NL-13-032, Technical Specification 5.6.8 - IP3 Steam Generator Tube Inspection Report - Spring 2013 Refueling Outage

Technical Specification 5.6.8 - IP3 Steam Generator Tube Inspection Report - Spring 2013 Refueling Outage

ML13235A047
Person / Time
Site:	Indian Point
Issue date:	08/15/2013
From:	Robert Walpole Entergy Nuclear Northeast
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-13-032
Download: ML13235A047 (8)

Text

Enteray Nuclear Northeast Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 SEntergy Buchanan, N.Y. 10511-0249 Tel (914) 254-6710 Robert Walpole Licensing Manager NL-13-032 August 15, 2013 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk 11555 Rockville Pike Rockville, MD 20852

SUBJECT:

Technical Specification 5.6.8 - IP3 Steam Generator Tube Inspection Report -

Spring 2013 Refueling Outage Indian Point Unit No. 3 Docket No. 50-286 License No. DPR-64 REFERENCE 1. Entergy letter to the NRC, NL-07-083 'Steam Generator Tube Inspection Report for Spring 2007 Refueling Outage', dated July 19, 2007

Dear Sir or Madam:

Entergy Nuclear Operations, Inc (Entergy) is providing in Enclosure 1 the Steam Generator (SG) Tube Inspection Report required by Indian Point 3 Technical Specification 5.6.8.

Technical Specification 5.6.8 requires that a report be submitted to the NRC 180 days after the initial entry into Mode 4 following completion of a steam generator program inspection. Entergy performed a SG inspection in March 2013 during unit 3 refueling outage 3R17. The unit initially entered MODE 4 following this inspection on March 25, 2013. This inspection was the 7th in-service inspection (ISI) following SG replacement in 1989 and the 1st of 2 scheduled inspections in the 2nd inspection period.

The scope of the inspection included all four steam generators, which are Westinghouse Model 44F with thermally treated Alloy 690 tubes. At the time of the inspection, the steam generators had approximately 205 effective full power months of operation since the first inservice inspection performed in 1990. All four steam generators were found to be in compliance with condition monitoring requirements. The previous inspection was performed during the spring 2007, Refueling Outage 3R14 (reference 1).

NL-13-032 Docket No. 50-286 Page 2 of 2 There are no new commitments contained in this letter. If you have any questions or require additional information, please contact me at 914-254-6710.

Sincerely, RW/mb/rd cc: Mr. William Dean, Regional Administrator, NRC Region 1 Mr. Douglas Pickett,, Senior Project Manager, NRC NRR DORL Mr. Peter Habighorst, Material Control and Accounting Branch, NRC IPEC NRC Resident Inspector's Office Mr. Francis J. Murray, President and CEO, NYSERDA Ms. Bridget Frymire, New York State Department of Public Service

ENCLOSURE 1 TO NL-13-032 Steam Generator Examination Program Results 2013 Refueling Outage (3R17)

ENTERGY NUCLEAR OPERATIONS, INC INDIAN POINT NUCLEAR GENERATING UNIT 3 DOCKET NO. 50-286

Indian Point Unit 3 Steam Generator Examination Program Results 2013 Refueling Outage Q3R17) 1.0 Introduction Indian Point Unit 3 Technical Specification (TS) 5.6.8 Steam Generator Tube Inspection Report, requires Entergy Nuclear Northeast to submit a report to the NRC within 180 days after initial entry into Mode 4 following a steam generator inspection performed in accordance with Technical Specification 5.5.8, Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SG,

b. Active degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism,

d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e. Number of tubes plugged during the inspection outage for each active degradation mechanism,

f. Total number and percentage of tubes plugged to date,

g. The results of condition monitoring, including the results of tube pulls and in-situ testing, and

h. The effective plugging percentage for all plugging in each SG.

Entergy performed a SG inspection in March 2013 during refueling outage 3R17. This inspection was the 7 th in-service inspection (ISI) following SG replacement in 1989 and the 1st of 2 scheduled inspections in the 2 nd inspection period. The unit initially entered MODE 4 following this inspection on March 25, 2013.

2.0 Steam Generator Background The original Westinghouse Model 44 SG's at Indian Point 3 were replaced in 1989 with Westinghouse Model 44F SG's. Each SG has 3214 tubes made from thermally treated Alloy 690. The nominal OD of each tube is .875 in. and the nominal tube wall is .050 in.

thick. At the time of the 2013 refueling outage (3R1 7) at Indian Point 3, the steam generators had accumulated approximately 205 effective full power months (EFPM) of operation since the first in-service inspection performed in 1990. Refueling outage 3R17 was the first of two refueling outages in the second inspection period of 108 EFPM as defined in section 5.5.8.d.2 of Technical Specifications and the first scheduled inspection of the period. The steam generator inspections were performed in March 2013.

Indian Point 3 Steam Generator Primary Inspection Plan Outage Year Cycle SG Inspection Sequential Notes EFPM Cumulative Period Inspection EFPM EFPM Period 3R07 1990 13 13 N/A n/a First ISI

3R08 1992 13.6 26.6 13.6 First 3R09 1997 18.6 45.2 32.2 3R10 1999 21.5 66.7 53.7 3R12 2003 39.7 106.4 93.4 Skip I outage 3R14 2007 44.7 151.1 138.1 Skip 1 outage 3R17 2013 67 218.1 67 Second Skip 2 outages 3R19 2017 45 est. 263 est. 112 est. Skip 1 outage 3R21 2021 45 est. 308 est. 45 est. Third Skip 1 outage 3R22 2023 22.5 est. 330 est. 67.5 est. Skip 1 outage 3.0 Required Report Content - The following information is provided as required in Section 5.6.8 of IP3 Technical Specifications a) Scope of Inspections The eddy current examination scope consisted of the acquisition of 50% of all tubes in all four steam generators full length with a bobbin probe, 60% of the hot leg 1 st span and 22% cold leg 1St span with an X-probe array coil probe, 50% of rows 1 and 2 u-bends with the single coil +Pt rotating probe. In addition, tubes containing indications from previous examinations were examined. All indications identified during the bobbin coil had a diagnostic examination performed with the +Pt coil. The data analysis scope included the bobbin coil data, the array coil data, all +Pt coil data, and AVB position verification.

In addition to the eddy current inspections, visual inspections were performed on all tube plugs and on the primary bowl drain area.

Sludge lancing and Foreign Object Search and Retrieval (FOSAR) was performed on the secondary side of all SG's. Secondary side Steam Drum and Top Tube Support Plate (TTSP) inspections were performed on 2 SG's.

b) Active Degradation Mechanisms No active degradation mechanisms were found during the SG inspection in 3R17.

c) Nondestructive Examination Techniques NDE Techniques used for Potential Degradation Mechanisms Technique EPRI ETSS Degradation Mechanism Bobbin 96004.1 (Rev 13) Wear at AVB supports, support plates & FDB Bobbin 27091.2 (Rev 0) Loose part volumetric wear, part not present Bobbin 27091.3 (Rev 1) Loose part volumetric wear, part not present

+Pt 10908.4 (Rev 1) Wear at AVB locations

+Pt 21998.1 (Rev 4) Freespan Volumetric

+Pt 27901.1 -

27907.1 (Rev 1) Freespan Volumetric

+Pt 96910.1 (Rev 10) Wear at supports and PLP wear part present Array 20200.1 (Rev 5) OD Circ indications (ODSCC) at TTS and Expansion Transitions

d) Service Induced Indications There were no service induced indications detected during 3R17 e) Number of Tubes Plugged by Mechanism for 3R17 No tubes were plugged during 3R17 f) Tubes Plugged to Date SG 31 SG 32 SG 33 SG 34 Total Total Number of Tubes 3214 3214 3214 3214 12856 Tubes Plugged Pre-Service 0 0 0 2 2 Tubes Plugged in Prior Outages 3 6 3 2 14 Tubes Plugged in 3R17 0 0 0 0 0 Total Tubes Plugged to Date 3 6 3 4 16

% of Tubes Plugged to Date 0.09% 0.19% 0.09% 0.12% 0.12%

g) Condition Monitoring Results No tubes were found with newly-formed degradation. There are eight small volumetric indications, created in 2001 as a result of sludge lance rail wear, that are still present, unchanged. The eight indications did not exceed condition monitoring limits. There was no detectable SG primary-to-secondary leakage during the previous operating period. Therefore, all of the steam generator performance criteria were met for the three previous operating cycles. Because the tube degradation found was sized less than the condition monitoring limit, in situ pressure testing was neither required nor performed. No tube pulls were performed. Tube plug and primary bowl drain inspections had no findings. Secondary side inspections were performed with all results nominal.

h) Effective Plugging Percentage Since there are no sleeves installed, the effective tube plugging is equivalent to the percentage of tubes plugged.

4.0 Attachments a) SG Tubesheet Map b) SG Landmark Information

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Steam Generator Location Landmarks AV2 AV3 AV1 AV4 6H 6C 5H 5C 4H 4C 3H 3C 2H 2C 1H 1C BPH BPC TSH TEH TEC Westinghouse Model 44F Steam Generator Ligand AV = Anti-Vibration Bar (AVB)

C = cold leg H = hot leg

1. = support plate (TSP)

BP = baffle plate (FDB)

TS = tubesheet TE = tube end