ML081200258

Condition Monitoring & Operational Assessment RFO-15 SG-SGDA-02-45 September 2003 Westinghouse Electric Co

ML081200258
Person / Time
Site:	Indian Point
Issue date:	09/15/2003
From:	Goldstein J, Ingram D Entergy Nuclear Operations
To:	Office of Nuclear Reactor Regulation
References
FMX-00323, Rev 0
Download: ML081200258 (48)

Text

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ATTACHMENT 9.11A IP-2 CALCULATION TRANSMITTAL AND DISTRIBUTION MEMORANDUM IP-2 MEMORANDUM Date: 9115/03 From: D. Curt lnqram To: Administrative Services (Document Control)

Subject:

Calculation No. FMX-00323 Revision 0.r Calculation Title Indian Point Unit 2 Condition Monitoring and Operational Assessment RFO-15 SG-SGDA-02-45 SeDtember. 2003 Westinahouse Electric Companv QA Classification A 0FP a MET 0NON-CLASS IP-2 Calculation Database Information Calculation Status: 0Preliminary 1Pending 0Voided As Built 0Superseded Transmitted herewith, please find the original approved documentation of the subject calculation for distribution, indexing in the IP-2 Calculation Database and retention in the Optical Imaging System.

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CZ-ENN NUCLEAR MANAGEMENT MANUAL QUALITY RELATED ADMINISTRATIVEPROCEDURE ENN-DC-126 Revision 1 INFORMATIONAL USE Page 3 of 45

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ATTACHMENT 9.2 CALCULATION P

COVERPAGE CALCULATION COVER PAGE Calculation No. FMX-00323 Revision 01 Sheet 1 of 46

Title:

Indian Point Unit 2 Condition Monitarinq and OPerational (XI QR Assessment RFO-15 SG-SGDA-02-45 SeDtember, 2003 Westinahouse Electric Cornpanv 0NQR

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Design Basis Calculation?

Discipline: Enaineerinq Proarams nYes NNo

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This calculation supercedes/voids calculation: NA Modification No./Task No/ER No:J/N R

No software used Software used and filed separately (Include Computer Run Summary Sheet)

Software used and filed with this calculation System No./Name: RCS Component No./Name: 2 1SG. 22SG,23SG,24SG (Attached additional pages if necessary)

Print / Sign 1 STATUS I OTHER REV# 1 (Prei.

Pend,A, 1 I REVIEWEWDESIGN VERIFIER REVIEWER)

DESIGN 1 APPROVER DATE VERI FlER D.Curt lngram Jeffrey Goldstein

ENN QUALITYRELATED ENN-DC426 Revision 1 lllr 25s NUCLEAR ADMINISTRATIVEPROCEDURE

--- EnteFgy MANAGEMENT MANUAL lNFORMATlONAL USE Page 3 of 41 ATTACHMENT9.6 CALCULATION RECORD P

OF REVISIONS RECORD OF REVISIONS Calculation Number: FMX-00323 Page 2 of 46 F--

Revision No. Description of Change Reason For Change Original Issue NA P-A conservative BOC indication of 20% was C R- IPZ-2003-05542 assumed in section VI.B.l I---

I A n

-hk?.f@fl ENN NUCLEAR MANAGEMENT MANUAL I

QUALITYRELATED ADMINISTRATIVE PROCEDURE lNFORMATlONAL USE I

ENN-DC-126 Page 28 Revision 1 of 45 I I I ATTACHMENT 9.4 CALCULATION

SUMMARY

PAGE CALCULATION

SUMMARY

PAGE Page3 of@

Calculation No. FMX-00323 Revision No. 0.1 CALCULATION OBJECTIVE: This report provides a summary of the Indian Point 2 steam generator tube integrity condition determined in the year 2002 refueling outage RFO 15 by NDE inspection and a projection by analysis of the tube integrity until the next planned steam generator inspection.

CONCLUSIONS: All indications in this inspection were below the calculated integrity limits and therefore met integrity requirements without further testing. Based on the inspection results, all four SGs were found in compliance with CM requirements provided in Ref 1 and 2. OA for an assumed operation duration Of 4.0 EFPY for Cycle 16 and 17 confirms that the SG tube structural and leakage integrity will be maintained until t h e next planned SG inspection.

ASSUMPTIONS: Calculated integrity limits, including consideration for appropriate uncertainties, burst and leak analytical correlations, material properties, and NDE technique and analyst uncertainties were provided in the degradation assessment report (Reference 5).

DESIGN INPUT DOCUMENTS: N/A AFFECTED DOCUMENTS: N/A METHODOLOGY: The inspection at RFO 15 is at the end of the first fuel cycle after replacement, therefore all four steam generators were inspected in the current inspection. A Condition Monitoring assessment was performed, on a defect specific basis, by demonstrating compliance with integrity criteria through comparison of RFO 15 NDE measurements with calculated burst and leakage integrity limits.

Westinghouse Electric Nuclear Services Executive Summary SGSGDA-02-45, Rev. I Page 2 of 43 EXECUTlVE

SUMMARY

This report provides a summary of the Indian Point 2 steam generator tube integrity condition, determined in the year 2002 refbeling outage (RFO 15) by NDE inspection, and a projection by analysis of the tube integrity until the next planned steam generator inspection. The next inspection is planned for RFO I7 in 2006, which is at the completion of two he1 cycles. All of the activities reported in this report have been conducted in accordance with NEI 97-06 Revision 1 (Reference 1 ) and associated guidelines (References 2 and 3).

The RFO 15 inspection occurred at the end of the first fuel cycle after the original steam generators were replaced, therefore all four steam generators were inspected. A Condition Monitoring assessment was performed, on a defect specific basis, by demonstrating compliance with integrity criteria through a comparison of RFO 15 NDE measurements with calculated burst and leakage integty limits. Calculated integrity limits, including consideration for appropriate uncertainties, burst and leak analytical correlations, material properties, and NDE technique and analyst uncertainties were provided in the RFO 15 degradation assessment report (Reference 5 ) .

All indications in this inspection were below the calculated integrity limits and therefore met integrity requirements without further testing. Based on the inspection results, all four steam generators were found in Compliance with Condition Monitoring requirements provided in Reference 1 and2.

An Operational Assessment for an assumed operation duration of 4.0 EFPY for Cycle 16 and 17 confirms that the steam generator tube structural and leakage integrity will be maintained until the next planned steam generator inspection.

Westinghouse Electric Nuclear Services Table of Contents SCSGDA-02-45, Rev. 1 Page 3 of 43 TABLE OF CONTENTS EXECUTIVE

SUMMARY

................................................................................................ 2 I INTRODUCTION .............................................................................................................. 5 I1 NDE

SUMMARY

................................................................................................................ 7 111 LOOSE PARTS ASSESSMENT......................................................................................... 8 N CONDITION MONITORING ........................................................................................... 10 A. Defect Specific Approach B. AVB Wear C. Loose Parts Wear D. Volumetric Indications V TUBE REPAIRS ................................................................................................................ 13 VI OPERATIONAL ASSESSMENT ..................................................................................... 14 A. Operational Assessment Overview B. Indication Specific Operational Assessment I. AVB Wear

2. Loose Parts Wear

3. Volumetric Indications

4. Potential Degradation VI1 CM/OA RESULTS AND CONCLUSIONS ............................ ..................................... 17 VI11 REFERENCES .................................................................................................................. I8 Appendix A RFO 15 NDE Inspection Plan. ............................................................................... 19 Appendix B RFO 15 NDE Indication Results Table ................................................................. 20 Appendix C RFO 15 Loose Parts Inventory .............................................................................. 22 Appendix D StabiIization Recommendations ............................................................................. 27 Appendix E NDE Noise Comparisons ....................................................................................... 29

Westinghouse Electric Nuclear Services Table of Conrents SGSGDA.02.45, Rev .1 Page 4 of43 LIST OF FIGURES Figure VI- I Model F Average AVB Wear Rates ...................................................................... 15 LIST OF TABLES Table II- I RFO 15 NDE Summary of Tubes with Indications ................................................. 7 Table lV-1 Wear in SG 2 1 ........................................................................................................ 10 Table Tv-2 Wear in SG 22 ........................................................................................................ 10 Table W-3 Wear in SG 2 .......................................................................................................... 11 Table IV-4 Wear in SG 24 ........................................................................................................ 11 Table IV-5 Volumetric Indications in SG 21 ........................................................................... 11 Table IV-6 Volumetric Indications in SG 23 ........................................................................... 12

Westinghouse Electric Nuclear Services Introduction SGSGDA-02-45, Rev.1 Page 5 of43

1. INTRODUCTION A steam generator tube integrity condition monitoring assessment and an operational assessment have been completed for the Indian Point Unit 2 (IP2) refueling outage RFO 15 and Operating Cycles 16 and 17, sespectively. The assessments have been conducted to meet the requirements and intent ofNEI 97-06 Revision 1 and EPRI guideline documents (References 1 - 3).

This report provides:

0 a summary of the inspection findings; 0 the defect specific condition monitoring assessment of the tube structural and leakage integrity at the end of Cycle 15; and 0 the operational assessment of the tube structural and leakage integrity for Cycles 16 and 17 for all steam generators.

These assessments are based on the RFO I5 inspection results. The inspection scope, as planned and expanded, is summarized in Appendix A.

The results of the steam generator tube inspection and secondary side inspection are summarized in Section II of this report. A list of tube indications detected by the primary side NDE is provided in Appendix B. The final official inspection results are provided in a separate Westinghouse report on the NDE inspection, Reference 4.The results of the secondary side search for loose parts are summarized in Section 111. A detailed list of objects retrieved from the SGs and of objects remaining in the SGs is provided in Appendix C A condition monitoring assessment was performed, on a defect specific basis, by demonstrating compliance with integrity criteria through comparison of reported NDE measurements with calculated pressure or leakage integrity limits as appropriate or the flaw indication type. The indication size calls by NDE were compared to defect-specific burst and leakage criteria specified in the degradation assessment (Rcference 5). All indications were below the calculated integrity limits. Based on RFO 15 inspection results, all steam generators were determined to be in compliance with condition monitoring integrity guidelines as provided by Reference 1. The details of the defect specific condition monitoring assessment are provided in Section IV of this report.

In order to assure that the integrity criteria will continue to be satisfied, all 13 tubes that had AVB wear and the 3 tubes with volumetric indications were plugged. This is a very conservative treatment since no indication depth exceeded 20% of the tube wall thickness.

An operational assessment was performed on a defect specific bask considering the indications detected and the tube repairs performed during RFO 15. The assessment analysis was performed in accordance with the requirements of Reference 1 and the supporting EPRI guidelines (References 2-3). The results of the assessment confinn that the steam generator tube structural and leakage integrity will be maintained until the next scheduled inspection at the end of operation cycle 17, which will be after two he1 cycles of 24 months each. The detaiIs of the operational assessment analysis are provided in Section VI of this report.

Westinghouse Electric Nuclear Services Introduction SGSGDA-02-45, Rev.1 Page 6 of 43 This report has been independently reviewed under the requirements of the Westinghouse Nuclear Services Policies and Procedures Manual, Revision 15, October 1,2002.

Westinghouse Electric Nuclear Services NDE Summary SGSGDA-02-45. Rev. 1 Page 7 of 43 II. NDE

SUMMARY

Appendix A contains the steam generator tube NDE inspection summary plan for RFO 15 and the scope expansions performed.

Appendix B contains the NDE indications found in RFO 15. Wear indications at AVBs were observed in all four steam generators. A total of thirteen tubes had indications of AVB wear. All wear indications were less than or equal to 20% through-wall (YoTW). Although loose parts on the secondary side were identified in all four steam gencrators, no loose part wear was found by eddy current examination. There were 3 freespan volumetric indications detected, These indications were also less than 20% through-wall. Bobbin probe examination of the wear indications measured 20% through-wall depth, but plus point examination showed that the wear was double sided, indicating that the 20% through-wall wear was the sum of shallower wear on two sides. Therefore the through-wall extent at any location was less than 20% through-wall, and wear projections based on the 20% through-wall value are conservative.

Table 11 RFO 15 NDE Summary of Tubes with Indications Data quality was assured by the site specific qualification of Reference 9. Data on ECT noise was obtained for all ECT techniques used and compared to the corresponding data from the preservice inspection and the EPRI ETSS database in Appendix E. The noise levels measured indicate that the detection capabilities of the inspection techniques are consistent with the detection capabilities defined in the corresponding EPRI ETSS.

Westinghouse Electric Nuclear Services Loose Parts Assessmnt SGSGDA-02-45. Rev. 1 Page 8 of 43 111. LOOSE PARTS ASSESSMENT Loose parts were detected in all four steam generators. The larger objects were removed, but many small objects were not retrieved. The list of objects identified and if they were retrieved is provided in Tables C1 through C4 in Appendix C. No tubes were repaired or degraded due to the presence of loose parts. The following is paraphrased from Reference 7.

During the current Fall 2002 (2R15) outage at Indian Point Unit 2, various loose objects were observed during FOSAR of the tubesheet region of the steam generators. Many of the small wires and sludge rocks in the bundle were left due to size, time and personnel exposure to retrieve. Tables C1 through C4 contain a hul listing of all of the objects found during the visual inspection of the steam generators with an indication if the object was removed. The objects that remain inside the S/G can be classified into three general types of objects as indicated below: I Object Type 1 - Sludge rocks/sludge pebbles - Diameters ranging from 118 to 314 inch.

Object Type 2 - Small wireshristles - 1/64 inch diameter with maximum length of 1.5 inches Object Type 3 - Metallic mass, oval shaped- -1/2 inch x 1/4 inch The putpose of Reference 7 is to document calculations performed to determine the effects of leaving these objects in the steam generators. These calculations have been performed specifically for the most limiting of each type of object, however the results of this limiting calculation would conservatively envelope the less limiting objects. The amount of time required for these objects to wear a tube down to a minimum, yet still acceptable, tube wall thickness has been determined. The minimum acceptable lube wall thickness is the same as the structural repair limit or tube plugging limit of 40% depth or 60% remaining.

The analysis has been completed assuming that objects are located at the tube that exhibits the limiting amplitudes of vibration and cross flow fluid velocity, and that this tube has existing 20%

through wall degradation. The evaluation also assumes that the objects rests upon a sludge pile approximatdy 6 inches deep (at this elevation, tube vibration amplitudes are larger than those at the top of the tubesheet). This is very conservative because the actual sludge pile is less than 1/4 inch deep. Additional conservative assumptions are made in that the object will remain in the same location (once tube wear begins), and that only the tubes will experience wear. With respect to the steam generator operating conditions, the effects of the 1.4% power uprate have also been considered.

As indicated above the analysis has been performed assuming that a small degree of undetected wear (20% depth which is very conservative or the limit of detection) has occurred on the Iimiting tube. The assumption of essentialIy no tube degradation has been verified via eddy current inspection of the steam generators.

The evaluation shows that the time required for the limiting sludge rock type object to wear a tube down to 40% through wall (assuming :20% initial hibe wear) is greater than 2 operating cycles.

Westinghouse Electric Nuclear Services Loose Parts Assessmnt SGSGDA-02-45, Rev.1 Page 9 of 43 The evaluation shows that the time required for the limiting wire bristle type object to wear a tube down to 40% through wall (assuming 20% initial tube wear) is greater than 2 operating cycles.

The evaluation shows that the time required for the metallic mass rock-like mass, object type 3, to wear a tube down to 40% through wall {assuming 20% initial tube wear) is greater than 2 operating cycles.

An analysis has also been performed to determine the effect on the tubes should the material begin to migrate and repeatedly contact the tubes (impacting only analysis). A conservative analysis was performed assuming that the maximum possible weight of the largest loose object is less than 0.1 Ibs. This weight conservatively bounds the weight of any of the known objects remaining inside the S/Gs. The analysis has determined that the energy that the postulated loose object would impart on a tube during repeated collisions is sufficiently low that significant deformation of the tubes due to impacting only is not expected.

In summary, the analysis has determined that continued SIG operation with the objects known to be present in the secondary side as described in Tables 1 through 4 will not adversely affect the steam generator for the next two 24 month operating cycles. At the end of this period of time, FOSAR should be performed to attempt to remove these objects from the steam generators if they are still in the same location in the steam generators, or eddy current testing should be performed to detect the presence of wear.

Westinghouse Electric Nuclear Services Condition Monitoring SGSGDA-02-45, Rev.1 Page 10 of 43 IV. CONDITION MONITORING A. Defect Specific Approach A defect specific approach based on EPRI Steam Generator Tube Integrity Report (Reference 2 ) was used to determine steam generator tube integrity. Consistent with the approach identified in the pre-outage degradation assessment report (Reference 5), an assessment based on NDE depth was used for wear degradation forms detected. The degradation assessment provided information for addressing a number of potential degradation types. The information provided in this report is focused on the degradation types actually detccted in the RFO 15 inspection.

B. AVB Wear AVB wear was detected in a11 four steam generators. The location and indication depth are shown in Tables IV- 1 through IV-4.

TabIe IV Wear in SG 21 TabIe IV Wear in SG 22

Westinghouse Electric Nuclear Services Condition Monitoring SGSGDA-02-45, Rev.1 Page 1 1 of 43 Table IV-3 -. Wear in SG 23 Table IV Wear in SG 24 The deepest wear indication is 20% of the tube wail thickness, which is well below the 63% condition monitoring limit developed in the degradation assessment. The plus point determination that the deepest indications were two-sided wear adds additional conservatism. Therefore all AVB wear indications meet the condition monitoring structural integrity and leakage criteria.

C. Loose Parts Wear Although loose parts were found as discussed in Section 111, no tube degradation due to the loose parts was found. Therefore there is no challenge to the tube integrity and leakage criteria due to loose parts.

D. Volumetric Indications Three freespan volumetric indications were detected and are summarized in Tables IV-5 and IV-6.

Table IV Volumetric Indications in SG 21

Westinghouse Electric Nuclear Services Condition Monitoring SGSGDA-02-45, Rev. 1 Page 12 of 43 Table IV Volumetric Indication in SG 23 All three volumetric indications are well below the condition monitoring limits for structural and leakage integrity established in the degradation assessment. Experience with other plants has shown that tubes with Alloy 600 TT material have a tedency to develop freespan indications during the first cycle of operation. At IP2 many signals were detected by bobbin that did not exist in the preservice inspection. All of these indications were tested with plus point and only three were reported as volumetric indications by plus point. There is no basis for suspecting that these indications are due to a corrosive mechanism. It is presumed that these volumetric indications were caused by a transient loose part, or perhaps were deep buff marks that became indications after the sustained heating of the operation cycle.

Westinghouse Electric Nuclear Services Tube Repairs SGSGDA-02-45, Rev.1 Page 13 of 43 V. TUBE REPAIRS All thirteen tubes with AVB wear indications were plugged even though the wear depth was significantly below the pIugging Iimit.It is recommended that the tubes with wear indications at all four AVB supports (SG 21 R45 C45 and SG 23 R41 C46) be stabilized. This stabilization does not need to be accomplished for at least two operationa1 cycles as discussed in Reference 8, which is presented in Appendix D.Wear projection analyses are required to provide a more quantitative assessment of the latest time to install the cable dampener in the two tubes at issue.

All three tubes with freespan volumetric indications were plugged even though the indication depth was significantly below the plugging limit.

Westinghouse Electric Nuclear Services Operational Assessment SGSGDA-02-45,Rev.l Page I4 of 43 VI. OPERATIONAL ASSESSMENT A. Operational Assessment Overview The operational assessment of the steam generators requires the consideration of growth of degradation to assess if the structural and leak integrity of the Indian Point 2 steam generators will be maintained during the next operation cycles. The only observed degradation with potential growth is the AVB wear. Operating condition changes for a potential power uprate of 1.4% would have negligible impact on degradation since AVB wear is not temperature dependent, and flow changes should be minimal. Since the tubes with wear indications were plugged, and the wear rate is expected to show a significant decrease with operating time, it is reasonable to expect that the Indian Point 2 steam generator tube structural and leakage integrity requirements will be maintained through the next two operating cycles.

The Operational Assessment process of evaluating the condition of the steam generator tubes during the next cycles of operation requires that the steam generator tubes meet specified performance criteria that provide reasonable assurance of adequate tube structural and leakage integrity at the end of the next inspection interval.

The operational assessment consists of four parts:

1. the size of the largest flaws of each type that are present at the beginning of the next operating cycle.

2. the estimation of the size of the largest flaws that will be present at the end of the next inspection interval.

3. the demonstration that the largest flaw of each type will meet the structural integrity criterion at the next inspection.

4. the demonstration that at the elid of the next inspection interval the total leakage at hlSLB conditiorls will meet the leakage criterion.

This Operational Assessment of the IP2 steam generators is based on the inspection results summarized in Section 11 of this report and provided in Appendix B. This assessment has been conducted in accordance with the requirements of NE1 97-06 Rev. 1 and EPRI Tube Integrity Assessment Guidelines (References I and 2). This report provides a demonstration based on the NDE data acquired that the structural and leakage criteria will be maintained in all steam generators throughout the estimated 4.0 effective h l l power years of Cycles 16 and 17.

B. Indication Specific Operational Assessment

1. AVB Wear The Tube Integrity Guidelines (Reference 2) suggest using the largest flaw left in service or the depth of flaw with a fraction detected < 95% from the appropriate ETSS. A11 tubes exhibiting AVB Wear were plugged. The fraction detected for

Westinghouse Electric Nuclear Services Operational Assessment SGSGDA-02-45, Rev. I Page 15 of 43 flaws in the 0% to 19% depth range in the ETSS 96004.1 is 100% with the minimum flaw depth in the 0% to 19% depth population given in the ETSS as 4%. Therefore a 4% depth could be used as the assumed beginning of cycle (BOC) indication. A conservative estimate of the maximum depth of wear scar left in service would be the maximum depth detected. The maximum wear detected was 20% (Tables IV-1-4) through-wall in one cycle. If the same rate would continue to occur in a tube with an initial depth of 20%, then the depth could be as large as 20% t- 20% x 2 = 60% through-wall at the end of the next two cycles which would be below the condition monitoring limit of 63% (which includes consideration of NDE uncertainties). Therefore either tube integrity guidehe option satisfies thc operational assessment requirement.

Wear rates, however, are known to decrease with time as shown in Figure VI- 1 (a similar figure was presented in Reference lo). Therefore the assumption of constant growth leads to a very conservative degradation prediction. In addition as discussed previously, the plus point showed that the wear was two sided, providing additional conservatism to the projection of through-wall wear.

Figure VI-2 Model F Average AVB Wear Rates 30 4

+ I

+ + y = - 6 7586Ln(x) + 23.601 0)

U

-Log. (Average) m 0

1 I j

0 5 10 15 Operation T i m e ( E F P Y )

In the Figure legend, Average means the average growth rate for a particular model F plant at the given operation time. The value of the model F plant growth rates may not be representative of IP2 growth rates, but the trends with operating time are expected to be similar because the AVB design and materials are similar.

The IP2 Vol points on the Figure show the prediction of wear rate for IP2 starting at the maximum measured value of 20% through-wall in the first cycle, and continuing in time as a constant volume wear rate. Both the model F data and

Westinghouse Electric Nuclear Services Operational Assessment SGSGDA-02-45, Rev. 1 Page I6 of 43 the assumption of constant volume wear rate indicate a significant reduction in growth rate with operating time. The similarity of the slope of the curve fitted to the model F data and the slope of the IP2 Vol points shows that the constant volume wear rate is a reasonable explanation for the observed reduction of wear rate with increasing operation time. This reduction of the growth rate with operating time provides additional assurance of the acceptability of operation until the next planned steam generator inspection. Appendix D addresses the issue of the potential for continued wear after plugging and the need to stabilize the tubes at a hture time.

2. Loose Parts Wear No wear was detected due to loose parts. The location of all possible Ioose part signals were reviewed from the secondary side on tape, or re- looked to veri@ no foreign object was on the tube. If no object was present the cause of the signal is likely to be sludge. If an object was observed it was removed if possible or left if it was unable to be retrieved. The wear that could result from the loose parts that were identified and remain in the steam generators was evaluated in Section Ill.

The result of that evaluation is that wear conservatively predicted will not reach the plugging limit for more than 2 cycles of operation. Therefore the operational assessmcnt integrity criteria will be satisfied for at least the next two cycles of operation.

3. VoIumetric Indications The tubes with freespan volumetric indications detected were plugged. Tubes with Alloy 600 TT material have a tendency to develop freespan indications during the first cycle of operation. At IP2 many signals were detected by bobbin that did nut exist in the pre-service inspection. All of these indications were tested with plus point and only three were reported as volumetric indications by plus point. There is no basis for suspecting that these indications are due to a corrosive mechanism. It is presumed that these volumetric indications were caused by a transient loose part. or perhaps were deep buff marks which became indications after the sustained heating of the operation cycle.

4. Potential Degyadation Potential degradation modes described in the Degradation Assessment, Reference 5 , were not detected and are not anticipated to develop during the next operating cycles. Therefore those degradation modes are not expected to impact the integrity of the steam generators.

Westinghouse Electric Nuclear Services CM/OA Results and Conclusions SGSGDA-02-45, Rev.! Page !7 of 43 VII. CM/OA RESULTS AND CONCLUSIONS A condition monitoring assessment for the 2002 refueling outage and an operational assessment have been completed.

A total of 16 tubes with voIumetric indications were detected by NDE including:

1. 13 tubes with Wear at AVBs

2. 3 tubes with Volumetric Freespan Indications All indications had measured depths well below their respective structural and leakage condition monitoring limit. Therefore all detected indications were determined to meet the performance criteria recommended in NE1 97-06, Revision 1. The Indian Point 2 Condition Monitoring for RFO 15 complies with the guidance of NE1 97-06, Revision 1, and demonstrates that the structural and leakage criteria for all steam generators are satisfied.

The operational assessment demonstrated that there is reasonable assurance that the two observed degradation modes will not lead to a predicted flaw of any type that would equal or exceed its respective structural or leakage limit at the end of the next two operation cycles. The Indian Point 2 Operational Assessment for all fbur steam generators complies with the guidance of NE1 97-06, Revision 1, and demonstrates that the Indian Point 2 Steam Generators are expected to continue to meet the leakage and struchual integrity performance criteria for the duration of Cycles 16 and 17.

Westinghouse Electric Nuclear Services References SGSGDA-02-45, Rev. 1 Page 18 o f 43 VIII. WFERENCES

1. NEI 97-06, Rev. 1, Steam Generator Program Guidelines, January, 200 1.

2. EPRI Report TR- 10762I, Rev. 1,Steam Generator Integrity Assessment Guidelines:

Revision 1, March 2000.

3. EPRI Report TR- 107569-VIR5, PWR SC Examination Guidelines: Revision 5, September 1997.

4. Westinghouse Report, Indian Point 2 RF015 Fall 2002 RSG Eddy Current Dataroom Summary, Rev. 0, I1/20/02.

5. Westinghouse Report SGSGDA-02-29, Rev.0, Steam Generator Degradation Assessment for Indian Point Unit 2 RFO 15, September 2002.

6. Westinghouse Field Service Report, Indian Point 2 RSG Pre-Service Inspection, Summer 2000.

7. Westinghouse Letter, LTR-SGDA-02-369, Revision 1, Evaluation of Foreign Objects in the Steam Generators at Indian Point Unit 2 During 2R15, November 22,2002.

8. Westinghouse Letter, LTR- SGDA-02-372, Revision I, Stabilization Recommendations to Address Indian Point Unit 2 AVB Wear Identified During RFOl5, November 15, 2002.

9. Westinghouse Report, MRS-TRC- 1291 , Rev. 0, Use of Appendix H Qualified Techniques at Indian Point Fall 2002, September 19,2002.

10. Extended Inspection Intervals, by H. Lagally and R. Lieder, EPRI Steam Generator NDE Conference, July 200 1.

Westinghouse Electric Nuclear Services Appendix A SGSGDA-02-45, Rev. 1 Page 19 of 43 APPENDIX A RFO 15 NDE Inspection Plan The planned scope for the IP2 RFO 15 outage was as follows:

100% bobbin in all 4 SIGs except U-bend portion of Row 1 and 2 100% Row 1 and 2 U-bend plus point RPC in all 4 S/Gs 20% H/L TTS +3 inches plus point RPC 100% H/L TTS periphery rt3 inches plus point RPC 20% H/L DNRs plus point RPC 40% Dings > 5 volts plus point RPC (consistent with ALARA & Schedule) 100% of bobbin I codes 4 selected tubes from PSI for plus point RPC (Reference 6):

SG 21: R28C50 VOL at TSC _. 1.81 SG 24: R22C80 VOL at 4C + 32.42 SG 2 1 : R9C89 IDV at 6H - 2.2 1 SG 23: R2C33 IDV at TSH - 2.28 RFO 15 NDE Inspection CompIeted Scope The completed scope for the IP2 RFO 15 outage was as follows (Detailed description is available in Reference IO):

100% bobbin in all 4 SIGs except Ubend portion of Row I and 2 100% Row 1 and 2 U-bend plus point JXPC in all 4 S/Gs 20% H/L TTS f 3 inches plus point RPC (26 - 27% when including periphery) 100% WL TTS periphery +_3 inches plus point RPC 100% H/L DNRs plus point RPC (Total of 9) 100% Dings > 5 volts plus point RPC (Total of 2 1) 991 special Irderest ( included 1074 bobbin I codes) o The IDVs from the PSI wcre selected because this code is not used in present guidelines and they were signals of interest or tracking purposes. All ended up as either NDF or reported as TRA (trackable anomalies).

5 1 TTS plus point RPC to bound Possible Loose Parts (PLP) 100% C/L TTS periphery +3 inches plus point RPC (In order to perform a more complete search for loose parts)

Wcstinghouse Electric Nuclear Services Appendix B SGSGDA-02-45,Rev.l Page 20of43 APPENDIX Is - NDE Indication Results Table Table B-l - Indication Results for SG 21

Westinghouse Electric Nuclear Scrvices Appendix B SC-SGDA-02-45, Rev. I Page 2 1 of 43 Table B Indication Results for SG 22 Table B Indication Results for SG 23 Table B Indication Results for SG 24

Westinghouse Electric Nuclear Services Appendix C SGSGDA-02-45, Rev. 1 Page 22 of 43 APPENDXX C Loose Part Inventory

Westinghouse Electric Nuclear Services Appendix C SGSGDA-02-45, Rev. 1 Page 23 of 43

. Appendix c Westinghouse Electric Nuclear Services SGSGDA-02-45, Rev.1 Page 25 of 43

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Westinghouse Electric Nuclear Services Appendix D SGSGDA-02-45, Rev.1 Page 27 of 43 APPENDIX D Stabilizitation Recommendations To: David J. Ayres Date: November 15,2002 cc: R. J. Sterdis P. R. Nelson From: Hermann 0. Lagally Your ref:

Ext: WIN 224-5082 Our ref: LTR-SGDA-02-372, Rev. 1 Fax: WIN 224-5889

Subject:

Stabilization Recommendations to Address Indian Paint Unit 2 AVB Wear identified During RF015 During the RFOl5 inspection of Indian Point 2, AVB wear was reported in thirteen tubes in the 4 SG as shown on the following table.

Indian Point eIected to administratively plug these tubes to maintain the option for an extended operating period until the next inspection of the SGs. Westinghouse recommended installation of a cable darlpener in tubes 2 1 -45/45 and 23-41/46, but noted that the installation of the dampener was optional at this time. The purpose of this letter is to discuss the rationale for this recommendation and to clarify the timing for installation of the cable dampener.

Tube wear at the AVBs continues after a tube is removed from service due to AVB wear. Deplugging of tubes at D.C. Cook, Millstone and Diablo Canyon some years after the original plugging, for reasons unrelated to AVB wear, has proved this by the discovery of water contained in the tube and inspection showing that the AVB wear had progressed to a throughwall condition. I t is possible that an aggressively vibrating tube could wear through the cross section of the tube sufficiently SO that the remaining section could break due to fatigue, and the tube ends become

Westinghouse Electric Nuclear Services Appendix D SGSGDA-02-45, Rev. 1 Page 28 of 43 damage propagation mechanisms to adjacent, active tubes. The criterion that is used by Westinghouse for stabilization recommendations is prevention of tube-to-tube contact. The potential for tube-to-tube contact from AVB wear is not expected to occur at IP2 during the next two operating cycles.

Westinghouse has performed Wear Projection analyses for a iiumber of operating SGs to determine the need for installation of cable dampeners in tubes that were repaired for AVB wear. The experience from these prior analyses shows that tubes that are recommended for stabilization are usually those that are plugged early in life. Further, the available operating data show that the characteristics of the most aggressively wearing tubes are similar - they have indications at all of the AVB contact points. Thus, the judgment was made that the two Indian Point tubes exhibiting wear at all four interscctions would eventually be idcntified as the tubes requiring installation o f a cable dampener.

From prior experience, the rate of the continuing AVB wear process allows for installation of a cable dampener in a subsequent outage, and the Westinghouse recoinmendation for stabilization does not suggest an immediate action.

Rather, the recommendation is intended to provide flexibility such that the dampener can be installed in a later outage if needed to allow for adequate planning time.

If a linear depth wear rate is conservatively assumed, an additional four operating cycles would be required to attain a 100% throughwall condition. Additional wear would be required to reach a through-section condition that would result in tube-to-tube contact. Thus, it is conservatively estimated that the current IP2 condition has significant margin to the critical condition where tube separation could be a concern. The potential for tubs-to-tube contact is not expected to occur during the next two operating cycles. Wear Projection analyses are required to provide a more quantitative assessment of the latest time to install the cable dampener in the two tubes at issue.

Author: Verifier; Herrnann 0 . Lagally Kim J. Romanko S/G Design & Analysis S/G Design & Analysis

Westinghouse Electric Nuclear Services Appendix E SGSGDA-02-45, Rev.1 Page 29 of 43 APPENDIX E NDE Noise Comparisons Noise measurements have been obtained to compare the noise in the eddy current signals in the current inspection to the noise in the preservice inspection and to the noise in the corresponding EPRT ETSS data. The following data tables illustrate the noise levels for the various inspection techniques. Reference 9 provides the preservice inspection data and the EPRl ETSS data. In all cases, the noise is essentially unchanged from the preservice inspection. In all cases except the U-bends, the average of the noise is less than the EPRI ETSS data. In the case of U-bends, the 95'h percentile is less than the 95thpercentile of the noise in the EPRI ETSS dataset. For Ubend data, however, an additional acceptance criterion was established whereby no row 1 or 2 U-bend could exceed 0.65 volts (vert. max.) at the apex. This criterion was conservativeIy based on the 95% confidence of 0.71 volts from the EPRl measurements and experience at Indian Point 2. In addition, if the signal at the Ubend was greater than 0.50 volt at the apex, the resolution analyst, as well as the primary and secondary analysts, was also required to analyze the entire Ubend.

The noise levels measured indicate that the detection capabilities of the inspection techniques are consistent with the detection capabilities defined in the corresponding EPRI ETSS.

Westinghouse Electric Nuclear Services Appendix E SGSGDA-02-45, Rev.l Page 30 of 43 2H 0.69 0.8 0.15 0.22 22 40 1I1 0.92 1.07 0.2 0.18 2H 0.59 0.63 0.17 0.74 -

21' 43 1H 0.49 0.61 0.22 0.25 2H 1.01 I .32 0.17 0.22 I

~ ~~~

AVG. 0.73 0 90 0.22 0.23 SD 0.25 0.26 0.03 0.02 95% 1.14 1.33 0.28 0.27

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Westinghouse Electric Nuclear Services Appendix E SGSGDA-02-45, RW.1 Page 32 of 43 Table E l C - Bobbin Noise at Supports: EPRI ETSS DATA 1#96008.1 1 SD I 0.73 I 0.53 I I I 95% I 2.72 I 1.65 I

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e Westinghouse Electric Nuclear Services Appendix E SG-SGDA-02-45, Rcv.1 Page 39 of 43 Table E 3 C - TODof Tubesheet and Freespan: EPRI ETSS DATA

~~

ETSS # NIA NIA 0.53 0.30 21410.1 ETSS # N/A 0.64 NIA 1.18 205 10.1 ETSS ## NIA 0.54 NA 1.41 20511.1 I I I I 1 1

• For the EPRI ETSS data, the max noise was recordcd wherever it occurred within the transition, not including flaw influence.

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Westinghouse Electric Nuclear Services Appendix E SGSGDA-02-45, Rev.1 Page 42 of 43 Table E S A - AVB Wear: Indian Point 2 Data 21 1 18 0.14 0.12 21 35 I 20 I AV3 I 0.18 0. I4 21 32 21 AV2 0.16 0.12 21 36 20 AV2 0.22 0.17 21 40 25 AV3 0.19 0.15 21 34 23 AV2 0.1 0.07 21 41 29 AV2 0.19 0.23 71  ?? 27 AV3 n 12 n11 21 35 28 AV2 0.12 0.12 21 40 28 AV3 0.1s 0.19 21 42 30 AV3 0.18 0.18 21 36 31 AVl 0.15 0.12 21 43 34 AV4 0.18 0.19 21 41 35 AV2 0.19 0.21 21 36 36 AV2 0.14 0. I4 21 40 36 AV2 0.2 1 0.24 21 43 39 AV3 0.19 0.17 21 45 41 AV2 0.14 0.18 21 43 47 AV3 0. I9 ~ ~

0.16 21 42 54 AV3 0.2 0.22 23 35 31 AV2 0.17 0.19 23 41 31 AV2 0.19 0.23 23 34 34 AV3 0.17 0.2 1 23 43 34 AV2 0.15 0.15 23 35 54 AV2 0.2 0.22 23 40 59 AV4 0.18 0.2 23 40 62 AVI 0.19 0.17 23 I 33 I 62 I A V ~ I 0.14 I 0.15 1 I I AVG. I 0.17 I 0.17

Westinghouse Electric Nuclear Services Appendix E SGSGDA-02-45, Rev. 1 Page 43 of 43 Table ESB - AVB Wear: EPRI ETSS DATA 11.4 Indian Point Unit 2 baseline data CaI Curve 12 0.3 Indian Point Unit 2 baseline data Cal Curve 15 0.4