleeve Past inspection Sleeve Future Inspection Notea Arkansas Nuclear One - 2 Nore (crosswound) 20% with Appendix H circ. crack 442 B&W kinedcal!y sleeved tubes presently in qualified technique. If detect,100%,
service. All sleeves installed in 1992 (2F92).
References 44 and 46.
Calvert Cliffs I N/A N/A No sleeves installed. References 58 and 60.
Calvert Cliffs 2 N/A N/A No sleeves installed. References 58 and 60.
Fort Calhoun N/A N/A No sleeves installed. References 56 and 79.
Maine Yankee N/A 20% C5 or PP If detect and Installed Westinghouse laser-weided sleeves associated with unique installation during 1995 outage. References 59 and 80.
condition,100% of sleeves with this condition. If detect and not associated with unique installation condition, 100 %.
Niisades N/A N/A No sleeves installed. References 68 and 81.
Palo Verde i N/A N/A No sleeves installed. References 36 and 82.
Palo Verde 2 N/A N/A No sleeves installed. References 36 and 82.
Palo Verde 3 N/A N/A No sleeves installed. References 36 and 82.
San Onofre 2 N/A N/A No sleeves installed. References 65 and 83.
San Onofre 3 N/A N/A No sleeves installed. References 65 arul 83.
NUREG-1604 6-22
-. -. - -... ~.
-..~.-..-
Table 6-5 Inspections at sleeve joints in CE plants (Part 2)
Fleet Sleeve Past le Sleeve Future I== par +6a=
Notes St. Lucie 1 N/A N/A No sleeves installed. References 67 and 84.
St. Luce 2 N/A N/A No sleeves installed. References 67 and 84.
Wawrford 3 N/A N/A No sleeves installed. References 85 and 86.
i l
i 6 23 NUREG-1604
l 7 WESTINGHOUSE STEAM GENERATORS 1
7.1 Design nozzle, flows through the U-tubes, and exits the lower plenum through an outlet nozzle. A divider The steam generators in currently operating plate in the lower plenum below the tubesheet i
Westinghouse units liave tubes that are made either separates the inlet and outlet primary coolant and
}
from alloy 600 or alloy 690. Dese steam generators directs the flow through the tubes. A cutaway view are further distinguished in that the alloy 600_ tubes of a typical RSG is provided in Figure 7-1, and a l
were either heat treated with a mill-annealing process conceptual illustration is provided in Figure 2-1.
j or a thermal treatment process, whereas all alloy 690 tubes received thermal treatmenti For ease of As of June 1995, there were 51 operating
]
reference, Appendix B contains several lists that Westinghouse units. Of these, 46 were operating provide information pertaining to the type of steam with steam generators with alloy 600 steam generator l
generators used at the various units. These lists tubes, while 5 (i.e., D.C. Cook 2, Indian Point 3, include an alphabetical listing by plant name, an North Anna 1 and 2, and Summer) had alloy 690 alphabetical listing by vendor, and an alphabetical tubes. Of the 46 units with alloy 600 tubes, 40 were listing by tube material and the method used to operating with the original steam generators, while 6 i
expand the tube into the tubesheet.
had replaced their steam generators. As of June
]
1995, a total of 11 Westinghouse units had j
This section describes the features of steam replacement steam generators (6 with alloy 600 tubes, generators in Westinghouse units with alloy 600 5 with alloy 690 tubes),
steam generator tubes in either the mill-annealed or i
thermally treated state. (Westinghouse plants with Between June 1995 and December 1996, three
]
alloy 690 steam generator tubes are discussed in additional Westinghouse units (i.e., Ginna, Catawba j
Section 8.) The first part of this section provides a 1, and Point Beach 1) replaced their steam generators generic description of the steam generator designs with steam generators with alloy 690 tubes. (Steam j:
used in Westinghouse plants, followed by generators with alloy 690 tubes are discussed in plant-specific design features that may affect a plant's Section 8.)
In addition, the licensee for Salem 1
susceptibility to circumferential cracking. Among the commenced replacement of the Unit I steam Westinghouse units with alloy 600 steam generator generators in 1996 with steam generators with j
tubes, there are a variety of steam generators, each thermally treated alloy 600 tubes.
1 with slightly different operating experience. As a result, depending on the location of the Each Westinghouse plant has two to four steam circumferential cracking being discussed, the generators depending on the plant design. The Westinghouse units are divided according to various number of tubes in a steam generator with alloy 600 design features, such as type of tube support plate tubes varies from approximately 3000 to 6000, and material, heat treatment of the tube, and/or the these tubes are arranged in a rectangular pattern as method used to expand the tube into the tubesheet shown in Figure 7-2.
(e.g., hydraulic).
A number of different Westinghouse steam generator 7.1.1 Generic Features designs exist in the United States. These designs include the model 27, 44, 51, D2, D3, D4, DS, E, The Westinghouse steam generators are recirculating and F steam generators. Some differences in these shell and U-tube heat exchangers with integral designs include the method and extent of tube moisture separation equipment. A large cylindrical expansion into the tubesheet, the type of heat vessel encloses an inverted U-shaped tube bundle treatment the tubes received, the type of tube support consisting of many thousands of individual tubes, plate material (i.e., carbon steel or stainless steel),
each welded to a thick plate (called a tubesheet) with the design of the preheater (if any), and the tube a hole for each tube end located near the bottom of diameter. In addition, differences exist among the the steam generator vessel. The reactor coolant steam generators of a given model. (For example, enters the hemispherical bottom head through an inlet the expansion transition for model 51 steam 7-1 NUREG 1604
. - -. - - - - -. -. _. -. - - - - - ~ ~ - - -. -. - ~ - -. - - -..... -
-. ~. - -.. -.
l 1
i generators can be one of several types.) Design treatment that a steam generator tube receives is very differences that are pertinent to steam generator tube important, since this treatment has an effect on the i
integrity (and particularly circumferential cracking) tubes susceptibility to corrosion. Most of the earlier are discussed in the following paragraphs.
In model steam generators have tubes that were addition, Appendix B contains information related to mill-==naaled at a relatively low temperature (i.e.,
i the steam generator design at a given plant.
Iow temperature mill-annealed). The residual stresses 1
and microstructure of the tube material are such that J
The tubes in the Westinghouse steam generators were these tubes are relatively susceptible to primary-and expanded into the tubesheet by a variety of methods.
secondary-side SCC (Ref. 24). Most of the later i
These tubes are expanded in both the primary coolant operating model steam generators (e.g., models DS, inlet region (hot leg) and the outlet region (cold leg).
E, F, and replacement steam generators) contain As a result, the expansion transitions are frequently tubes that were thermally treated. These thermally referred to as the hot-leg and cold-leg expansion treated tubes generally have an improved transitions, as appropriate. Typically, only one microstructure as a result of the precipitation of method was used to expand the tubes into the chromium carbides at the grain boundaries. In the tubesheet.
thermal treatment process, the chromium diffused from the grain interiors to the chromium-depleted Early Westinghouse steam generators were partially regions near the grain boundaries, thereby preventing mechanically rolled into the tubesheet (i.e., the tubes sensitization (Ref. 24).
were only expanded a few inches into the tubesheet).
t L.ater Westinghouse steam generators contained tubes h tubes in most Westinghouse steam generators that were mechanically (i.e., roll) expanded for the with alloy 600 tubes are supported with plates at a entire length (i.e., thickness) of the tubesheet. As a number of fixed axial locations along the tube bundle, l
result, the expansior transition, where the tube and have variously shaped bars and/or plates in the changes from being fully expanded against the U-bend region of the tube bundle. Most of the tubesheet to its nominal size, was located at, or near, earlier model steam generators (i.e., models 27,44,'
the top of the tubesheet. (Illustrations of partial-and 51 D2, D3, and D4) have horizontally oriented tube full-depth expanded tubes are provided in Figure supports that are drilled hole carbon steel plates.
6-2.)
Later model steam generators (i.e., models DS, E, F, and non-alloy 690 replacements) typically have in addition to mechanically expanding tubes into the horizontally oriented tube support plates which are tubesheet, tubes in some steam generators were constructed of stainless steel. Dese plates are either explosively expanded into the tubesheet using the of the drilled hole or the broached-quatrefoil design Westinghouse Explosive Tube Expansion (WEXTEX)
(refer to Figure 6-3).
process. The resultant expansion transition was located at, or near, the top of the tubesheet. Most The support structures in the U-bend portion of the later-model Westinghouse steam generators, and tubes are typically referred to as AVBs and are replacement steam generators, contain tubes that were diagonally oriented in the tube bundle. The AVBs in i
hydraulically expanded the full depth of the Westinghouse steam generators stiffen the tubes tubesheet.
and/or prevent tube-to-tube contact (Ref. 24).
In summary, the types of expansion transitions in The construction material and design of the tube i
Westinghouse model steam generators include support structures play an integral role in determining partial-depth roll expansions, full-depth roll the degradation mechanisms that may affect a tube, expansions, WEXTEX expansions, and hydraulic For example, carbon steel plates are more susceptible expansions.
The relative susceptibility of the to corrosion than stainless steel plates in the steam expansion transition to cracking depends, in part, on generamr operating environment. The corrosion of the type of expansion transition.
the plates can lead to tube denting and can limit the flow past the tube, resulting in the accumulation of in addition to the different tube expansion methods, corrosive impurities.
the different steam generator models have tubes that received different heat treatments.
The heat NUREG-1604 7-2
i 1
l l
2 Depending on the steam guerator model, the degradation was identified during these inspections secondary system water (i.e., feedwater) may pass (Ref. 87).
through a separate preheater region in the steam j
- generator. Feedwater for many of the earlier model Tube wear has been observed in the outer tube rows
[
steam generators (i.e., models 27, 44, and 51) does near the secondary side inlet nozzle in the preheater j
not pass through such a preheater. In these steam region in several Westinghouse steam generators.
generators, the feedwater is generally fed through a This wear has been observed both in the split-flow i
feedwater nozzle, to a feedring, into the downcomer, preheaters (D2 and D3) and in the counterflow where it mixes with recirculating water draining from prhrs (D4, DS, and E). The degradation was i
j the moisture separation equipment. Later-model determined to be caused by large flow velocities and t
i steam generators typically have a separate economizer turbulence, as well as insufficient tube restraint. b l
section (i.e., preheater) near the cold-leg outlet. b wear / fretting problem in the D2 and D3 steam i
feedwater flows into the preheater through a nozzle generators was addressed by redistributing feedwater
{
located in the lower part of the vessel, and auxiliary flow between the primary'and auxiliary feedwater feedwater is injected through a separate nozzle in the inlets to reduce the flow into the preheater through j
upper part of the vessel. Two different types of the primary inlet, and by incorporating a preheater i
preheaters have been used in the Westinghouse steam manifold to reduce cross-flow vibration. For the
{
generators.
These include split-flow preheaters counterflow preheaters, the problem was addressed (models D2 and D3) and counterflow preheaters by expanding the tubes within the tube baffle plates j
(models D4, DS, and E). In the model F steam at certain id.c.:s locations, in effect changing the i
generators, which represent some of the newer steam tube's natural frequency. In addition, the feedwater generators, preheaters were not included in the flow was also split between primary and auxiliary i
design. Figure 7-3 illustrates a typical Westinghouse inlets on the D4 and D5 steam generators (Ref. 24).
j steam generator with a preheater, h stresses associated with these transitions (i.e., the 4
expanded tubes within the baffle plates of the
}
7.1.2 Plant-Specific Features preheater) may lead to corrosion cracking, although j
this is less likely than other transitions (such as the There are unique features that affect steam generator hot-leg expansion transition) because of the lower tube integrity in several of the Westinghouse RSGs.
temperature in the preheater region (compared to the h following paragraphs describe some of the more hot leg).
i pertinent features that may affect a tube's j
susceptibility to circumferential cracking or its At Byron 1 and Braidwood I, several tubes were i
inspectability, expanded both above and below various tube support j
plate elevations. These expansions were performed To field test advanced steam generator tubing to lock the tube support plates in place to permit the l
materials, a few plants installed implant sections use of higher-voltage limits than those described in (implant tubes) with sleeve-like inserts in selected GL 95-05 (Ref.1) for predommantly axially oriented tubes during initial steam generator fabrication.
ODSCC. b stresses associated with the transitions Diablo Canyon 1 is one of the few Westinghouse at the tube support plate elevations of these expanded model 51 steam generator plants which installed these tubes may lead to corrosion cracking, although this is sleeve-like inserts. To install the implant tubes at less likely than other transitions because these tubes Diablo Canyon 1, approximately 76 cm (30 inches) are at a lower temperature since they were removed of tubing from the hot-leg side of 16 steam generator from service by plugging.
tubes was removed, and the 76 cm (30-inch) implant tubes were installed and joined to the respective 7.2 Locations Suscentible to Circumferential parent tubes by a 2.5 cm (1-inch) sleeve-like insert, Crackms which was then welded. The licensee for Diablo Canyon I has not identified the inserts as being At the time of the original responses to GL 95-03, susceptible to circumferential cracking; however, the the Westinghouse Owners Group (WOG) identified licensee did inspect the insert region of the implant three regions as having exhibited circumferential tubes with a rotating pancake coil probe during the cracking. Specifically, these regions included the sixth refueling outage in 1994. No indication of expansion transition area located within, or near, the 7-3 NUREG-1604
=
1 i
i tubesheet, the Row I and 2 U-bend area, and the these plants were mechanically rolled for only a few dented tube support plate intersections.
inches into the tubesheet starting at the primary face Circumferential indications had also been detected in of the tubesheet. As a result, a crevice exists sleeve joints.
between the tube and the tubesheet for the portion of the tube within the tubesheet that is not expanded.
7.2.1 Expansion Transition and/or Top of Tubesheet Since the tube was mechanically Mlal into the Region tubesheet, the expansion transition is commonly referred to as a roll transition. Plants with steam Tubes in Westinghouse damm generators have been generator tubes that are only roll expanded for a expanded into the tuba =haat using a variety of small distance within the tubesheet are referred to as methods (e.g.,
mechanically, explosively, and partial-depth hardroll plants.
All partial-depth hydraulically) and the tubes were expanded either a hardroll plants have mill-annealed alloy 600 tubes, partial distance into the tubesheet (typically 2.5 to 7.6 which are supported by carbon steel tube support cm (1 to 3 inches)) or for the full length of the plates along the straight span portion of the tubes.
tubesheet (approximately 61 cm (24 inches)) as depicted in Figure 6-2.
The older Westinghouse The crevice region has been a site for corrosion that steam generators typically have tubes that were has led, in part, to extensive sleeving campaigns in partially expanded. into the tubesheet, whereas the tubesheet region. In addition, degradation in the later-model Westinghouse steam generators all have crevice and/or at the roll transition has led some tubes that were fully expanded throughout the length plants to re-roll the tubes above the original rolled of the tubenheet. Partial-depth tube expansions were region to provide a new pressure boundary free of only used in steam generators that were mechanically any detectable defects. By so doing, the licensees expanded (i.e., rolled) into the tubesheet. These have been able to apply alternate repair criteria, such plants are commonly referred to as partial-depth as the F-star criteria, in which tubes with degradation hardroll plants. Mechanically rolling the tube the full a certain specified distance below the top of the length of the tubesheet was also used in several tubesheet or the bottom of the roll transition plants, which are commonly known as full-depth (whichever is lower), may remain in service hardroll plants.
Other Westinghouse steam regardless of the depth of degradation. Tubes generators have tubes that were either explosively or accepted for continued service on the basis of the hydraulically expanded the full length of the F-star criteria have structural and leakage integrity tubesheet. The explosive tube expansion process is consistent with regulatory criteria. In summary, the referred to as the WEXTEX process, and plants with expansion transition (or roll transition in this case)
WEXTEX expansion transitions are referred to as for steam generators with partial-depth expansions is WEXTEX plants. Plants with steam generator tubes typically located well below the top of tubesheet; that have been hydraulically expanded the full length however, the possibility exists that licensees with of the tubesheet are commonly referred to as such steam generators may eventually re-roll the tube hydraulically expanded plants. Most of the newer the entire length of the tubesheet, steam generators have tubes that have been hydraulically expanded into the tubesheet.
The plants with partial-depth hardroll steam generators are among the oldest PWRs in this Since the circumferential cracking operating country. The steam generators for these plants are of experience at the expansion transition varies with the various models, including model 27 steam generators type of transition, the operating experience for each (Haddam Neck), model 44 steam generators (Ginna, type of transition is discussed separately in the Indian Point 2, and Point Beach 2), and model 51 following sections.
steam generators (D.C. Cook 1, Kewaunee, Prairie Island I and 2, and Zion I and 2). As of December 7.2.1.1 Partial-Depth Hardroll Steam Generators 1996, Ginna had replaced its steam generators with a more advanced design, and Point Beach 2 was in the Of the Westinghouse plants operating in June 1995, process of a similar replacement. The replacement 10 plants had steam generators with expansion steam generators at Ginna and Point Beach 2 have transitions located well within the tubesheet region tubes that have been hydraulically expanded the full (refer to Appendix B). The steam generator tubes for depth of the tubesheet.
NUREG-1604 74
i In general, plants with steam generators that have located at the expansion transition region. The only been panially expanded into the tubeeheet have indications at Indian Point 2 are believed to be exhibited very little circumferentially oriented closely spaced axial cracks, although no tube pulls j
degradation either at the expansion (i.e., roll) have been performed to confirm this, and the j
transition or at the top of the tubesheet. Some of the indications at D.C. Cook I are attributable to cellular more pertinent experience is summarized in the corrosion rather than distinct circumferential cracks.
following paragraphs. In general, most of these i
plants have exhibited axially oriented PWSCC at the 7.2.1.2 Full Depth Hardroll Steam Generators
)
roll transition, axially oriented ODSCC in the tubesheet crevice region, and/or intergranular attack Of the Westinghouse plants operating in June 1995, j
in the tubesheet crevice region.
11 plants had steam generator tubes that were j
mechanically expanded (i.e., rolled) for the entire
{
{
In 1992, the licensee for D.C. Cook I detected length (i.e., thickness) of the tubesheet (refer to l
j indications of a circumferential nature in several Appendix B). As a result, the crevice that existed i
steam generator tubes using a rotatmg probe with a between the tube and the tubesheet in partial-depth j
pancake coil (i.e., a rotating pancake coil probe) at expanded tubes, no longer exists, and the expansion i
the top of the tubesheet region. Four tubes were transition is located at or near the top of the l
pulled to characterize the degradation being observed.
tubesheet. As with the partial-depth hardroll plants.
l The destructive examinations revealed that these the expansion transition is commonly referred to as a j
indications were attributable to outside diameter roll transition, since the tube was mechanically rolled
. initiated cellular corrosion. In cellular corrosion, into the tubesheet. Plants with steam generator tubes f
axial cracking dominates the crack morphology; that have been roll expanded for the entire thickness however, short circumferential crack components of of the tubesheet are referred to as full-depth hardroll
)
lesser depth than the dominant axial cracking can plants.
All full-depth hardroll plants have cause linkage between the axial cracks.
The mill-annealed alloy 600 tubes supported by carbon 1
l degradation was attributed to localized denting of the steel tube support plates along the straight span l
tube (Ref. 88).
portion of the tubes. Full-depth hardroll plants have j
tubes with outside diameters of either 22.22 mm (7/8 j
In 1995, circumferentially oriented indications in the inch) or 19.05 mm (3/4 inch).
partial-depth expansion transition region were reported at Indian Point 2. Although no tube samples
'Ihe steam generators for full-depth hardroll plants j
were obtained to verify the degradation morphology, are of various models. That is, there are model 51 the indications are believed to be bands of closely and SIM steam generators (Farley 2 and Beaver spaced axial cracks, rather than distinct Valley 2, respectively), model D2 steam generators circumferential cracks (Ref. 88). Axial primary (McGuire 1), model D3 steam generators (Catawba j
water stress corrosion cracks at the roll transition 1, McGuire 2, and Watts Bar 1), model D4 steam j
have been detected at Indian Point 2.
generators (Braidwood 1. Byron 1, Comanche Peak 1, and Shearon Harris 1), and model E steam i
In 1991, indications of circumferential cracking were generators (South Texas Project 1). As of December l
detected at the hot-leg roll transitions at Ginna.
1996, Catawba i had replaced its steam generators i
Subsequent inspections before steam generator with a more advanced design that has full-depth replacement identified additional indications of hydraulically expanded tubes. Farley 2 and Beaver j
circumferential cracking at the hot-leg roll transitions.
Valley 2 (model 51 and 51M steam generators) are i
Inspections of the cold-leg roll transitions were the only two plants that use 22.22 mm (7/8-inch)
~
performed, and no circumferential indications were outside diameter tubing with full-depth mechanical observed.
expansions.
- t
{
In general, panial-depth hardroll plants have In general, plants with steam generators that have exhibited very little circumferential cracking at the been mechanically expanded the full length of the j
expansion transition. For partial-depth hardroll plants tubesheet have exhibited circumferentially oriented operating in June 1995, the eddy current indications hgradation primarily at the expansion (i.e., roll)
{
at Ginna are perhaps the only circumferential cracks vansition. Some of the more pertinent experience is 1
7-5 NUREG-1604 i
l
l summarized below. In addition to circumferential of both circumferential and axial cracking. Beaver cracking at the expansion transition, most of these Valley peoned before commercial operation (Ref. 91),
plants have exhibited primarily axially oriented ODSCC at the tube support plate elevations.
Subsequent to June 1995, a number of circumferential indications were detected at both Beaver Valley 2 and Of the 11 operating plants with steam generators that Farley 2. The licensee for Beaver Valley 2 reported have full-depth mechanical expansion transitions,7 identifying 21 circumferential indications at the had exhibited circumferential cracking at the roll expansion transition during their fall 1996 outage transition as of June 1995. These seven plants (Ref. 92).
At Parley 2, 192 cimumferential include Braidwood I, Byron 1, Catawba 1, Farley 2, indications were identified during their fall 1996 McGuire 1 and 2, and South Texas Project 1. b outage. Most of the circumferential indications at four units that did not exhibit circumferential Farley 2 were attributed to ODSCC. The licensee indications include two units that just recently for Farley 2 attributed the increase in the number of commenced commercial operation (Watts Bar 1 and detected indications at the expansioe transition, in Comanche Peak 1) and one unit that has 22.22 mm part, to the first time use of the plus-point probe (7/8-inch) (outside diameter) tubing (Beaver Valley (Ref. 93).
2).
Several tube specimens were removed from the steam Peening of the expansion transitions was performed generators at plants with full-depth hardroll expansion at all of the plants with steam generator tubes that transitions to characterize circumferentially oriented were full-depth hardroll expanded into the tubesheet.
tube degradation at the roll transition. These tube Both rotopeening and shotpeening have been specimens. include both 22.22 mm (7/8-inch) and performed. In some instances, the peming was done 19.05 mm (3/4-inch) outside diameter tubes. The
^
before commercial operation while in other instances results from several of these specimens are discussed it was performed after commencing commercial below.
operation. The shotpeening and rotopeemag process impart a thin layer of compressive stress on the tube Several tube specimens were removed from Farley 2 i
inside diameter which is intended to inhibit the (which has 22.22 mm (7/8-inch) outside diameter L
initiation of PWSCC (Ref. 89).
Nonetheless, tubes) for destructive examination of indications at circumferential cracks have been found in these the tube support pine elevations; however, plants, regardless of when the peemng was information regarding the integrity of the expansion 1
performed. However, the predommant mode of transition region was obtained. For example, one circumferential cracking in units with full-depth tube pulled from Farley 2 in 1989 had some expansion transitions has been initiated on the cimumferentially oriented PWSCC at the roll secondary side rather than the primary side, transition, while the other tube had no detectable suggesting that the peening application may be degradation at the roll transition. Additional tube successful at mitigating the initiation of PWSCC, pulls performed by the licensee for Farley 2 (for particularly when applied before operation (Refs. 89 reasons not related to circumferential cracking at the and 90). Peening may delay the onset of primary expansion transition) did not reveal any significant j
side cracking, although no formal assessment was roll transition degradation (Ref. 90). The licensee for j
performed as part of this review.
Farley 2 also removed several tube specimens during their 1996 outage to characterize the cracking (axial At Beaver Valley 2 and Farley 2 (which have and circumferential) occurring at the expansion 22.22 mm (7/8-inch) outside diameter tubes), a low transition (Ref. 93). The results from the destructive rate of occurrence of circumferential cracking was examination of these specimens were not available at observed through June 1995. As of June 1995, the time this report was prepared.
circumferential indications at the expansion transition had not been observed at Beaver Valley 2, and only The licensee for McGuire removed a tube from one one circumferential indication had been detected at of the McGuire I steam generators for destructive
)
Farley 2.
As with the other full-depth hardroll examination in January 1990. McGuire I has steam j
l plants, both of these units peened the hot-leg generators with 19.05 mm (3/4-inch) outside diameter expansion transition region to mitigate the occurrence tubes. The pulled tube was burst tested and exhibited NUREG-1604 76 j
~
a burst pressure of 648 bar (9400 pounds per square cracks. b corrosion was contained to at least a 7.6 inch (psi)) and bent over before bursting. The crack mm (0.3-inch) band at the roll transition, and there initiated on the outside of the tube and was were ligament areas of non-corroded material intergranular in nature. The crack was through-wall between the stress corrosion cracks. Some of the for 90 degrees and less than 40-percent through-wall cracks extandad through the tube wall, and some of ODSCC for the remaining part of the circumference the outside diameter cracks had an axial orientation.
(Refs. 52 and 90).
Furthermore, there were some inside diameter initiated axial cracks (i.e., axially oriented primary In 1991, a tube was pulled from Catawba 1 (19.05 water stress corrosion cracks). ' These inside diameter mm (3/4-inch) outside dia== ear tubes) for a initiated cracks did not interact with the outside circumferential indication at the top of the tuba =haae.
diamater initiated cracks. The licensee's conclusions This tube had a burst pressure of 634 bar (9200 pai).
from the destructive exammation of the pulled tubes b corrosion for this tube initiated from the outside is that the eddy current indications are difficult to size of the tube b degradation was characterized as a (in terms of average and maximum depth and 360-degree band of circumferential intergranular circumferential arc length) to quantitatively assess the attack which was less than 15-percent through the structural and leakage integrity of the tubes; however, tube wall (Ref. 52).
the tubes have high structural integrity for the same reason the indications are difficult to size (i.e., the Tubes that were pulled from the hot leg of the steam presence of ligaments of non-corroded material) generators at South Texas Project I (19.05 mm (Ref. 95).
(3/4-inch) outside diameter tubes) to characterize the nature of circumfemotial indications at the expansion Whereas most full-depth hardroll plants have transition indicate that the morphology of the observed circumferential indications at the expansion crackmg is ODSCC. The locations of the tubes with transition, the number of tubes with these indications eddy current indications indicate anhancad varies from plant to plant. At Braidwood I and-susceptibility in the central portion of the tube bundle Byron 1 (which have 19.0f e (3/4-inch) outside on the hot leg. b licensee concluded that this diameter tubes), a @ M et number of aah=arad susceptibility is most likely the result of circumfematial indicatione % been detected at the thermal hydraulic conditions in this area, which affect roll transition (approximately 300 affected tubes at the local hot leg top-of-tubesheet environment Byron I and approximately 1000 tubes at Braidwood (Ref. 94). Mapping the location of the detected 1). The licensees for the other full-depth hardroll circumferential indications reveal a preference for plants have not observed the number of reported hardened sludge piles concentrated around circumferential indications at the expansion transition the hot-leg T-slot (Ref. 90). The T-slot is an that Braidwood and Byron 1 have.
"untubed' region in the tube bundle, which is in the shape of a 'T". This region contains steam generator As a result of the structural and leakage integrity of blowdown piping at the top of the tubesheet and was the steam generator tube bundle, two full-depth designed with the objective of enhancing sludge hardroll units have performed midcycle steam removal from the steam generator.
generator tube inspections. At Braidwood I, a midcycle inspection commenced in October 1996 as Approximately 12 tube sections were removed from a result of circumferential cracking at the roll the steam generators at Byron 1 (19.05 mm transition. At Byron 1, a midcycle inspection was (3/4-inch) outside diameter tubes) during the performed in November 1995 as a result of axially 1994-1996 time-frame to address circumferential oriented ODSCC at the tube support plate elevations cracking at the expansion (i.e., roll) transition. b and, in part, because of circumferential cracking at destructive exammation revealed that the morphology the roll transition. These midcycle inspections were of the degradation was consistent with tubes pulled warranted on the basis of the structural and leakage from other steam generators. b circumferential significance of the degradation in the tube bundle (not indications at Byron I were attributed primarily to simply on the number ofindications detected). That oukide diameter initiated stress corrosion cracks.
is, from a safety standpoint, the significance of the b crack network consisted of multiple initiation indications detected (e.g., length and depth) is more sites for short, discrete, non-coplanar circumferential important than the number of indications.
At 7-7 NUREG-1604
ls A
k Braidwood and Byron 1 (as of December 1996), all 22.22 mm (7/8-inch), with the exception of of the circumferential indications detected were Comanche Peak I which has 19.05 mm (3/4-inch) capable of withstanding the required pressure outside diameter tubes. (Comanche Peak I only has loadings (although the operating interval was limited).
a limited number of WEXTEX expansion transitions; t
the majority of the expansion transitions are In summary, plants with full-depth hardroll expansion full-depth hardroll expansion transitions.)
At transitions have exhibited circumferential cracking at Comanche Peak 1,
to facilitate a stayrod the expansion (i.e., roll) transition as evidenced modification,3839 tubes were removed in the shop, through tube inspections and the destructive When the tubes were replaced, the WEXTEX process examination of pulled tubes. The vast majority of the was used.
indications have been located at the hot-leg roll transition; however, circumferential indications at the With the exception of Comanche Peak 1, the steam cold-leg roll transition have been observed in at least generators for WEXTEX plants are all model 51 one unit (i.e., Byron 1) (Ref. 96). As of December steam generators (Beaver Valley 1, Diablo Canyon 1 1996, the extent of the cracking in terms of the and 2, Farley 1, Salem 1 and 2, and Sequoyah 1 and number of tubes affected was more severe in plants 2). The steam generators at Comanche Peak I are with 19.05 mm (3/4-inch) outside diameter tubes than model D4 steam generators. As of 1996, plans in plants with 22.22 mm (7/8-inch) outside diameter currently exist to replace the Salem I steam tubes; however, all full-depth hardroll plants with generators with the steam generators originally alloy 600 mill-annealed tubes are susceptible to purchased for the cancelled Seabrook unit (Ref. 97).
circumferential cracking regardless of tube size. The severity of the circumferential indications at several In general, plants with steam generators that have plants has resulted in the performance of mideycle WEXTEX transitions have exhibited eircumferentially steam generator tube inspections.
oriented degradation primarily at the expansion (i.e.,
roll) transition.
Some of the more pertinent 7.2.1.3 WEXTEX Steam Generators experience is summarized below. In addition to circumferential cracking at the expansion transition, Of the Westinghouse plants operating in June 1995, most of these plants have exhibited primarily axially nine plants had steam generator tubes that were oriented ODSCC at the tube support plate elevations.
expanded the entire length (i.e., thickness) of the tubesheet using the Westinghouse Explosive Tube In the 1991-1992 time-frame, the WEXTEX owners Expansion (i.e.,
WEXTEX) process (refer to issued guidance for a top-of-tubesheet (i.e.,
Appendix B). The WEXTEX tube expansion process WEXTEX transition) inspection plan.
This uses an explosive charge to produce tube-to-tubesheet inspection plan divides the tubesheet into four distinct contact throughout the tubesheet region. Compared zones. Zone 4 represents the central region of the to the mechanical rolling process, WEXTEX tube tubesheet and is coincident with the low cross-flow expansion generally produces lower residual stresses velocity region where most sludge accumulation within the expanded-to-unexpanded tube transition occurs. Additionally,about 95 percent of allindustry region.
The WEXTEX process has been WEXTEX cracking has been identified in Zone 4.
implemented only in alloy 600 mill-annealed tubing The WEXTEX inspection plan uses an initial sample (Ref. 90). As a result of the WEXTEX process, the size of 50 percent of the active Zone 4 tubes (700 expansion transition is located at or near the top of tubes minimum, which is approximately 20 percent of the tubesheet.
the total number of steam generator tubes) utilizing the motorized rotating pancake coil probe for the Plants with steam generator tubes that were detection of circumferential indications at the top of explosively expanded into the tubesheet by the the tubesheet. All steam generators are inspected Westinghouse process are referred to as WEXTEX with this minimum sample size during each refueling plants. All WEXTEX plants have mill-annealed alloy outage. The WEXTEX transition region is defined as 600 tubes supported by carbon steel tube support the area from the bottom of the WEXTEX transition plates along the straight span portion of the tubes.
to 12.7 mm (0.5 inch) above the bottom of the Furthermore, all plants with WEXTEX expansion WEXTEX transition. (The bottom of the WEXTEX transitions have tubes with outside diameters of transition is the first point of contact between the tube NUREG-1604 7.g l
i
~\\
I l
f and the tubesheet, when traversing the tube from the performed after commencing commercial operation.
unexpanded portion to the expanded portion.) h As discussed earlier, the peening process imparts a j
inspection focuses on the region of the tube bundle thin layer of compressive stress on the tube's inside with the largest potential for finding indications, and diameter in order to inhibit the initiation of PWSCC thus results in the highest likelihood for requiring an (Ref. 89). Regardless of when, or if, the peening j
expansion of the sample size (Refs. 87 and 91).
was performed, circumferential indications have been i
j found at the expansion transition.
Unlike the W criteria for expanding the initial sample size for experience for full-depth hardroll plants, the WEXTEX steam generator tubes are predicated on circumferential cracking in units with WEXTEX i
the number of indications found in the WEXTEX transitions appears to predominantly initiate from the 3;-
transition region and their circumferential extent.
primary side rather than from the secondary side The expansion criteria require an additional 8-percent (i.e., the outside diameter). However, as mentioned increase in the sample size in the affected steam earlier, the peening process may delay the onset of generator for the first indication identified, and then the cracking (although no formal assessment was l
a stepwise increase in the sample size of at least performed as part of this review).
l 10-percent in the affected steam generator for each subsequent circumferential indication found.
Several tube specimens were removed from the steam l
l Expansion of the sample to include 100 percent of the generators at WEXTEX plants to characterize tubes in the affected steam generator would be circumferentially oriented tube degradation at the i
performed if more than seven circumferential expansion transition. The results from several of
(
indications were found (as summed over all inspected these specimens are discussed below.
tubes in the initial or expanded samples) or if any circumferential or axial indication exceeds the Circumferential cracking in WEXTEX transitions was j'
structural criteria (Ref. 87).
Additional sample first observed at North Anna 1 in 1987. Tube pull I
expansion criteria exist; however, the above criteria results from two tubes removed from North Anna 1 indicate the general program.
revealed circumferentially oriented PWSCC.
i 1
Non-corroded ligaments were detected in the crack l
Of the nine operating plants with steam generator network of both tube pull specimens. The measured tubes that have WEXTEX transitions (including burst pressures for these tubes exceeded 620 bar Comanche Peak 1, which only has a limited number (9000 psi), which is well above the structural criteria 3
of WEXTEX transitions), seven had exhibited contained within Reference 33 (Ref, 90).
circumferential cracking at the expansion transition as l
of June 1995. These seven plants include Beaver Other tubes removed from WEXTEX plants were i
Valley 1, Diablo Canyon 2, Farley 1, Salem 1 and 2, removed primarily to characterize predominantly i
and Sequoyah I and 2. The two units that had not axially oriented ODSCC at the tube support plate j
exhibited circumferential indications include elevations, l.aboratory analysis of the expansion Comanche Peak 1 (which is a relatively new plant transition portion of these tubes revealed degradation that has only 3839 tubes with WEXTEX transitions) in several instances (Refs. 90 and 91). In other and Diablo Canyon 1. However, subsequent to June instances, no circumferentially oriented degradation
]
1995, a circumferential indication was detected at the was observed at the expansion transition (Ref. 99).
expansion transition in Diablo Canyon 1 (Ref. 98).
7.2.1.4 Hydraulic Steam Generators Peening of the expansion transitions was performed 1
in at least seven of the nine plants with steam Of the Westinghouse plants operating in June 1995, 4
generator tubes that have WEXTEX transitions.
17 had steam generator tubes that were hydraulically Neither Beaver Valley I nor Farley 1 indicated that expanded the entire length (i.e., thickness) of the f
peening had been performed in the steam generators tubesheet (refer to Appendix B).
Compared to at their plants. The steam generator tubes at the mechanical rolling or WEXTEX processes, the i
other WEXTEX units were peened using the hydraulic tube expansion process generally reduces
]
shotpeening process. In one instance (i.e., Comanche residual stresses within the expanded-to-unexpanded l
Peak 1), the peening was done before commercial tube transition region (Ref.100). As a result of the
}
operation, while in the other instances it was hydraulic process, the expansion transition is located I
i 79 NUREG-1604 i
l at or near the top of the tubesheet, Plants with steam above, Callaway is one of two plants with generator tubes that were hydraulically expanded into -
mill-annealed tubes with hydraulic expansion the tubesheet are referred to as hydraulic plants. All transitions; the other is South Texas Project 2. In alloy 600 hydraulic plants have tubes supported by April 1995, the licensee for the Callaway plant stainless steel tube support plates along the straight identified circumferential indications at the hot-leg span portion of the tubes. De alloy 600 tubes for all expansion transition in 10 tubes. In addition,15 but two of these plants were thermally treated rather tubes were identified with axial indications at the than mill-annealed. The alm-ter of the steam expansion transition, including 2 tubes with both axial generator tubes varies from plant to plant. In and circumferential indications. This was the first general, however, the diameters used include 19.05 outage during which any evidence of cracking was mm (3/4-inch) (e.g.', Comanche Peak 2), 22.22 mm observed. His indicates that axial indications need (7/8-inch)(e.g., Point Beach 1, Surry 1 and 2), and not be observed before finding circumferential 17.46 mm (11/16-inch)(e.g., Callaway, Millstone 3, indications. (That is, other forms of degradation Seabrook I, Vogtle 1 and 2, and Wolf Creek 1).
need not precede the detection of circumferential indications.) It is important to note that only the The steam generators for hydraulic plants are of mill-annealed tubes at Callaway exhibited indications various models, including model 44F (Point Beach 1, of cracking during the 1995 mspection (i.e., no Robinson 2, and Turkey Point 3 and 4), model 51F cracking was observed in the thermally treated tubes (Surry 1 and 2), model DS (Braidwood 2, Byron 2, during the 1995 inspection).
l Catawba 2, and Comanche Peak 2), model E (South l
Texas Project 2), and model F (Callaway, Millstone At the Callaway plant, additional indications of tube 3, Seabrook 1, Vogtle 1 and 2, and Wolf Creek 1).
degradation at the hot-leg expansion transition were The steam generators currently in use at Point observed in 1996. During the 1996 mspection
[
Beach 1, Robinson 2, Surry 1 and 2, and Turkey outage, both axial and circumferential indications l
Point 3 and 4 are replacement steam generators, were identified (along with volumetric indications).
l Two 'of these plants (Callaway and South Texas However, unlike the 1995 inspection, indications l
Project 2) have mill-annealed tubes, while the others were found both in thermally treated -and have thermally treated tubes. Callaway, which is one mill-annealed tubes. Although the licensee has not l
of the oldest plants with hydraulic expansion removed any tubes for destructive examination to
~
l transitions, is unique in that the first 10 rows of the characterize the nature of these indications, one can steam generators contain thermally treated tubes, speculate that these indications at the expansion while the remaining rows contain mill-annealed transition are possibly attributed to SCC.
tubing. Both Callaway and South Texas Project 2 have shotpeened the hot-leg expansion transition In 1990, two tubes were pulled from the Surry 1 l
region. South Texas Project 2 also shotpeened the replacement steam generators to characterize l
cold-leg expansion transition region.
circumferentially oriented degradation detected in the field at the expansion transition.
Dese steam j
In general, plants with steara generators that have generators contain thermally treated alley 600 tubes, hydraulic transitions have not exhibited which were hydraulically expanded tire full depth of l
circumferentially oriented degradation at any location.
the tubesheet. For one of the pulled tubes, the The one exception is Callaway, which detected licensee concluded that the poorly defined rotating circumferential indications at the hydraulic expansion pancake coil probe signal was similar to that of a transitions in 1995 and in 1996. De relatively good ding or mechanical deformation. Upon destructive operating experience at these plants is a result of a testing of this tube, no SCC was detected on either combination of design and operational improvements.
the inside or outside diameter. The source of the l
The following paragraphs summarize some of the eddy current signal was therefore attributed to probe i
more pertinent experience.
liftoff in the expansion transition and nwchanical imperfections in the tube resulting from the tube Of the 17 operating plants with steam generator tubes installation process. For the other pulled tube, the that have hydraulic expansion transitions, only maximum diameter occurred approximately 15 mm l
Callaway had exhibited circumferential indications at (0.6-inch) above the top of the tube 4heet.
A the expansion transition as of June 1995. As noted 70-degree mechanical groove was found on the i
I NUREG-1604 7-10
~ ~,, - -
-a,
~
outside diameter of the tube and the licensee from rolling of the tube into the tubesheet, tube attributed this groove to the interaction of the tube ovality, and residual stresses resulting from the with the edge of the tubesheet during the expansion manufacturing process (Ref.104). Tube ovality can process (Refs.101 and 102). Although the hydraulic result from the manufacturing process and from tube expansion process was designed to locate the denting. Tube denting, if excessive, can lead to tube transition slightly below the top of the tubesheet, the support plate deformation and eventually to closure of licensee concluded that the tube was overexpanded the support plate flow slots (i.e., a row of rectangular
]
above the top of the tubesheet. Several other units slots located in the tube support plate between the hot also have steam generator tubes that were and cold-leg side of the steam generator). Closure of j
overexpanded (i.e., expanded above the top of the the upper support plate flow slots can induce bending i
tubesheet), including Turkey Point 3 and 4.
One of the tube and ovality at the apex of the inner row licensee indicated that the overexpansion of these U-bends (Ref. 30).
j tubes can produce residual stresses in the affected j
locations, which could make them more susceptible In addition to the stresses on a tube, tube to cracking than areas that are not overexpanded microstructure affects a tube's susceptibility to SCC.
{
(Ref.103).
Tube microstructure can vary from tube-to-tube and
)
includes properties such as grain size, carbide J
{
As previously noted, the relatively good merating distribution, chemistry, and hardness. Since many of j
experience for these plants can be at'.ribute( to these factors (stresses and microstructure) are plant-j several factors including the method usel to expand and/or tube-specific, it is difficult to predict the the tubes and the heat treatment that the tutes relative SCC susceptibility of any particular steam received. A study performed by Westinghouse led generator or any given tube in a steam generator.
i the manufacturer to conclude that the residual stress levels present in hydraulically expanded tubes at the The tubes of certain model Westinghouse steam expansion transition are lower than in either generators with small-radius U-bends (i.e., the tubes i
mechanically or explosively expanded tubing in rows 1 and 2) have exhibited axial and j
(Ref. 90). Since crack growth rate and time to crack circumferential cracking. The vast majority of the i
initiation depend, in part, on the stress level, lower these small-radius U-bend cracks have occurred in i
stresses may result in lower crack growth rates tubes with mill-annealed alloy 600 tubes with at least
{
and/or longer time before crack initiation. The one reported instance of possible U-bend cracking in thermal treatment process used for most of the steam alloy 600 thermally treated tubing (the reported l
generator tubes in the hydraulically expanded plants cracking in the alloy 600 thermally treated tube was is intended to provide =h=M corrosion resistance axial in orientation). These cracks generally initiate i
over the mill annealing process used in the steam from the primary side of the tube and occur at or 3
generators at other plants (Ref. 90). Operating near the apex of the U-bend and at the tangent j
experience indicates that thermally treated tubes are transition between the U-bend and the straight span of less susceptible to SCC than mill-annealed tubes since the tube (refer to Figure 7-4).
One licensee i
no significant amount of cracking has yet to be characterized the root causes of U-bend PWSCC as j
detected in thermally treated tubes.
being the result of (1) high residual stresses in a i
susceptible material, (2) excess ovality during the 7.2.2 Small-Radius U-Bends tube bending process (apex location), (3) irregular transitions formed during the tube bending process 3
Susceptibility to circumferential SCC (and SCC in (i.e., a bulge at the tangent point), and (4) top tube l
general) depends, in part, on the stresses exerted on support plate denting (Ref.105),
j the tube at the location of interest, as well as the j
microstructure of the alloy. For the case of SCC in Small-radius tube U-bend cracking has led some 4
the U-bend, the stresses at this location can be plants with mill-annealed alloy 600 tubes to i
attributed to several sources including operating preventively plug tubes in this region (primarily the j
tempe.ature and pressure, tolerances on tube hole row I tubes). Additionally, some plants have applied I
position at the top support plate, hot-to-cold-leg an in situ heat treatment to the U-bend region. The temperature variations, differential expansion of the intent of this in situ U-bend heat treatment (UBHT) l nbe and support plate, twist of the tube resulting i
I 7-11 NUREG-1604 4
i
4 is to reduce the susceptibility of the tube to axial and
,circumferential cracking in the U-bend region is circumferential SCC, by altering its grain structure minimal. De tubes in row I have been preventively and reducing the residual stresses introduced during plugged at H=Mam Neck (in two of the four steam the bending process (Ref. 30). His in situ UBHT generators), Indian Point 2, and Zion 1.
For the was performed in some units before commercial remammg steam generators with partial <lepth operation while in other units it was performed after expansion transitions, with the exception of the steam conanencing commercial operation. In addition, generators at Zion 2, circumferential indications at some plants have removed the plugs from this location have not been observed. At Zion 2, a small-radius tubes that were preventively plugged, in circumferential crack at the tangent point in a row I situ heat treated the U-bend region of these tubes, tube, which exhibited primary-to-secondary leakage, and returned these tubes to service (Ref. 90),
was detected in 1989. b crack was located on the cold-leg side approximately 9.4 cm (3.7 inches)
As discussed above, indications in the U-bend region above the top of the seventh (uppermost) tube support of tubes with small-bend radii have been observed in plate. b portion of the straight leg of the tube that mill-annealed and thermally treated tubes. b extends above the top tube support plate is 8.4 cm i
following paragraphs contain a summary of some of (3.3 inches); therefore, this crack was located the more pertinent operating experience related to approximately 1.0 cm (0.4 inches) ab.e the tangent cracking in the U-bend region of tubes with of the U-bend (generally on the intrados of the tube),
small-bend radii.
h operating experience is b ts.be was not removea ror destructive i
organized based on the type of expansion transition at examination (Ref.104), b tubes in i ww I at Zion the plant (e.g., full-depth hardroll plants) since plants 2 were preventively plugged during ths fall 1996 with similar expansion transitions tend to be of the outage after detecting 64 tubes with row I U-bend same vintage. As a result, many of the factors that indications (Ref.106).
influence a tube's susceptibility to cracking are considered to be similar at these plants.
As of June 1995, no plants with full-depth hardrol.1 expansion transitions reported circumferential As discussed above, the factors that influence a tube's indications in the U-bend region of small-radius tubes susceptibility to cracking include the tube material, (Ref. 90) even though these locations were monitored the tube microstructure, and the stresses exerted on at many of these plants (refer to Table 7-7). All of the tube. The tube material for all the plants these plants have performed in situ heat treatment in discussed in this section is alloy 600. De tube the U-bend region of the small radius tubes to reduce microstructure, which is partially determmed by the the susceptibility of the tubes to this degradation heat treatment that the material receives, differs from mechanism. A few of these plants performed this plant-to-plant; however, the steam generators in this heat treatment before commencing commercial section can be distinguished based on whether the operation, while most of the plants performed the tubes received a thermal treatment or were UBHT after commencing commercial operation. The mill-annealed. In general, most of the older plants licensee for South Texas Project 1, a full-depth (i.e., the partial-depth hardroll plants, full-depth hardroll plant, does not believe their small-radius hardroll plants, and WEXTEX plants) have tubes that U-bend tubes are susceptible to circumferential were mill-annealed and the newer plants have tubes cracking since the model E steam generators have that were thermally treated. De stresses exerted on larger bend radii than other Westinghouse steam the tubes arise from many sources including the generators and the tubes were in situ heat treated, design, fabrication, and operation of the steam Nonetheless, the licensee takes the precaution to generator. In general, improvements in each of these monitor these locations to provide additional areas were made as the steam generator design assurance of tube integrity. (South Texas Project 1 evolved (from mechanical and explosive expansion of is the only full-depth hardroll plant with model E the tubes into the tubesheet to hydraulic expansion of steam generators.) Subsequent to June 1995, at least the tubes).
one plant with steam generators with full-depth hardroll expansion transitions (i.e., Byron 1) has For plants with partial-depth expansion transitions, exhibited circumferential indications in the U-bend the number of small-radius tubes affected by region of tubes with small-bend radii (Ref.105).
NUREG-1604 7-12
For plants with WEXTEX transitions, all of which The remaining plants with hydraulic expansion have model 51 steam generators (with the exception transitions have thermally treated tubes.
As of Comanche Peak I which has model D4 steam previously noted for the expansion transition region, generators), in situ UBHT was performed in row I the steam generator operating experience with and/or row 2 (refer to Table 7-11). Most of these thermally treated tubes has thus far been much better plants performed this UBHT of the tubes after than that for steam generators with mill-annealed commencing plant operation.
The exception, tubes. As of June 1995, no indication of cracking Comanche Peak 1,
in situ heat treated the had been found in the U-bend region of any of the small-radius U-bend tubes before commencing small-radius tubes (i.e., rows 1 and 2) in these steam commercial operation. In some instances (e.g.,
generators. This relatively good operating experience Farley I and Sequoyah 2), tubes in row I were has been attributed, in part, to the thermal treatment preventively plugged and then subsequently recovered which is intended to improve the resistance to and placed back in service following the UBHT. In cracking compared to mill-annealed tubes (Refs. 90 other instances (e.g., Salem 1 and 2), the row I tubes and 108). Other features to which this relatively were preventively plugged and never retumed to good operating experience have been attributed service.
Several of these plants hr.ve exhibited include features similar to those at South Texas circumferential cracking in the U-bend region. Such Project 2 (i.e., larger bend radii, stress-relieved plants include Diablo Canyon 1, which performed U-bends). These features are present in some of the UBHT in 1988 and subsequently detected four plants with thermally treated tubes and hydraulic circumferential indications in 1W2 (Ref. 107),
expansion transitions.
Farley 1, and Sequoyah I and 2.
For Farley 1, most, if not all, of the circumferential indications For plants with model F steam generators (all of detected were present in the tubes when the tubes which have thermally treated tubes), the design is were recovered and heat treated rather than having intended to reduce the potential for tube cracking in been initiated after the UBHT.
Circumferential the U-bend location. Actions taken in model F steam indications have been detected in heat-treated and generators specifically to teduce the potential tor non-heat-treated U-bends in the small-radius tubes of crack initiation include using thermally treated alloy WEXTEX plants. In addition to circumferential 600 tubes, as well as relieving stress in the U-bend indications, axial indications have been detected in region of severtl rows (e.g., rows 1 through 10) of small-radius U-bend tubes (e.g., at Farley 1, axial tubes (Refs.108 and 109). Starting with the model indications were detected even after UBHT F steam generators, a set of U-bend manufacturing (Ref. 35)).
geometric controls were implemented.
These controls included requirements on tube ovality, For plants with steam generator tubes with hydraulic U-bend-to-leg flatness, and leg spacing.
These expansion transitions, only Callaway and South Texas improved manufacturing requirements are intended to Project 2 do not have steam generators with all ensure U-bend consistency, which in turn translates thermally treated tubes. Of these two plants, South to consistent stresses. In summary, the geometric Texas Project 2 is the only plant with steam generator controls help to eliminate localized stress j
tubes with hydraulic expansion transitions that does discontinuities (Ref.100).
not have thermally treated tubes in the small-radius tubes. (Callaway has thermally treated tubes in the One licensee has indicated that the residual stresses in first 10 rows of the tube bundle.) At South Texas the tube-to-tubesheet hydraulic expansion transition Project 2, however, the bend radii for the low row region have been estimated to be as high as 2758 bar tubes are greater than other Westinghouse steam (40,000 psi). The lack of degradation in this region generators with mill-annealed tubes, and these tubes (i.e., the expansion transition) suggests that other were in situ beat treated. These features (i.e., larger areas of the tube, operating at lower residual stresses bend radii and UBHT) are intended to reduce the and lower temperatures (such as the U-bend region of susceptibility of the U-bend region of the small-radius small-radius tubes), would be even less likely to tubes to cracking.
exhibit degradation (Ref.100).
Alth%gb she operating experience with small-radius (i.e., row I and 2) tubes in thermally treated alloy 7-13 NUREG-1604
i l
600 tubes has been relatively good to date, several frequently located in the tube free span (i.e., the indications have been detected in these tubes. At region of the tube between tube support structures).
Callaway,' one undefined indication was identified in These dents can be introduced during the manufacture a row 2 tube. This indication was not identifier l with and maintenance of the steam generators.
In j
the bobbin coil and was located just above the addition, mechanically induced dents on the steam seventh cold-leg support plate.
'Ihe licensee generator tubes can arise from loose parts introduced concluded that this indication was an anomaly since during maintanance activities. All steam generators no degradation mechanism had been identified in this are susceptible to this type of denting, and many region. In addition, a senior eddy current analyst _
plants, if not all, have observed mechanically induced (i.e., a level III analyst) judged this indication to be dents.
a distorted signal caused by its location in the U-bend transition (this tube was not removed for destructive Corrosion induced dents typically occur at locations examination) (Ref.110). Recently (i.e., after June where the tube is in close proximity to a tube support i
1995), the licensee for Braidwood 2, which has steam structure fabricated from carbon steel and at the top generators with alloy 600 thermally treated tubes, of the tubesheet where sludge tends to accumulate reported an axial indication in the U-bend region of near and around the tube. Corrosion-%ced denting one small-radius (i.e., Row 1) tube. The licensee at tube support structures (e.g., carbon steel tube concluded that the most likely cause for this support plates) is typically the result of corrosion of indication is PWSCC (this tube was not removed for the support structure. In the case of the carbon steel destructive examination) (Ref.105). 'Ihis licensee tube support plates, corrosion results in the buildup j
also reported that at least one other plant with alloy of corrosion products (typically magnetite) in the 600 thermally treated tubes and stress relieved crevice between the tube and the tube support plate.
U-bends has experienced axially oriented indications Since the corrosion products (e.g.,
magnetite) in row I and 2 (Ref.105).
typically occupy a larger volume than the base material (i.e., carbon steel), the buildup of magnetite In summary, axial and circumferential cracking of the (i.e.,- iron oxide) leads directly to the mechanical
' U-bend region of small-radius tubes has occurred in deformation of tubes where they meet the tube i
plants that have in stru stress relieved, as well as support plate; when it is extensive, denting can lead i
plants that have not in situ stress relieved the U-bend to the deformation and cracking of the tube support l
region of these tubes.
The majority of the plates themselves. Corrosion induced dents have small-radius U-bend indications have been in plants primarily been observed at the intersection of the :ube with alloy 600 mill-annealed tubes; however, and carbon steel tube supports.
indications have also been detected.in alloy 600 thermally treated tubes. The indications reported The denting process is a concern because it can to-date in alloy 600 thermally treated tubes were axial produce high levels of stress on both the tube and the in nature rather than circue6e.tial. "Ihe root cause support structure. In addition, it is of concern of the indications tsiag detected in the thermally because reliably analyzing and interpreting the treated tubes has not been confirmed through the inspection data ob:sined with a bobbin coil from a destructive examination of pulled tubes.
dented tube is morn difficult than for a non-dented 1
tube.
To reliably mspect dented locations for J
7.2.3 Dented Locations indications of degrrdation, many licensees currently (i.e.,1995) use ritating probes with pancake or Denting is the plastic deformation (constriction or plus-point coils as discussed below.
mechanical deformation) of the steam generator tubes, and it has resulted from both corrosion and The stresses imparted on the tube as a result of the mechanical processes Mechanically and corror, ion denting process can lead to axially and/or induced dents have been observed at a number of circumferentially oriented SCC. Furthermore, the locations in steam generators.
stresses imparted on the tube support structure as a result of the denting process can result in cracking of W hanically-induced dents generally occur this structure. Although both of these conditions
~
tnroughout the steau generator tube bundle and are have been observed (as discussed below), it is important to note that a tubes susceptibility to NUREG-1604 7 14 I
I.-
i 4
a l
j cracking at a dent depends on many factors including generators with tubes that were hydraulically the magnitude of the dent (i.e., the level of stress),
expanded into the tubesheet) wen, designed and l
the temperature at the dented location, - the operated to reduce the likelihood of corrosion-induced microstructure of the maamal, and the operstmg denting.
i environment.
l Denting has led to SCC of the tubes in Westinghouse Corrosion-induced denting was fi:st identified and steam generators. In the 1970s, the denting and SCC
+
became prevalent in the early to taid 1970s. In the were so severe at some plants that the steam mid to late 1970s, excessive decting of tubes at the generators were replaced (as discussed above).
i tube support structures was reporteA, resulting in (1)
Currently, however, the dents being observed in j
primary-to-secondary leaks (as a restit of SCC which steam generators are less severe than those previously initiated primarily "com the inside (primary side) observed (i.e., in the pre-replacement stea m l
surface of the dented tube), (2) cracking of the tube generators) as a result of steam generator design and j
support structures (specifically horizontally oriented operational improvements. 'Ihis is evidenced by the j
4 4
tube support plates), and (3) inability to pass relatively limited number of tubes that are found in j
i standard-size inspection probes through the tubes. In which the standard size eddy current inspection some instances. steam generators were replaced as a probes can not pass through the tubes. Even though i
result of cdasive denting. As discussed above, the denting is minor, axially and circumferentially 1
operating g&.ts with carbon steel tube support oriented SCC continues to occur.
The axially l
structures au maceptible to corrosion-induced tube oriented SCC generally initiates from either the inside sienting.
or outside diameter of the tube and the i
circumferentially oriented SCC generally initiates 1
l Corrosion-induced denting of the steam generator from the outside diameter of the tube, although, in a l
tubes can be controlled in operating steam generators few ' instances, circumferentially oriented SCC has
{
through proper water chemistry controls and a been reported to initiate from the inside diameter of i
secondary system design that minimizes the likelihood the tube.
of introducing impurities that contribute to denting.
The corrosion-induced denting currently being Since the majority of the circumferential indications obsemd in the field is relatively minor compared to being detected at dents in Westinghouse steam tim koting observed in the 1970s. (That is, the generators occur at corrosion induced dents at tube axte at of tube deformation is minor, and standard-size support plate intersections, the following paragraphs, I
proces can typically be passed through the tubes.)
which summarize the operating experience at dented j
1mprovements in water chemistry, which have limited locations, emphasize degradation observed at these the impact of denting, include more restrictive limits locations. Circumferential indications that propagate 1
j on impurity levels. Improved secondary system os a result of fatigue and are attributable to denting design measures include replacing copper-bearing are discussed in Section 7.2.3.1.
Circumferential j
components, replacing condenser tubes to reduce the indications that develop as a result of non-fatigue j
potential for leakage, and (in the newer steam related mechanisms (e.g., SCC) are discussed in generators) using tube support structures of different Section 7.2.3.2.
The non-fatigue related designs (e.g., broached hole, lattice grids) and circumferential indications occur primarily at constructed from stainless steel, a more corrosion induced dents. Corrosion-induced denting l
corrosion resistant material.
primarily occurs at carbon steel tube support l
structures rather than stainless steel tube support in summary, both mechanically and corrosion structures since stainless steel is more corrosion induced dents can lead to increased levels of stress on resistant in the steam generator operating i
the tuks. As a result, degradation, particularly environment.
As a result, Section 7.2.3.2 is SCC, can occur at either mechanically or corrosion organized based on the type of expansion transition ir.Juced dents.
Most of the denting-related since most of the earlier model Westinghouse steam degradation, however, has primarily occurred at generators (i.e., in steam generators with corrosion induced dents rather than at mechanically partial-depth hardroll, full-depth hardroll, or induced dents. Most of the newer steam generators WEXTEX expansion transitions) have carbon steel (e.g., replacement steam generators and steam tube support plates and most of the later model 7-15 NUREG-1604
_ _ _ ~..._ _ _ _. ~.
Westinghouse stea m generators - (e.g.,
steam exceed 1.0, the potentially affected tube was either generators with hydraulically expanded tubes) have plugged and stabilized or plugged using a stainless steel tube support plates. b Westinghouse leak-limiting sentinel plug (Ref. 90).
Steam replacement steam generators, the model F, and some generators with stainless steel support plates and model D and E steam generators are fabricated with broached (quatrefoil and trefoil) tube holes were no:
stainless steel tube supports.
expected to experience this phenomenon (Ref. 90).
Nonetheless, some Westinghouse plants with steam 7.2.3.1 Fatigue Cracks at North Anna 1 and Indian generator tube support plates constructed from Point 3 stainless steel (e.g., Catawba 2) have been analyzed for high cycle fatigue by evaluating the position of Only a limited number of fatigue cracks that are the AVBs (Ref. 52).
attributable, in part, to the denting phenomena have been observed in Westinghouse steam generators.
In addition to the North Anna 1 fatigue crack, a The limited number of fatigue cracks is attributable, fatigue crack was also observed at Indian Point 3 in in part, to the corrective actions taken by the licensee 1988. As discussed below, this crack occurred under as a result of these events. The followingparagraphs slightly different circumstances. In October 1988, a summarize the two circumstances in which fatigue primary-to-secondary leak occurred at the Indian cracks associated with the denting phenomena were Point 3, as detailed in NRC Information Notice observed.
88-99, " Detection and Monitoring of Sudden and/or Rapidly increasing Primary-to-Secondary Leakage" In 1988, the NRC staff published Bulletin 88-02, (Ref.19). The steam generators at Indian Point 3 at
" Rapidly Propagating Cracks in Steam Generator the time of this leak were Westinghouse Model 44 Tubes" (Ref. 23), which addressed the 1987 North steam generators. These steam generators had alloy Anna 1 tube rupture resulting from high cycle 600 mill-annealed tubes, which were partially rolled fatigue. In 1987, the steam generators at North Anna on each end at the bottom of a 56 cm (22-inch) thick were Westinghouse model 51 steam generators with tubesheet. The tubes passed through drilled-hole tube WEXTEX tube expansions, carbon steel tube support support plates that were 19.05 rem (0.75 inch) thick plates, and alloy 600 tubes that had been mill and constructed of carbon steel.
annealed. The 1987 rupture location was at the top support plate on the cold-leg side of the tube in row ARet replacing the model 44 steam generators at 9 column 51, a relatively small-radius tube. Bulletin Indian Point 3 in a subsequent outage, the leaking 88-02 cited three conditions that could lead to a tube, which had been plugged, was removed for rapidly propagating fatigue failure similar to the destructive examination. The leak was attributed to occurrence at North Anna. These three conditions a circumferential crack 1.52 mm (0.06 inches) above were (1) denting at the upper support plate, (2) the top edge of the sixth tube support plate, the fluid-elastic stability ratio approaching that for the uppermost plate, on the hot-leg side of the tube in tube that ruptured at North Anna, and (3) absence of Row 45 Column 50.
This tube'is a peripheral effective AVB support. The WOG indicated that (large-radius U-bend) steam generator tube with locally elevated steam velocities could be observed U-bend region supports at four locations (two hot-leg because of nonuniform AVB insertion depths and two cold-leg locations) by AVBs. The crack (Ref. 90). The bulletin was issued to holders of extended 250 degrees around the circumference and operating licenses or construction permits for was through-wall over its entire length. It was a Westinghouse-designed nuclear power reactors with single crack with no branching or major secondary steam generators having carbon steel support plates.
crackr, with the exception of two short parallel secondary cracks. Profilometry of the tube indicated As ~ a result of Bulletin 88-02, all domestic that the tube was dented at the sixth hot-leg support Westinghouse steam generators with carboo steel tube plate from 0.267 to 0.381 mm (0.0105 to 0.015 support plates were analyzed using a methodology inches) radially. Field eddy current data indicated approved by the NRC to identify the potential to that this tube was also dented at its other support experience high cycle fatigue at the uppermost tube plate locations in both the hot and cold legs. The j
support plate. The WOG indicated that, in cases destructive examination indicated that the j
where the analysis indicated that fatigue usage could circumferential crack had only transgranular, fatigue NUREG-1604 7 16 l
type features (other than minor, typically 0.051 mm denting. At the time of the original GL 95-03 (0.002 inches) deep, intergranular corrosion that responses (i,c.,
June 1995), this cracking sporadically occurred at the outside diameter edge).
phenomenon had not been observed at other b licensee concluded that the crack initiated on the Westinghouse units; however, as discussed below, outside diameter surface at a location where the circumferential cracks at dented tube support plate maximum outside diameter stresses from tube elevations have been observed at other plants since displacement perpendicular to the U-bend plane are June 1995. b circumferential indications detected believed to occur (Ref.111).
at North Anna 1 and 2 were in the pre-replacement steam generators.
b North Anna 1 and 2 b fatigue failure of the tube in row 45 column 59 pre-replacement steam generators were Westinghouse at Indian Point 3 indicates that large-radius steam model 51 steam generators which had alloy 600 generator tubes in Westinghouse steam generators mill-annealed tubes, carbon steel drilled tube support may be susceptible to fatigue failure at the uppermost plates, and WEXTEX transitions (Ref. 90).
tube support plate, even though they are supported by AVBs. b failure at North Anna 1 involved a Several tubes were pulled from the North Anna 1 relatively small-radius tube (i.e., row 9), which did steam generators to characterize the degradation at not have any AVB support. Both tubes (i.e., the the tube support plates from 1985 until the steam tubes at North Anna 1 and Indian Point 3) were generators were replaced in 1993. In 1985, two dented at the uppermost tube support plate near the dented tubes were pulled to examine distorted eddy location of the circumferential crack. Denting, per current indications at the edges of the tube support Bulletin 88-02, was considered to include evidence of plates. b tube pulls revealed axially oriented upper tube support plate corrosion and the presence PWSCC extending outside the confines of the tube of magnetite in the tube-to-tube support plate support plate on both tubes. One tube showed two crevices, regardless of whether there was detectable axially oriented primary water stress corrosion cracks distortion (i.e., mechanical deformation) of the tubes.
separated by about 180 degrees. These cracks were located at the minor diameter of the ovalized tube 7.2.3.2 Non-Fatigue Cracks (i.e., the tube was dented). The primary water stress corrosion cracks obx-ved on bd tubes showed As of December 1996, circumferential indications multiple microcracks compridng the macrocracks.
had been reported at dented locations a: a number of ODSCC within the tube support plate crevices was Westinghouse plants.
Most of the reported also found during the tube examinations.
No circumferentially oriented indications occurred either circumferential indications were reported for either of at dents at the top of the tubesheet for partial-depth these tubes. The maximum amount of deformation hardroll plants as discussed in Section 6.2.1.1 or at ranged from 4-percent ovality to 7-percent ovality dents at the tube support plate elevations; however, depending upon the technique used to measure the at least one circumferentially oriented indication has deformation (Ref.112).
I occurred at a free-span dent (i.e., a dent not in the sludge pile or at a tube support structure). In In 1987, two additional tube specimens were removed addition, both inside and outside diameter initiated from North Anna 1 for destructive examination of circumferential indications have been observed at circumferential indications in the WEXTEX dented tube support plate elevations. The inside transition. For one of these specimens, the licensee diameter initiated circumferential indications were obtained the portion of the tube located at a tube recently observed (i.e., 1996).
The following support plate elevation. The crevice region of this paragraphs summarize recent operating experience tube support plate intersection specimen had a 0.025 with respect to circumferential indications at dented to 0.051 mm (0.001-to 0.002-inch) radial dent. An outside diameter initiated, circumferentially oriented locations.
crack network was foundjust below the tube support Circumferential cracks were detected and reported at plate top edge, which consisted primarily of dented tube support plate locations in the field at circumferentially oriented microcracks.
He North Anna 1 in 1991 and then at North Anna 2 in microcracks formed a 90-degree long macrocrack.
1992. These circumferential indications were the The morphology of these outside diameter initiated first reported in recent times for tubes with minor cracks was IGSCC with some intergranular attack.
NUREG-1604 7-17
Axially oriented PWSCC was also found on this probe. In addition,5 tubes were found to have both specimen extending from the top edge of the tube circumferential and axial indications outside the same support plate to the mid-support plate region. Near edge of the tube suppost plate. On the basis of the mid-support plate region, the cracking was axially previous tube pull results and the eddy current data, oriented; near the support plate top edge, the the licensee concluded that these indications were cracking was tilted 45 degrees to the major tube axis azimuchally separated by approximate?y 90 degrees.
(Ref.112).
Dese indications were found at highly ovalized, dented locations. The licennee for North Anna 1 in 1991, a portion of the tube in row 11, colurun 14, postulated that the cimumferentially oriented ODSCC in steam genemtor B of North Anna I was pulled to indications are located at the major diameter of the assess circumferential indications found with 8xl and ovalized tubes and the PWSCC axial cracks at the rotating pancake coil probe inspections at the first minor diameter (Ref.112).
tube support pee elevation. Circumferential cracks (predominantly ODSCC with some intergranular in addition to identifying tube suppert plate elevations attack components) were found at locations with both axial and circumferential indications during corresponding to the top and bottom edges of the tube the 1991 mapection, the licensee for North Anna 1 support plate. At the top edge, a 360-degree crack also identified tube support plate elevations with was present; however, the cracking was most multiple circumferential indications.
Multiple significant (i.e., greater than 40-percent through the circumferential indications are indications composed tube wall) for two areas of 124 degrees and 113 of typically two circumferential cracks at nearly the degrees with smaller crack depths between the two same elevation on the tube but circumferentially deeper circumferential cracks. At the bottom edge, separated by a ligament of tube material that is not circumferential cracks of 135 degrees and 112 cmcked.
degrees were found with no significant crack depth (i.e., less than 40-percent through the tube wall)
As a result of the 1991 refueling outage inspection between these cracks. The cracks were centered near results, the NRC stipulated that the licensee for North the apex of the ovalized dented tube (0.711 mm Anna 1 must either implement a midcycle inspection (0.028-inch) ovality). The field eddy current data of the steam generator tubes or provide the additional evaluation for the crack at the bottom edge of the information necessary to justify a full cycle of tube support p', did not result in the identification operation.
De licensee provided information of an approx.tuly ll2-degree portion of the intended to justify a full cycle of operation and the indication, which had a maximum depth of 61 percent staff requested add tional information from the and an average depth of 51 percent through the tube licersee. As a resnt of the staff's and licensee's wall. b circumferentially oriented ODSCC on this evaluation of tube stegrity at North Anna 1, the tube was very similar to the cracking found in the licensee shut down North Anna 1 in late December tube pulled in 1987.
In summary, the 1991 1991 for a midcycle steam generator tube inspection, destructive examination indicated that the eddy h shutdown was a result of a review of the 1991 current indication was attributable to circumferentially refueiing outage eddy current data, which identified oriented ODSCC consisting of macrocracks formed additional potential defects. b eddy current data by lineup of individual microcracks followed by re-review was requested in early December 1991 by corrosion of the separating ligaments (Ref.112).
the NRC. This request was a result of the licensee's identification of additional circumferential cracks 1991 was the first time circumferential indications (during the destructive examination of a tube) that were reported in the field at dented tube support plate were not identified in the field by the eddy current elevations at North Anna 1. Earlier inspections with data analysts. b tube had been pulled during the the 8xl probes had discovered degradation at dented 1991 refueling outage. Since some of these potential tube support plate elevmions, but these were thought indications identified during the data re-review could to be volumetric in character. Approximately 110 exceed the tube repair criteria in the technical circumferential indications at dented tube support specifications, the licensee declared the steam plate elevations were found during the 1991 generators inoperable after detailed discussions with inspation. All of these were detected by the 8xl the NP.C (Ref.113).
probe and confirmed with a rotating pancake coil NUREG-1604 7-18 1
During the January / February 1992 midcycle stes to be a circumferential crack approximately 22.1 cm i
l generator tube mspection outage, the licensee's (8.7 inches) above the uppermost tube support plate mspection included a 100-percent rotatmg pancake on the cold-leg side of the steam generator near, or coil examination at the hot leg WEXTEX transitions slightly above, the U-bend tangent. The indication and 100-percent rotating pancake coil exammation at was associated with a dent (Ref.115). Although the tube support plate elevations.
A similar other circumferential indications have been detected examination at the WEXTEX transitions was in the U-bend region of steam generator tubes as performed during the 1991 refueling outage discussed in Section 7.2.2, the indication detected at
]
(i.e.,100-percent rotating pancake coil exanunation).
McGuire 1 is different than those discussed in Section The 100-percent rotating pancake coil examination of 7.2.2 since it was not in a tube with a small bend the tube support plate elevations was a result of a radii.
comparison test of the 8x1 arrayed pancake probe and
]
l the rotating pancake coil probe for indications at the As a result of industry experience regarding axially tube support plates. This test showed that it was and circumferentially oriented tube degradation at necessary to make conservative and possibly false dented locations, many plants have implemented a 8x1 calls to achieve more than 90-percent detection sample plan for inspecting such locations. His plan of the signals identified by rotating pancake coil generally includes inspecting dents using techniques inspection. His was attributed to the inability to capable of reliably detecting axially and discriminate effectively between lifitoff signals at the circumferentially oriented tube degradation when the i
dented tube support plate intersections and crack bobbin coil voltage associated with the dented j
indications at the tube support plate edges. His was location exceeds a specified voltage. De bobbin coil the first time a 100-percent rotating pancake coil test probe can identify dented regions; however, it cannot of the tube support plate elevations was performed at reliably detect all forms of degradation that may j
North Anna 1. The 1991 refueling outage inspection potentially occur at dented regions.
In many primarily consisted of 8x1 probe examinations matances, this inspection plan includes a 100-percent l
followed by further characterization with the rotating exammation of all dents with bobbin voltages of 5 pancake coil probe. Some random rotating pancake volu or greater. Since the bobbin coil voltage coil exams were performed during the 1991 refueling corresponding to a dent depends, in part, on the outage inspection (Ref.114).
calibration procedure of the bobbin coil, it is important to know the calibration procedure to ensure The steam gererators at North Anna 1 were replaced that the sample plan and results from one plant can in the subsequent outage, which took place in early be compared to another plant. He bobbin voltage 1993. Similar circumferential indications at dented associated with a dented region also depends, in part, tube support plate elevations were detected at North on the magnitude and geometry of the dented region Anna 2. He North Anna 2 steam generators were of the tube. He resultant bobbin voltage, therefore, replaced in 1995.
may or may not be representative of the local stresses in the tube at the dented location. As a result, the The operating experience of the pre-replacement susceptibility of a given dented location to cracking steam generators at North Anna 1 and 2 indicate that can not be determined based solely on the bobbin circumferential cracks can occur at dented tube voltage; however, it can be used as a guide for support locations; however, circumferential performing examinations. Since the voltage threshold indications can also occur at dents not associated with for inspecting dents is to some extent arbitrarily i
a tube support structure.
At McGuire 1, a determined, a comprehensive sample plan for dents j
circumferential indication was detected near a dent in should include sample expansion criteria both for the U-bend region of the tube located in row 41 dented locations above the voltage threshold (when a column 43 in steam generator B in the 100-percent examination is not performed) and for August / September 1993 time-frame. This tube was dented locations below the voltage threshold.
identified during an outage which co=W as a result of a primary-to-secondary leak. Although the Many plants implement the calibration procedure and majority of the leakage was coming from another mspection strategy specified in GL 95-05 (Ref.1) for tube, observable leakage was detected from this tube.
examining dented locations (specifically dented tube ne source of the leak for this tube was determined support plate elevations). For plants with many NUREG-1604 7-19
1 thousand dented locations, a 100-percent sample of Anna 1 and 2 steam generators, and the dents greater than 5 volta may not need be performed circumferentially oriented tube degradation observed in order to determine whether degradation associated in the dented U-bend region at McGuire 1.
j with dents is occurring. In these matances, even j
though a smaller initial sample in terms of the Since June 1995, axial and circumferential indications percentage of dented locations exammed is at dented tube support plate elevations were observed performed, the large number of locations to be at a few other plants including Diablo Canyon I and inspected can be sufficient to detect the onset of Sequoyah I which are both WEXTEX plants. b I
axially and circumferentially oriented degradation at results from the Diablo Canyon I steam generator dented locations. In addition, expansion of the tube mspection in the fall of 1995 revealed sample to include dented locations above and below approximately 77 axially oriented primary water 5 volts provides added assurance that tubes containing stress corrosion cracks at dented tube support plate axially or circumferentially oriented tube degradation elevations. Of these axial indications, 61 were
)
at dented locations are identified (Ref.116).
located in dents less than 5 volts using the calibration procedure specified in GL 95-05 (Ref.1).
In As of June 1995, plants with steam generators addition,17 of these axial indications were not j
currently in service that have partial-depth expansion identified with the bobbin coil.
Three transitions (as opposed to plants that had replaced circumferentially oriented primary water stress similarly designed steam generators) had not reported corrosion cracks were identified at dented tube-any circumferential indications at dented locations support plate elevations during this outage, other than at dents near the top of the tubesheet.
These indications were attributed to cellular corrosion As a result of these findings, the licensee for Diablo as evidenced by tube pull data from D.C. Cook 1 as Canyon I removed portions of four tubes for discussed in Section 6.2.1.1. Many of these plants destructive examination. Three of the four tubes have monitored these dented locations; these plants contained axial indications, and one of the pulled include two with severe denting by today's standards tubes had a circumferential indication.
b (i.e., some dents do not permit the passage of the circumferential indication that was removed for
{
nominal sized eddy current inspection probes for the destructive exanunation was at a tube support plate tube size). Similar to the partial-depth expanded elevation that had a 28 voit dent. b destructive plants, the plants currently operating with full-depth examination revealed two circumferential bands of or WEXTEX expansion transitions had not reported inside diameter initiated degradation; one band any occurrences of circumferential cracking at dents contained a single macrocrack, v$ile the second band at tube support plate locations as of June 1995 was a narrow band of microcracks (Ref. 98).
although most plants with dents had monitored these locations.
At Sequoyah 1, in the fall of 1995, axially and i
circumferentially oriented tube degradation was Although no circumferential cracking at dented tube observed at dented tube support plate elevations. In support plate locations had been identified as of June addition, portions of three tubes were removed for 1995 at plants currently operating with steam destructive examination. These included four tube generator tubes with partial-depth hardroll, full-depth support plate intersections from two of the three hardroll, and WEXTEX transitions, these plants are pulled tubes with eddy current indications attributed susceptible to circumferential cracking at dented to axially oriented ODSCC and circumferential j
locations regardless of the location of the dent (i.e.,
cracking, h dented tube support plate intersection free-span dent, dent in sludge pile, dent at tube with a circumferential indication was determined support structure). In general, operating experience througb destructive examination to be attributable to has shown that dented locations are susceptible to circumferentially oriented ODSCC (Ref.117).
circumferentially oriented tube degradation. For example, the circumferentially oriented tube in summary, alloy 600 mill-annesled tubes have been degradation found at D.C. Cook I (which was shown through operating experience to be susceptible discussed in Section 6.2.1.1), the circumferentially to circumferentially oriented tube degradation at oriented tube degradation observed at dented tube dented locations. The circumferential degradation support plate locations in the pre-replacement North can occur at dents at various tube support structures NUREG-1604 7-20 l
.___._m._
e i
i
\\
i
[
(e.g., tube support plates), in the sludge pile region 7.2.4 Sleeve Joints
]
near the top of the tubesheet, and in the tube free l
span. Furthermore, the cracking may be inside or As discussed in Appendix A, various types of sleeve outside diameter initiated although to date the designs exist and some have exhibited majority of the circumferential indications have been circumferentially oriented degradation.
In outside diameter initiated.
Westinghouse steam generators with partial-depth i
hardroll expansion transitions, several sleeve designs Several Westinghouse plants have steam generator are used including the B&W brazed, B&W kinetic, 4
tube support plates that are constructed from stainless Westinghouse hybrid expansion joint (HEJ), and CE steel. All plants with steam generator tubes that have TIG-welded sleeves. The B&W brazed sleeves and l
hydraulic expansion transitions have stainless steel the B&W kinetic sleeves were only used in the i
tube support plates. These support plates have either pre-replacement Ginna steam generators.
drilled holes (e.g., South Texas Project 2) or Consequently, of the currently operating quatrefoil broached holes (e.g., model D5 and F Westinghouse plants with partial-depth expansion steam generators). In addition, most of these plants transitions, the only sleeve designs being used as of 4
have thermally treated alloy 600 tubes (Callaway and December 1996 are the Westinghouse HEJ sleeves l
South Texas Project 2 are the exceptions).
and the CE TIG-welded sleeves.
j Corrosion-induced denting is not as likely at these plants (i.e., stainless steel tube support plate plants)
In Westinghouse steam generators with full. depth as it is for plants with carbon steel tube support hardroll expansion transitions, only Farley 2 had l
plates, since the stainless steel tube supports are not sleeves installed and inservice as of June 1995. The i
expected to corrode and form nonprotective magnetite sleeves in service at Farley 2 are Westinghouse as does carbon steel. Furthermore, for plants with laser-welded sleeves. Sleeves had been in service at steam generator tube support plates that have other full-depth hardroll units (e.g., Catawba 1, quatrefoil hole designs, the area of tube to tube McGuire 1 and 2) but were removed from service by support contact is reduced (compared to drilled plugging the tube in subsequent outages as discussed holes), tliereby allowing greater flow past the tube at in Appendix A (Refs. 118 and 119).
The the suppert gap and minimizing the buildup of preventively plugged sleeves at these other units were corrosive deposits (Refs.108 and 109). If denting B&W kinetically welded sleeves. Subsequent to June were to occur in these plants (i.e., plants with steam 1995, additional full-depth hardroll units had installed generator tubes that were hydraulically expanded into sleeves. These units include Braidwood and Byron 1.
the tubesheet and have stainless steel tube supports),
ne sleeves installed at Byron I were Westinghouse the thermally treated tubes usal in the vast majority laser-welded sleeves and CE TIG-welded sleeves of these plants are generally considered to be less while the sleeves being installed at Braidwood 1 L susceptible to tube cracking than mill-annealed tubes.
the fall of 1996 were Westinghouse laser-welded (Mill-annealed tubing is only used at Callaway and sleeves.
South Texas Project 2.)
In Westinghouse steam generators with WEXTEX Even though several design features exist to limit the transitions, only Farley I had sleeves installed as of potential for denting in the hydraulic plants, dents June 1995. He sleeves in service at Farley 1 are still occur in these steam generators. %is denting Westinghouse laser-welded sleeves. Subsequent to can produce residual stresses in the affected locations June 1995, additional units may have installed which could make them more susceptible to cracking sleeves.
than non-dented areas (Ref.103). De dents in these steam generators are generally attributable to steam No sleeves have been installed in the Westinghouse generator manufacture, maintenance, or the presence steam generators with hydraulic expansion transitions of loose parts in the steam generator. nese areas as of June 1995. Subsequent to June 1995, Callaway are sometimes referred to as either dents, dings, or installed Westinghouse laser-welded sleeves in some expansion anomalies (Ref.109). For the purposes of of the steam generator tubes. He Callaway steam this report, these terms are considered synonymous.
generators have some thermally treated tules and some mill-annealed tubes.
NUREG-1604 7-21
7.2.5 Preheater Expansion Transitions steam generator that were hydraulically expanded at two baffle plates to mitigate flow induced vibration.
As discussed in Sectson 7.1.2, W=' f _ model ne licensee for these plants has indicated that D4, DS, and E steam generators have tubes that were 20-percent of the expanded preheater tubes in one
)
expanded into the tube support plates in the preassar steam generator will be inspected with a rotating j
regma as a result of concerns about tube wear. De pancake coil during the next outage (i.e., the next j
steam generston that have expanded tubes in the outage per GL 95-03) (Ref.120). Similarly, the
{
p.J = region include steam generators with either licensee for Shenron Harris 1 indicated that it would alloy 600 mill-annealed or alloy 600 thermally treated inspect 20-percent of the expanded preheater tubes (at
{
tubes. Furthermore, the expansion transitions at both expansion zones) in one steam generator.
j these plants are a result of either a full-depth During this inspection of the mill-annealed alloy 600 hardrolling or hydraulic expansion process. (Note preheater tubes at Shearon Harris 1,
no i
that Comanche Peak I has some WEXTEX expansion circumferential cracks were detected (Ref.121).
i transitions.)
Byron and Braidwood 2 have _ model D5 steam The preheater is on the cold-leg side of the steam generators that contain approximately 132 cold-leg J
generator. As such, the susceptibility to cracking preheater tubes per steam generator that were would be expected to be lower as a result of the hydraulically expanded at two baffle plates to mitigate lower operating temperature, all other pc.ic.1.4 flow induced vibration. The tubes in the steam (e.g., stresses) being equal. One licensee indicated generators at these plants are thermally treated. The that corrosion tests of hydraulic tubesheet expansions licensee for these plants has indicated that 20-percent l
in highly caustic environments have shown that of the prebester tube expansions in one steam corrosion has been arrested at a temperature of 292 generator will be inspected with a technique qualified degrees Celsius (557 degrees Fahrenheit), while to Appendix H of the Electric Power Research j
increasing corrosion rates were found at temperatures Institute report NP-6201, "FWR Steam Generator of 302 to 313 degrees Celsius (575 to 595 degrees Examination Guidelines" (Ref.
122).
If Fahrenheit). That licensee also indicated that their circumferential indications are detected during this preheater expansions are hydraulic expansions and are 20-percent sample, the licensee indicated that
{
located in areas with a primary side temperature of 100-percant of preheater tube expansions will be
{
approximately 288 degrees Celsius (550 degrees mapected in all four steam generators. The licensee Fahrenheit), which is below the temperature at which indicated that it considers the potential for corrosion was arrested in the corrosion tests circumferential cracking at this location to be low for (Ref.120). Consequently, the susceptibility of these the reasons cited above (i.e.,
relatively low
}
locations to cracking, including circumferential temperature at this location, lower stresses associated cracking, is expected to be low as a result of the with hydraulic expansion) and the ime of thermally lower temperature.
In addition, since the treated tubes rather than mill-annealed tubes
{
3 susceptibility to circumferential cracking is a function (Ref.120).
of temperature as well as stress and operating time, the susceptibility to circumferential cracking of Catawba 2 has model D5 steam generators with 141 hydraulic expansions would be expected to be lower tubes expanded at two locations into the tube support
'since the stresses induced as a result of the hydraulic plate to prevent vibration in the preheater section of expansion process are significantly lower than the each steam generator. The licensee indicated that stresses associated with a hardroll process.
other locations are more susceptible to cracking than the expanded preheater locations as a result of Although the susceptibility to circumferential cracking temperature and residual stress considerations. As a at the preheater expansions is considered low (for the result, since no cracking had been observed at any reasons discussed above), licensees inspect this region other location in the Unit 2 steam generators, this to ensure that comprehensive exammations are being area was not believed to be susceptible to cracking performed. Byron 1 and Braidwood I both have (Ref.123).
model D4 steam generators with milbannealed alloy 600 tubes.
Dese steam generators have Comanche Peak I and 2 have model D4 and DS j
approximately 132 cold-leg preheater tubes in each steam generators, respectively. In the preheater l
NUREG-1604 7-22
region of these steam generators, several tubes have cracking at the expansion transition even though they been expanded at the "B" and "D" baffle plates to are some of the oldest steam generators in the reduce the potential for vibration induced wear. The country and have mill-annealed alloy 600 tubes.
licensee indicated that the expanded tubes in the Circumferential indications at the expansion transition preheaur region may be susceptible to have been detected at only two of these plants (i.e.,
circumbrential cracking. As a result, the licensee Ginna and Indian Point 2), and the indications at one inspected 20-percent of the expanded intersections at of these plants (Indian Point 2) are believed to be the
'B' and "D" baffle plates in Unit I with a from closely spaced axial cracks rather than from rotating pancake coil technique. For Unit 2, the circumferential cracks.
licensee plans to inspect a sample of the expanded intersections at the "B" and *D" baffle plates during Circumferential indications have also been detected at their December 19% outage (Ref.124).
other locations in partial-depth hardroll plants.
Circumferential indications have been observed in the Preheater expansions have been performed at South U-bend region of small-radius tubes, at dented Texas Projects 1 and 2 in nominally 160 tubes per locations at the top of the tubesheet, and in sleeve steam generator to reduce the potential for vibration joints. The circumferential indications observed at induce wear. The licensee for South Texas Projects the top of the tubesheet have been attributed to I and 2 concluded that the original qualification cellular corrosion rather than distinct circumferential program for the expansion process indicated that no cracks. The indications detected in the sleeve joints definitive increase in outside or inside diameter (in Westinghouse HEJ sleeves and CE TIG-welded stresses result from the maximum expected field sleeves (fabrication related)) are discussel in expansions. The licensee for South Texas Projects 1 Appendix A. Althoughno circumferentialindications and 2 has examined process records from the field have been found at dented tube support plate regions implementation of the process at South Texas Projects in currently operating partial-depth hardroll plants, I and 2 and noted no examples of tubes that were this area is considered susceptible to such degradation overexpanded. Also, the licensee indicated that other since it has been observed in similarly designed steam operating Model E steam generators, which have generators (i.e., steam generators with mill-annealed operated several more cycles than South Texas tubes and carbon steel tube support plates).
Projects 1 and 2, have been examined at these expansions with no circumferential cracking being 7.2.6.2 Full-Depth Hardroll Steam Generators noted during the inspections. As a result, the licensee for South Texas Projects 1 and 2 censiders in general, plants with steam generators that have the tube areas expanded in the preheater not to be tubes which are roll expanded the full length of the susceptible to circumferential cracking. Nonetheless, tubesheet have exhibited circumferential cracking.
the licensee has monitored a small sample of these All of the tubes in these steam generators are preheater expansions for cracking of all types with a mill-annealed alloy 600.
The circumferential rotating pancake coil to provide additional assurance cracking is predominantly associated with the of tube integrity (Ref.125).
expansion transition. In some instances the number of circumferential indications detected at the 7.2.6 Summary expansion transition has been extensive (e.g.,
thousands of indications). A limited number of Sections 7.2.6.1, 7.2.6.2, 7.2.6.3, and 7.2.6.4, circumferential indications have been detected in the contain a brief summary of the circumferential U-bend region of small-radius tubes in these steam cracking operating experience of partial-depth generators and at dented locations. The sleeves hardroll, full-depth hardroll, WEXTEX, and currently in service at these units are welded sleeves, hydraulic plants, respectively.
and no service-induced circumferential cracking has been detected in these sleeves. Several of these units 7.2.6.1 Partial-Depth Hardroll Steam Generators have preheaters which have expanded tubes. These locations have not exhibited any circumferential Plants with steam generators that have tubes which cracking.
are only partially expanded within the tubesheet region have had a low incidence of circumferential l
7-23 NUREG-1604
7.2.6.3 WEXTEX Steam Generators and operation with respect to the other plants which tend to be older. Like the full-depth hardroll plants, In general, plants with steam generators that have several of the hydraulic plants have preheaters which tubes which have been expanded the fulllength of the have expanded tubes. These locations have not tubesheet with the WEXTEX process have exhibited exhibited any circumferential cracking.
circumferential cracking. All of these plants have mill-annealed tubes. He circumferential cracking 7.3 Justification for Continued Operation has occurred at the expansion transition, in the U-bends of small-radius tubes, and at dented The majority of circumferential indications in the locations. The sleeves currently in service at these Westinghouse units with alloy 600 steam generator units are welded sleeves, and no service-induced tubes occur at the expansion transition. The severity circumferential cracking has been detected in these of these indications has led to midcycle inspections at sleeves.
a few plants (e.g., Braidwood 1). A limited number and typically less severe circumferential indications 7.2.6.4 Hydraulic Steam Generators have been detected at dented locations, in the U-bend region of small-radius tubes, and in sleeve joints.
Plants with steam generators that have tubes which have been expanded the full length of the tubesheet ne staff evaluated each of the GL 95-03 responses with a hydraulic process can be broken into two submitted by Westinghouse plant owners with alloy categories:
mill-annealed and thermally treated.
600 steam generator tubes to confirm that the plants Only two hydraulie plants have mill-annealed tubes in could safely operate until the next scheduled steam the steam generators (i.e., Callaway and South Texas generator tube inspection outage.
The staff Project 2). Circumferential indications have been concluded that all of these units could operate until observed at the expansion transition in the their next scheduled steam generator tube inspection.
mill-annealed tubes at Callaway although tube pulls The staff based this conclusion on the following have not been performed to confirm the nature of factors:
these indications.
No other circumferential
)
indications have been detected at any other location (1) scope and results of the prior inspection i
in hydraulic plants with mill-annealed steam generator including the experience at other similarly l
tubes.
designed units - In steam generators that do not exhibit signals that interfere with eddy l
For the hydraulic plants with thermally treated tubes, current inspections (e.g., copper, denting),
circumferential indications have been detected at the inspections may have been more effective in expansion transition at Callaway. (Callaway has identifying tube degradation.
l thermally treated tubes in rows I through 10.) No other circumferential indications attributed to (2) preventive measures taken (e.g., peening of the service-induced degradation have been observed at expansion transition, U-bend heat treatment, any other location in these steam generator tubes.
etc.)- nese measures are expected to reduce No tubes have been removed for destructive the rate of cracking from PWSCC (Ref. 94).
examination from the Callaway steam generators to characterize the nature of the indications at the (3) removal of tubes for destructive examination-expansion transition in the thermally treated tubes. In Data from steam generator tubes removed for addition to the circumferential indications detected at destructive examination provide useful the expansion transition in the thermally treated information concerning the causal effects and tubes, it is important to note that axial indications morphology of the degradation.
In many have been observed in thermally treated tubes in the instances, tube pull data provide information U-bend region of small-radius tubes and at the that tubes with circumferential indications are expansion transition, able to withstand the pressure loadings specified in Reference 33. Because of the inability to The relatively good operating experience for the reliably size and characterize the degradation, hydraulic plants (including those with mill-annealed tube pulls are sometimes necessary to confirm tubes) can be attributed to improvements in design tube integrity and degradation morphology.
NUREG-1604 7 24
(4) in sfru pressure test data indicating that the identification and isolation of a steam generator circumferential indications are capable of that has a tube that has degraded to the point of withstanding the pressure loadings specified in leakage.
Improved integrated leak rate Reference 33 - When all circumferential monitoring programs provide added confidence indications are removed from service before an that tube integrity will be maintamed. These operating interval (i.e., the time between tube programs feature administrative limits on inspections) and the most severe indications primary-to-secondary leakage. ney also use identified at the end of this operating interval equipment and procedural upgrades to enable have adequate integrity based on in situ testing, plant operators to detect and respond to changes it would be reasonable to conclude that the in steam generator primary-to-secondary steam generator could be safely operated for a leakage, and to shut down the unit before a similar operating interval.
' significant leak or steam generator tube rupture if tube degradation should exceed expected (5) operating conditions at the plant (e.g., bot-leg values.
water chemistry operating temperature,_
practices) - Stringent water chemistry control (8) use of procedures. including emergency
(
in accordance with industry standards designed operating procedures) to diagnose and address to prevent uncontrolled tube degradation (i.e.,
steam generator tube leaks and ruptures -
' mitigate the initiation and propagation of SCC)
Procedures, equipment, and training programs and transport of impurities into the steam that are in place to identify and mitigate the generator provides confidence that tube consequences of failed tubes provide confidence integrity will be maintained consistent with that, in the event of a loss of tube integrity, the previous observations. Sludge lancing and plant can be safely operated. Dese programs chemical cleaning - to reduce sludge include simulator training on steam generator accumulation provides confidence that tube leaks and ruptures for control room impurities that may affect the rate of crack operators.
initiation and growth are controlled.
In addition, sludge lancing and chemical cleaning (9) design enhancements present in the steam can be effective at removing corrosion products generators which are intended to reduce the i
These at the top of the tubesheet and at controlling the susceptibility to cracking build up of conductive deposits which might enhancements include using thermally treated interfere with the eddy current data analysis tubes rather than mill-annealed tubes, using (Ref. 52).
stainless steel tube support plates rather than carbon steel tube support plates, using larger Since SCC is a thermally activated process, bend radii in the small-radius tubes, using reducing the hot-leg operating temperature hydraulic expansion techniques rather than takes advantage of the temperature niva4ce others, and using tube support plates with of SCC growth and initiation rates (i.e., the broached quatrefoil holes rather than drilled growth and initiation would be slower, with all holes.
other parameters being equal).
(10) the location of the expansion transition - If (6) operating time until the next steam generator severe circumferentially oriented degradation tube inspections were to be conducted were postulated to occur in the expansion transition region of a partial-depth expanded (7) requirement to monitor primary-to-secondary tube, the consequences would be different from leakage and to shut down the plant when ler.k a plant where the expansion transition is near rate limits are exceeded - leakage monitoring the top of the tubesheet. In steam generators is a defense-in-depth operating practice that can with partial-depth expansion transitions, the provide operators with a timely indication of a postulated circumferential crack would be well steam generator tube leak or tube rupture, below the top of the tubesheet. The presence Nitrogen 16 monitors can permit faster of the tubesheet essentially precludes the possibility of tube burst. In addition, if severe 7-25 NUREG-1604
degradation were to occur, the tube would not already been completed as a result of the time taken be able to move significantly either in the axial to prepare this document for publishing.
or radial direction thereby restricting the flow of primary coolant when compared to a tube h inspection summary tables are divided into four with similar degradation at the top of the plant groupings on the basis of the type of method tubesheet. 'Ihis is a result of the proximity of used to expand the tube into the tubesheet. These the tube outside diameter and tubesheet hole four groupings are padial-depth hardroll plants, inside diameter.
In addition, sludge full-depth hardroll plants, WEXTEX plants, and accumulation in the tube-to-tubesheet crevice, hydraulic plants. For each of these groupings of if preneut, could also act to restrict any plants, tables are provided for each location which potential leakage, has exhibited circumferential cracking (i.e.,
expension transition and top of tubesheet region h tubesheet also provides lateral support (partial-depth hardroll plants), U-bend region of which restrains bending of the tube during small-radius tubes, dented locations, and sleeve pressurization.
s joints). It should be noted that grouping of plants by other parameters may provide a better comparison of (11) the risk and potential consequences of a range operstmg experience and inspection scope (e.g., for of steam generator _ tube rupture events as dented locations, it may be better to divide the plants discussed in Reference 27 - In Reference 27, into two groupings such as mill-annealed and the staff estimated the risk contribution due to thermally treated and/or mill-annealed with carbon the potential for single and multiple steam steel tube support structures, mill-annealed with generator tube ruptures. In addition, this study stainless steel support structures, and thermally examined the expected consequences of steam treated with stainless steel support structures). h generator tube rupture scenarios, including staff elected to group the plants by expansion beyond design basis situations, such as the transition type since plants with similar expension potential for release due to containment bypass transitions typically are of the same vintage.
via failed tubes concurrent with a breach of Appendix B can be used to determine other pertinent secondary system integrity.
facts regarding the steam perevi tubes at these plants.
'7.4 Tube Inspections Tables 7-1 through 7-5 summarize the scope of the GL 95-03 requested,-in part, a safety assessment past and future inspections at the expansion transition, justifying continued operation on the basis of post at the top of the tubesheet, in the U-bend of inspection results and a summary ofinspection plans small-radius tubes, dented locations, and sleeve for the next scheduled steam generator tube joints, respectively, for partial-depth hardroll plants, mspection outage as they pertain to the detection of Tables 7-6 through 7-9 provide the scope of the past circumferential cracking. h inspection plans were and future inspections at the expension transition, in to consist of both an initial scope and sample the U-bend of small-radius tubes, dented locatious, expansion criteria. For the Westinghouse units with and sleevejoints, respectively, for full depth hardroll alloy 600 steam generator tubes, the staff summarized plants, Tables 7-10 through 7-13 provide this some of the information provided by the licensees information for WEXTEX plants, and Tables 7-14 with respect to the previous and next inspection for through 7-17 provide this information for hydraulic each of the areas identified as being potentially plants. Tube inspections performed using a technique susceptible to circumferential cracking.
h not capable of reliably detecting circumferentially designation of " previous
- refers to inspections oriented degradation were recorded as 'None" in performed before issuing or responding to GL 95-03.
Tables 7-1 through 7-17.
In instances where the The designation of "next" (and/or " future") refers to results of the tube inspections were readily available, an inspection performed after issuing or responding the results were included in the tables, as appropriate.
to GL 95 03. b phrase. *if detect", is used to Acronyms and abbreviations used in the tables are describe the inspection expansion criteria when a explained in Appendix C.
circumferential indication is detected.
In many instances, the next (and/or future) mspections have NUREG-1604 7-26 l
e As can be seen from evaluating the data in Tables 7-1 through 7-17, there are plant-specific differences in the inspection plans (e.g., probe type, scope of examination). %ese differences in the inspection plans were considered along with other plant-specific circumstances (e.g., preventive measures taken) in evaluating the acceptability of a licensee's response as discud in Section 7.3. For example, even though a licensee may have implemented a smaller initial inspection scope than another licensee, this may have j
been considered acceptable if the cumulative l
operating time for the plant was less than that of the other plant (all other parameters being equal).
The staff has reviewed the submissions provided by 4
the licensees that have Westinghouse steam generators with alloy 600 tubes and has concluded that they contain the information requested in GL 95-03, i
General conclusions regarding the responses are l
discussed in Section 9.
l l
i 7 27 NUREG-1604 L r
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- Steam outlet to turbine generator L
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Swirlvane moisture separator Anti-vibration bars N Tube bundle Tube supports l
Lower shell Divider plat F
erl e no ad Support foot Primary coolant outlet (Cold Leg)
Primarycoolant inlet (Hot Leg)
Figure 7-1 Cutaway view of a typical RSG NUREG 1604 7-28 i
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Preheater outlet Primary coolant inlet (hot leg) i Figure 7 3 Typical Westinghouse steam generator with a preheater (Model D 2)
NUREG-1604 7 30
i
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i Hot-Leg Cold-Leg Figure 7-4 U-bend portion of a small-radius tube 7-31 NUREG.1604
Table 7-1 Inspections at the expansion transition region in Westingbouse partial-depth hardroll plants l
Plant k.xpaaslam Transition E=PA Transition Future Notes Past laspection Inspection D.C. Cook I 39 RPC exams to apply 100% C5.
Rerolled tubes in 1994. References 88,126, and F* criteria
- 127, Otana 100% RPC of unsleeved None (SG replacernent).
Sleeves installed within tubesheet region, j
tubes References 128,129, and 130.
j Haddarn Neck 754 P probe exams arri 100% combo probe in HL. 20%
The P probe contains 2 bobbin cotts separated 254 Combo probe exarns combo probe in CL If detect in CL, by a fixed distance which accounts for ficid inspect an additional 20% in that SG.
spread and is used to determitie if I' of sound roll If detect in CL expansion, expand to is present. 'the co nbo probe contains a bobbin, 100% in affected 50.
pancake, and PP coil and performs the same function as the P probe and is used to characterize the indication (e.g., cracklike).
References 109,131, and 132.
Ind an Point 2 100% C5 100% C5 in HL and 20% C5 in CL. If 114 I 48 MNed circumferential indications (66 Scis, detect.100% in CL.
). These indications are believed to be 1
closely spaced axial cracks. Rerolled tubes to i
apply P criteria. References 133 and 134.
Kewaunee 100% RPC of unsleeved 100% RPC of unsleeved HL.
Inspect from tube end to 3 inches above TTS.
HL tubes References 135 and 136.
Point Beach 2 None None.
Historically performed RPC inspection of distorted bobbin indications. For next inspection, will use PP to inspect distorted bobbin ind cations. References 137 and 138.
Prairie Island i 100% RPC in HL 100% PP or equivalent in HL.
Did not shotpeen or rotopeen since unit was already in service. References 139 and 140.
Prairie Island 2 100% PP in HL 100% PP or equivalent in HL.
Did not shotpeen or rotopeen since unit was already in service. References 139 and 140.
i Zion 1 300 to 700 RPC exams 20% in HL with Appendix H cire.
Next impections primarily with RPC unless '
{
cracking technique. If detect,100%.
dented. References 141,142, and 143.
If widespread cracking, sample CL.
Zion 2 300 to 700 RPC exams 20% in HL with Appendix H cire.
Next inspections primarily with RPC unless cracking technique. If detect,100%.
dented. References 141,142, and 143.
If widespread cracking, sarnple CL.
1 NUREG-1604 7-32 i
l
~
~. -. _.
Table 7-2 Inspections at top of-tubesheet ' dented locations in We=*!=i x partial-depth hardroll plants Plant Othee Deuted lacations Othee Dented Imcations (e.g.1TS)
Notes (e.g.,1TS) Past Fatare laspection Inspection D.C. Cook i 100% RPC atITS 100% C3 at Tfs.
OD cellular corrosion at1TS prunarily associated with dents. References 88,126, and 127.
Gmna 1TS inspections not None (50 replacmeno.
References 128,129, and 130.
addressed Haddam Neck 81 RPC exams focused in 20 HL exams with Appendix H cire.
Dents at the 1TS are smaller than at the tube the sludge pile region -
cracking technique plus all distorted support plates. Expect cracking at tube support bobbin uxlications and all flaws plates (large dents), roll transition, and in small cacceding the repair limit. If detect, radius U-bend tubes before cracking at TTS expand to 100% of affected (* critical')
(temperature and stresses taken into area in that 50 and 20% in other SGs.
consideration). References 109,131, and 132.
Indan Point 2 100% C5 through ist TSP 100% HL and 20% CL C5. If detect, References 133 and 134.
100% in CL.
l Kewaunee 100% RPC of unsleeved 100% RPC of unsleeved HL at1TS.
Inspect from tube end to 3 inches above the 1TS.
j HL at TTS References 135 and 136.
Point Beach 2 None None.
Historically performed RPC inspection of distorted bobbin indications. For next inspection, will use PP to inspect distorted bobbin indications. References 137 and 138.
1 Prairie Island i 100% RPC in HL at1TS 100% PP or equivalent n HL at TTS.
Performed as a result of D.C. Cook findings.
References 139 and 140.
Prairk Island 2 100% PP in HL at 1TS 100% PP or equivalent in HL at TTS.
Performed as a result of D.C. Cook findmgs.
References 139 and 140.
Zion 1 N/A N/A No denting at TTS. If denting develops, will evaluate need for inspecting for circumferential indications. References 141,142, and 143.
Zion 2 N/A N/A No denting at 17S. If denung develops, will evaluate need for inspecting for circumferential indications. References 141,142, and 141 7-33 NUREG-1604
Table 7-3 Inspections in the U-bend region of small-radil tubes in WM=- :x: partial-depth hardroll plants f
Plant U4end Past lespection U4eed Future Inspectico Notes 1
D.C. Cook i None 20% RPC in R1 and R2 in i SG. If References 88,126, and 127, detect,100% RPC in all SGs. [No 1
degradation detected.)
Genna 100% RPC in R1 and R2 None (SG replacement).
References 128,129, sixi 130.
l Haddam Neck 95% RPC of inner row 20% of inner row U-bends (R2 in SGs R1 tubes in SGs 1 and 2 were preventively i
U-bends (R2 in SGs 1 and I and 2; R1 in SGs 3 and 4) with plugged in 1983 due to flow slot hourglassing.
2: R1 in SGs 3 and 4)
Appendix H circ. cracking technique.
References 109,131, and 132.
1 Expand to 100% and 20% of next row.
i Indian Point 2 None 100% C5 or RPC.
R1 plugged durity construction. Performed RPC
)
exams of bobbin indications only. References i
133 and 134.
l Kewaunce 1005 RPC in R1 and 100% RPC in R1 and 30% RPC in R2.
References 135 and 136.
30% RPC in R2 If detect in R2,100% in R2.
Point Beach 2 None None.
Historically performed RPC inspection of distnrted bobbin indications. For next inspection, will use PP to inspect distoried bobbin indications. References 137 and 138.
Prairie Island i 100% RPC in Rt and R2 100% PP in R1 and R2. [No References 139 and 140.
indications detected.)
Prairie Island 2 100% RPC or PP in RI 100% rotating coil in R1 and R2.
References 139 and 140.
and R2 Zian 1 25 segmented pancake coil 100% in R2 with Appendix H cire.
Preventively plugged R1 in RFO 6 (1982). Next exams per SG cracking technique.
,nspections primardy with RPC. References 141, 5
142, and 143.
Zion 2 100% segmented pancake 100% in R1 and R2 with Appendix H R1 U-bend leak in 1989. Next inspections coil exams in R1 and 25 cire. cracking technique.
primarily with RPC. References 141,142,and exams in R2 per 50 143.
NUREG-1604 7 34
.~
Table 7-4 Inspections at dented locations in Westinghouse pahpth hardroll plaats Plant Dested TSP Past Deuted TSP Future laspection Notes inspection D.C. Cook 1 100% RPC of dents > SV GL 9545 criteria.
GL 9545 cahbradon. References 88,126. and 127.
Ginna RPC sarnpling None (SG replacement).
References 128,129, and 130.
Haddam Neck 200 girnbeled PP exarns of 200 tubes with Appendix H circ.
Used gimbated PP. If find cracking at denu. will largest dents cracking technique. Expand until perform evaluation relating dent size with tube smallest dent size where cracking cracking. Expansion primarily based on would occur is determined.
i non results. Dent size determined by size of in probe that can successfully pass through the location and by bobbin cod voltage.
Cahbration: 20 volts for each 0.001-inch of denting for a 0.012 inch radial dent at 600 kilz.
References 109,131, and 132.
Indian Point 2 100% C5 in SGs 1/2,1st 100% HL and 20% CL C5. If detect, Emanuned dents with restrictions that prevented TSP dents in SGs 3/4 100% in CL.
" adequate" examinadon with a 0.680 or 0.700 j
probe. References 133 and 134.
Kewaunce 100% RPC of dents > SV GL 95-05 criteria.
GL 9545 calibration References 135 and 136.
Point Beach 2 None None.
Historically performed RPC inspection of distorted bobbin indications. For next inspection, will use PP to inspect distorted bobbin indications. References 137 and 138.
Prairie Island 1 Not addressed 100% PP of dents > SV [No GL 9545 calibradon. References 139 and 140.
indications detected.]
Prairie Island 2 Not addressed 100% PP of dents > !*
GL 95-05 cahbration. References 139 and 140.
Zion 1 None 20% RPC of dents > SV concentradng 10 dents > SV. Cabbration: 0.001 inch dent is at lowest TSP. If detect,100% of all set to IV on the 400 kHz absolute channel.
References 141,142, and 143.
Zion 2 None 20% RPC of dents > $Y concentraung I dent > SV. Calibradon: 0.001-inch dent is set at lowest TSP. If detect.100% of all to 1V on the 400 kHz absolute channel.
HL TSPs with dents > $V.
References 141,142, and 143.
l 7-35 NUREG-1604 w
l Table 7-5 Inspections at sleevejoints in Westinghouse partial-depth hardroll plants Plant Sleeve Past laspection Sleeve Future lyw Notes i
l D.C. Cook 1 Mm (crosswound) 100% C5. [No degradation detected.)
1840 Wesunghouse HEl sleeves instaued in 1992.
References 88,126, and 127.
Ginna 20% sarnpling None ($O replacement).
Probe type used dependent on of sleeve inspected (e.g., PP used for O welded sleeves). Sleeve types used: B&W brazed.
B&W explosive, B&W kinetic, CE T10 welded.
References 128,129, and 130.
Haddarn Neck N/A N/A No sleeves installed. References 109,131,and 132.
l l
Irdian Point 2 N/A N/A No sleeves installed. References 133 and 134.
I Kewaunee 100% PP of upper joint of 100% PP of upper joint of HEJ sleeves.
Westinghouse HEJ sleeves insulted in 1988, HEJ sleeves; 100% PP of 1989, and 1991. CE welded sleeves installed in the upper weld of CE 1992. Majority of indications at top of lower sleeves hardroll transition of HEJ sleeves. References 135 and 136.
Point Beach 2 100% C5 20% PP or C5 Expansion based on HEJ cracking. 222 indicauons in 3001 sleeved TS.
tubes. Nearly all indications at HEJ hardroll lower transition. A few indications at the upper ard lower hydraulic transitions. References 137 and 138.
l Prairie Island 1 319 sleeves with I-cod 100% PP. [No service induced 680 CE 'nO welded sleeves were installed as of cracking.]
1996 (all in 5012). Non-service induced circumferential indications detected were attributed to weld oxide inclusions and lack of fusion caused by inadequate cleaning of the parent tube prior to welding. References 139 amt 140.
Prairie Island 2 N/A N/A No sleeves installed. References 139 and 140.
Zion 1 None (crosswound) 100% PP.
47 HEJ and 759 CE TIO welded sleeves installed.
References 141,142, and 143.
Zion 2 100% PP 100% PP.
414 CE TIO welded sleeves installed. References 141,142, and 143.
i i
NUREG-1604 7 36
l l
1 l
Table 7-4 tr;- +n at the exp===ta= transition region in Westinghouse fuu-depth hardroll plants Fineg r--
Trwh F-
- Transition Future Noten het 6p=*6
!aspection Beaver Valley 2 20% RPC 20% RPC. If detect 100% RPC.
Sixxpeeped prior to conunercial service in 1987.
References 91 and 144.
Braidwood i 100% RPC in HL
, 100% in HL with Appendix H circ.
Shotpeened HL and CL pror to operation. Next cracking technique. If widespread inspections primarily with RPC unleas dented.
cracking, s.aple CL References 120,141, and 143.
Byron i 100% RPC in HL 100% in HL with Appendix H circ.
Shotpcened HL in RFO 1 and CL in RFO 3.
cracking technique. If widespread Next inspections primarily with RPC unless cracking, aampic CL.
dented. References 120,141, and 143.
Catawba 1 100% RPC in HL None (SG replacement).
Shot peened HL in 1987 and CL in 1991. CL ia=a-ted in 1992. References 50,52, and 123.
Cornanche Peak i 27% RPC in 2 of 4 SGs 20% in HL with Appendix H circ.
89.5 % of tubes are hardroll expanded (14473 including 100% of the cracking echnique.. If detect,100%.
tubes). The remainder (3839) have WEXTEX WEXTEX tubes in these 2 SGs expansions. Shotpeened transitions prior to operation. Anomalous readings dunng 1995 inspection. References 89,124, and 145.
Farley 2 100% RPC in HL 100% RPC in HL 20% RPC in CL.
Shotpeenrxl HL in 1987. Only I circumferential Expansion based on inspecuan results.
indication detected to-date. References 35 and 146.
j McGuire i 100% RPC in HL 100% RPC in HL 20% RPC in CL Shotpeened HL in 1986 and CL in 1990. 100%
RPC in CL in 1991. References 50,52, and 147.
McGuire 2 100% RPC in HL 100% RPC in HL 20% RPC in CL Shotpeened HL in 1987 arat CL in 1989. 100%
l RPC in CL in 1992. References 50,52, and 147.
Shearon Harns 1
-400 RPC exams in HL 20% RPC in HL If detect,100%
Rosopeened HL prior to service. Shotpeened CL in 2 of the 3 SGs RPC. [Actually performed 100% in in RFO 3. References 121 and 148.
HL. Detected 3 cire, indications.]
South Texas Project i 100% RPC in HL 100% RPC in ifL. If detect sufficient Rotopeened HL and CL prior to comnercial amount of cracking,20% RPC in CL service. Have identified some rollmg anomalies -
these tubes are monitored more frequently. Cire.
indications located near terminal ends of the blowdown pipe. References 94,125, and 149.
Watts Bar i N/A 20% RPC, or equivalent, in HL If Rotopeened HL and shotpeened CL prior to detect,100% in aflected SG. If commercial service. Expect plant to commence widespread cracking,20% RPC in CL commercial operation in late 1995. Reference in all SGs.
- 150, 7-37 NUREG-1604
_ _ ~ -_ _
Table 7-7 Inspections in the U-bend region of asiall-radil tubes in Westinghouse full-depth hardroll plants Plass U4eed Past %w U4end Future Inspect 6ao Notes Beaver Valle.T 100% RPC in R1 and R2 100% RPC in R1 and R2.
UBift in R1 and R2 prior to commercial operation in 1987. References 91 and 144.
Braidwood i None 20% in Rt and R2 with Appendix 11 UBHT in RI and R2 in RFO 1. 100% RPC in cire, cracking technique. If cetect, R1 in RFO 2. Next inspection primarily with 100% in R1 ard R2.
RPC, References 120,141, and 143.
Byroni None 20% in R1 and R2 with Appendia H UBifT in R1 and R2 in RFO 2. Next inspecuon cire. cracking technique. Ir detect, primarily with RPC. References 120,141, and 100% in R1 and R2.
143.
Catawba 1 1005 RPC in R1, R2, and None (SG replacerners).
UBiff in R1 and R2 in 12/88. References 50, R3. 20% RPC in R4 52, and 123.
Comanche Peak i None 20% in R1 and R2 with Appendix H UBiff in R1 and R2 prior to operation.
technique. If detect 100% in Rt and References 89,124, and 145.
R2.
Farley 2 100% RPC in R1 and R2 100% RPC m R1 and R2 in 1 SG. If Preventively plugged low row U-bends in 1980s.
in one 50 detect, 2005 in R1 ard R2 in all sos.
Recovered R1 in 1990 (2R7). UBlfT in 1990.
No cire. cracking since UBIfT. References 35 and 146.
McGuire i 100% RPC in R1, R2 and 100% RPC in R1, R2, and R3; 20%
UBIfr in 10/88. References 50,52, and 147.
R3. 20% RPC in R4 RPC in R4.
McGuire 2 100% RPC in R1, R2 and 100% RPC in R1, R2, and R3; 20%
UBifT in 6/88. References 50,52, and 147.
R3. 20% RPC in R4 RPC in R4.
Shearon Harris i None 20% RPC in R1 and R2. If detect, UBirt in R1 and R2 prior to servce. References 100% RPC in RI and R2. [No circ.
121 and 148.
cracking detected.]
South Texas Project i Some RPC sampling None.
Larger bend radii than other Westinghouse SGs.
UBHT in R1 and R2. Not susceptible to circ.
cracking in this region. Monitoring of this region done for conservatism. References 94,123, and 149.
Watts Bar i N/A 20% RPC, or equivalent, in R1 and UBHT in Rt and R2 in 1986 (prior to commercial R2. If detect, luu% in R1 and R2 in servce). Expect p' ant to commence commercial affected 50.
operation in late 1995. Reference 150.
NUREG-1604 7-38
~-
-._---.. -._ ~
1 Table 7-8 Inspections at dented locations in Westinghouse full-depth hardroll plants Plant Dested 'I3P Past Deated TSP Futuce laspection Notes Inspection Beaver Vdey 2 7% RPC of dems > SV 100% RPC of dents > SV. Expand Denting not prevalem. Most dents / dings present based on dependence of indications on since preservice Dents at uppermost hot and i
voltage.
cold leg TSP inspected for Bulledn 88-02 phenomena. Considered a dent if voltage exceeds SV. Calibration: 4-20% holes set to 4V peak-to-peak on all channels. References 91 and 2;
144.
1 j,
Braidwood i 100% RPC of dents > SV GL 9545 criteria, GL 9545 calibration. Next inspection will primarily be with RPC. References 120, 141 and 143.
Byron 1 100% RPC of dents > SV GL 9545 criteria.
GL 9545 calibradon. Next inspet: tion will primarily be with RPC. References 120, 141, and 143.
4 Catawba 1 100% RPC of dents > 4V None (SG replacernent).
GL 95-05 cahbraten. References 50,52, and 123.
Comanche Peak i None 100% of dents > $V atlowest HLTSP Cahbration: 4-20% holes set to 2.75V on 4
with Appendiz H technique. If detect, 550/130 mix. References 89.124, and 145.
j 100% of dents > SV in all SGs.
1 Farley 2 100% RPC of dents > SV GL 95-05 criteria.
GL 9545 calibration. References 35 and 146.
McGuire 1 100% RPC of dents > 4V 100% RPC of dents > 4V.
GL 9545 calibradon. For freespan dents, the dent voltage is measured using the 400 kHz differential channel. References 50,52, and 147.
McGuire 2 100% RPC of dents > 4V 100% RPC of dents > 4V.
GL 9545 calibration, For freespan dents, the dent voltage is measured using the 400 kHz differential channel. References 50,52, and 147.
Shearon Harris i None 20% of dents > SV with Appendix H Calibration: 20% holes set to 4V on 550 kHz i
circ. cracking technique. If detect, differendal channel. Record dents at SV.
j 100% of HL dess in affected SG and References 121 and 148.
20% of CL dents in affected SG. [No circ. cracks detected.]
South Texas Project i None RPC exam of the one dent that may be less than 50 dents. Most dents are traceable i
service related.
back to manufacturing artifacu with the exception i
of one possible service induced dent Record dents at 3V. GL 95-05 calibration. References 94,125, and 149.
Watts Bar 1 N/A 20% RPC, or equivalent, of dents >
A few fabrication related dents were observed SV at TSPs 1 and 2. If detect,100%.
during the preservice inspection. Expect plant to comrnence commercial operadon in late 1995.
Reference 150.
7-39 NUREG-1604
~
wr
Table 7-9 Inspections at sleeve joints in WM' -i _ full-depth hardroll plants Plant Sleeve Pad Inspectiess Sleeve Futuee Ingection Notes Deaver Valley 2 N/A N/A No sleeves installed. References 91 and 144.
Braidwatul 1 N/A if installed, Appendix H technique.
No sleeves installed. Next inspections, if any, will prirnarily be with PP or Cecco. References 120,141, and 143.
Byron 1 N/A If installed, Appendix H technique.
No sleeves installed. Next inspections. if any, will primarily be with PP or Cecco. References 120,141, and 143.
Catawba 1 N/A N/A No sleeves in service. Plugged all previously sleeved tubes. References 50,52, aruf 123.
Cananche Peak i N/A N/A No sleeves installed. References 89,124, and 145.
Farley 2 100% C5 in 150 100% CS, Only laser welded sleeves installed. First installation in 1992. All sleeves have been heat treated. References 35 and 146.
McGuire i N/A N/A No sleeves in service. Plugged all previously sleeved tubes. References 50,52, and 147.
McGuire 2 N/A N/A No sleeves in service. Plugged all previously sleeved tubes. References 50,52, and 147, I
Shearon Harris 1 N/A N/A No sleeves installed. References 121 and 148.
South Texas Project 1 N/A N/A No sleeves installed. References 94,125, and 149.
e l
Watts Bar i N/A N/A No sleeves installed. Reference 150.
]
i i
i 1
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NUREG-1604 7-40
i Table 710 Inspections at the expension transition region in Westinghouse WEXTEX plants Plang Frp=* Transition Expansion Trandtion Future Notes Past laspection laspection Beaver Valley 1 100% RPC 100% wnh Appendix H circ. cracking Only I circumferential indication detected, technique References 91 and 144.
Comanche Peak 1 27% RPC in 2 of 4 SGs 20% in HL with Appendtx H circ.
89.5% of tubes are hardroll expanded (14473 including 100% of the cracking technique. If detect 100%.
tubes). The remainder (3839) have WEXTEX WEXTEX tubes in these 2 expansions. Shotpeened transitions prior to SGs operation. Anomalous readings during 1995 inspection. References 89 and 145.
Diablo Canyon 1 22% RPC in HL in SGs 50% PP in zone 4. Sample outside Shotpeened HL in IRS (1992). References 87, 11,12 and 14:23% RPC zone 4. If detect, expand per WOG 101, and 151.
in HL in SG 13 WEXTLX guidelines.
Diablo Canyon 2 28% RPC in HL in SG 50% PP in zone 4. Sample outside Shotpeened HL in 2RS (1993). I circumferential 21; 22% RPC in HL in zone 4. If detect, expand per WOG Indication detected in 1993 and I circumferential SGs 22 and 23; 46% RPC WEXTEX guidelines.
Indication detected in 1994. References 87,107, in HL in SG 24 anxi 151.
Farley 1 100% RPC in HL 100% RPC in HL. 20% RPC in CL.
References 35 and 146.
i If detect, expand per WOG WEXTEX guidelines, j
i Salem 1 39% RPC in HL in SG 100% C5 or PP in HL zone 4. 200 Shotpeened HL in 1993. 4 circumferential i1; 21% RPC in Illin tube sample with C5 or PP in HL zones indications detected. Previous exams SGs 12 and 14; IJO%
1,2, and 3. If detect, expand per concentrated in zone 4. References 152 and 153.
RPC in HL in SG 13 WOG WEXTEX guidelines. (Actually inspected 100%.]
Salem 2 1005 RPC in HL in SG 100% C5 or PP in HL zone 4. 200 Shotpecned HL in 1994. I circumferential 21; 21% RPC in HL zone tube esmple with C5 or PP in HL zones indication.tetected. Previous exams concentrated 1
4 in SGs 22 and 23; 27%
1,2, and 3. If detect, expand per in zone 4. References 152 and 153.
RPC in HL zone 4 in SG WOG WEXTEX guidelines.
24
)
Sequoyah1 50% RPC in SGs I and 2; 100% RPC, or equivalent, in zone 4.
Shotpeened in Cycle 5 RFO (10/91). References 61% RPC in SG 3; 100%
If detect, expand per WOG WEXTEX 99 and 154.
RPC in SG 4 guidelines.
Sequoyah 2 45% RPC in SGs I and 4; 100% RPC, or equivalent, in zone 4.
Snotpeened in Cycle 5 RFO (1992). References 100% RPC in SG 2; 63%
If detect, expand per WOG %IXTEX 99 and 154, RPC in SG 3 guidelines.
1 7-41 NUREG-1604
i e
Table 711 7 ^' : in the U-bend region of small-radii tubes in WM' f:r: WEXTEX plants Plant U-bena Past laspecties U-bend Futere laspectice Notes 1
Beaver Valley 1 1005 RPC in R1 and R2 100% in R1 and R2 with Appendix H UBiff in R1 and R2 in 1995. References 91 and cire. cracking tecimique 144.
l Comanche Peak i None 20% in Rt and R2 with Appendis H UBifT in Rt and R2 pror to operauon.
technique. If detect,100% in R1 and References 89 and 145.
R2.
i DistNo Canyon i 100% RPC in Rt and R2 100% RPC in R1 and R2.
UBHT in R1 ar.d R2 in 1R2 (1988). 4 circumferential uidications detected in R1 in 1992. Refercrus 87,107, and 151.
J l
Dieth Canyon 2 100% RPC in Al and R2 100% RPC in Rt and R2.
UBirr in Rt and R2 in 2R1 (1986). References 87,107, and 151.
Farley 1 1005 RPC in R1 and R2 100% RPC in R1 and R2.
Prevenuvely plugged low R/a U bends in 1980s.
Recovered R1 in 1991 (ifJJ). UBiff in R1 and R2 in 1991. 3/91 iruficatiys were present prior to UBiff(i.e., recovered tubes). No l
circumferential indications detected since UBiff.
References 35 and 146, i
Salem i Not addressed 20% RPC in R2. If detect,100% in Preventively plug 8ed Rt U-bends in 1989.
R2.
UBifT in R2 in 1991. References 152 and 153.
Salern 2 Not addressed 20% RPC in R2. If detect.100% in Prevenovely plugged R1 U-bends in 1988.
R2.
UBitT in R2 in 1990. References 152 and 153.
Sequoyah1 100% RPC in Rt. RPC 20% RPC in Rt and R2. If detect.
UBifT in R1 and R2 during cycle 3 RFO (1987).
sampling in R2 100% RPC in R1 and R2.
References 99 and 154.
N-iah 2 100% RPC in R1 and R2 20% RPC in R1 and R2. If detect, Preventively plugged R1 in 4/88. Deplugged RI 100% RPC in RI and R2.
in Cycle 6 RFO (1994). UBifT in R1 and R2 in Cycle 6 RFO (1994). References 99 and 154.
NUREG-1604 7 42
Table 7-12 Inspections at dented beh in Wu-N WEXTEX plants I
Plant Deuted TSP Past Dested TSP Future laspectica Notes Ing=esta=
i Beaver Valley 1 100% RPC of dents > SV GL 95-05 criteria.
Dentmg not prevalent. GI 954)5 calibradon.
References 91 and 144.
Comanche Peak i Nonc 100% of dents > SV atlowest HLTSP Calibration: 4 20% holes set in 2.75V on with Appendix H technique. If detect.
550/130 mix. References 89 and 145.
100% of dents > SV in all SGs.
Diablo Canyon i 10% to 26% RPC of dents 20% PP of HL dents > SV. If detect, GL 95-05 cahbration. Exams will be focused at
> 3V 100% PP of dents > $V at affected dents at the lower hot les TSP elevations. Will TSP clevation and the lower TSP perform 100% PP of dents > 5V at IH,2H, and elevations and 20% PP of dents > SV 3H. If more exams are needed to obtain a 20%
at the next higher TSP elevation.
sanple, additional exams above 3H will be performed. References 87,107, and 151, Diablo Canyon 2 100% RPC of dents > SV 20% PP of HL dents > 5V. If detect, i GL 95-05 calibtation. Exams will be focused at 100% PP of dents > SV at affeued dents at the lower hot leg TSP elevations.
TSP elevadon and the lowr; l'SP References 87,107, and 151.
elevations and 20% PPor dents > 5V at ea next higher TSP clevation.
Farley 1 100% RPC of dents > 5V GL 9545 criteria.
GL 95-05 calibration. References 35 and 146.
Salem 1 580 RPC exams of dents 100% C5 or PP of dess > $V at TSPs No circumft:rential indications detected; however,
> SV 1,2, and 3. 20% C5 or PP of dents >
5 axial indications detected. Calibration: 20 %
SV at TSP 4. If detect.100% C5 or ASME defect set to 4V on 400 kHz differential PP of dents > SV at affected TSP channel and saved to all channels. Record dents elevation and 20% C5 or PP of dents at SV from 400/100 kHz differential mix.
> SV at next higher TSP elevation.
References 152 and 153.
Salem 2 64I RPC exams of dents 20% C5 or PP of dents > SV at TSP No circumferential indications detected.
> SV
> SV at affected TSP clevation and kHz differential channel and saved to all 20% C5 or PP of dents > SV at next channels. Record dents at SV from 400/100 kHz higher TSP elevation.
differential mix. References 152 and 153.
Sequoy s i 100% RPC of dents > $V 1005 RPC, or equivalent, of dents >
SGs 3 and 4 have a large number of dents > SV, through HL TSP 5 SV at HL TSPs 1 and 2 in SGs 3 and 4; GL 95-05 calibration. Expansion cri:eria 20% RPC, or equivalent, of dents >
specified in technical specification amendment SY at HL TSP 3 in SGs 3 and 4; 100%
related to *oitage-based tube repair criteria RPC, or equivalent, of HL dents > SV (10-11-95). References 99 and 154.
In SGs 1 and 2. If detect, expand based on results.
Sequoyah 2 100% RPC of dents > 5V 100% RPC, or equivalent, of dents >
Small number of dents > 5V. GL 954)5 through HL TSP 3 5V at HL TSPs 1 and 2; 20% of dents calibration. References 99 and 154.
> SV at HL TSP 3. If detect 100%
RPC, or equivalent, of dents > SV at affected HL TSP elevation and 20% of dems > SV at next higher HLTSP elevation.
7-43 NUREG-1604
l Table 7-13 Inspections at sleeve joints in WH' f 1_ WEXTEX plants -
Plant Sleeve Past laspection Sleeve Funere :. ;i--
Notes Beaver Valley i N/A N/A No sleeves insulled References 91 and 144.
j Comanche Peak i N/A N/A No sleeves installed. References 89 ami 145.
)
1 l
Diablo Canyon 1 N/A N/A No sleeves installed. Sleeve like inserts insuued in 16 tubea. Inserts periodically tested with RPC probe. References 87,107, and !$1.
l Diablo Canyon 2 N/A N/A No sleeves installed. References 87,107, and 151.
Farley I None (crosswound) 100% C5.
Only laser wekled s!ceves insulled. ist I
installadon in 1992. All free span j*oiras have received a post weld stress relief. References 35 and 146.
i Salem i N/A N/A No sleeves installed. References 152 and 153.
Salem 2 N/A N/A No sleeves installed. References 152 and 153.
Sequoyah i N/A N/A No sleeves installed. References 99 and 154.
I Sequoyah 2 N/A N/A No sleeves installed. References 99 and 154.
1 l
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NUREG-1604 7-44
Table 7-14 Inspections at the expansion transition region la Wd 7x: hydraulic plants (Part 1)
Plang F-.
Trundtion W
Treadtion Futuee Notes Past laspecanon Inspection Braidwood 2 10% RPC in HL in 2 of 4 20% in 11L with Appendix H circ.
Next inspections primarily with RPC unless SGs cracking technique in each SG dented. References 120,141, and 143.
inspected. If detect,100% in HL in all SGs. If widespread cracking, sampic CL.
Byron 2 20% RPC in HL in 1 of 4 20% in HL with Appendix H circ.
Next inspections primarily with RPC unicss SGs cracking technique in each SG dented. References 120,141, and 143.
inspected. If detect,100% in HL in all SGs. If widespread cracking, sample CL.
)
Callaway 100% RPC in HL MA 100% in HL with Appeadix H Shotpeened HL in 1992. TT tubes in rows 1 tubes; 8.3% RPC in HL technique.
through 10 (1214 tubes per steam generator).
TT tubes Remaining tubes are MA. Previous exams concentrated in sludge zone. References 110 and 155.
Cauwba 2 100% RPC in HL
> 50% RPC in HL. If detect,100%
References 50,52, and 123.
RPC.
Cnmm-he Peak 2 359 RPC exams in HL in 40% with Appendix 11 cire. cracking References 89 and 145.
3 of 4 SGs technique in each SG inspected If detect,100% in all SGs.
l Millstone 3 500 C5 exams in 2 of 4 500 tubes with Appendix H cire.
Will concentrate inspections in sludge pile.
SGs cracking technique in 2 of 4 SGs. If References 109,131, and 132.
detect,100% in affected SG in critical area and 20% ouuide critical area plus 20% in critical area in all other SGs.
Point Beach i None None.
References 137 and 138.
Robinson 2 None 40% rotating probe in 1 $0.
References 156,157, and 158.
Seabrook1 None 500 tubes with Appendix H cire.
Will concentrate inspections in sludge pile. If cracking technique in 2 of 4 SGs. If find cires. outside sludge pile 100% of tubes in detect,100% in sludge pile and 20%
all 4 SGs will be inspected. References 89 and outside sludge pile in all 4 SGs.
132.
South Texas Project 2 21% RPC in HL 100% RPC in HL. If detect sufficient Shoipeened IIL and CL in ist and 2nd RFO.
amount of cracking,20% RPC in CL.
References 94,125, and 149.
Surry i 3% RPC in 1 of 3 SGs 3% of all tubes in 1 SG with Appendix Some tubes overexpanded above the TTS.100%
H technique. If detect,6% of all tubes RPC exam in 1990, inspections focused in in affected SG. Further sample sludge pile area. References 100 and 101.
expansions based on results.
Surry 2 3% RPC in I of 3 SGs 3% of all tubes in 1 SG with Appendix Some tubes overexpanded above the 1TS.
H icchnique. Itdetect,6% of all tubes Inspections focused in sit 4ge pile area.
in affected SG. Further sampic References 100 aid 101.
expansions based on results.
7-45 NUREG-1604
Table 7-14 Jopections at the expansion transition region in W# : bydraulic plaats (Part 2)
Plant f~=p==da= Transkies F y=ada= Transkies Future Neses Past Imapartion Imapardam Turkey Point 3 100% RPC of 20% RPC in HL overexpanded tuues in Several tubes expanded above the tutwstact.
overenpanded tubes in 2 1 or more SGs. If detect,100% RPC targer expansions receive prionry in deiermining of 3 SGs in HL overenparuled tubes in affected which tubes to exarnine. References 103,159, SG, sample in CL,20% RPC in HL 160,161, and 162, expanded tubes in at least one other SG, and sample of ;ioreoverexpanded tubes.
Turkey Poirs 4 12% RPC of 20% RPC in HL overexpanded tubes in Several tubes expasuled above the tubeabset.
overenpanded tubes in i I or more SGs. If detect,100% RPC larger *xpansions receive priority in determining of 3 SGs in HL overexpanded tubes in affected which tubes 80 examine. Overexpanded tubes in SG, sample in CL,20% RPC in HL other SGs were sampled in previous inspections.
expanded tubes in at least one other Reference + 501 159,160,161, and 162.
SG, and sample of non-overespanded tubes.
Vogtle i None 20% RPC in HL in 2 o(4 SGs with 20% sample includes 420 tubes (7.5 %) in sludge Appendix H cire. cracking technique.
pile. Sludge pile surrounded by rows 1 and 20, if detect outside sludge pile,100% in columns 50 and 70. References 102 and 163.
affected SG. If detect inside sludge l
pile, inspect sludge pile in other 2 SGs armiinspect 3 columns and rows of tubes outside defined sludge pile in the
- 50. Condnue until bounded.
Vogtle 2 None Same as for Vogde Unit i except that See Vogde Unit 1. References 102 and 163.
EPRI sample sixs guidelines may be used. Expansion criscria is the same as for Unit 1.
Wolf Creek 1 11% RPC in HL in 1 of 4 10% in HL in 2 of 4 SGs with Focus exams in sludge pile. References 164 and SGs App % in s!! 4 SGs.endix H technique. If vetect, 165.
luu I
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NUkEG-1604 7-46
T Table 7-15 laspections in the U-bend region of small-radii tubes in Wd "- _ c Hydraulic plants (Part 1)
Fleet U4end Pisst laspecties U4eed Future lampact6es Notes Braidwood 2 None 20% in RI and R2 with Appendix H Rows 1 through 7 received an addauonal heat circ. cracking technique in each SG treatment after bendm inspected. If detect,100% in all SGs.
primarily with RPC, g. Next inspection References 120,141, and 143.
Byron 2 None 20% in R1 and R2 with Appendix H Rows 1 through 7 received an additional heat cire. cracking technique in each 50 treaunent after bending. Next inspecuan inspected. If detect,100% in all SGs.
prirnarily with RPC. References 120,141, and 143.
l l
l Callaway None 100% in R1 with Appendix H Tubes in rows 1 through 10 were thermally technique in SG C.
treated after bending. Found i undefined i
indication in a R2 tube in 1992 which was considered by the licensee to be an anomsty.
References 110 and 155.
Catawba 2 Not Addressed 20% RPC in R1 and R2. If detect, References 50,52, and 123.
100% RPC in Rt and R2.
- - - ' Peak 2 None 40% in R1 and R2 with Appendix il Rows 1 through 9 were heat treated prior to SG technique in each SG selected. If assembly References 89 and 145.
detect,100% in all SGs.
Millstone 3 None 20% in RI with Appendix H circ.
Rows I through 10 were stress reheved.
cracking technique in 2 of 4 SGs. If References 109,131, and 132.
detect,100% in RI and R2 in all SGs.
Point Beach i None None.
References 137 and 138.
Robinson 2 None 20% rotating probe in Rt and R2 in i References 156,157, and 158.
SG.
Seabronti None 20% in RI with Appendia H cire.
Rows i through 10 were stress reheved.
cracking technique m 2 of 4 SGs. If References 89 and 132.
detect,100% in R1 and R2 in all SGs.
South Texas Project 2 Some RPC sampling Some RPC sampling.
1.arger bend radii than other Westinghouse SGs.
UBHT in R1 and R2. Not susceptible to circ.
cracking in this region. Monitorms of this region done for conservatism. References 94,125, and
- 149, Surry 1 None None.
Rows 1 through 8 were stress relieved after bending. References 100 and 101.
Surry 2 None None.
Rows 1 through 8 were stress relieved after bending. References 100 and 101.
I 1
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7-47 NUREG-1604 t
Table 7-15 Insg+:;'--- la the U-bend region of small-radG psbes in W+d'--7:_ : Hydraulic plants (Part 2)
N at t%eed Pad Iew 14eed Futwe laspwnes Notes Turkey Point 3 None None.
References 103,159,160,161, and 162.
Turkey Point 4 None None.
References 103,159,160,161, and 162.
Yogue i None None.
Thermally treated tubes. References 102 and 163.
~Vogde 2 None None.
Thermaliy treated tubes. References 102 and i
163.
Wolf Creek i None 20% in R1 with Appendit H technique R1 U bend radius slighdy larger than older rnodel in 2 of 4 SGs. If detect,100% of R1 Westinghouse SGs. References 164 and 165.
and R2 in these 2 SGs. If detect in expansion, expand to higher rows in original 2 SGs and inspect in RI and R2 in other 2 SGs. Continue until bounded.
4 NUREG-1604 7-48
= -.
~ -. - -
Table 716 laspections at dented lar=elaas in Wd--i c: Hydraulic plants (Part 1)
Plant Dented TSP Past Deuted TSP Futace laspection Notes lampacties Braidwood 2 None 20% of dems > SV at HL TSPs with OL 9545 calibrauon. References 120,141,and Appendia H eire, cracking iechnique in 143.
each 30 irupected. If detect.100% of dents > SV at HL and CL TSPs in all SGs.
Byron 2 None 20% of dents > SV at HL TSPs with GL 9545 calibration. References 120,141,and Appendia H cire. cracking technique in 143.
each SG 'mspected. If detect.100% of dents > SV at HL and CL TSPs in all SGs.
Callaway None None.
No service induced denting. References 110 and 155.
Catawba 2 N/A N/A No denung. References 50,52, and 123.
l Commanche Peak 2 None 100% of dents > 5V atlowest HLTSP Calibradon: 4-20% holes set to 2.75V on with Appendix H technique. If detect.
550/130 mix. References 89 and 145.
100% of dents > SV in all SGs.
Millstone 3 100 exams with Appendia None.
Calibration: 4 20% holes set to 4V on 630/150 H technique during kHz differential mix. Dents (TSPs, sludge pile) prevmus 2 outages reported from 630/150 kHz mia if voltage > SV.
Dings (freespan) reporied from 630 kHz differential channel with no voltage threshold.
No corrosion induced deming has been observed, References 109,131, and 132.
Point Beach i None None.
References 137 and 138.
Robinson 2 None 100% totaung probe of HL dents >
Calibradon: 4-20% holes set to $V on 400 kHz 2V.100% rotating probe of dents >
channel and saved to other channels. Not 2V at a HL structure or > 10V in experiencing an active denting mechanism. Dents freespan in 1 SG.
in freespan (not at a structure or at TTS) recorded at 10V from 400 kHz channel. Dents at structures recorded at 2V from 400/100 kHz mis channel. References 156,157, and 158.
Seabrook i 12 rotating probe exams None.
Calibrauon: 4-20% holes set to 4V on 630 kHz during previous 2 outages differential channel Record dents at SV from 630/150 kHz differential mix. Dents in freespan are called dings. Dings are recorded from 630 kHz with no voltage threshold. References 89 and 132.
South Texas Project 2 None None.
lass than 50 dents. Most dents are traceable back to manufacturing artifacts. Staltdess steel TSPs make unit resistant to denting and, j
therefore, not susceptible to circ. cracking in the 13P area. References 94,125, and 149.
Surry 1 None None.
Denting not expected due to stamiess steel TSPs.
i References luu and 101.
Surry 2 None None.
Denting not expected due to stainless steel TSPs.
References luu and 101, i
1 7-49 NUREG-1604
Table 716 Inspections at dented locations in Westinghouse Hydraulic plants (Part 2) i Plant Dested TSP Past Dested TSP Future laspection Notes Inspatka Turkey Poim 3 2 % RPC in i of 3 SGs 20% RPC of HL denis in 1 or more Caltrauon: 4-20% holes set to 5V on 400/100 SGs. If detect,100% RPC of HL dents kHz mix. Denting is from manufacturing in sffected 50, sample in CL. 20%
praese. Criteria for calling a dent not specirud.
RPC of HL denu in at least one other utrger dents receive priority in determanmg 50, and sample of non<lented tubes.
which denu to exanww. References 103, 159, 160,161, and 162.
Turkey Pomt 4 54% RPC in 1 of 3 SGs 20% RPC of HL derns in 1 or more Calibration: 4-20% holes set to SV on 400/100 50s. If detect,100% RPC of HL denu kHz mia. Denting is frorn manufacturing in affected SG, sample in CL,20%
process. Criteria for calling a dent not specified.
RPC of HL dents in at least one other larger dents receive priority in determmug SO, and sample of non<lented nabes.
which dems to examine. References 103,159, 160,161, sul 162.
Vogtle i N/A N/A No denting. References 102 and 163.
Vogtle 2 N/A N/A I
No denting. References 102 and 163.
Wolf Creek i N/A N/A No denting. Refererces 164 and 165.
i r
I t
NUREG-1604 7-50
l l
Table 7-17 Iz;- ^'- at sleeve joints im We'--f-
- bydraulic plants (Part 1)
Pimmt SeevePastl e Seeve Fatare IPL Notes Brsidwood 2 N/A N/A No sleeves insulted. References 120,141, and 143.
Byron 2 N/A N/A No sleeves installed. References 120,141, and 143.
CaJiaway N/A N/A No sleeves installed. References !10 and 155.
Catawba 2 N/A N/A No sleeves in servi::e. References 50,52, and 123.
C --* Peak 2 N/A N/A No sleeves installed. References 89 and 145.
Millstone 3 N/A N/A No sleeves installed. References 109,131, and 132.
Point Beach i N/A N/A No sleeves installed References 137 and 138.
Robinson 2 N/A N/A No siceves installed. Refererces 156,157, and 158.
Seabrooki N/A N/A No sleeves installed. References 89 and 132.
South Texas Project 2 N/A N/A No sleeves installed. References 94,125, and 149.
Surry 1 N/A N/A No sleeves installed. References 100 and 101.
Surry 2 N/A N/A No sleeves installed. References 100 and 101.
7-51 NUREG-1604
.m..
l Table 7-17 Inspections at sisevejoints la W =' 7 -__ bydraulic plaats (Part 2)
Plant 81erve Past '- ; '
sleeve Future Ivw Neses l
l Turkey Point 3 N/A N/A No sleeves nstalled. References 103, 159, 160, i
l l
161 and 162.
-Turkey Pains 4 N/A N/A No sleeves installed. References 103, 159, 160, 161, anr1162.
i Vogtle i N/A N/A No saceves installed. References 102 and 163.
j Vogtle 2 N/A N/A No sleeves installed. References 102 and 163.
Wolf Creek 1 N/A N/A No sleeves installed. References 164 and 165.
\\
l NUREG-1604 7-52 4
i i
I 8 STEAM GENERATORS WITH ALLOY 690 TUBES 8.1 Den and Emh S w ible to limit the susceptibility of the alloy 690 tubes to Circumferential Crackmg degradation at these locations are described below.
Because of steam generator tube degradation, a The alloy 690 material used in these replacement number of plants have replaced their original steam steam generators is intended to be less susceptible to generators (refer to Appendix B).
For the vast degradation than alloy 600 material. In addition, the majority of these replacement steam generators, thermal treatment of the alloy 690 tubes is intended l
changes in the design and materials of consudon to further reduce the potential for tube cracking were made to reduce the susceptibility of the steam compared to mill-annealed tubes. To date, thermally generator tubes to degradation. As of June 1995,13 treated alloy 690 tubes in operating steam generators
. plants had replaced their steam generators. Between have not experienced cracking although the operating June 1995 and December 1996, an additional 3 plants experience has been limited (Ref.109). In summary, had replaced their steam generators. A fourth plant, thermally treated alloy 690 is intended to provide a Salem 1, began replacing their steam generators in substantial increase in the resistance to corrosion December 1996. He steam generators that have compared to mill-annealed tubing and compared to been replaced are presented in Table B-6.
alloy 600 thermally treated tubing (Ref.100).
Steam generators have been replaced at Westinghouse The susceptibility of the expansion transition region and CE units. No B&W steam generators have been to cracking is expected to be lower in steam replaced to date. The tube material in use in the generators with alloy 690 tubes. This is a result of i
l steam generators at the 16 units which have replaced the hydraulic process used in these steam generators l
their ateam generators include mill annealed alloy 600 to expand the tubes within the tubesheet. This tubes (1 plant), thermally treated alloy 600 tubes hydraulic process is expected to result in residual i
I (6 plants), and thermally treated alloy 690 (9 plants).
stresses at the expansion transition which are lower Steam generators currently (1990s) being put into than for tubes expanded within the tubesheet with service in the United States generally have alloy 690 other techniques (e.g., hardrolling, explosive) thermally treated tubes. For the 6 units with steam (Ref.100). De licensee for D.C. Cook 2 concluded
'1 generators with alloy 690 tubes at the time GL 95-03 that the inherent corrosion resistance of the steam 1
was issued, one unit (Indian Point 3) installed generator tubes (i.e., alloy 690 thermally treated j
Westinghouse model 44F steam generators, three tubes) coupled with the low design stress levels in the units (D.C. Cook 2, North Anna 1 and 2) installed full. depth hydraulically expanded tube-to-tubesheet Westinghouse model 54F steam generators, one unit joint effectively preclude concerns about (Millstone 2) installed steam generators manufactured circumferential cracking in the tubesheet region
]
by Babcock and Wilcox, Canada (BWC), and one (Ref. 88).
unit (Summer) installed Westinghouse model Delta 75 steam generators.
The design of the tube support structures in the replacement steam generators with alloy 690 tubes is
%e design of the replacement steam generators with intended to limit the potential for tube denting since alloy 690 tubes can vary from plant-to-plant.
degradation has been associated with dented L
However, there are some similarities particularly with locations.
Specifically, the replacement steam I
respect to susceptibility, or lack of susceptibility, to generators in use at the plants with alloy 690 tubes degradation. In general, the materials and design of have tube supports made from stainless steel rather these steam generators are intended to reduce the than from carbon steel. As discussed in Section potential for tube degradation. Since circumferential 7.2.3, carbon steel tube supports can lead to cracking has occurred in alloy 600 mill-annealed corrosion-induced tube denting. The stainless steel tubes at the expansion transition, in the U-bend tube supports are not expected to corrode and form region of small-radius tubes (i.e., rows 1 and 2), at nonprotective magnetite as does carbon steel (Ref.
i dented locations, and in sleeve joints, design and 109). Furthermore, the design of the tube supports operational features of the steam generators which has been modified to reduce the area of tube to 8-1 NUREG 1604 i
support contact which permits greater secondary 8.2 Justification for Caadan~i Operation water flow past the tube-to-tube support gap. This greater secondary water flow can limit the buildup of he staff evaluated the GL 95-03 responses submitted corrosive deposits which may lead to tube by plant owners with alloy 690 steam generator tubes degradation. h design of these tube supports varies to confirm that the plants could anfely operate until from plant-to-plant, ne designs include a lattice the next scheduled steam generator tube inspection grid system at Millstone 2 (Ref.109) and quatrefoil outage. De staff concluded that all of these units i
broached holes at D.C. Cook 2 (Ref. 88) and North could operate until their next scheduled steam Anna 1 and 2 (Ref.100) and trefoil holes at Summer generator tube mapection. The staff based this (Ref.166).
conclusion on the following factors:
No appreciable corrosion-induced denting has been (1) the scope and results of the prior inspection observed at any of the units with alloy 690 steam including the experience at other similarly a
generator tubes. However minor geometry variations designed us.its (i.e., dents) have been observed at a few plants.
These variations were introduced before'the steam (2) the alloy 690 thermally treated tube material, generators were put into operation.
Geometry which has shown excellent resistance to variations placed in the tubes during the steam primary side cracking on the basis oflaboratory generator manufacturing process could result in areas testing of higher stress which could potentially increase the susceptibility of the tube to degradation (Ref.109),
(3) the design enhancements cited in Section 8.1 At North Anna 1, some dents have been noticed in the steam generator tubes in the tube free span (Ref.
(4) the short amount of time that these steam 100).
generators have been in service j
i The design of the replacement steam generators with 8.3 Tube aan~ tions r
alloy 690 tubes is also intended to reduce the likelihood of cracking in the U bend region of GL 95-03 requested, in part, a safety assessment small-radius tubes (i.e., rows 1 and 2).
b justifying continued operation on the basis of past measures taken at a specific plant to reduce the inspection results and a summary of inspection plans likelihood of U-bend cracking vary. For example, at for the next scheduled steam generator tube Millstone 2, the inner row U bends were designed inspection outage as they pertain to the detection of such that the tubes in rows 1 through 4 have circumferential cracking, h inspection plans were essentially the same bend radius. The licensee for to consist of both an initial scope and sample this plant expects the stresses in these tubes to be expansion criteria. For the units with alloy 690 substantially lower than those present in the inner steam generator tubes, the staff summarized some of row tubes of other steam generator designs (Ref, the information provided by the licensees with respect 132). At D.C. Cook 2 the potential for U-bend to the previous and next inspection for each of the cracking was addressed in the design of the steam areas historically identified in alloy 600 mill-annealed generators by increasing the minimum bend indius tubes as being susceptible to circumferential cracking.
and heat treating the U-bends after bending (Ref. 88).
h designation of ' previous
- refes to inspections At North Anna 1 and 2, the tubes in rows 1 through performed before issuing or resproding to GL 95-03.
8 received a supplemental thermal treatment (stress h designation of "next' (ancor " future") refers to relieving) after bending (Ref.109). This thermal an inspection performed after issuing or responding treatment is intended to reduce the residual stresses in to GL 95-03. b phrase, "if detect", is used to the U-bend region. In addition, as discussed in describe the mspection expansion criteria when a Section 7.2.2, a set of U-bend manufacturing circumferential indication is detected.
In many geometric controls were implemented to reduce the instances, the next (and/or future) inspections have likelihood of cracking in the U-bend region for the already been completed as a result of tne time taken North Anna 1 and 2 steam generators.
to prepare this document for publishing.
i NUREG-1604 8-2
Tables 8-1 through 8-3 provide the scope of the past and future inspections at the expansion transition, in the U-bend of small-radius tubes, and dented locations, respectively for plants with alloy 690 steam guerstor tubes. No sleeves have been installed in any of these units. Tube inspections performed using a technique not capable of reliably detectag circumferentially oriented degradation were recorded as 'None' in Tables 8-1 through 8-3.
The past inspection scope and future inspection plans provided by the licensees were considered in evaluating the acceptability of a licensee's response. Acronyms and abbreviations used in the tables are explained in Appendix C.
The staff has reviewed the submissions provided by the licenaaan with alloy 690 steam generator tubes and has concluded that they contain the information requested in GL 95-03.
General conclusions regarding the responses are discussed in Section 9.
(
i 8-3 NUREG-1604
I l
Table 81 laspections at the expansion transition region in plants with alloy 690 steam generator tubes Plant Espaamon Translaine Expnasion Transition Future Notes Past lampaction Inspection D C. Cook 2 None 20% C5 in HL in 1 SG. If detect, References 88,126, and 127, 100% C5 in affected SO and 20% in remaining SGs. If detect in expanded 20% sample,100% in all SGs.
Indian Point 3 None Iniual scope and expansion criteria per References 167 and 168.
Technical Specificanons with an Appendix 11 cire. cracking technique.
Mdistone 2 None None 0.007 inch daarnetric expansion. References 109, 131, and 132.
Nore Anna 1 None 3 % of all tubes in I of 3 SGs with References 100 and 101.
Appendia H cire. cracking technique.
If detect,6% of all tubes in affected
)
SG. Further expansions based on j
results.
i North Anna 2 None 3% of all tubes in 1 of 3 SGs with Past inspecdon was the preservice inspecdon.
Ap.pendia H circ. cracking technique.
References 100 and 101.
If detect,6% of all tubes m affected SG. Purther espansions based on results.
Summer 100% RPC in ill and CL 100 RPC exams.
Past inspection was the preservice inspecton.
References 166,169, and 170.
1 I
i l
l NUREG-1604 8-4 l
l
8 Table 8-2 Inspecticos la the U-bene region of small-radil tubes in plants with alloy 690 steam generator tubes Plane U-bend Past Wh
. UM Future laspect6en Notes D.C. Cook 2 None None Increased R1 radius and heat treated low row U-bends after bending. References 88,126, and l
127.
Indan Point 3 None initial scope and expanson criteria per References 167 and 168.
Technical Specifications with an Appendas H circ. cracking technique.
Mdistone 2 None None Tubes in R1 through R4 have a radius of 3.9 inches which should reduce suscepubility to cracking. References 109,131, and 132.
North Arum 1 None None Rows I through 8 were stress relieved after bending. References 100 and 101.
North Anna 2 None None Rows 1 through 8 were stress relieved after bending. References 100 and 101.
l 1
Surnmer None 100 RPC exams.
Past inspection was the preservice inspection.
Small radii U-bend tubes were stress relieved during manufacture. References 166.169, and 170.
I l
1 i
l i
8-5 NUREG-1604
Table 8-3 Inspections at dented locations in plants with ll I~~ Pleas a oy 690 steman generator tubes
~
Dested TSP Past '
IP6 Deused TSP Futurelaspecalas~
hC. Cook 2 Neses None R
None References 88,126, and 127.
Indtaskint 3 None Technical Specificatum widt aninnial scope and expansion criteris p
' No voltage W i Mdtstorw 2 Appendin h cire. cracking technique.
w ll be used to determum deras of concern for sample sir.e considerations ~
None Refereacce 167 and 168.
None differential. Dents GSPs, sludge pile) re from 350/130 min if volta freespan are called dings.ge > SV. Dents in North Anna i dentmg has been observed.No corrosion induced None None 131, and 132.
Refenncec 109, Referentes 100 and 101, 1
iNorth Anna 2
-_None--
None References 100 and 101.
Sumrner
~
N/A N/A
- No denti
- however,100'RPC exans will be u-1 perf 169, and 170.at TSP intersections. References 166,~
1 j
NUREG-1604 8-6
Table 8 2 Inspections la the U-bend region of smau-radii tubes in plants with anoy 690 steam generator tubes Phmt U-bend Past lampactico U-bend Future lampacties Notes D.C. Cook 2 None None increased R1 radius and heat ueated low row U-bends after bending. References 88,126, asu!
127.
Trutian Point 3 None Initial scope and capansion crueria per References 167 and 168.
Technical Specificatunis with an Appendix H circ. cracking technique.
Mdistone 2 None None Tubes in R1 through R4 have a radius of 3.9 inches which should reduce susceptibility to cracking. References 109,131. and 132.
North Anna 1 None None Rows 1 through 8 were stress relieved after bending. References 100 and 101.
North Anna 2 None None Rows I through 8 were stress relieved after bending. References 100 and 101.
Summer None 100 RPC exarna.
Past irspection was the preservice inspection.
Srnall radii U-bend tubes were stress relieved during nuriufacture. References 166,169, and 170.
1 l
8-5 NUREG-1604 1
1 I
Tabie 8 3 I --;- "- at dented locations la plaats with ahoy 690 steaan generater tubes Plant Dessed TSP Past Dested TSP Future Ieta=
Neses syi D.C. Cook 2 None None References 88,126, and 127.
Indian Pom 3 None initial scope and expansion criteria per No voltage threshold will be used to determine Technical Specifications with an dents of concern for sample sise considerations Appendia il circ. cracking technique.
References 167 and 163.
Mdistone 2 None None Calibration: 4-100% holes set to 6V on 550 km differential. Dents (T5Ps, sludge pile) reported frorn 550/110 mix if voltage > $V. Denis in freespan are called dings. No corrosion induced denting has been observed. References 109, 131, and 132.
North Anna i None None References 100 and 101.
North Anna 2 Nona None References 100 and 101.
l Surnrner N/A N/A No denting; however,100 RPC exams will be performed at TSP truersections References 166, 169, and 170.
4 j
i I
i NUREG-1604 8-6 i
i l
1
9 CONCLUSIONS Mill-annealed alloy 600 steam generator tubes are thermally treated alloy 600 tubes. He nature of susceptible to circumferential cracking. For B&W these axial indications is uncertain since these tubes plants (all of which have alloy 600 mill-annealed were not removed for destructive examination. Dese tubes), circumferential indications have been observed results, however, indicate that thermally treated alloy in the tubesheet ares (associated with intergranular 600 tubes are potentially susceptible to cracking, attack), in the lane / wedge region, and at dented
. locations. Although service-induced circumferential In addition to the locations mentioned ' above, indications have not been observed at the expansion circumferential indications associated with sleeve transition in B&W plants, the expansion transition is joints have occurred in a number of sleeve designs.
considered to be potentially susceptible to this form Service-induced eircumferential indications have been of degradation since axially oriented SCC has been observed in the joints of B&W kinetically welded observed at this location. For CE plants with alloy sleeves and Westinghouse HEJ sleeves. Non-service j
600 mill-annealed tubes, circumferential indications induced circumferential indications have been have been detected on both the hot-and cold-leg sides detected in CE TIG-welded sleeve joints.
In of the steam generator at the expansion transition, at addition, circumferential indications have been
)
dented locations, and at non-dented drilled hole tube observed in B&W mechanical sleeve joints although support plate locations (primarily in the chord no tube pulls have been performed to confirm the region). In addition,circumferentialindicationswere nature of these indications.
observed in the pre-replacement Millstone 2 and Palisades steam generators at various locations in the On the basis of the staft's review of the GL 95-03 upper portion of the tube bundle on both the hot-and responses and from discussions with various PWR cold leg sides of the steam generator. Denting and licensees during public meetings, inspections, and tube bowing are believed to have contributed to these telephone conversations in which steam generator indications. The circumferential cracking operating mapection scope and results were discussed, the NRC experience for Westinghouse plants with alloy 600 identified potential weaknesses in a few steam mill-annealed tubes is summarized in Section 7.2.6.
generator tube integrity programs.
The staff The results from the B&W, CE, and Westinghouse considered these weaknesses along with other factors plants indicate that alloy 600 mill-annealed tubes are (e.g., other elements of a licensee's tube integrity potentially susceptible to circumferential cracking at program, pertinent operating experience, etc.) in a variety of locations on both the hot-and cold-leg determining the acceptability of a licensee's response sides of the steam generator.
to GL 95-03. Ahhough weaknesses were identified in several licensee's responses to GL 95-03, the staff The operating experience for steam generators with believes these weaknesses have been and will thermally treated tubes has been relatively good when continue to be addressed in the short-term through compared to steam generators with mill-annealed ongoing regulatory oversight performed in tubes. However, the length of service for some of headquarters and in the regions. In the long-term, the steam generators with thermally treated tubes is revisions of the regulatory framework, which are limited when compared to the steam generators with currently being developed, will provide more direct alloy 600 mill-annealed tubes. No circumferential oversight of licensee programs to ensure that steam indications have been observed in thermally treated generator tube integrity is maintained, alloy 690 tubes, although a few circumferential indications have recently been identified in the alloy 600 thermally treated tubes at Callaway. These indications were located at the expansion transition.
No tube pulls were performed to confirm the nature of these indications.-
In addition to these circumferential indications, axial indications have been observed in the U-bend region and at the expansion transition region of a limited number of 91 NUREG-1604
i APPENDIX A: STEAM GENERATOR TUBE SLEEVES j
%e installation of steam generator tube plugs For the steam generators in service as of December removes the heat transfer surface of the plugged tube 1996, the majority of the inservice sleeves (i.e.,
i from service and leads to a reduction in the primary non-plugged sleeves) are one of five major designs:
l coolant flow available for core cooling. To prolong Westinghouse hybrid expansion joint (HEJ) sleeves, the life of severely degraded steam generator tubes, B&W kinetically welded sleeves, B&W mechanical some utilities, with prior NRC approval, have sleeves, CE TIG-welded sleeves, and Westinghouse i
repaired defective tubes by sleeving since steam laser-welded sleeves. With respect to circumferential i
generator tube sleeves do not gastly affect the heat cracking, one major difference between the sleeves.
transfer capability of the tube being sleeved and after (besides the material of the sleeve) is the method of l
sleeving, the repaired tube may remain in service. A attaching the sleeve to the parent tube. Each of the i
large number of sleeves can be installed without sleeves in use are described below along with significantly affecting primary flow rate, relevant operating experience.
4 I
The tube sleeving procedure involves inserting a tube ne Westinghouse HET sleeves have upperjoints that j
of smaller diameter and length (a sleeve) inside the were formed by first hydraulically expanding a j
tube to be repaired. %e sleeve is positioned to span portion of the sleeve into the parent tube and then by the degraded portion of the original tube (i.e., the rnechanically roll expanding the central portion of this parent tube) and the ends of the sleeve are secured to hydraulically expanded region. Thisjoint is typically 4
the parent tube forming a new pressure boundary and referred to as a HEJ and is depicted in Figure A-1.
structural element between the attachment points. As The intent of the initial hydraulic expansion is to l
a result, there are at least two joints in a sleeve: one lessen the residual stress levels in the joint ares at the top of the sleeve and the other at the bottom of (Ref. 90). Although the lowerjoint of an HEJ sleeve the sleeve. Sleeves vary in length and are typically consists of a hydraulic expansion and hardroll as attached to the parent tube by a mechanical seal (e.g.,
well, it is not referred to as an HEJ joint. An hydraulic expansion, roll expansion, and/or explosive illustration of the entire Westinghouse HEJ sleeve is expansion) or weld (Ref.171). After installation of depicted in Figure A-2. Westinghouse HEJ sleeves
]
the sleeve, stress relief of thejoints can be performed are currently (1995) in service at Kewaunee, Point to relieve residual stresses. In general, most sleeves Beach 1, D.C. Cook 1, and Giana. The Point Beach installed recently have been stress relieved.
A 1 and Ginna steam generators were replaced in 1996 variety of sleeve designs exist, and the names for the (Ref.171).
various types of sleeves typically reflect the method by which one or more of the sleeve ends is secured Circumferential crack-like indications in the parent to the parent tube, tube associated with the upper joint of Westinghouse HEJ sleeves were first identified at Kewaunee in Sleeving repairs to restore primary coolant boundary 1994. Additional indications were detected in a integrity have been performed at several plants.
subsequent inspection outage (i.e., 1995).
The Rese repairs are typically performed on the straight majo-ity of the indications detected were in the lower portion of tubing degraded by such mechanisms as hardroll transition of the upper joint (i.e., the HEJ) wastage, intergranular attack, and SCC. Severely (refer to Figure A-1).
As a result of these dented locations have generally not been sleeved.
indications, the licensee for Kewaunee removed 3 Tubesheet, expansion transition zone, and tube sleeved tubes for destructive exanunation. As a support plate sleeve designs exist. Currently, most result of the destructive examination, the licensee for sleeves are hydraulically expanded into the tube and Kewaunee concluded that the eddy current indications then welded (laser or TIG-welded) to ensure were attributable to circumferentially oriented, inner stmetural and leakage integrity. Sleeves made from diameter initiated, intergranular stress corrosion alloys 600 and 690 have been used throughout the cracks on the parent tube. The crack network was industry. Currently, the material of choice for segmented and the crack initiation sites were scattered sleeves is alloy 690.
in elevation (i.e., non-coplanar). No indication of A-1 NUREG-1604
4 1
1 corrosion or cracking were detected on the sleeves tube and 50-percent through the wall for the (Ref.172).
ramaimag 90 degrees. De crack had initiated from the inside of the parent tube and was characteristic of The B&W kinetically welded sleeve design generally PWSCC (Refs.11 and 174). hee tubes had been
{
consists of at least one kinetic expansion joint and stress relieved; however, the stress relief temperature may have a mechanical joint depending on the did not assure a long service-life for this tube, installation location. In general, tube support plate Additional circumferential indications were detected sleeves (Figure A-3) have two kinetic expansion at McGuire ! in January 1994, as a result of a tube joints whereas tubesheet sleeves (Figure A 4) can leak event. His tube leek event, and uncertainty in i
either have one kinetic expansion joint and one the root cause of the sleeve failures, led the licensee mechanicaljoint or two kinetic expansionjoints. The to preventively plug the remaining sleeved tubes in free span joints of both the tubesheet and tube support this unit (Ref.118). Indications have recently been plate sleeves are performed by a kinetic welding detected in the ANO-2 sleeves. b nature of these i
j process. In this process, the sleeve is expanded into indications was described as non-quantifiable parent l
the parent tube by detonating a kinetic weld device, tube indications (Ref. 62).
l h kinetically expanded region is narrow (axially) and it results in the outer sleeve wall being fused to De B&W mechanical sleeve design has only been the inner tube wall. Dese kinetically expanded free used in B&W OTSGs and has three mechanical joints span joints are stress relieved to improve corrosion as depicted in Figure A 5.
The sleeve joints are characteristics (i.e., reduce residual stresses to make made by roller expanding the ends of the sleeve into material less susceptible to corrosion). b lower the tube. The upper joint is typically located in the joint (i.e., tubesheet joint), if not kinetically UTS region. He other twojoints are located in the expanded, can be formed by mechanically rolling the tube free span region and are in close proximity to sleeve into the tube (Refs.173 and 174). B&W each other and to the lower end of the sleeve. All kinetically welded sleeves are currently (1995) in B&W OTSGs have B&W mechanical sleeves installed service at ANO-2. These sleeves had been used in (Refs 43 and 177). Dese sleeves are typically several other plants; however, in these instances, the located in the lane / wedge region of the steam sleeves were either removed from service by generator, which is discussed in Section 5.2.1.
plugging (e.g., Catawba 1 and McGuire 1 and 2), the steam generators were replaced (e.g., Summer),
Circumferential indications have recently been and/or the plant was shut down permanently (e.g.,
detected in the joints of B&W mechanical sleeves at Trojan).
ANO-1 although no tube pulls have been performed to identify the nature of the degradation. The Degradation of B&W kinetically expanded joints has licensee believes the majority of the indications been observed at a few plants. As a result of a tube detected are associated with the parent tube rather leak that resulted in a plant shutdown, multiple than the sleeve itself. The degradation has been indications in the parent tube material in the area of observed at both the upperjoint (within the tubesheet) the lower sleeve-to-tube kinetic weld were observed and the lowerjoints (in the tube free span) (Ref. 42).
in a B&W kinetically welded tube support plate sleeve at Trojan in November 1992. De indications There are 3 basic types of CE TIG-welded sleeves spanned a region estimated to exceed 180 degrees of that may be installed in various combinations within the tube circumference. The cause of the leak was a steam generator tube. Each of the sleeve types determined to be the failure to properly stress relieve includes a chamfer at both ends to prevent hang-up of the tube after the. sleeve was kinetically welded equipment used to install or inspect the sleeve or (Ref.175). Circumferentialindicationswere detected tube. The three types of sleeve are the expansion in B&W kinetically welded sleeves in 1993 at transition zone (or roll transition zone) sleeve, the McGuire 1 as a result of a tube leak (Ref.176). A tubesheet sleeve, and the tube support plate sleeve.
tube pull from the McGuire I confirmed a circumferential crack in the parent tube approximately h CE TIG-welded expansion transition zone sleeve 2.5 mm (0.1 inch) above the apex of the upper weld spans the expansion transition zone at the top of the thatjoined the tube and sleeve. b circumferential tubesheet to a maximum height of 10 cm (4 inches) crack was through-wall for 270 degrees around the above the first support. This sleeve configuration is NUREG-1604 A-2
depicted in Figure A-6. This sleeve is welded near tube surface prevents wetting of the joint faying the top end of the sleeve, which is above the surfaces during welding and blocks coalescence.
secondary face of the tubesheet, and rolled near the Sleeve OD suckback is a term given to a rounded 1
bottom end of the sleeve, wtuch is near the neutral cavity formed on the edge of the weld. The rounded axis of the tubesheet. Before the welding and rolling, nature of this type of discontinuity indicates that it is the sleeve is hydraulically exp=adad into the sleeve to due to the evolution of a gas within the joint.
hold the sleeve in place. The sleeve has a band of nickel at one end to improve sealing of the sleeve The licensee attributed both the incomplete fusion and when the lower end is hardrolled, and it has a band OD suckback to an anomaly associated with the of chromium oxide, which has a rough surface, to sleeve installation process. The licensee attributed provide a strong mechanical joint.
both forms of tacomplete fusion to insufficient removal of the tube surface oxide from the inside b CE TIG-welded tubanhaat sleeve, illustrated in diameter of the parent tube before sleeve installation Figure A-7, spans the tubesheet and a portion of the (i.e., improper cleaning of the tube before sleeve tube to a maximum height of approximately 23 cm (9 installation). The sleeve OD suckback occurred inches) above the tubesheet. b sleeve is welded predominately on the upper (non-pressure boundary near each end. Before welding the upper end of the portion of the weld, and when it did occur below the sleeve, the upper end of the sleeve is hydraulically weld it was localized and limited to less than expanded into the tube. The lower end of the sleeve 15-percent through the sleeve wall. The licensee is tapered so as to limit the insertion of the sleeve to indicated that the source of the gas forming the OD the proper elevation during installation, to suckback regions could be a result of contamination temporarily hold the sleeve in place, and to provide of the tube surface or from moisture behind the a tight contact with the tube for welding. The lower sleeve. No service-induced propagation, including end of the sleeve is positioned flush with the bottom environmental degradation, of any type was observed of the steam generator tube.
during the destructive examinations.
Most all inclusion termination points, many of which were As its name implies, the CE TIG-welded tube support rounded pores, were oxide filled. In addition, the plate sleeve, illustrated in Figure A-8, spans a tube licensees comparison of previous outage eddy current support plate, b upper and lower ends of the volumetric and eircumferential signals exhibited no or sleeve are hydraulically expanded into the tube to very minor change (Ref.179). In summary, the hold the sleeve in place for welding and to provide licensee concluded that the eddy current indications the sleeve to tube fit-up necessary for welding. This observed in these sleeves were fabrication related sleeve is welded near each end (Ref.178).
rather than service-induced. The NRC staff agreed that the indications in the CE sleeves appeared to be CE TIG-welded sleeves are currently (1996) in installation induced rather than service-induced service at a number of plants including Prairie Island (Ref.180). Similar eddy current indications to those 1, Zion 2. ANO-2, Kewaunee, and Byron 1. As a observed at Prairie Island I have been observed at result of identifying circumferential and volumetric other plants that have installed CE sleeves.
indications in the sleeve joints, the licensee for Prairie Island I removed several sleeve specimens Post weld heat treatment can be performed on these from their steam generators for destructive sleeves to reduce the residual stresses present in the examination in early 1996. From the metallurgical sleeve / tube joint without significantly affecting the examinations, the licensee concluded that the eddy microstructure of the materials. Reducing residual current indications were a result of the radial stresses is desirable since SCC is dependent to a large component of either or both of two weld conditions extent on these stresses; therefore, a reduction in the termed incomplete fusion and sleeve outside diameter residual stress level in the sleeve welds will enhance (OD) suckback. Incomplete fusion manifests itselfin the corrosion resistance of thejoints.
two forms: (1) the formation of refractory oxides in the weld nugget creating laminar, non-linear, Like the CE TIG-welded sleeve, there are several inclusions emanating from the intersection of the basic types of Westinghouse laser-welded sleeves that sleeve and tube faying surfaces; and (2) a lack of may be installed in various combinations within a fusion flaw where the indigenous oxide layer on the cteam f;entor tube. The types of Westinghouse A-3 NUREG-1604
8' laser-welded sleeves include the fulllength tube 6heet developed to reduce the residual stresses present in sleeve (FLTS), the elevated tubesheet sleeve (ETS),
the sleeve /tubejoint without significantly affecting the and the tube support sleeve (TSS).
- nicrostructure of the material. Reducing residual stresses is desirable since SCC is dependent to a large b FLTS, illustrated in Figure A-9, extends from extent on residual stresses; therefore, a reduction in the tubenheet primary face to above the tubesheet the residual stress level in the laser sleeve welds will secondary face. b lower joint of this sleeve is anhance the corrosion resistance of the weldedjoints formed at the bottom of the tubesheet and thisjoint is (Ref.181).
made by first hydraulically expanding the sleeve into the parent tube and then rolling a section of the Westinghouse laser-welded sleeves have been used at sleeve within the hydraulically expanded region into a number of plants including Farley I and 2, Maine the parent tube. b upper joint of the sleeve is Yankee, and Byron 1. b sleeves at Farley I and 2 made by first hydraulically expanding a portion of the have been in service the longest, they were installed sleeve into the tube to achieve the proper fit-up in 1992 (Ref. 90). To date, the operating experience geometry for welding and then by performing an with Westinghouse laser-welded sleeves has been autogenous weld between the sleeve and tube using good (although limited) with no reports of the laser welding process. Thisjoint is located above degradation to the staff's knowledge.
the tubesheet and occurs in the tube free span. For the lower joint, a seal weld can be added, as an Other sleeve designs l' ave been used throughout the
- option, industry; however, che extent to which they are i
currently being naed is minimal. For example, in the The ETS, illustrated in Figure A-10, extends over pre-replacement Ginna steam generators, B&W approximately one-third of the tube length within the brazed sleevra and B&W explosive sleeves were used tubesheet and is joined to the tube approximately 36 (Ref.128). In addition, at Oconee 1, there are a few cm (14 inches) above the tubesheet bottom (i.e.,
(approxinu.tely 4) sleeves which have been lower joint) and above the tubesheet secondary face hydraulically expanded into the tube (Ref. 38).
(i.e., upperjoint). This type of sleeve allows greater radial coverage of the bundle (i.e., closer to the In addition to these other sleeve types, other forms of bundle periphery) than the FLTS since it is smaller sleeve degradation have occurred; however, the and is less likely to interfere with the primary number of reported occurrences has been minimal.
channel head during installation. The upper and For example, bulging of the sleeve has been observed lower joints of the ETS are identical to the joints in in at least one plant (e.g., Farley 2) (Ref.146).
the FLTS (i.e., the upper joint is hydraulically expanded and then welded; the lower joint is In summary, most types of sleeves currently in hydraulically expanded then rolled and can be service have exhibited either fabrication induced or laser-welded as an option). b ETS is similar to the service-induced degradation. This degradation is FLTS in that it is designed to address degradation at normally associated with the parent tube and is or near the top of the tubesheet; however, they are located at the point of attachment of the sleeve to the dissimilar in that the ETS is not designed to address tube (i.e., the sleeve joint). Due, in part, to the degradation in the remainder of the tube within the relatively good operating experience to date, most tubesheet.
sleeves currently (1996) being installed are welded (laser-or TIG-welded) to the parent tube. After the The tube support plate sleeve, illustrated in Figure welding, the resultantjoint is frequently heat treated A II, can be used to span degradation at the tube to reduce the residual stresses in thejoint. Currently, supports (i.e., flow distribution baffle plates, drilled the material of choice for sleeves is alloy 690.
plates, and eggerates) or in the tube free span. Each end of the sleeve has a hydraulic expansion region within which the weld is placed.
Post weld heat treatment can be performed for each of the three Westinghouse laser-welded sleeves discussed above. Post weld heat treatment was NUREG-1604 A-4
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NUREG-1604 A-10
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APPENDIX B: PLANT LISTINGS Information related to the steam generators at plants in operation in June 1995 is provided in Tablee B-1 through B-6. The following information is included in these tables.
(1) Table B-1: Alphabetical listing of plants indicating the steam generator vendor and model number as well as the steam generator tube material, heat treatment, and outside diameter.
(2) Table B-2: Alphabetical listing of the plants on the basis of tube material and heat treatment (e.g., plants with alloy 600 mill-annealed tubes).
(3) Table B-3: Alphabetical listing of the plants based on nuclear steam supply system vendor (e.g., B&W plants).
(4) Table B-4: Alphabetical listing of Westinghouse plants on the basis of tube material and expansion method (e.g., partial-depth hardroll plants with alloy 600 tubes).
(5) Table B-5: Alphabetical listing of Westinghouse plants on the basis of steam generator model.
(6) Table B-6: Listing of plants that have replaced their steam generators indicating the number of steam generators and the pre-and post-replacement steam generator manufacturer and model number. The table is ordered based on the approximate completion date for the steam generator replacement project.
B-1 NUREG-1604
Table B-1 1%it Listing by Name Plant Narne Vendor */Model #
Tube Material lleat Treatmert Outside Diarneter (inch)
Arkansas Nuc! car One - 1 B&W 600 MA - sensitized 0.625 Arkansas Nuclear One - 2 CE 600 MA 0.750 Beaver Valley 1 Westinghouse /51 600 MA 0.875 Beaver Valley 2 Westinghouse /51M 600 MA 0.875 Braidwood I Westinghouse /D4 600 MA 0.750 i
Braidwood 2 Westinghouse /D5 600 "IT 0.750 i
Byron 1 Westinghouse /D4 600 MA 0.750 Byron 2 Westinghouse /D5 600 TT 0.750 Callaway Westinghouae/F 600 MAnTS 0.6875 Calvert Cliffs 1 CE 600 MA 0.750 Calvert Cliffs 2 CE 600 MA 0.750 Catawba l' Westinghouse /D3 600 MA 0.750 Catawba 2 Westinghouse /D5 600 1T 0.750 l
Comanche Peak 1 Westinghouse /D4 600 MA 0.750 Conunche Peak 2 Westinghouse /D5 600 TT 0.73 0 Crystal River 3 B&W 600 M A - sensitized 0.625 Davis-Besse B&W 600 MA - sensitized 0.625 D.C. Cook I Westinghouse /51 600 MA 0.875 D.C. Cook 2' Wesiinghouse/54F 690 1T 0.875 Diablo Canyon i Westinghouse /51 600 MA 0.875 Diablo Canyon 2 Westinghouse /51 600 MA 0.875 Farley 1 Westinghouse /51 600 MA 0.875 Farley 2 Westinghouse /51 600 MA 0.875 Fort Calhoun CE 600 MA 0.750 Ginna' Westinghouse /44 600 MA 0.875 Haddam Neck Westinghouse /27 600 MA 0.750 Indian Point 2 Westinghouse /44 600 MA 0.875 Indian Point 3*
Westinghouse /44F 690 1T 0.875 Kewaunee Westinghouse /51 600 MA 0.875 Maine Yankee CE 600 MA 0.750 McGuire !
Westinghouse /D2 600 MA 0.750 McGuire 2 Westinghouse /D3 600 MA 0.750 Millstone 2*
BWC 690 TT 0.750 Millstone 3 Westinghouse /F 600 TT 0.6875 North Anna l' Westinghouse /54F 690 TT 0.875 North Anna 2*
Westinghouse /54F 600 TT 0.875 Oconee 1 B&W 600 MA - sensitized 0.625 Ocones 2 B&W 600 MA - sensitized 0.625 Oconee3 B&W 600 MA - sensitized 0.625 i
Palisadet CE 600 MA 0.750 l
l NUREG-1604 B-2
Table B-1 Plant Listing by Nane (continued)
Plant Name Vendd/Model #
Tube Meterial Heat Treatmest Outside Diameter (inch)
Pelo Wade 1 CE 600 MA 0.750 Pelo Wade 2 CE 600 MA 0.750 Pelo Wade 3 CE 600 MA 0.750 Point Beach l' W '-f =_m144F 600 TT 0.875 Point Beach 28 Westinghouse /44 600 MA 0.875 Prairie 1 eland I Weetanshouse/51 600 MA 0.875 Prairie taland 2 Westinghouse /51 600 MA 0.875 Robinson 2*
W '-f -144F 600 TT 0.875 Salem 1 W-f -----151 600 MA 0.875 Salem 2 Westinghouse /51 600 MA 0.875 San Onot.s 2 CE 600 MA 0.750 San Onofre 3 CE 600 MA 0.750 Seabrook1 W--
f--/F 600 TT 0.6875 SequoyahI W ^ f - "$1 600 MA 0.875 Sequoyah 2 Westinghouse /51 600 MA 0.875 Shearon Harris 1 Westarsboum/D4 600 MA 0.750 South Texas Project I W-f =.m/E 600 MA 0.750 South Toxas Project 2 Weatinghmm/E 600 MA 0.750 St. Lucie 1 CE 600 MA 0.750 St. Lucie 2 CE 600 MA 0.750 Summer
- Westinghoum/deka 75 690 TT 0.6875 Surry 1*
W-f =2s/51F 600 1T 0.875 Surry 2*
Westinghoues/51F 600 TT 0.875 nroe Mile taland i B&W 600 MA - eenaitized 0.625 Turkey Poins 3*
Westanghouse/44F 600 TT 0.875 Turkey Point 4' Westmshouse/44F 600 TT 0.875 Vogtle 1 Westinghouse /F 600 TT 0.6875 Vogtle 2 Westinghouse /F 600 TT 0.6875 Waterford 3 CE 600 MA 0.750 Wette Bar 1 Westinghouse /D3 600 MA 0.750 Wolf Creek l Westinghouse /F 600 TT 0.6875 f
Zion 1 Westinghouse /51 600 MA 0.875 Zion 2 Westinghoum/51 600 MA 0.875 l
i B&W = Babcock and Wilcox; CE = Combustion Engineering 8
MA = mill-enneeled;TT = thermelly treated De tubes la rowe I through 10 in each steam generator are thermelly treated. He remainder are mill-annealed.
Replacement steam generatore l
s Steam generators were replaced during 1996.
B-3 NUREG-1604 I
[.
Table B-2 Plant Listing by Tube Material ABoy 600 MB aamenled l
Arkansas Nuclear One - 1 Oconee2 Arkansas Nuclear One 2 Oconee 3 Beaver Valley 1 Palisades Beaver Valley 2 Palo Verde 1 Braidwood 1 Palo Venie 2 Byron 1 Palo Verde 3 8
Ca!!away Point Beach 2' Calvert Cliffa 1 Prairie Island 1 Calvert Cliffs 2 Prairie Island 2 Catawba 18 Salem 1 Comanche Peak 1 Salem 2 Crystal River 3 San Onofre 2 Davis-Besse San Onofre 3 D.C. Cook 1 Sequoyah1 Diablo Canyon i Sequoyah 2 Diablo Canyon 2 Shearon Harris 1 Farley 1 South Texas Project 1 3
Farley 2 South Texas Project 2 Fort Calhoun St. Lucie i Ginna' St. Lucie 2 Haddam Neck Three Mile Island 1 Indian Point 2 Waterford 3 Kcwaunce Watta Bar 1 Maine Yankee Wolf Creek 1 McGuire 1 Zion 1 McGuire 2 Zion 2 Oconee!
ABoy 600 Thennany Treeted Braidwood 2 Seabrooki Byron 2 Surry 1 Callaway Surry 2 Catawba 2 Turkey Point 3 Comanche Peak 2 Turkey Point 4 Mi!! stone 3 Vogtle i Poirs Beach 1 Vogtle 2 Robinson 2 Wolf Creek i ABoy 690 2hermaDy Treated' D.C. Cook 2 North Anna l Indian Point 3 North Anna 2 Millstone 2 Summer
- Steam generators were replaced during 1996. Replacement steam generators have alloy 690 thermally treated tubes.
8 Rows 11 and higher are mill-annealed. Rows I through 10 are thermally treated.
- Catawbs I, Ginna, and Point Beach 2 replaced their original steam generators in 1996 with steam generators that have alloy 690 thermally treated tubes.
NUREG-1604 B-4
Table B-3 Plant Listing by Vendor Babcock and Wilces Plants Arkanssa Nuclear One 1 Oconee2 Crystal River 3 Oconee3 Davie-Deane Three Mile Island i Oconee1 r h n Fage ing ytants Arkansas Nuclear One - 2 Palo Vesde 2 Calvert CliNs 1 Palo Verde 3 Calvert Cliffs 2 San Onofs 2 Fort Calhoun San Onofre 3 Maine Yankee St. Lucie 1 Millstone 2' St. Imcio 2 Paliandes Waterford 3 Palo Verde 1 Wm.el=gha.== Plands Beaver Valley 1 North Anna 2 Beaver Valley 2 Point Beach i Braidwood 1 Point Beach 2 Braidwood 2 Prairie Island i Byron 1 Prairie tal==d 2 Byron 2 Robinson 2 Callaway Salem 1 Catawba 18 Salem 2 Catawba 2 Seabrook1 Comanche Peak i SequoyahI Comanche Peak 2 Sequoyah 2 D.C. Cook i Shearon Harria 1 D.C. Cook 2 South Tenas Projsct i Diablo Canyon i South Texas Project 2 Diablo Canyon 2 Sumnwr Farley 1 Surry i Farley 2 Surry 2 Turkey Poid 3 Ginna8 Haddam Neck Turkey Point 4 Indian Point 2 Vogtle 1 Indian Point 3 Vogtle 2 Kawaunee Watta Bar 1 McGuin 1 Wolf Creek 1 McGuins 2 Zion 1 Millstone 3 Zion 2 North Anna 1 8 Millstone 2 la a CE plant with replacement steam generators fabricated by B&W Canada.
8 Catawba 1 and Ginna are Westmshouse planta which replaced their original steam generators in 1996 with steam generators fabricated by B&W Canada.
B.5 NUREG 1604
i Table B-4 Westingbouse Plant Listing by Tube Expansion Type and Material EXPANSION TYPE /TLDE PLANT NAME MATERIAL Partial-Depsk Herdroll-D.C. Cook i Point Beach 2' Alloy 600 Ginna' Prairie taland 1 Haddam Neck Prairie I= tad 2 Indian Point 2 Zion 1 Kewaunee Zion 2 Full-Depth Hardmil -
Beaver Volley 2 McGuire 1 Alloy 600 Braidwood 1 MeOuire 2 Byron i Shearon Harris I Catawba l' South Texas Project I Comanche Peak 18 Watta Bar 1 Farley 2 WEXs r.a Alloy 600 Beaver Valley i Salem 1 Comanche Peak 18 Salem 2 Diablo Canyon i Sequoyahi Diablo Canyon 2 Sequoyab 2 Farley i Full-Depth Hydraulic -
Braidwood 2 South Texas Project 2 Alloy 600 Byron 2 Surry 1 Callaway Surry 2 Catawba 2 Turkey Point 3 Comanche Peak 2 Turkey Poird 4 Millstone 3 Vogtle i Point Beach I Vogtie 2 Robinson 2 Wolf Creek 1 Seabrooki Full Depth Hydraulic -
D.C. Cook 2 North Anna 2 Alloy 690' Indian Point 3 Summer North Anna !
8 Catawba 1, Ginna, and Point Beach 2 replaced their odginal steam generators la 1996 with steam gensretors that have alloy 690 thermally treated tubee that have been hydraulically expanded the fbli length of the tubesheet, 8
89.5-percent of the tubes have fbil depth hardroll expansion transitions (14,473 tubes).
i The remaining tubes have WEXTEX transitions (3839 tubes).
NUREG-1604 B-6
Table B-5 Westinghouse Plant Listing by Steam Generator Model STEAM GENERATOR PLANT NAME MODEL Westsaghouse anodal 27 Haddam Neck W'f
- model 44 Gians' Point Beach 2' Indian Point 2 W-
- andel51 Beaver Vaney 1 Prairie laland 2 D.C. Cook 1 Salem i Diablo Canyon 1 Salem 2 Diablo Canyon 2 Sequoyah!
Farley I Sequoyah 2 Farley 2 Zion i Kewaunse Zion 2 Prairie Island I W
.'== anodel SIM Beaver Vauey 2 Westinghoum anodel D2 McGuire 1 Womarh=== model D3 Catawba l' Watta Bar 1 McGuire 2 Westinghouse model D4 Braidwood 1 Comanche Peak I Byron i Shearon Harris I Westinghouse model D5 Braidwood 2 Catawba 2 Byron 2 Comanche Peak 2 Westinghouse model E South Texas Project 1 South Texas Project 2 Westanghouse model F Calloway Vogtle ! -
Millstone 3 Vogtle 2 Seabrook 1 Wolf Creek I W:'1== niodel 44F Indian Point 3 Turkey Point 3 Point Beach I Turkey Point 4 Robinson 2 W-f ---- model 51F Sc.<y t Surry 2 Westinghouse model 54F D.C. Cook 2 North Anna 2 Nosth Anna 1 Westeghouse deka 75 Sumawr Westinghouse deha 47 Point Beach 2' B&W Canada Ginna' Catawba l' Ginns replaced their original Wutinghouse model 44 steam generators in 1996 with steam generators fabricated by B&W Canada.
8 Point Beach 2 replaced their original Westinghouse model 44 steam generators in 1996 with Westinghouse deka 47 steam generators.
Catawba i replaced their original Westinghouse model D3 steam generators in 1996 with steam generators fabricated by B&W Canada B7 NUREG-1604
Table B-6 Plants with Replacement Steam Generators (December 1996)
No. of Steam SG Manufacturer /Model' Approximate Plant Name Generators Completion Original New p.g.
Surry 2 3
W/51 W/51F 9/80 Surry 1 3
W/51 W/51F 7/81 Turkey Point 3 3
W/44 W/44F 4/82 Turkey Point 4 3
W/44 W/44F 5/83 Point Beach 1 2
W/44 W/44F 3/84 Robinson 2 3
W/44 W/44F 10/84 D.C. Cook 2 4
W/51 W/54F 3/89 Indian Point 3 4
W/44 W/44F 6/89 Palisades 2
CE CE 3/91 Millstone 2 2
CE BWC 1/93 North Anna 1 3
W/51 W/54F 4/93 Sununer 3
W/D3 W/D75 12/94 North Anna 2 3
W/51 W/54F 5/95 Ginna 2
W/44 BWC 6/%
Catawba 1 4
W/D3 BWC 9/96 Point Beach 2 2
W/44 W/D47 On-going 8
W-Westir.ghouse; CE = Combustion Engineering; BWC = Babcock and Wilcox, Canada; D75 = delta 75; D47 = delta 47 NUREG-1604 B-8
APPENDIX C: ACRONYMS ANO Arkan=== Nuclear One App. H Appendix H of Reference 122 AVB anti-vibration bar B&W Babcock and Wilcox BWC Babcock and Wilcox, c=a d=
CL cold leg cm centimeter CS Cecco 5 CE Combustion RapHog CFR Code of Federal Regulations cire circumferential ECT eddy current testing EOC end-of<ycle EPRI Electric Power Research Institute ETS elevated tubesheet sleeve FLTS full-length tubesheet sleeve GL generic letter HEJ hybrid expansionjoint HL hot leg l
ID inside diameter IGA intergranular attack IGSCC intergranular stress corrosion cracking IN information notice kHz kilohertz LTS lower tubesheet LTSF lower tubesheet secondary face MA mill-annealed MAI multiple axial indication mm millimeter NRC U.S. Nuclear Regulatory Commission
'ODSCC outside diamater strees corrosion crackmg OTSO once-through steam generator r
REG-1 W C-1 i
i i
t
APPENDIX C: A CRONYM.S (cont'd)
PP plus-point psi pounds per square inch PVNGS Palo Verde Nuclear Generating Station PWR pressunzed-water reactor PWSCC primary water stress corrosion cracking RAI request for additional information RPC rotating pancake coil RSG recirculating steam generator T.x row "x" SAI single axial indication SCC stress corrosion cracking SG steam generator SGMP Steam Generator Strategic Management Program TIG tungsten inert gas TMI Three Mile Island TSP tube support plate TSS tube support sleeve TT thermaPy treated TTS top of tubesheet UBHT U-bend heat treatment UTS upper tubesheet UTSF upper tubeabeet secondary face V
volt WEXTEX Westinghouse Explosive Tube Expansion WOG Westinghouse Owners Group NUREG-1604 C-2
APPENDIX D: REFERENCES 1.
U.S. NRC Generic letter 95-05, " Voltage-Based Repair Criteria for WWhal== Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Creding," August 3,1995.
2.
U.S. NRC Generic letter 95-03, 'Circumferential Cracking of Steam Omerator Tubes," April 28,1995.
3.
U.S. NRC Generic Imeter 91-04, " Changes in Technical Specification Surveillance Intervals to Acco==~'-ea a 24-Month Fuel Cycle," April 2,1991.
4.
U.S. NRC Information Notice %-38, "Results of Steam Generator Tube Examinations," June 21,1996.
5.
U.S. NRC Information Notice %-09, Supplement 1, " Damage in Foreign Steam Generator Inte,rnals,*
July 10,1996.
6.
U.S. NRC Information Notice 96-09, " Damage in Foreign Steam Generator Internals," February 12, 1996.
7.
U.S. NRC Information Notice 95-40, " Supplemental InfounsLion L.: Generic letter 95-03,
'Circumferential Cracking of Steam Generator Tubes'," September 20,1995.
8.
U.S. NRC Information Notice 94-88, " Inservice Inspection Deficiencies Result in Severely Degraded Steam Generator Tubes," December 23,1994.
9.
U.S. NRC Information Notice 94-62, " Operational Experience on Steam Generator Tube Leaks and Tube Ruptures," August 30,1994.
10.
U.S. NRC Information Notice 94-43, " Determination of Primary-to-Secondary Steam Generator leak Rate," June 10, 1994.
11.
U.S. NRC Information Notice 94-05, " Potential Failure of Steam Generator Tubes Sleeved With Kinetically Welded Sleeves," January 19, 1994.
l 12.
U.S. NRC Information Notice 93-56, ' Weaknesses in Emergency Operating Procedures Found as a Result of Steam Generator Tube Rupture,' July 22,1993.
13.
U.S. NRC Information Notice 93-52, " Draft NUREG-1477, ' Voltage-Based Interim Plugging Criteria for Steam Generator Tubes'," July 14, 1993.
J 14.
U.S. NRC Information Notice 92-80, " Operation With Steam Generator Tubes Seriously Degraded,'
Ds.a~r 7,1992.
15.
U.S. NRC Information Notice 91-67, " Problems With the Reliable Detection of Intergranular Attack (IGA) of Steam Generator Tubing," October 21,1991.
16.
U.S. NRC Information Notice 91-43, "Recent Incidents favolving Rapid increases in Primary-to-Secondary leak Rate," July 5,1991.
17.
U.S. NRC Information Notice 90-49, "Stresa Corrosion Cracking in PWR Steam Generator Tubes,"
August 6,1990.
NUREG-1604 D1
18.
U.S. NRC Information Notice 89-65, " Potential for Stress Corrosion Cracking in Steam Generator Tube Plugs Supplied by Babcock and Wilcox," September 8,1989.
19.
U.S. NRC Information Notice 88-99, " Detection and Monitoring of Sudden and/or Rapidly Increasing Primary-to-Secondary Leakage," December 20,1988.
20.
U.S. NRC Bulletin 89-01, Supplement 2, " Failure of Westinghouse Steam Generator Tube Mechanical Plugs," June 28,1991.
21.
U.S. NRC Bulletin 89-01, Supplement 1, " Failure of Westinghouse Steam Generator Tube Mechanical Plugs," November 14, 1990.
22.
U.S. NRC Bulletin 89-01, " Failure of Westinghouse Steam Generator Tube Mechanical Plugs," May 15, 1989.
23.
U.S. NRC Bulletin 88-02, " Rapidly Propagating Cracks in Steam Generator Tubes," Febnsary 5,1988.
24, U.S. NRC, " Steam Generator Tube Failures," NUREG/CR-6365 (INEL-95/0383), April 1996.
25.
U.S. NRC, " Steam Generator Operating Experience, Update for 1989-1990," NUREG/CR-57%,
December 1991.
26.
U.S. NRC, " Steam Generator Operating Experience, Update for 1987-1988,' NUREG/CR-5349 (SAIC-89/Ill3), June 1989.
27.
U.S. NRC, "NRC Integrated Program for the Resolution of Unresolved Safety Issues A-3, A-4, and A-5 Regarding Steam Generator Tube Integrity," NUREG-0844, September 1988.
28.
U.S. NRC, ' Steam Generator Operating Experience, Update for 1984-1986,' NUREG/CR-5150 (SAIC-87/3014), June 1988.
29.
U.S. NRC, " Steam Generator Operating Experience Update for 19821983," NUREG 1063. June 1984.
30.
U.S. NRC, " Steam Generator Tube Experience," NUREG-0886, February 1982.
31.
U.S. NRC, " Summary of Tube Integrity Operating Experience with Once-Through Steam Generators,"
NUREG-0571, March 1980.
32.
U.S. NRC, " Summary of Operating Experience with Recirculating Steam Generators," NUREG-0523, January 1979.
33.
U.S. NRC, " Bases for Plugging Degraded PWR Steam Generator Tubes," Regulatory Guide 1.121, 34.
Letter from C.S. Welty, Jr., Manager, Steam Generator Program, Electric Power Research Institute (EPRI), to the NRC, dated March 31,1995, " Guidance on Circumferential Crack Detection and Length Sizing."
35.
Letter from D. Morey, Vice President, Farley Project, Southern Nuclear Operating Company, to the NRC, dated June 26,1995, " Joseph M. Parley Nuclear Plant; Response to Generic Letter 95-03.*
NUREG-1604 p.2 l
l
~
I I
- 36. I.etter from W.L. Stewart, Executive Vice President, Nuclear, Arizona Public Service, to the NRC, dated June 27,1995, "Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2, and 3; Docket Nos. STN 50-528/529/530; Response to Generic Letter 95-03, Circumferential Cracking of Steam Generator Tubes."
- 37. Ietter from P.M. Beard, Jr., Senior Vice President Nuclear Plant Operations, Florida Power Corporation, to the NRC, dated July 25,1996, " Generic letter 95-03, Request for AdditionalInformation."
- 38. letter from J.A. Selva, B&W Owners Group Steering Committee Chairman, The B&W Owners Group, to the NRC, dated June 13,1995, "B&W Owners Group Generic Response to GL 95-03."
- 39. Ietter from J. Knubel, Vice Presxtent and Director, TMI, General Public Utilities Nuclear Corporation, to the NRC, dated October 30,1995, "Three Mile Island Nuclear Station, Unit 1 (TMI 1); Operating License No. DPR 50; Docket No. 50-289; AdditionalInfornetion in Response to Generic letter (GL) 95-03, 'Circumferential Cracking of Steam Generator Tubes'."
- 40. letter from J.R. Marshall, Manager, Licensmg, Arkanama Power A Light Company, to the NRC, dated September 15,1983, " Arkansas Nuclear One - Unit 1; Docket No. 50-313; License No. DPR-51; ANO-1 OTSG Mid-Cycle Inspection & Tube Pull Examination Results."
- 41. Letter from J.P. Stetz, Vice President - Nuclear, Davis-Besse, Centerior Energy, to the NRC, dated June 3,1996, " Revision of Toledo Edison's Responses to Generic letter 95-03, 'Circumferential Cracking of Steam Generator Tubes * (TAC No. M92238)."
- 42. Letter from D.C. Mims, Director, Nuclear Safety, Entergy Operations, Inc., to the NRC, dated February 5,1997, ' Arkansas Nuc. ear One - Unit 1: Docket No. 50-313; License No. DPR-51; Additional l
Information Related to Inservice Inspection of Once Through Steam Generator Tubes (TAC No.
M97485)."
- 43. Letter from J.P. Stetz, Vice President - Nuclear, Davis-Besse, Centerior Energy, to the NRC, dated February 1,1996, " Response to NRC Request for AdditionalInformation Regarding the Toledo Edison Response to Generic letter 95 03, 'Circumferential Cracking of Steam Generator Tubes' (TAC No.
M92238)."
- 44. Letter from D.C. Mims, Director, Licensing, Entergy Operations, Inc., to the NRC, dated October 12, 1995, " Arkansas Nuclear One - Units 1 and 2; Docket Nos. 50-313 and 50-368; License Nos. DPR-51 and NPF-6; AdditionalInformation in Response to Generic Letter 95-03 (TAC Nos. M92220 and M92221)."
- 45. Meeting summary from D. Scaletti, Senior Project Manager, NRC, dated April 16,1996 " Summary of Meeting Held with the B&W Owners Group Steam Generator Committee.'
- 46. Letter from D.C. Mims, Director, Licensing, Entergy Operations, Inc., to the NRC, dated June 27,1995,
" Arkansas Nuclear One - Units 1 and 2: Docket Nos. 50-313 and 50-368; License Nos. DPR-51 and NPF-6; Response to Generic Letter 95-03, 'Circumferential Cracking of Steam Generator Tubes'."
- 47. Letter from N.C. Kazanas, General Manager, Tennessee Valley Authority, to the NRC, dated June 27, 1995, "Bellefonte Nuclear Plant (BLN) - Response to NRC Generic letter 95-03
'Circumferential Cracking of Steam Generator Tubes'."
- 48. letter from G.L. Boldt, Vice President Nuclear Production, Florida Power Corporation, to the NRC, dated June 22,1995, " Response to Generic letter 95-03, 'Circumferential Cracking of Steam Generator Tubes'."
NUREG-1604 O.3 1
, ~ - - -
c
Letter from J.P. Stetz, Vice Pasident - Nuclear, Davis-Besse, Caterior Energy, to the NRC, dated June
- 49.
23,1995, " Toledo Edison Response to Generic Letter 95-03, 'Circumferential Cracking of Steam Generator Tubes'.*
50.
Letter from M.s. Tuckman, Senior Vice Pressdent, Nuclear Generation, Duke Power Company, to the NRC, dated October 25,1995, "McGuire Nuclear Station Docket Nos. 50-369,370; Catawba Nuclear Station Docket Nos. 50-413,414; Oconee Nuclear Station Docket Nos. 50-269,270,287; Response to Request for Additione S'armation Concermag Generic latter 95 03."
51.
Letter from M.S. Tuckman, Senior Vice President, Nuclear Generation, Duke Power Company, to the NRC, dated October 2,1995, 'Oconee Nuclear Station Units 1, 2, and 3; Docket Nos. 50-269,270, 287; Response to Request for AdditionalInformation Concerning Generic latter 95-03."
52.
Letter from M.S. Tuckman, Senior Vice President, Nuclear Generation, Duke Power Company, to the NRC, dated June 27,1995, 'McGuire Nuclear Station Docket Nos. 50-369,370; Catawba Nuclear Station Docket Nos. 50-413,414; Oconee Nuclear Station Docket Nos. 50 269,270,287; Response to Generic letter 95-03: Circumferential Cracking of Steam Generator Tubes."
53.
Letter from J. Knubel, Vice President and Director, TMI, General Public Utilities Nuclear Corporation, to the NRC, dated November 21,1995, 'Three Mile Island Nuclear Station, Unit 1 (TMI 1); Operating License No. DPR-50; Docket No. 50-289; AdditionalInformation in Response to Generic Letter (GL) 95-03, 'Circumferential Cracking of Steam Generator Tubes'."
- 54. Letter from T.G. Broughton, Vice President and Dimetor, TMI, General Public Utilities Nuclear Corporation, to the NRC, dated June 20,1995, "Thrse Mile Island Nuclear Station, Unit 1 (TMI-1);
Operating License No. DPR-50; Docket No. 50-289; GPU Nuclear Response Generic Letter (GL) 95-03
'Circumferential Cracking of Steam Generator Tubes'."
55.
Letter from W.F. Conway, Executive Vice President, Nuclear, Arizona Public Service Company, to the NRC, dated July 18,1993, "Palo Verde Nuclear Generating Station (PVNGS) Unit 2; Docket No. STN 50-529; Steam Generator Confirmatory Action Letter; File: 93-056-026."
56.
Letter from T.L. Patterson, Division Manager, Nuclear Operations, Omaho Public Power District, to the NRC, dated July 31,1996, " Response to Request for Additional Information (RAI) for Generic Letter (GL) 95-03, 'Circumferential Cracking of Steam Generator Tubes' (TAC No. M92243)."
- 57. Letter from P.W. Richardson, Assistant Project Manager, CE Owners Group Project Office, Combustion Engineering Owners Group, to Mr. C. Callaway, Nuclear Energy Institute, dated May 25,1995, *CEOG Experience Summary Regarding Circumferential Cracking of Steam Generator Tubes." locluded as an attachment to Reference 51.
58.
letter from R.E. Denton, Vice President, Nuclear Energy, Baltimore Gas and Electric Company, to the NRC, dated June 27,1995, 'Calvert Cliffs Nuclear Power Plant, Unit Nos.1 & 2; Docket Nos. 50-317
& 50-318; Response to NRC Generic Letter 95-03: Circumferential Cracking of Steam Generator Tubes."
- 59. Letter from G.D. Whittier, Vice President, Licensing & Eng. Support Dept., Maine Yankee Atomic Power Company, to the NRC, dated June 27,1995, "NRC Generic Letter 95-03: Circumferential Cracking of Steam Generator Tubes.*
NUREG-1604 D-4
i
- 60. Letter from R.E. Denton, Vice President, Nuclear Energy, Baltimore Gas and Electric Company, to the NRC, dated October 5,1995, "Calvert Cliffs Nuclear Power Plant, Unit Nos.1 & 2; Docket Nos. 50-317
& 50-318; Response to NRC's Request for AdditionalInformation Concerning Baltimore Gas and Electric Company's Response to NRC Generic Letter 95-03, 'Circumferential Cracking of Steam Generator Tubes,'-(Units 1 & 2 TAC Nos. M92229 & M92230)."
- 61. Meeting summary from P.F. McKee, Director, Project Directorate I-3, Division of Reactor Projects -
I/II, Office of Nuclear Reactor Regulation, NRC, dated June 1,1995, " Summary of May 11,1995, Meeting on Maine Yankee Steam Generator Inspections and Plans for Repair."
- 62. Letter from D.C. Mims, Director, Nuclear Safety, Entergy Operations, Inc., to the NRC, dated February 29,1996, " Arkansas Nuclear One - Unit 2; Docket No. 50-268 [368]; License No. NPF-6; 1995 Annual Report of Steam Generator Tubing Inservice Inspections."
- 63. Ietter from J.A. Stall, Vice President, St. Lucie Plant, Florida Power & Light Company, to the NRC, dated October 24,1996, "St. Lucie Unit 1; Docket No. 50-335, Steam Generator Run Time Analysis for Cycle 14."
- 64. Letter from J.J. Fisicaro, Director, Nuclear Safety, Entergy Operations, Inc., to the NRC, dated October 10,1996, "Waterford 3 SES; Docket No. 50-382; License No. NPF-38; Reporting of Special Report."
- 65. Letter from W.C. Marsh, Manager of Nuclear Regulatory Affairs, Southern California Edison Company, to the NRC, dated November 29.1995, " Docket Nos. 50-361 and 50-362; AdditionalInformation -
Generic Letter 95-03, 'Circumferential Cracking of Steam Generator Tubes' (TAC Nos. M92271 and M92272); San Onofre Nuclear Generating Station Units 2 and 3."
- 66. Letter from J.A. Stall, Vice Presirkat, St. Lucie Plant, Florida Power & Light Company, to the NRC, dated December 17,1996, "St. Lucie Unit 1; Docket No. 50-335; Outage SL1-14 Steam Generator Tube Inspection; NRC Information Request Response."
- 67. Letter from D.A. Sagar, Vice President, St. Lucie Plant, Fionida Power & Light Company, to the NRC, dated December 11,1995, "St. Lucie Units 1 and 2; Docket Nos. 50-335 and 50-389; Supplemental Information Generic Letter 95-03."
- 68. Letter from R.W. Smedley, Manager, Licensing, Consumers Power, to the NRC, dated October 25, 1995, " Docket 50-255 - License DPR Palisades Plant; Response to Request for Additional l
Information - Generic letter 95 Circumferential Cracking of Steam Generator Tubes."
- 69. Letter from B.D. Johnson, Senior Licensing Engineer, Consumers Power Company, to the NRC, dated July 9,1982, " Docket 50-255 - License DPR Palisades Plant - Steam Generator Tube Inspection Report."
- 70. Ietter from D.J. VandeWalle, Director, Nuclear Licensing, Consumers Power Company, to the NRC, dated April 19,1984, " Docket 50-255 - License DPR Palisades Plant - 1983/1984 Steam Generator Evaluation and Repair Report.'
- 71. letter from D.M. Crutchfield, Chief, Operating Reactors Branch #5, Division of Licensing, NRC, to D.J.
VandeWalle, Consumers Power Company, dated June 11,1984, *1983-84 Steam Generator Inspection."
- 72. Letter from J.L. Kuemin, Staff Licensing Engineer, Consumers Power, to the NRC, dated February 8, 1988, " Docket 50-255 - License DPR Palisades Plant - 1987 Steam Generator Inservice Inspection Report."
D.5 NUREG-1604
- 73. Ietter from K.W. Berry, Director, Nuclear Licensing, C--_. Power, to the NRC, dated November 3,1989, " Docket 50-255 - License DPR 20 - Palisades Plant - 1989 Maintenance Outage Steam Generator Inspection Results (TAC NO 69344)."
- 74. Letter from E.J. Mroczka, Senior Vice President, Nor a==* Utilities, to the NRC, dated February 7, e
1990, " Millstone Nuclear Power Station, Unit No. 2 Steam Generator Inspection."
- 75. Meeting summary from G.S. Vissing, Senior Project Manager, Project Directorate I-4, Division of Reactor Projects - I/II, Office of Nuclear Reactor Regulation, NRC, dated May 29,1990, " Summary of Meeting with Representatives of Northeast Utilities Concermag the Steam Generator Tube Integrity of Millstone Unit 2 - May 11,1990."
- 76. Letter from E. J. Mroczka, Senior Vice President, Northeast Utilities, to the NRC, dated July 3,1990,
' Millstone Nuclear Power Station, Unit No. 2 Steam Generator Inspection."
Letter from O.S. Vissing, Senior Project Manager, Project Directorate I-4, Division of Reactor Projects 77.
- I/II, NRC, dated September 23,1991, " Summary of Meeting with Representatives of Northeast Utilities Concerning the Assessment of the Steam Generators at Millstone 2, August 28,1991.*
- 78. Letter from E.C. Wenzinger, Chief, Projects Branch No. 4, Division of Reactor Projects, NRC, to J.F.
Opeka, Northeast Nuclear Energy Company, dated March 9,1992, " Millstone Combined Inspection 92-04."
- 79. Letter from T.L. Patterson, Division Manager, Nuclest Operations, Omaho Public Power District, to the NRC, dated June 23,1995, " Response to Generic beter (GL) 95-03, Circumferential Cracking of Steam Generator Tubes.*
- 80. Letter from J.R. Hebert, Manager, Licensing & Engineenng Support Department, Maine Yankee, to the NRC, dated December 3,1996, " Response to Generic Letter 95-03 'Circumferential Cracking of Steam
[ Generator] Tubes' Request for AdditionalInformation.'
- 81. Letter from K.M. Haas, Plant Safety and Licensing Director, Consumers Power, to the NRC, dated June 27,1995, " Docket 50-255 - License DPR Palisades Plant; Response to Generic Letter 95 Circumferential Cracking of Steam Generator Tubes.'
- 82. Ietter from W.L. Stewart, Executive Vice President, Nuclear, Arizona Public Service, to the NRC, dated February 8,1996, "Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2, and 3; Docket Nos. STN 50-528/529/530; Response to the Request for AdditionalInformation Regerding Generic letter 95-03."
83.
Letter from W.C. Marsh, Manager of Nuclear Regulatory Affairs, Southern California Edison Company, to the NRC, dated June 27,1995, " Docket Nos. 50-361 and 50-362; Response to NRC Generic letter 95-03, 'Circumferential Cracking of Steam Generator Tubes'; San Onofre Nuclear Generating Station Units 2 and 3.*
- 84. Letter from D.A. Sagar, Vice President, St. Lucie Plant, Florida Power & Light Company, to the NRC, dated June 23,1995, "St. Lucie Units 1 and 2; Docket Nos. 50-335 and 50-389; Generic letter 95-03 Response."
- 85. Ietter from R.F. Burski, Director, Nuclear Safety, Waterford 3, Entergy Operations, Inc., to the NRC, dated October 5,1995, 'Waterford 3 SES; Docket No. 50-382; License No. NPF-38; Request for AdditionalInformation Regarding NRC Generic letter 95-03.*
NUREG-1604 D-6
l
- 86. Letter from R.F. Burski, Director, Nuclear Safety, Waterford 3, Entergy Operations, Inc., to the NRC, dated June 27,1995, "Waterford 3 SES: Docket No. 50-382; License No. NPF-38; NRC Generic letter 95-03, 'Circumferential Cracking of Steam Generator Tubes'."
- 87. Letter from G.M. Rueger, Senior Vice President and General Manager, Nuclear Power Generation, Pac;iic Gas and Electric Company, to the NRC, dated October 2,1995, " Docket No. 50-275, OL-DPR-80; Docket No. 50-323, OL-DPR-82; Diablo Canyon Units I and 2; Response to NRC Requests for AdditionalInformation Related to Generic Letter 95-03, 'Circumferential Cracking of Steam Generator Tubes'."
- 88. Ietter from E.E. Fitzpatrick, Vice President, Indiana Michigan Power Company, to the NRC, dated June 27,1995, " Donald C. Cook Nuclear Plant Units I and 2; Generic Letter 95-03 Response Circumferential Cracking of Steam Generator Tubes."
- 89. 12tter from C.L. Terry, Group Vice President, TU Electric, to the NRC, dated June 27,1995,
" Comanche Peak Steam Electric Station (CPSES); Docket Nos. 50-445 and 50-446 Units 1 and 2; Response to Generic Letter 95-03, 'Circumferential Cracking of Steam Generator Tubes'."
- 90. Letter fmm T.V. Greene, Vice Chairman, Westinghouse Owners Group, to the NRC, dated June 26, 1995, " Westinghouse Owners Group; Circumferential Cracking of Steam Generator Tubes GL-95 03 -
WOG Report; ' Operating Experience Data and Safety Assessment,' MUHP-4101."
- 91. Letter from T.P. Noonan, Division Vice President, Nuclear Operations, Duquesne Light Corppany, to the NRC, dated June 23,1995, " Beaver Valley Power Station, Unit No. I and No. 2; BV-1 Docket No.
50-334, License No. DPR-66; BV-2 Docket No. 50-412, License No. NPF-73, Generic Letter 95-03 Response; Circumferential Cracking of Steam Generator Tubes."
- 92. Letter from S.C. Jain, Division Vice President, Nuclear Services, Nuclear Power Division, Duquesne Light Company, to the NRC, dated October 9,1996, " Beaver Valley Power Station, Unit No. 2; Docket No. 50-412, License No. NPF-73, Special Report."
- 93. Letter from D. Morey, Vice President, Farley Project, Southern Nuclear Operating Company, to the NRC, dated December 6,1996, " Joseph M. Farley Nuclear Plant - Unit 2; Licensee Event Report Number 96-003-00; Steam Generator Tube Degradation and Tube Status."
- 94. Letter from T.H. Cloninger, Vice President, Nuclear Engineering, Houston Lighting & Power, to the NRC, dated June 27,1995, ' South Texas Project, Units 1 and 2; Docket Nos. STN 50-498; STN 50-499; Response to NRC Generic letter 95-03: 'Circumferential Cracking of Steam Generator Tubes'."
- 95. Letter from K.L. Kaup, Site Vice President, Braidwood Generating Station, Commonwealth Edison Company, to the NRC, dated February 23,1996, "AdditionalInformation on the Braidwood Unit 1 Interim Inspection; Braidwood Nuclear Power Station Unit 1; NRC Docket Nos. 50-456."
- 96. Note from K.J. Karwoski, Materials Engineer, NRC, to E.J. Sullivan, NRC, dated December 3,1996,
" Steam Generator Tube Circumferential Indications at the Cold Leg Expansion Transition at Byron 1.*
- 97. Meeting summary from L.N. Olshan, Senior Project Manager, Project Directorate I-2, Division of Reactor Projects - IIII, Office of Nuclear Reactor Regulation, NRC, dated June 19,1996, " Summary of May 28,1996, Meeting to Discuss Steam Generators (TAC Nos. M94797 and M94798)."
D-7 NUREG 1604
- 98. Meetag summary from S.D. Bloom, Project Manager, Project Directorate IV-2 Division of Reactor Projects III/IV. Office of Nuclear Reactor Regulation, NRC, dated July 26,1996, " Summary of Meeting Held on February 23,1996, with Pacific Gas and Electne and Westinghouse to Discuss Diablo Canyon Unit 1 Steam Generator Tube Inspection Results and Tube lategrity Analyses.'
- 99. Letter from R.H. Shell, Manager, SQN Site Licensing, Tennessee Valley Authority, to the NRC, dated June 27,1995, 'Sequoyah Nuclear Plant (SQN) - Response to NRC Generic letter (GL) 95-03,
'Circumferential Cincking of Steam Generator Tubes'."
100. Latter from J.P. O'Hanlon, Senior Vice President - Nuclear, Virginia Electric and Power Company, to the NRC, dated October 5,1995, " Virginia Electric and Power Company; Surry Power Station Units 1 and 2; Nodh Anna Power Station Units 1 and 2; Response to NRC Request for AdditionalInformation Regarding Our Response to Generic letter 95-03; Circumferential Cracking of Steam Generator Tubes."
101. Letter from J.P. O'Hanlon, Senior Vice President - Nuclear, Virginia Electric and Power Company, to the NRC, dated June 27,1995, " Virginia Electric and Power Company; Surry Power Station Units 1 and 2; North Anna Power Station Units 1 and 2; Response to NRC Generic Letter 95-03; Circumferential Cracking of Steam Generator Tubes."
102. Letter from C.K. McCoy, Vice President, Nuclear, Vogtle Project, Georgia Power Company, to the NRC, dated June 27,1995, "Vogtle Electric Generating Plant, Response to NRC Generic Letter 95-03.*
103. Letter from T.F. Plunkett, Vice President, Turkey Point Plant, Florida Power and Light Company, to the NRC, dated October 11,1995, " Turkey Point Units 3 and 4; Docket Nos. 50-250 and 50-251; Response to Request for Additional Information - Generic Letter 95-03, 'Circumferential Cracking of Steam Generator Tubes','
104 letter from G.E. Trzyna, Nuclear Licensing Administrator, Zion Station, Commonwealth Edison Company, to the NRC, dated March 2,1989, " Zion Nuclear Power Station, Unit 2; License No. DPR-48; NRC Docket No. 50-304; Evaluation of Degraded Tube in 2A Steam Generator."
105. Meeting summary from G.F. Dick, Jr., Project Manager, Project Directorate III-2, Division of Reactor Projects - III/IV, Office of Nuclear Reactor Regu*ation, NRC, dated July 3,1996, " Summary of Meeting Discussing Steam Generator Tube U-Bend Flaws and Length of Byron 1 Operating Cycle - June 20, 1996."
106. Ietter from J.B. Hosmer, Engmeermg Vice President, Commonwealth Edison Company, to the NRC, dated January 30,1997, " Additional Information Pertaming to Zion Unit 2 Steam Generator Inspection; Zion Nuclear Power Station Unit 2; NRC Docket Number: 50-304."
107. Letter from W.H. Fujimoto, Vice President-Diablo Canyon, Operations and Plant Manager, Pacific Gas and Electric Company, to the NRC, dated June 29,1995, " Docket No. 50-275, OL-DPR-80; Docket No.
50-323, OL-DPR-82; Diablo Canyon Units 1 and 2; Response to NRC Generic Letter 95-03,
'Circumferential Cracking of Steam Generator Tubes'."
108. Letter from T.C. Feigenbaum, Senior Vice President & Chief Nuclear Officer, North Atlantic Energy Service Corporation, to the NRC, dated October 23,1995, " Response to Request for Additional Information Related to Generic Letter 95-03.*
109. letter from F.R. Decimo, Vice President, Northeast Utilities System, to the NRC, dated February 16, 1996, "Haddam Neck Plant; Millstone Nuclear Power Station, Unit Nos. 2 and 3; Generic Letter 95-03,
'Circumferential Cracking of Steam Generator Tubes'; Response to Request for AdditionalInformation."
NUREG-1604 D-8 I
110. Letter from D.P. Schnell, Senior Vice President, Nuclear, Union Electric, to the NRC, dated i
1996, "Callaway Plant; Docket Number 50-483; Circumferential Cracking of Steam Generator Tubes."
111. Letter from R.E. Beedle, Executive Vice President, Nuclear Generation, New York Power A the NRC, dated May 24,1991, ' Indian Point 3 Nuclear Power Plant; Docket No. 50-286; Analysis Steam Generator Tube Crack."
112. Letter from W.L. Stewart, Senior Vice President - Nuclear, Virginia Electric and Power Compa the NRC, dated August 30,1991, " Virginia Electric and Power Company; North Anna Power Station U 1; Steam Generator Operating Cycle Evaluation; Transmittal of the Westinghouse WCAP Reports 113. Letter from G.E. Kane, Station Manager, Virginia Electric and Power Company, to the NRC, date January 21,1992, " Licensee Event Report (LER): Unit Shutdown Due to Indeterminate Status of Stea Generators Following Eddy Current Data Re-Review.'
114. letter from W.L. Stewart, Senior Vice President - Nuclear, Virginia Electric and Power Comp the NRC, dated May 1,1992, ' Virginia Electric and Power Company; North Anna Power Station Mid-Cycle Steam Generator Inspection Results and Steam Generator Operating cycle Evaluation."
115. Letter from C.A. Julian, Chief, Engineering Branch, Division of Reactor Safety, NRC, to T.C.
McMeekin, Duke Power Company, dated September 29,1993, "NRC Inspection Report Nos.
50-369/93-19 and 50-370/93-19.*
116. Letter from D.E. IaBarge, Senior Project Manager, Project Directorate 11-3, Division of Re
- I/I[I}, Office of Nuclear Reactor Regulation, NRC, to O.D. Kingsley, Jr., Tennessee Valley Au dated October 11,1995, ' Issuance of Technical Speciiication Amendment for the Sequoyah Nuclear Plant Unit 1 (TAC No, M92%1)(TS 95-15)."
117. Meeting summary from D.E. LaBarge, Senior Project Manager, Project Directorate II 3, Division of Reactor Projects - I/II, Office of Nuclear Reactor Regulation, NRC, dated January 16,1996, " Summary of the December 11, 1995, Meeting on the Unit 1 Steam Generator Tube Inspection Results."
118. Letter from C.A. Julian, Chief, Engineering Branch, Division of Reactor Safety, NRC, to T.C.
McMeekin, Duke Power Company, March 9,1994, 'NRC Inspection Repoit Nos.
50-369/94-05 and 50-370/94-05."
119. Letter from C. A. Julian, Chief, Engineering Branch, Division of Reactor Safety, NRC, to D.L. Re Duke Power Company, dated December 50-413/93-32 and 50-414/93-32).*
30,1993, " Notice of Violation (NRC Inspection Peort No.
120. Letter from D.M. Saccomando, Senior Nuclear Licensing Administrator, Commonwealth Edison Company, to the NRC, dated October 13,1995, " Response to Request for AdditionalInformation Regarding GL 95-03, 'Circumferential Cracking of Steam Generator Tubes'; Byron Station Units I Braidwood Station Units 1 and 2; NRC Docket Numbers:
50-456 and 50-457."
50-454, and 50-455; NRC Docket Numbers:
121. Letter from W.R. Robinson, Vice President, Harris Nuclear Plant, Carolina Power & Light Com the NRC, dated October 11,1995, 'Shearon Harris Nuclear Power Plam; Docket No. 50-400/ License No NPF-63; Generic Letter 95 03, Circumferential Cracking of Steam Generator Tubes, Request for Additional Information."
D-9 NUREG-1604
122. Electric Power Research Institute report NP-6201, Revision 3, "PWR Steam Generator Examination j
Guidelines: Revision 3," Novesaber 1992.
123. Letter from M.S. Tuckman, Senior Vice President, Nuclear Generation, Duke Power Company, to the NRC, dated September 19,1995, " Catawba Nuclear Station Units 1 & 2; Docket Nos. 50-413,414; Response to Request for AdditionalInformation Concerning Generic letter 95-03."
124. Letter from C.L. Terry, Group Vice President, TU Electric, to the NRC, dated December 13, 1996,
' Comanche Peak Steam Electric Station (CPSES) - Units 1 and 2; Docket Nos. 50-445 and 50-446; Response to Request for AdditionalInformation on CPSES Response to Generic letter 9543,
'Circumferential Cracking of Steam Generator Tubes' (TAC Nos. M92233 and M92234)."
i 125. Letter from T.H. Cloninger, Vice President, Nuclear Engineering, Houston Lighting & Power, to the NRC, dated October 2,1995, ' South Texas Project Electric Generating Station, Units 1 and 2; Docket Nos. STN 50-498, STN 50-499; AdditionalInformation Regarding NRC Generic Letter 95-03:
'Circumferential Cracking of Steam Generator Tubes'."
126. Letter from E.E. Fitzpatrick, Vice President, Indiana Michigan Power Company, to the NRC, dated December 19,1995, " Donald C. Cook Nuclear Plant Units 1 and 2; Generic Letter (GL) 95-03:
Circumferential Cracking of Steam Generator Tubes; Request for AdditionalInformation."
l 127. Letter from E.E. Fitzpatrick, Vice President, Indiana Michigan Power Company, to the NRC, dated October 13,1995, " Donald C, Cook Nuclear Plant Units 1 and 2; Generic Letter 95-03 Response Circumferential Cracking of Steam Generator Tubes; Request for AdditionalInformation (RA1)."
128. btter from R.C. Meeredy, Vice President, Nuclear Operations, Rochester Gas and Electric Corporation, to the NRC, dated July 25,1996, " Response to Request for AdditionalInformation (RAI) Concerning Circumferential Cracking of Steam Generator Tubes (TAC No. M92244); R.E. Ginna Nuclear Power Plant; Docket No. 50-244."
129. Letter from R.C. Meeredy, Vice President, Nuclear Operations, Rochester Gas and Electric Corporation, to the NRC, dated March 22,1996, " Generic Letter (GL) 95-03 dated April 28,1995, 'Circumferential Cracking of Steam Generator Tubes'; Response to NRC Request for.*dditionalInformation (RAI); R.E.
Ginna Nuclear Power Plant; Docket No. 50-244."
130. letter from R.C. Meeredy, Vice President, Nuclear Operations, Rochester Gas and Electric Corporation, to the NRC, dated June 27,1995, " Response to Generic Letter 95-03, 'Circumferential Cracking of Steam Generator Tubes,' dated April 28,1995; R.E. Ginna Nuclear Power Plant; Docket No. 50-244."
131. Letter from E.A. DeBarba, Vice President, Northeast Utilities System, to the NRC, dated January 2, 1996, "Haddam Neck Plant; Millstone Nuclear Power Station, Unit Nos. 2 and 3; Generic Letter 95-03,
'Circumferential Cracking of Steam Generator Tubes' - Request for Additional Information."
132. Letter from J.F. Opeka, Executive Vice President, Northeast Utilities System, to the NRC, dated June 27, 1995, "Haddam Neck Plant; Millstone Nuclear Power Station, Unit Nos. 2 and 3; Seabrook Station; Response to Generic Letter 95-03, 'Circumferential Cracking of Steam Generator Tubes'."
133. Letter from S.E. Quinn, Vice President, Consolidated Edison Company of New York, Inc., to the NRC, dated January 12,1996, " Response to NRC's Request for Additional Information (RAI), Response to Generic Letter 95-03, Circumferential Cracking of Steam Generator Tubes, Indian Point Nuclear Generating Station Unit No. 2 (TAC No. M92247)."
NUREG-1604 D 10
l 134. lettee from S.E. Quinn, Vice President, Consolidated Edison Company of New York, Inc., to the NRC, dated June 26,1995, ' Generic Letter 95-03.*
135. Letter from C.R. Steinhardt, Senior Vice President - Nuclear Power, Wisconsin Public Service Corporation, to the NRC, dated January 12,1996, " Docket 50-305; Operating License DPR-43; Kewaunee Nuclear Power Plant; Response to Request for AdditionalInformation Regarding Generic letter 95-03."
136. Letter from C.R. Stembardt, Senior Vice President - Nuclear Power, Wisconsin Public Service Corporation, to the NRC, dated June 27,1995, " Docket 50.305; Operstmg License DPR-43; Kewaunee Nuclear Power Plant; Response to NRC Generic letter 95-03 'Circumferential Cracking of Steam Generator Tubes'."
137. letter from B. Link, Vice President, Nuclear Power, Wisconam Electric Power Company, to the NRC, dated October 6,1995, ' Dockets 50-266 and 50-301; Request for Additional Information Regarding Generic Letter 95 03, Circumferential Crackmg of Steam Generator Tubes; Point Beach Nuclear Plant, Units 1 and 2."
138. Letter from B. Link, Vice President, Nuclear Power, Wisconsin Electric Power Company, to the NRC, dated June 26,1995, ' Dockets 50-266 and 50-301; Response to Generic letter 95-03; Circumferential Cracking of Steam Generator Tubes; Point Beach Nuclear Plant, Units 1 and 2."
139. Letter from M.D. Wadley, Plant Manager, Prairie Island Nuclear Generating Plant, Northern States Power Company, to the NRC, dated April 8,1996, " Prairie Island Nuclear Generating Plant; Docket Nos.
50-282 and 50-306; License Nos. DPR 42 and DPR-60; Response to Request for AdditionalInformation, Prairie Island Nuclest Generating Plant, Units 1 and 2, Generic Letter 95-03, 'Circumferential Cracking of Steam Generator Tubes' (TAC Nos. M92266 and M92267)."
140. Letter from R.O. Anderson, Director, Licensing and Management Issues, Northern States Power Company, to the NRC, dated June 27,1995, " Prairie Island Nuclear Generating Plant; Docket Nos.
50-282 and 50-306; License Nos. DPR-42 and DPR-60; Response to Generic Letter 95-03,
'Circumferential Cracking of Steam Generator Tubes'."
141. Letter from D.M. Saccomando, Senior Nuclear Licensing Administrator, Commonwealth Edison Company, to the NRC, dated November 10,1995, " Response to Request for AdditionalInformation Pertaining to GL 95-03 'Circumferential Crack [ing) of Steam Generator Tubes'."
142. Letter from G.K. Schwartz, Station Manager, Zion Station, Commonwealth Edison Company, to the NRC, dated September 28,1995, " Zion Response b NRC Roquest for Additional Information for Generic l
letter 95-03, 'Circumferential Cracking of Steam.bator Tubes'; Zion Nuclear Power Station Units 1 and 2; NRC Dockets 50-295 and 50-304.*
143. letter from M.J. Vonk, Generic Issues Adminishator, Nuclear Regulatory Services, Commonwealth Edison Company, to the NRC, dated June 27,1995, "Braidwood Station Units 1 and 2; Byron Station Units 1 and 2; Zion Station Units 1 and 2; Comed Response to NRC Generic Letter 95-03:
'Circumferential Cracking of Steam Generator Tubes'; NRC Dockets 50-456 and 50-457; NRC Dockets 50-454 and 50-455; NRC Dockets 50-295 and 50 304."
144. letter from G.S. Thomas, Division Vice President, Nuclear Services, Nuclear Power Division, Duquesne Light Company, to the NRC, dated January 10,1996, " Beaver Valley Power Station, Unit No. I and No.
2; BV-1 Docket No. 50-334, License No. DPR-66; BV 2 Docket No. 50-412, License No. NPF-73, Response to Request for AdditionalInformation Dated December 7,1995.*
D-Il NUREG-1604
145. Ietter from C.L. Terry, Group Vice President, TU Electric, to the NRC, dated January 18, 1996,
" Comanche Peak Steam Electric Station (CPSES); Docket Nos. 50-445 and 50-446 Units 1 and 2; Response to Request for AdditionalInformation on CPSES Response to Generic Letter 95-03,
'Circumferential Cracking of Steam Geowntor Tubes' (TAC Nos. M92233 and M92234)."
146. Ietter from D. Morey, Vice President, Farley Project, Southern Nuclear Operating Company, to the FRC, dated September 26,1995, " Joseph M. Farley Nuclear Plant; Request for AdditionalInfonnation Concerning Generic Letter 95-03."
147. Letter from M. S. Tuckman, Senior Vice President - Nuclear Generation, Duke Power Company, to the NRC, dated October 9,1995, "McGuire Nuclear Station Units 1 & 2; Docket Nos. 50 369,370; Response to Request for AdditionalInformation Concerning Generic Letter 95-03."
148. Letter from W.R. Robinson, Vice President, Harris Nuclear Plant, Carolina Power & Light Company, to the NRC, dated June 27,1995, "Shearon Harris Nuclear Power Plant; Docket No. 50400/ License No.
NPF-63; Generic letter 95-03, Circumferential Cracking of Steam Generator Tubes."
149. Leuer from S.E. Thomas, Manager, Design Engineering, Houston Lighting & Power, to the NRC, dated October 23,1995, " South Texas Project Electric Generating Station, Units 1 and 2; Docket Nos. STN 50-498, STN 50-499; AdditionalInformation Regarding NRC Generic Letter 95-03: 'Circumferential Cracking of Steam Generator Tubes'."
150. Letter from R.R. Baron, Nuclear Assurance and Licensing Manager (Acting), Tennessee Valley Authority, to the NRC, dated June 27,1995, " Watts Bar Nuclear Plant (WBN) - NRC Generic Letter (GL) 95 Circumferential Cracking of Steam Generator Tubes."
151. Letter from W.H. Fujimoto, Vice President-Diablo Canyon, Operations and Plant Manager, Pacific Gas and Electric Company, to the NRC, dated November 28,1995, ' Docket No. 50-275, OL-DPR-80; Docket No. 50-323, OL-DPR-82; Diablo Canyon Units 1 and 2; Clarification of Eddy Current Probes Used for Steam Generator Inspections.*
152. Letter from E.C. Simpson, Senior Vice President - Nuclear Engineering, Public Service Electric and Gas Company, to the NRC, dated October 17,1995, " Response to NRC Request for AdditionalInformation; Generic letter 95 Circumferential Cracking of Steam Generator Tubes; Salem Generating Station Units 1 and 2; Facility Operating License Nos. DPR-70 and DPR-75; Docket Nos. 50-272 and 50-311."
153. Letter from J.J. Hagan, Vice President - Nuclear Operations, Public Service Electric and Gas Company, to the NRC, dated July 17,1995, ' Response to Generic Letter 95-03; Circumferential Cracking of Steam Generator Tubes; Salem Generating Station Units 1 aad 2; Facility Operating License Nos. DPR-70 and DPR 75; Docket Nos. 50-272 and 50-311."
154. Ietter from R.H. Shell, Manager, SQN Site Licensing, Tennessee Valley Authority, to the NRC, dated December 6,1995, "Sequoyah Nuclear Plant (SQN) - Units I and 2 - Response to NRC Request for AdditionalInformation Regarding Generic Letter (GL) 95-03."
155. Letter from D.F. Schnell, Senior Vice President, Nuclear, Union Electric, to the NRC, dated June 27, 1995, "Callaway Plant; Docket Number 50-483; Circumferential Cracking of Steam Generator Tubes."
156. Ietter from R.M. Krich, Manager - Regulatory Affairs, Carolina Power & Light Company, to the NRC, dated July 31,1996, 'H.B. Robinson Steam Electric Plant Unit No. 2; Docket No. 50-261/ License No.
DPR-23; Response to Request for AdditionalInformation Regarding Response to Generic Letter 95-03,
'Circumferential Cracking of Steam Generator Tubes'."
NUREG-1604 p.12 l
157. Letter from R.M. Krich, Manager - Regulatory Affairs, Carolina Power & Light Company, to the NRC, dated July 22,1996, "H.B. Robinson Steam Electric Plant, Unit No. 2; Docket No. 50-261/ License No.
DPR-23; Response to Request for Additiona1Information Regarding Response to Generic Letter 95-03,
'Circumferential Cracking of Steam Generator Tubes'."
158. Letter from R.M. Krich, Manager - Regulatory Affairs, Carolina Power & Light Company, dated June 27,1995, "H.B. Robinson Steam Electric Plant, Unit No. 2 Docket No. 50-261/ License No. DPR-23; Response to NRC Genenc beter 95-03, 'Circumferential Cracking of Steam Generator Tubes'."
159. Letter from R.P. Croteau, Project Manager, Project Directorate II 1, Division of Reactor Projects - I/II, Office of Nuclear Reactor Regulation, NRC, to T.F. Plunkett, Flonda Power and Light Company, dated March 18,1996,
- Turkey Point Units 3 and 4 - Oceenc Imtter 95 Circumferential Cracking of Steam Generator Tubes (TAC Nos. M92284 and M92285)."
160. beter from R.J. Hovey, Vim President, Turkey Point Plant, Florida Power and Light Company, to the NRC, dated Febru M. " Turkey Point Unit 4; Docket No. 50-251; Generic Letter 95 'Circumferentic' eng of Steam Generator Tubes'."
161. Letter from R.J. Hovey, Vice President, Turkey Point Plant, Florida Power and Light Company, to the NRC, dated January 16,1996, " Turkey Point Unit 4; Docket No. 50-251; Generic Letter 95 'Circumferential Cracking of Steam Generator Tubes'."
162. Letter from W.H. Bohlke, Vice President, Nuclear Engineering and Licensing, Florida Power and Light Company, to the NRC, dated June 22,1995, " Turkey Point Units 3 and 4; Docket Nos. 50-250 and 50-231; Response to Generic Letter 95-03, Circumferential Cracking of Steam Generator Tubes."
163. Letter from C.K. McCoy, Vice President, Nuclear, Vogtle Project, Georgia Power Company, to the NRC, dated December 22,1995, "Vogtle Electric Generating Plant, Response to Request for Additional Information; Generic Letter 95-03, 'Circumferential Cracking of Steam Generator Tubes'."
I 164. Letter from N.S. Carns, Chairman, President, and Chief Executive Officer, Wolf Creek Nuclear Operating Corporation, to the NRC, dated January 9,1996, " Docket 50-482: Response to Request for Additional Information for Generic letter 95-03."
165. Ietter from R.C. Hagan, Vice President Engineenng, Wolf Creek Nuclear Operating Corporation, to the NRC, dated June 23,1995, " Docket 50-482: Response to Generic Letter 95-03."
166. Letter from G.J. Taylor, Vice President, Nuclear Operations, South Carolina Electric & Gas Company, to the NRC, dated February 1,1996, " Virgil C. Summer Nuclear Station; Docket No. 50/395; Operating License No. NPF-12; ' Request for Additional Information, Generic letter 95-03, Circumferential Cracking of Steam Generator Tubes'."
167. Letter from W.J. Cahill, Jr., Chief Nuclear Officer, Nuclear Generation, New York Power Authority, to the NRC, dated January 17,1996, " Indian Point 3 Nuclear Power Plant; Docket No. 50-286; Response to Request for Additional Information (RAI), Response to Generic letter (GL) 95-03, Circumferential Cracking of Steam Generator Tubes."
168. letter from W.J. Cahill, Jr., Chief Nuclear Officer, Nuclear Generation, New York Power Authority, to the NRC, dated June 27,1995, " Indian Point 3 Nuclear Power Plant; Docket No. 50-286; License No.
DPR-64; Response to NRC Generic letter 95-03: 'Circumferential Cracking of Steam Generator Tubes'. "
p.13 NUREG-1604 l
l
169. Letter from G.J. Taylor, Vice President, Nuclear Operations, South Carolina Electric & Gas Company, to the NRC, dated August 16,1995, " Virgil C. Sumaw Nuclear Station (VCSNS); Docket No. 50/395; Operstag License No. NPF-12; 'Circumferential Crackmg of Steam Generator Tubes Generic Letter 95-03,' Supplement 1 Notarization."
x 1~0. Letter from G.J. Taylor, Vice President, Nuclear Operations, South Carolina Electric & Gas Company, to the NRC, dated June 27,1995, " Virgil C. Summer Nuclear Station: Docket No. 50/395; Operating License No. NPF-12; 'Circumferential Cracking of Steam Generator Tubes Generic IAtter 95-03'."
171. Ietter from B. Link, Vice President, Nuclear Power, Wisconam Electric Power Company, to the NRC, dated August 26,1994, " Dockets 50-266 and 50-301; Technical Specifications Change Request 175; Modifications to Section 15.4.2, 'In-Service Inspection of Safety Class Components'; Point Beach Nuclear Plant, Units 1 and 2.*
172. Letter from C.R. Seamhardt, Senior Vice President - Nuclear Power, Wisconsin Public Service Corporation, to the NRC, dated May 1,1996, " Docket 50-305; Operstag License DPR-43; Kewaunee Nuclear Power Plant; Proposed Amendment 136a to the Kewaunee Nuclear Power Plant Technical Specifications; Pressure Boundary Redefinition for Westinghouse Hybrid Expansion Joint Sleeved Tubes."
173. Letter from J.H. Taylor, Manager, Licensing Services, B&W Nuclear Technologies, to the NRC, dated
- i___M 21,1992, " Accepted Versions of Topical Report BAW-2045P, Rev. I and BAW-2045, Rev.1,
' Recirculating Steam Generators Kinetic Sleeve Qualification for 3/4 Inch O.D. Tubes'."
174. Meeting summary from V. Nerses, Project Manager, Project Directorate II-3, Division of Reactor Projects - I/II, Office of Nuclear Reactor Regulation, NRC, dated October 29,1993, " Summary of Meeting with Duke Power Company."
175. Letter from W.R. Robinson for J.E. Cross, Vice President and Chief Nuclear Officer, Portland General Electric Company, to the NRC, dated November 26,1992, " License Change Application (LCA) 227 -
Deferral of Unscheduled Steam Generator Inservice Inspection."
176.12tter from J.H. Taylor, Manager, Licensing Services, B&W Nuclear Technologies, to the NRC, dated Septsmber 29,1993, "Information to Utilities Regarding McGuire Tube Leak."
177. Letter from LT. Enos, Manager, Licensing, Arkanama Power & Light Company, to the NRC, dated October 22,1984, "Arkana== Nuclear One - Unit 1; Docket No. 50-313; License No. DPR-51; Steam Generator Surveillance Technical Specification Change Request."
178. Letter from M.D. Wadley, Plant Manager, Prairie Island Nuclear Generating Plant, Northern States Power Company, to the NRC, dated August 9,1996, " Prairie Island Nuclear Generating Plant; Docket Nos. 50-282 and 50-306; License Nos. DPR-42 and DPR-60; Steam Generator Tube Sleeves Metallurgical Examination Results."
.179. Letter from M.D. Wadley, Plant Manager, Prairie Island Nuclear Generating Plant Northern States Power Company, to the NRC, dated June 27,1996, " Prairie Island Nuclear Generating Plant; Docket Nos. 50-282 and 50-306; License Nos. DPR-42 and DPR 60; January 1996 Steam Generator Sleeving Issues Ninety Day Response letter."
180. Meeting summary from B. Wetzel, Project Manager, Project Directorate III-1, Division of Reactor Projects - Ill/IV, Office of Nuclear Reactor Regulation, NRC, dated June 26,1996, " Meeting with NSP to Discuss Issues Pertaining to the CE Steam Generator Sleeves Installed at Prairie Island Unit 1."
NUREG 1604 p.14
l 181. Letter from G.M. Leitch, Vice President Operations, Maine Yankee Atomic Power Company, to the NRC, dated April 14,1995, " Proposed Technical Specification Change No.190: Maine Yankee Steam Generator Tube Sleeving."
f D-15 NUREG-1604
NRC FORM 33s u.S. NUCLEAR REGULATORY C000husStoN
- 1. REPORT NUMBER (2 89)
(Assigned by NRC, Add Vol., supp., Rev.,
NRcu 1102, and Addendum Numbers, W eny.)
201. 32o2 BIBLIOGRAPHIC DATA SHEET (seeirmnceone on e,e re,or )
- 2. TITLE AND SUBTITLE Circumferential Cracking of Steam Generator Tubes 3.
DATE REPORT PUBLISHED l
YEAR MONTH e
April 1997
- 5. AUTHOR (S)
- 6. TYPE OF REPORT K:nneth J. Karwoski T W iul
- 7. PERIOD COVERED (Indusve Dates)
December 31,1996
- 8. PERFORMING ORGANIZATION. NAME AND ADDRESS (# NRc, provase Dueen, onics or Repon. U S. Nw.Aser Reputefory comtrwsmon, and meeng ack*ess; a contractor i
j provide name and means ad&ses)
Division of Engineering Office of Nuclear Reactor Regulation i
U.S. Nuclear Regulatory Commission W:shington, DC 2055 50001 4
- 9. SPONSORING ORGANIZATION NAME AND ADDRESS (#NRc. type "Same se ecove* acontrocsor, prowde NRc Dwoon. Onice or Repon. U S. N*csoar Reguratory commismon, and manno ad&ess }
Same as above
-l
^
- 10. SUPPLEMENTARY NOTES 1
i
- 11. ABSTRACT (200 eords or sess)
On April 28,1995, the U.S. Nuclear Regulatory Commission (NRC) issued Generic Letter (GL) 95-03,"Circumferential Cracking of Steam Generator Tubes." GL 954)3 was issued to obtain information needed to verify licensee compliance with exisun regulatory requirements regarding the integrity of steam generator tubes in domestic pressurized-water reactors (PWRs This report briefly describes the design and function of domestic steam generators and summarizes the staff's assessment of the responses to GL 95-03. The report condudes with several observations related to steam generator operating experience.
This report is intended to be representative of significant operating experience pertaining to circumferential cracking of steam generator tubes from April 1995 through December 1996. Operating experience prior to April 1995 is discussed throughout the report, as necessary, for completeness.
13 AVAN.ABit.m SIAMENT
- 12. KEY WORDS/DESCRIPTORS (t. set eerde orpaveses ther est assmt researchers a besono the report.)
Unlimited pressurized-water reactor, steam generator, steam generator tubes, stress corrosion cracking,
". secuRITV ctAssificATION circumferential cracking, eddy current, degradation mechanisms, inspection, surveillance, tube integrity, Generic Letter 95-03 (nns rees)
Unclabified (nus Report)
Unclassified
- 15. NUMM OF PAGES
- 16. PRICE NRc ?ORM 335(249)
The form was encaronmeny prockmed by Etae Federal Forms. Inc.
Printed on recycled paper Federal Recycling Program
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