ML082740268

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Transmittal of Steam Generator Examination Report for the R.E. Ginna Nuclear Power Plant Conducted in 2008
ML082740268
Person / Time
Site: Ginna Constellation icon.png
Issue date: 09/23/2008
From: Joseph Pacher
Constellation Energy Group, Nuclear Generation Group, Ginna
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML082740268 (38)
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{{#Wiki_filter:Joseph E. Pacher Manager, Nuclear Engineering Services 0 Constellation Energy Nuclear Generation Group R.E. Ginna Nuclear Power Plant, LLC 1503 Lake Road Ontario, New York 14519-9364 585.771.5208 585.771.3392 Fax joseph.pacher@constellation.com September 23, 2008 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION:

SUBJECT:

Document Control Desk R.E. Ginna Nuclear Power Plant Docket No. 50-244 Transmittal of Steam Generator Examination Report for the R.E. Ginna Nuclear Power Plant Enclosed is a copy of the R.E. Ginna Nuclear Power Plant Steam Generator Examination Report for the refueling outage conducted in 2008. This report is submitted as specified by R.E. Ginna Technical Specification 5.6.7. There are no new commitments being made in this submittal. If you should have any questions regarding this submittal, please contact David F. Wilson at (585) 771-5219. e truly yours, seph E. Pacher

Attachment:

(1) R.E. Ginna Nuclear Power Plant Steam Generator Examination Report cc: S. J. Collins, NRC D. V. Pickett, NRC Resident Inspector, NRC k4cP7 ,oog2O(7

bcc: C. W. Fleming, Esquire D. Skolnik COMMITMENTS IDENTIFIED IN THIS CORRESPONDENCE: None UFSAR/TSB Revision Required? No

Attachment (1) R.E. Ginna Nuclear Power Plant Steam Generator Examination Report

Constellation Energy Nuclear Generation Group R.E. Ginna Nuclear Power Plant Steam Generator Examination Report April 2008 End Refueling Outage of Cycle 33 Prepared by: Reviewed by: Approved by: Michael Shie ds E Steam ge rator Program Engineer ord Ver -n Principal Engineer Gnarells E General Supervisor Engineering Programs late late late I

I I TABLE OF CONTENTS 1.0 Acronyms 2.0 Examination Summary 2.1 Eddy Current Data Evaluation and Data Management 2.2 Summary of Base Scope and Special Interest Examinations 3.0 Eddy Current Bobbin and Rotating Coil Examination Results Summary 3.1 Bobbin coil specifics 3.2 Rotating coil specifics 3.3 Cold leg denting 4.0 R.E. Ginna Nuclear Power Plant Technical Specification Summary Item A The scope of inspections performed on each SG Item B Active degradation mechanisms found Item C Nondestructive examination techniques utilized for each degradation mechanism Item D Location, orientation (if linear), and measured sizes (if available) of service induced indications Item E Number of tubes plugged during the inspection outage for each active degradation mechanism Item F Total number and percentage of tubes plugged to date Item G The results of condition monitoring, including the results of tube pulls and in-situ testing 2

Q2 2008 Refueling Outage R.E. Ginna Nuclear Power Plant Steam Generator Examination Report 1.0 Acronyms R.E. Ginna Nuclear Power Plant Eddy Current Acronyms CATEGORY I: NO FURTHER ACTION REQUIRED Deposit (MRPC Only) DEP Sludge (Bobbin Only) SLG Plug PLG Indication not recordable (Bobbin and MRPC) INR No Detectable degradation (Bobbin and MRPC) () indicates a blank Previous Bobbin call (current outage) (Re-run test) PBC Previous Rotating call (current outage) (Re-run test) PRC CATEGORY II: RETEST REQUIRED Retest Bad Data (Bobbin and MRPC) RBD Retest Fixture (Bobbin and MRPC) RFX Retest tube questionable encode (Bobbin and MRPC) RNC Retest incomplete(MRPC) RIC Retest with wear scar standard (Bobbin and MRPC) RWS Retest no data (Bobbin and MRPC) RND Retest restricted tube (analyze data with extent and retest with a smaller RES probe) (Bobbin and MRPC) Retest Analyst discretion (MRPC Only with RIC exams) RAD CATEGORY III: DIAGNOSTIC REQUIRED Absolute Drift Indication (Bobbin Only) ADI Dent with Indication (Bobbin Only) DNI Distorted Tubesheet Indication (Bobbin Only) DTI Distorted Support Indication (Bobbin Only) DSI Non Quantifiable Indication (Bobbin Only) NQI CATEGORY IV: Post DIAGNOSTIC TEST Absolute Drift Signal (Bobbin only) ADS Circumferential extent measurement (MRPC only) ARC Dent signal (Bobbin Only) I[ DNS Distorted Support Signal (Bobbin Only) DSS Distorted Tubesheet Signal (Bobbin) DTS Length measurement (MRPC only) LEN Non Quantifiable Signal (Bobbin Only) NQS Possible loose part signal (MRPC only) PLP Tube to Tube proximity Signal (MRPC Only) PRS Volumetric within 1" of historical MBM (MRPC only) MBC No Defect Found (MRPC) NDF 3

Qý 2008 Refueling Outage R.E. Ginna Nuclear Power Plant Steam Generator Examination Report 1.0 Acronyms (continued) R.E. Ginna Nuclear Power Plant Eddy Current Acronyms (cont'd) CATEGORY V: REPAIR, ARC, OR ENGINEERING EVALUATION Multiple Axial Indication (MRPC Only) MAI Single Axial Indication (MRPC Only) SAI Multiple Circumferential Indication (MRPC Only) MCI Single Circumferential Indication (MRPC Only) SCI Mixed Mode Indication (MRPC Only) MMI Single Volumetric Indication (MRPC Only) SVI Obstructed (Bobbin and MRPC) OBS Positive Identification (Resolution only) PID Tube to Be Repaired (Resolution only) TBP CATEGORY VI: N/A CATEGORY VII: REVIEW INDICATION, REVIEW HISTORY, DIAGNOSTIC SAMPLING, OR ENGINEERING EVALUATION Volumetric Indication (MRPC Only) VOL Bulge (Bobbin only) BLG Ding (Bobbin Only) DNG Dent (Bobbin Only) DNT Lead Analyst Review (Bobbin and MRPC) LAR Manufacturing Burnish Mark (Bobbin Only) MBM Over Expansion (Bobbin Only) OXP Permeability Variation (Bobbin and MRPC) PVN Possible Loose Part (Bobbin Only) PLP Tube to Tube Proximity (Bobbin only) PRO Geometric Indication (MRPC U bend only) GEO Indication Not Found (Bobbin and MRPC) INF Retest (Bobbin Probe with wear Std.) MRPC code WAR 4

2008 Refueling Outage R.E. Ginna Nuclear Power Plant Steam Generator Examination Report 2.0 Summary Report This summary report describes the Constellation Energy Group Inservice Examination (ISI) of the steam generator (S/G) tubing at the R. E. Ginna Nuclear Power Plant. The ISI was performed during the April 2008 refueling outage. The steam generator tube ISI Examination was performed in accordance with The American Society of Mechanical Engineers (ASME) Section XI Code "Inservice Examination of Nuclear Power Plant Components" prescribed by Title 10 of the Code of Federal Regulations, Part 50, Section 50.55a(g). Therefore, the ISI is required to meet the 1995 Edition of the ASME Code Section XI with 1996 addenda. The R.E. Ginna Nuclear Power Plant design contains two (2) recirculating design steam generators designed and fabricated by Babcock and Wilcox International (BWI) of Cambridge Ontario (Canada). The nomenclature used for fabrication and subsequent in-service Examinations is BWI #34 (SG-A) and BWI #35 (SG-B). Each BWI steam generator was designed to contain 4765 tubes per S/G. One tube in each steam generator was removed from service during the fabrication by means of a shop welded 1-690 plug. S/G A contains 4764 inspectable tubes, and S/G B had four (4) tubes plugged in end of cycle 32 due to a loose part that brings the total to 4760 tubes. The tubing material is thermally treated Inconel Alloy 690 having a nominal outer diameter (OD) of 0.750 inch and a nominal wall thickness of 0.043 inch. The nominal thickness of the tube sheet is 25.25 inches, with a full depth hydraulic expansion of all the tubes into the tube sheet material. 2.1 Eddy Current Data Evaluation and Data Management All eddy current data acquired was subjected to two (2) separate independent analyses. The primary, secondary, and resolution analysis functions were all performed at the R.E. Ginna Nuclear Power Plant site. Westinghouse served as the primary analysis team and performed manual evaluation of all probe types. Zetec represented the secondary analysis team and utilized Computerized Data Screening (CDS) for bobbin coil data and provided manual data analysis for all rotating coil data. A primary / secondary compare and subsequent resolution of discrepancies was performed by personnel representing both Westinghouse and Zetec. An independent qualified data analyst (IQDA) reviewed all I-codes which were changed to a non-pluggable indication, or were dispositioned to a No Detectable Degradation (NDD) code. In addition, the Independent Qualified Data Analyst IQDA reviewed a sample of NDD results for tubes in each S/G. 5

2008 Refueling Outage R.E. Ginna Nuclear Power Plant Steam Generator Examination Report 2.2 Summary of Base Scope and Special Interest Examinations Bobbin Coil - Base Scope All tubes were examined full length from tube end to tube end with one exception. In steam generator A, R9C121 would not pass a 0.620" or 0.610" diameter probe. This tube was tested with a rotating coil single plus point (+PT) probe in the restricted area. The cold leg of this tube did not receive a bobbin examination during 2008 but was inspected full length in RFO 2005 with a different robot. This tube has a manufacturing dent at this location and the use of a lighter robot has proven difficult. The tube R9C 121 was ultimately tested full length by a combination of a bobbin coil and plus point coil. The nominal probe diameter attempted for all bobbin examinations was 0.620"beaded probe design. A 0.610" beaded probe design was used on an as-needed basis for restrictions or retests approved for this probe diameter. Both were manufactured by Zetec. S/G Scheduled Completed Obstructed Existing Plugs Exams Exams Tubes A 2763 2762 1* 1 (pre-operational) 1 (pre-operational) B 2804 2804 0 4 (end of cycle32 2005)

  • Tube was obstructed with a bobbin coil, but was ultimately tested full length with a bobbin coil and plus point coil combination.

Rotating Coil, Inlet Tubesheet Region, - Base Scope A 25% sample hot leg top of tube sheet program was scheduled in each steam generator. The minimum test extent identified was continuous data collection from 3" below the top of tube sheet, through the expansion, ending at 3" above the transition. The nominal probe diameter utilized for all straight section examinations was 0.610", manufactured by Zetec. S/G Scheduled Completed Obstructed Pluggable Exams exams Tubes Indications A 1302 1302 0 0 B 1245 1245 0 0 6

2008 Refueling Outage R.E. Ginna Nuclear Power Plant Steam Generator Examination Report Rotating Coil, Low Row U-Bend Region - Base Scope A 20% sample program of low row U-Bends was scheduled in each steam generator. The test extent identified was from the upper tube support hot leg to the upper tube support cold leg. The nominal probe diameter utilized for all low row U-Bend examinations was 0.580", manufactured by Zetec. S/G Scheduled Completed Obstructed Pluggable Exams Exams Tubes Indications A 26 26 0 0 B 24 24 0 0 Rotating Coil, Proximity Tubes in the U-Bend Region - Base Scope Previously identified U-Bend areas, having tube to tube proximity signals, were again scheduled for examination. This test was to verify the proximity values had not changed, as well as screen for any tube to tube wear in this region. The test extent identified was from the upper tube support of either the hot leg or the cold leg to a specified fan bar location in the U-Bend area. The nominal probe diameter utilized for all proximity examinations was a 0.580". S/G Scheduled Completed Damaged Pluggable Exams Exams Tubes Indications A 13 13 0 0 B 12 12 0 0 7 Ip*

2008 Refueling Outage R.E. Ginna Nuclear Power Plant Steam Generator Examination Report Rotating Coil, "box-in" of PLP locations - Special interest Low frequency responses, which could possibly represent the presence of a foreign object, were identified during the base scope bobbin and/or tube sheet foreign object search and removal examination (FOSAR). Previous results and data evaluations have tracked these signals for multiple examinations with no tube wear detectable. In each case, the surrounding tubes were examined to determine if in fact multiple tube contact was present. The standard 3-coil probe, mentioned above was used for this supplemental testing. In concert with the eddy current examinations for loose parts, the FOSAR detected small low mass parts. Tubes that were identified by FOSAR were reviewed for an eddy current response. If those tubes were not tested, they were then scheduled for a bounding examination. Tubes with an eddy current possible loose part (PLP) response were then reviewed by FOSAR and determined if a loose part was present, or if an area of hard sludge was present. All tubes were bounded with either FOSAR or an eddy current response. S/G Scheduled Completed Damaged Pluggable Exams Exams Tubes Indications A 131 131 0 0 B 135 135 0 0 Rotating Coil, Outlet Tube Sheet Region In S/G B for Dents (Added Scope) During the bobbin examination using the nominal probe diameter of 0.620"beaded probe design, several tubes in the cold leg of S/G B had "Turbo" three-frequency mix responses. After the sludge lancing of S/G B a sample of these tubes with the largest amplitude signals were retested with the same technique at a slower 24"/sec. speed to assure that no probe mechanical influences were contributing to the dent signal. Upon further review of this bobbin data, it was determined that denting was occurring at the top-of-tube sheet. This was later verified by the plus point probe, and discussed further in section 3.3. S/G Scheduled Completed Damaged Pluggable Exams Exams Tubes Indications B 80 80 0 0 8

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  • The restriction is with a 0.620" diameter.
    • 20 tubes with 23 PLP indications
      • Denotes % through wall (TW) depth sizing of tube wear indications.
        • Only one tube tried twice and OBS to 0.610" diameter.

The area of restriction was located in the U-bend portion of the tube and was tested with single +PT probe achieving a full length inspection. 19

2008 Refueling Outage R.E. Ginna Nuclear Power Plant Steam Generator Examination Report Steam Generator B Bobbin Tube Sheet Tube Proximity U-Bend Special Straight Section Hot+Point +Point Interest Special Interest N/A~~ N/AN/ ARC N I*A. BLG 6 DEP 10 DNG 20 DNT 92* DSS 2 INF 11 INR 80 MBC N/A. MBM 271 NDF N/A NQS 1 PLP 4 PBC 29 PRS N/A PRC N/A RBD 16 RIC LEN

  • 80 dents from B cold leg Top of Tubesheet (TTS)
    • 23 tubes with 25 PLP indications
      • Denotes % through wall (TW) depth sizing of tube wear indications.

20

2008 Refueling Outage R.E. Ginna Nuclear Power Plant Steam Generator Examination Report 3.1 Bobbin Coil Specifics BLG - Bulges identified within the tube sheet expansion meeting the reporting criteria of 18 volt peak to peak (Vpp) on the 550kHz absolute channel were identified. DEP - Low frequency responses indicative of deposits were identified as such. The intent is to track these during future Examinations. In the event a previously identified DEP signal was not present in the current Examination, INF was used to address these locations. DNG, DNT - Tubing deformations were identified as ding in the freespan, and as dent at support structures. The reporting criteria was >=2.0 Vpp in the freespan and supports area and >=5 Vpp in the U-Bend area. Previous signals identified as a ding or dent meeting the reporting criteria would be reported. DNT/DNG that was less than the reporting voltage will be reported as an 1NR. All signals not reported in past examinations, were verified as present in historical data. In the event these were misclassified (tangent point or probe snap) or reported at the wrong elevation during a previous outage, INR may have been used to re-address these locations. The only active denting is in the cold leg top of tube sheet region. DSS - A signal identified as a "distorted support signal" can result from a historical or present outage indication that was shown to change from history. Based upon this change, a diagnostic +point exam was performed which could produce a benign indication (NDD) or a wear indication +point result. The database would maintain a DSS bobbin coil result in the database for each case as well as the appropriate diagnostic +point indication. Steam generator A has one DSS bobbin coil wear indication that was left in-service, see response to item d. Steam generator B has one historical bobbin coil DSS indication that was proven to be benign by post diagnostic +point exam, as well as one DSS bobbin coil wear indication that was left in-service, see response to item d. These indications types have been depicted earlier in the report. These tubes have not been plugged and will remain in-service based upon sizing. These tubes will be considered for re-inspection during future outages. INF - Signals identified in previous Examinations, not present in 2008 were classified as 'indication not found". A historical tube comparison was performed to ensure the correct location was tested. Once the tube has been verified to be the same, a historical review (HR) is placed in the utility 1 field. INR - In the event a signal was misclassified (tangent point, probe snap, etc) or was reported at the wrong elevation during a previous outage, INR may have been used to re-address these locations. 21

2008 Refueling Outage R.E. Ginna Nuclear Power Plant Steam Generator Examination Report INF / INR - SPECIFICS: In S/G A, two tubes were reported as INF due to being mis-located in 2002. One DNT was not found this outage at an FOl + 10.39 inches. The balance of the INF's reported are for DEP signals that had been reported in 1999 or 2002 and have either moved on or was not a DEP. When previous outages identified a DEP as either INR or INF no further action was required this outage. When a previous DEP had not been addressed a review of this outage data was performed and INF was identified with the location of previously reported DEP signal. MBM - Manufacturing Burnish Mark type signals represented the majority of all signals reported in both steam generators. The reporting criteria used was >=2.5 Vpp on channel P6 with consideration given to a response on the P 1 mix channel. Previous MBM signals have been reported on channel 6 (140 kHz) absolute, in 2008 the reporting channel for MBM signals is P6 due to the normalization of voltage. P6 channel is set to 4.42 Vpp on the 4-100% holes as in the past. A direct comparison of voltages could be made by the resolution analysts from channel 6 to P6. S/G-A had 265 reported. S/G-B had 271 reported. All signals observed were compared to historical data to look for signal 'change' for possible degradation at these locations. No changes have been noted at these manufacturing anomalies to date. NQS - A differential bobbin coil non-quantifiable signal. In previous examinations a NQI signal was called and a rotating coil examination was then performed. The results of the rotating coil were "no defect found" (NDF). Future examinations can compare the bobbin signal response of current outage to previous outages and the NQS will be used if they appear the same. There was one repeat NQS from 2005 in S/G B. OBS - Obstructed This means that a downsized probe was used and attempted to test a restricted condition. If the tube will not pass this downsized probe the operator puts another message in indicating that it is either restricted again or obstructed with that size probe. S/G A has R9C121 that will not pass a 0.610" diameter probe from the hot leg side. There were two separate attempts made to test this tube with a 0.610" diameter probe with no success from the hot leg side. That is the smallest bobbin probe approved for this outage. This tube was successfully examined in the restricted area with a 0.580" diameter +PT probe. The tube did not get a bobbin test on the cold leg side, however the tube was tested full length in 2005 with a different robot. PBC - Previous Bobbin call. This code is used when a tube is run more than once. Using this code will eliminate multiple entries of the same indication. S/G B has twenty nine (29) of these calls. PLP - Possible loose part signal. S/G A reported two (2) bobbin PLP calls and S/G B reported four (4) PLP calls. These tubes were put on a bounding list to be rotated with 22

2008 Refueling Outage R.E. Ginna Nuclear Power Plant Steam Generator Examination Report either the +PT U-Bend probe or the 3 coil hard body probe. The bobbin probe also detected PLP signals that were confirmed with Visual examination from the secondary side. RES - Restricted tube. When the operator puts in a message that a tube is restricted the analyst has no other choice-than to put RES in the % column and record what extent was tested. The tube will then get re-scheduled using a smaller probe size. There is one tube in S/G A R9C121 that a 0.620" diameter probe will not pass. RBD - Retest bad data. Bobbin analysis in S/G A had four (4) tubes that were called RBD while S/G B had 16 tubes that were called RBD. RIC - Retest Incomplete Bobbin analysis in S/G A had two (2) tubes that were called RIC while S/G B had one (1) tube called RIC. 3.2 Rotating Coil Specifics There were no axial or circumferential (crack-like) indications reported in any steam generator. There were two +PT indications representing wear from a loose part (loose part is not present now) reported in two supports along with a tapered wear signal in one of the supports. Details of those have been discussed in Item d section of this report. Some of the same codes are used for bobbin and rotating coil examinations. ARC - For each indication, the arc (width) is determined by using the C-Scan plot routine of Eddynet software. Section Item d of this report has applicable data for arc. LEN - Each indication reported is measured for length. The C-Scan plot routine of Eddynet software is used to determine the length. Section Item d of this report has applicable data for length. MBC - Manufacturing Burnish Mark confirmed by rotating coil. MBC is the code used when a Burnish Mark has been tested by rotating coil. A 20% sample of reported MBM's over 5 volts on the hot leg were selected for rotating coil examinations. S/G A tested six (6) MBM's with the 3 coil and all confirmed to be burnish marks with a volumetric response and no degradation. S/G B tested for five (5) burnish marks one in the U-Bend and four in the straight section. All confirmed to be burnish marks with a volumetric response and no degradation. NDF - No defect found. A planned 20% sample of hot leg dents were scheduled for a rotating coil examination to determine if degradation exists. S/G A had four (4) in the U-bend region and all reported NDF. S/G B had five (5) in the U-bend region with three (3) as part of the sample in the straight section. S/G B also had eighty (80) cold 23

2008 Refueling Outage R.E. Ginna Nuclear Power Plant Steam Generator Examination Report leg top of tube sheet dents that were tested with the 3 coil rotating probe. All reported NDF's. Percent (%)- Percent was used on three (3) indications. S/G A has one (1) tube that reported a 29% indication. S/G B has two (2) reported % calls one at 18% and second at 10%. The 10% is a tapered wear call. The 29% and 18% are from loose parts that have done damage and moved on. Eddy current also confirmed by low frequencies that no part was present where the damage has occurred. PLP - Possible Loose Parts S/G A has 20 tubes with 23 PLP's called. S/G B has 23 tubes with 25 PLP's called. Bounding (boxing in) would also occur when a PLP indication was called. In addition, a FOSAR examination was performed to confirm if foreign material was present or if sludge was the source of indication. When a visual PLP was called this would also require a bounding by the rotating coil examination. PRC - Previous rotating coil. This code is used when multiple runs have been made on the same tube and indications called are not repeated. S/G A had none while S/G B had three (3) tubes with PRC code. PRS - Tube to tube proximity has been tracked for subsequent Examinations after a manufactures notice was issued. S/G-A had 13 tubes with 26 calls, in S/G-B had 12 tubes with 22 calls, which were inspected to ensure no changes in proximity and/or tube to tube wear was occurring. There were no changes with respect to historical proximity regions and no associated wear indications were detected in this area. RBD - Retest Bad Data S/G A has three (3) top of tube sheet examinations while S/G B has one (1) tube called. RIC - Retest Incomplete S/G B had four (4) tubes reported RIC in the top of tube sheet examination. S/G A has two (2) reported tubes in the proximity examination and one (1) tube in the special interest U-Bend examination. 24

2008 Refueling Outage R.E. Ginna Nuclear Power Plant Steam Generator Examination Report 3.3 Cold Leg Denting During the 2008 Eddy Current (ET) Examination of S/G B several Distorted Tubesheet Indications (DTI) were identified by the 0.620" diameter nominal size bobbin coil at the cold leg top-of-tubesheet expansion region by both the Primary and Secondary analysis team. Since the tubing in the Steam Generators is Inconel 690 TT, which has no history of cracking susceptibility, it was initially thought by the Resolution team that these signals could be from a change in tube geometry. Upon further investigation of present outage data, with comparison to the historical bobbin data, it was noticed that some minor denting has occurred at the top-of-tubesheet secondary side interface. This was evidenced by the differential and absolute lissajous responses which indicated additional geometrical distortion in the area of the tube expansion and corresponding small depth crevice area per manufacturing design which are both located in the vicinity of the tubesheet secondary face. Since these indications were new, a decision was made to move the inspection robot to the cold leg of S/G B so that this dent population could be interrogated with the surface riding +PT 3-coil MRPC technique which is specifically designed for minimizing the effects of geometry such as tube expansions or tube denting. The examination was the standard +3/-3 inches from the top of tube sheet interface. The minor dents initially identified by the bobbin probe were confirmed by the surface riding +PT 3-coil MRPC technique. These small dents are located at the top of tube sheet interface and within the crevice elevation. The results of the surface riding +PT 3-coil MRPC technique Examination had no detectable degradation. These indications will be tracked in the database and considered for inspection in future examinations. The following attachments show the steam generator A and B examination scope each examination category. 25

SG - A Percent Through-Wall Indications RE. Ginna 1 33RFO RGE 20080401 05/02/2008 15:54:49 +--- t --....- +....------------...

+

-+-. ----- I INSPDATE ROW COL VOLTS DEG IND PER CHN LOCN INCHI INCH2 I CRLEN CRWIO CEG BEGT ENDT PDIA PTYPE CAL L COH[ +--------------------------------- +--------- 91 51 .83 133 DSS P1 05H .47 TEC TEH .620 ZBABP 23 HI I I 91 51 .34 0 LEN 4 05H .50 05H 05H .610 ZPSMN 39 HI I I 91 51 .43 0 PCT 29 4 05H .50 .34 58 05H 05H .610 ZPSHN 39 HI I 91 51 .38 58 ARC 4 05H .50 05H 05H .610 ZPSHN 39 HI I 1 20082/3/01 91 51 NOD TSH TSH .610 ZPSWR 38 HI I 1 2002/03/01 91 51 NOD TEC TEH .620 ZBALC 50 HI I 1 1997/10/01 91 51 NOD TEC TEH .620 ZBALC 65 HI I I 1995/11/01 91' 51 PSL 6 02H .13 TEH TEC .620 ZBALC 1603 CI I I 1995/11/01 91 51 1.06 82 MBM 6 06H 3.62 TEH TEC .620 ZBALC 1603 CI I I INSPDATE ROW COL VOLTS DEG IND PER CHN LOCN INCHI INCH2 I CRLEN CRWID CEG BEGT ENDT PDIA PTYPE CAL L COMl ,.----- + 4 I ---- I --------- +- 4 I-...... + 26 Tubes: 1 Records: 9

SG - B Percent Through-Wall Indications R.E. Ginna 1 33RFO RGE 20080401 05/02/2008 16:42:01 +....-- -+ -÷ -+-- I INSPDATE ROW COL VOLTS DEG IND PER CHN LOCN INCHi INCH2 I CRLEN CRWID CEG BEGT ENDT POIA PTYPE CAL L COMI +.... +......... + + +.------ + -+ -÷ 78 24 .31 130 DSS P1 01H 1.08 TEC TEH .620 ZBABP 27 HI I 78 24 1.88 179 INR P1 08C 1.52 TEC TEH .620 ZBABP 27 HI I 78 24 .28 0 PCT 19 4 01H 1.00 .33 40 01H 01H .610 ZPSHN 39 HI I I 78 24 .33 0 LEN 4 01H 1.00 8 010H .610 ZPSHN 39 HI I 78 24 .26 40 ARC 4 01H 1.00 01H 01H .610 ZPSHN 39 HI I 78 24 .15 0 PCT 10 P4 01H 1.29 1.22 25 01H O1H .610 ZPSHN 39 HI I 78 24 1.22 0 LEN P4 O0H 1.29 01H 01H .610, ZPSHN 39 HI I 78 24 .16 25 ARC P4 OIH 1.29 01K 01H .610 ZPSMN 39 Ni I 78 24 NOD TSH TSH .610 ZPSMN 40 HI I 78 24 .23 121 PRS 3 F03 6.90 16.00 F03 FOS .580 ZPUN8 42 H[ I 78 24 .32 98 PRS 3 F04 .00 15.00 F03 F05 .580 ZPUN8 42 HI I 2005/03/01 78 24 2.29 177 DNT P1 08C 1.52 TEC TEH .620 ZBABP 4 HI I 2005/03/01 78 24 .17 121 PRS 3 F03 6.90 15.00 F06 08H .580 ZPUIC 47 HI I 2005/03/01 78 24 .29 106 PRS 3 F04 .00 15.00 F06 08H .580 ZPU1C 47 HN I 2002/03/01 78 24 2.91 177 DNT P1 08C 1.51 TEC TEH .620 ZBALC 19 HI I 2002/03/01 78 24 RBD F04 06H .580 ZPU2C 31 HI I 2002/03/01 78 24 4.92 172 PRS P2 F03 6.95 14,90 F05 08H .580 ZPU2C 32 HI I 2002/03/01 78 24 7.54 292 PRS P2 F04 .00 14.19 F05 08H .580 ZPU2C 32 NI 1 1999/03/01 78 24 .00 0 INF FO1 6.00 17.00 TEC TEH .620 ZBALF 1 HI I 1999/03/01 78 24 .00 0 INF F02 .00 16,00 TEC TEH .620 ZBALF 1 HI I 1999/03/01 78 24 NOD F03 08H .580 ZRAGR 37 HI I 1999/03/01 78 24 .00 0 INF 5 F03 .00 16.00 F06 F02 .580 ZRAGR 68 HI I 1999/03/01 78 24 8.23 277 PRS P 6 F04 .00 8.16 F06 F02 .580 ZRAGR 68 HI I 1997/10/01 78 24 NOD TEC TEH .620 ZBALC 62 HI I 1997/10/01 78 24 .48 283 PRS 5 F01 6.00 17.00 F06 F01 .560 ZRAGR 90 HN I 1997/10/01 78 24 .62 285 PRS 5 F02 .00 16.00 F06 F01 .560 ZRAGR 90 HI I 1996/05/01 78 24 NOD TEC TEH '.620 ZBALC 5 HI I 1995/11/01 78 24 NOD TEH TEC .620 ZBALC 9 CI I +.----+---- + +-------.-.--....... +......... ÷. +-+ + I INSPDATE ROW COL VOLTS DEG IND PER CHN LOCN INCHI INCH2 I CRLEN CRWID CEG BEGT ENDT PDIA PTYPE CAL L COMI +-...... +........ +-.................. 27 1 Tubes: 1 Records: 28

2008 Refueling Outage R.E. Ginna Nuclear Power Plant Steam Generator Examination Report 4.0 Technical Specification Summary The following information satisfies the requirement of the R.E. Ginna Nuclear Power Plant Technical Specification TS 5.5.8 Steam Generator program and TS 5.6.7 Steam Generator Inspection Report. The TS 5.6.7 reporting requirements are within 180 days after initial entry into Mode 4 following completion of an inspection performed in accordance with TS 5.5.8. Information required by TS 5.6.7 Items a through g are provided. Mode 4 entry for the end of cycle (EOC) 33 R.E. Ginna Nuclear Power Plant refueling outage (RFO) occurred on May 8, 2008. Item a. The scope of inspections performed on each SG. Primary side examination type [ S/G A S/G B "As - found" visual Examination of primary bowl, sub 100% 100% components and existing tube plugs. Number of tubes inspected Full Length Bobbin coil 2762 (58%) 2804 (59%) Number of tubes inspected H/L Top of Tubesheet +point 1302 (27%) 1245 (26%) Number of tubes inspected row 1 & 2 (118 total) U-bend 26 (22%) 24 (20%) +point Number of tubes inspected for tube to tube proximity on outer 13 (100%) 12 (100%) radius tubes +point (indications consistent since 1997 RFO) Special interest including a planned sample of manufacturing 100% 100% anomalies, bobbin coil indications (I codes), tubesheet overexpansion' s, "As - left" visual examination of primary bowl, and sub 100% 100% components A Degradation Assessment (DA) was written prior to the EOC 33 2008 RFO. This document identified existing and potential damage mechanism and recommended locations for eddy current examinations. In addition, owner-elected supplemental examinations were performed. The necessary examination techniques and areas of applicability were documented and compliant to the EPRI "Pressurized Water Reactor Steam Generator Examination Guidelines" Revision 6 #1003138. The eddy current examination base scope consisted of the following planned and expanded examinations in each steam generator. Bobbin Coil: Full length from Tube-end Inlet to Tube-end Outlet. All tubes not inspected in the 2005 examination. This was an approximately 50% sample. All peripheral tubes (2 tubes deep) including tube lane locations (rows 1,2,3 and 4) Previous signals of interest, MBM, DEP, DNT, PRS (previous S-codes identified) The total is approximately 58% sample in each steam generator. 28

2008 Refueling Outage R.E. Ginna Nuclear Power Plant Steam Generator Examination Report Three Coil Straight Body Design Rotating Probe: Top-tube-sheet (TSH or TSC+/- 3") Special Interest (SI) An approximately 25% sample top of tube sheet hotleg at the expansion interface. There were 80 tubes tested in the cold leg of S/G-B for denting at the top-of-tube sheet. All identified over-expansions from the baseline examination were included within this sample as well as one baseline nonconformance in S/G B from TSH-TEH. Single Coil U-bend +PT Rotating Coil A 20% sample of rows 1 & 2 U-Bend region. Any special interest in U-Bend region. Scheduled supplemental testing and those as a result of Examination findings. 0 Rotating coil of all I-codes identified by bobbin coil. 0 Bobbin coil and/or RPC testing of tubes surrounding a potential loose part signal. 20% sample of reported dings/dents, > 2 Vpp degressing from largest voltage 8H-TEH set as priority. 20% sample of reported MBM's, > 5 Vpp digressing from largest voltage, 8H - TEH set as priority area. 100% of previous PRS calls. S/G A has 13 tubes and S/B has 12 tubes. Secondary side examination type S/G A S/G B Pre & post sludge lance visual examination of top of tubesheet 100% 100% annulus, no tube lane and sub components Pre & post sludge lance visual examination sampling of top of 100% of 100% of tubesheet inner bundle (tri pitch tube lanes) to confirm ET detected detected signals and general information, indications indications as well as a as well as a general general sampling sampling Feed ring visual examination YES NO Upper Internals visual examination including moisture NO YES separators, structural components, welds, tube supports, etc.(rotating frequency) _____1 29

2008 Refueling Outage R.E. Ginna Nuclear Power Plant Steam Generator Examination Report Item b. Active degradation mechanisms found. No active degradation mechanisms were reported during the end of cycle 33 2008 RFO steam generator tube examinations. R.E. Ginna Nuclear Power Plant has not detected active degradation since steam generator replacement in 1996. Item c. Nondestructive examination techniques utilized for each degradation mechanism. For the end of cycle 33 2008 RFO steam generator tube examinations the bobbin coil was used as the primary method of degradation detection except in the areas of the tube (low row U-bends, tube expansions) where the bobbin coil has not been demonstrated as being qualified for use. In those areas of the tube, plus point probes were used. Plus point probes were also used to re examine any ambiguous bobbin coil signals. All examination techniques utilized during the 2008 RFO are qualified for detection for potential and relevant degradation mechanisms listed in the R.E. Ginna Nuclear Power Plant 2008 RFO degradation assessment. In addition owner-elected supplemental examinations were performed with qualified techniques. This qualification is in accordance with the EPRI Steam Generator Examination Guidelines, Revision 6. Examination techniques for detection of degradation are tabulated below. 30

2008 Refueling Outage R.E. Ginna Nuclear Power Plant Steam Generator Examination Report Inspection Degradation Techniques Mechanisms EPRI ETSS # Tube Location Probe Type Wear: 96004.1 Tube support intersections Bobbin Fan Bar 96910.1 Tube support intersections Plus Point TSP 27091.2 Wear: Wear ETSS for Categories Sizing Tube Location Probe Type Loose Parts Circumferential 27901.1/2 Above tubesheet Plus Point Groove Tube-to-tube: Previous Axial Groove 27902.1/2 Above tubesheet Plus Point "PRS" (13 Tapered 27903.1/2 Above tubesheet Plus Point tubes in A-S/G Football and 12 tubes in Tapered Round 27904.1/2 Above tubesheet Plus Point B-S/G) Hole and any Flat 27905.1/2 Above tubesheet Plus Point new "PRO" Tapered 27906.1/2 Above tubesheet Plus Point proximity tubes 450 Tapered 27907.1/2 Above tubesheet Plus Point Alternate 21998.1 Above tubesheet Plus Point Method 31

2008 Refueling Outage R.E. Ginna Nuclear Power Plant Steam Generator Examination Report Degradation Inspection Techniques Mechanisms EPRI ETSS # Tube Location Probe Type MBM 96010.1 Free span Bobbin (manufacturing Anomaly) 21409.1 A 21410.1 C Free span Plus Point 20511.1 AID OXP 20510.1 C ID Tubesheet region Plus Point 24013. 1, <5V Free span Bobbin DNG (Freespan) 21409.1 A 21410.1 C Free span Plus Point DNT 21409.1 A 21410.1 C Tube support intersections Plus Point PWSCC: Row 1 and Row 2 U-96511.2 U bend region Plus Point Bend 21409.1 A OD TubeshCet 21410.1 C OD Tubesheet region Plus Point Expansion 20511.1 A ID ExpansC=ionum 205a10.1 C IDl A =axial; C = circumferential Item d. Location, Orientation (if linear), Service-induced Indications. and Measured Sizes (if Available) of R.E. Ginna Nuclear Power Plant detected the following three service induced indications in two tubes during the 2008 RFO. These service induced indications were the first indications of degradation since steam generator replacement in 1996. Attached is the final report printout for each tube with inspection history for each. Steam Generator A Volumetric wear was detected at one tube support plate intersection in SG A. This was identified in tube R91 C51 at support 05H. The depth was 29% sizing. The length of the indication was 0.34 inches, and the circumferential extent or ARC is measured at 0.38". Eddy current analysis indicated that the wear observed was caused by the presence of a loose part that had been located between the tube and the support grid, but had since moved elsewhere. The low frequency eddy current response was closely scrutinized at the support structure, this information confirmed the nominal tube support design with no additional foreign material. Steam Generator B Volumetric wear was detected at one tube support plate intersection in SG B. This was identified in tube R78 C24 at support 01H. Two wear locations were identified on opposite sides of the 32

2008 Refueling Outage R.E. Ginna Nuclear Power Plant Steam Generator Examination Report tube. On one side was an indication measured to be 18% deep. This is called Indication 1. The indication had a length in the axial direction of 0.33 inches, and the circumferential extent or ARC is measured at 0.26". Eddy current analysis indicated that the wear observed was caused by the presence of a loose part that had been located between the tube and the support grid, but had since moved elsewhere. The low frequency eddy current response was closely scrutinized at the support structure, this information confirmed the nominal tube support design with no additional foreign material. The second indication measured to be 10% deep tapered wear. This is called Indication 2. The length of the indication was 1.22 inches, and the circumferential extent or ARC is measured at 0.16". ECT analysis indicated that the wear observed did conform to expectations of support wear. It was hypothesized to be caused by the presence of a loose part that had been located between the tube and the support grid on the opposite side of the tube described above in indication 1. This is based upon the foreign material wear indication and tube support wear indication being at the same elevation with no tube wear history from historical data. Tube support wear will reside directly over a tube support structure, where loose part wear will be offset from a support structure. The R.E. Ginna power plant lattice grid tube support design provides a point of contact on a tube and provides a good basis for this decision. 33

2008 Refueling Outage R.E. Ginna Nuclear Power Plant Steam Generator Examination Report Item e. Number of Tubes Plugged During the Inspection Outage for Each Active Degradation Mechanism. There were no tubes removed from service by plugging during 2008 RFO Inservice examination. Item f. Total Number and Percentage of Tubes Plugged to Date. The following is a historical tube plugging summary since steam generator manufacture. R.E. Ginna Nuclear Power Plant Plug History A Steam Generator B Steam Generator Outage Year Number of % Bobbin Number of % Bobbin Plugs Inspected Plugs Inspected Baseline 1996 1 100% 1 100% 19971SI 0 100% 0 100% 19991SI 0 >50% 0 0 >50% 200251SI 0 >50% 0 >50% 20051ISI 0 >50% 4 >50% 2008 ISI 0 >50% 0 >50% Total 1 5 Item g. The Results of Condition Monitoring, Including the Results of Tube Pulls and In-situ Testing. In accordance with the R.E. Ginna Nuclear Power Plant Steam Generator Program, a condition monitoring evaluation was conducted. As indicated above in Item e, there were no indications of degradation exceeding the R.E. Ginna Nuclear Power Plant Technical Specification repair limits or the R.E. Ginna Nuclear Power Plant administrative plugging criteria identified. All steam generator performance criteria were satisfied. No tube pulls or in-situ pressure testing were required based upon the examination results. The applicable requirements of the EPRI Steam Generator Integrity Assessment Guidelines Revision 2, Section 10.5 were applied to all fosar examinations. Two foreign objects were successfully removed from the top of the tubesheet, one piece of flexitallic gasket was removed from the A steam generator, and one piece of wire was removed from the B steam generator. Plus point examinations revealed no additional objects or evidence of tube degradation. 34

2008 Refueling Outage R.E. Ginna Nuclear Power Plant Steam Generator Examination Report As noted in section 3.0 additional PLP indications on the top of the tubesheet were detected by a combination of both eddy current and visual examinations. All PLP indication locations were confirmed through visual examination and determined whether foreign material existed. These locations were bound with plus point and determined to have no degradation associated with these small foreign materials. An engineering assessment was performed for these specific foreign materials, steam generator locations, and R.E. Ginna Nuclear Power Plant specific operating conditions. Based upon this evaluation, and the fact that no wear had occurred it is reasonable to conclude that the required conditions to promote wear do not exist. These locations will be tracked in future inspections. 35}}

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