NL-17-131, Relief Requests IP3-ISI-RR-11 and IP3-ISI-RR-12 Associated with Extension of Fourth Interval Reactor Vessel and Piping Weld Inspections

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Relief Requests IP3-ISI-RR-11 and IP3-ISI-RR-12 Associated with Extension of Fourth Interval Reactor Vessel and Piping Weld Inspections
ML17297A461
Person / Time
Site: Indian Point Entergy icon.png
Issue date: 10/18/2017
From: Vitale A
Entergy Nuclear Northeast
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-17-131
Download: ML17297A461 (14)
v • d • e


Text

  • *~*Entergx Entergy Nuclear Northeast Indian Point Energy Center 450 Broadway, GSB P.O. Box249 Buchanan, N.Y. 10511-0249 Tel (914) 254-6700 Anthony J Vitale Site Vice President NL-17-131 October 18, 2017 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk 11555 Rockville Pike, OW,FN-2 F1 Rockville, MD 20852-2738

SUBJECT:

Relief Requests IP3-ISl-RR-11 and IP3-ISl-RR-12 Associated with Extension of Fourth Interval Reactor Vessel and Piping Weld Inspections Indian Point Unit 3 Docket No. 50-286 License No. DPR-64

REFERENCE:

Entergy letter NL-17-021 to NRC, "Notification of Permanent Cessation of Power Operations," dated February 8, 2017(ML17044A004)

Dear Sir or Madam:

Pursuant to 10 CFR 50.55a(z)(1 ), Entergy Nuclear Operations, Inc. (Entergy) requests Nuclear Regulatory Commission (NRC) approval of the attached Relief Requests IP3-ISl-RR-11 and IP3-ISl-RR-12 for Indian Point Unit 3 (IP3). The Code of Record for the current fourth 10 year interval is the 2001 Edition with Addenda through 2003 of the American Society of Mechanical Engineers (ASME)Section XI Boiler and Pressure Vessel (B&PV) Code, "Rules for lnservice Inspection of Nuclear Power Plant Components."

Specifically, Entergy proposes a one time extension of the current lnservice Inspection (ISi) 10 year interval associated with the reactor vessel shell and cold leg weld inspections from the Spring 2019 refueling outage (RFO 20) to the Spring 2021 refueling outage (RFO 21). This interval extension is being requested with deference to the recent decision to permanently cease power operations of IP3 by April 30, 2021 (Reference). The attached relief requests will allow the subject weld inspections, which involve removal of the reactor vessel core barrel, to be scheduled concurrently during RFO 21. These inspections would only be completed if the current plan to permanently cease operation of IP3 is changed in the future. Not having to perform these inspections will result in a substantial personnel radiation dose savings.

Entergy requests approval of the proposed alternative by October 1, 2018 to facilitate scheduling of the IP3 RF0-20 activities.

J ~

NL-17-131 Docket No. 50-286 Page 2 of 2 If you have any questions or require additional information, please contact Mr. Robert Walpole, Manager, Regulatory Assurance at (914) 254-6710.

Sincerely, AJV/cdm Attachments: 1. Indian Point Unit 3 Relief Request No. IP3-ISl-RR-11, Reactor Vessel Weld lnservice Inspection Frequency Extension

2. Indian Point Unit 3 Relief Request No. IP3-ISl-RR-12, ASME Code Case N-770-2 Weld Inspection Frequency Extension cc: Mr. Richard Guzman, Senior Project Manager, NRC NRR DORL Mr. Daniel H. Dorman, Regional Administrator, NRC Region 1 NRC Resident Inspectors Office Ms. Alicia Barton, President and CEO, NYSERDA Ms. Bridget Frymire, New York State Dept. of Public Service

NL-17-131 ATTACHMENT 1 INDIAN POINT UNIT 3 RELIEF REQUEST NO. IP3-ISl-RR-11 REACTOR VESSEL WELD INSERVICE INSPECTION FREQUENCY EXTENSION ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 DOCKET NO. 50-286

NL-17-131 Docket No. 50-286 Attachment 1 Page 1 of 5 Indian Point Unit 3 Fourth 10 Year ISi Interval Relief Request No. IP3-ISl-RR-11 Reactor Vessel Weld lnservice Inspection Frequency Extension Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

Alternative Provides Acceptable Level of Quality and Safety

1. ASME Code Component(s) Affected The affected component is the Indian Point Unit 3 (IP3) reactor vessel (31 RV). Specifically, the following are the applicable American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code,Section XI, (Reference 1) examination categories and item numbers affecting the reactor pressure vessel (RPV). These examination categories and item numbers are from paragraph IWB-2500 and Table IWB-2500-1 of the ASME BPV Code,Section XI.

Examination Category Item No. Description B-A B1 .11 Circumferential Shell Welds B-A B1.12 Longitudinal Shell Welds B-A B1.21 Circumferential Bottom Head Welds B-A B1.22 Meridional Bottom Head Welds B-A B1.30 Shell-to-Flange Weld B-A B1.40 Head-to-Flange Weld B-D B3.90 Nozzle-to-Vessel Welds

.B-D B3.100 Nozzle Inside Radius Section B-N-2 813.60 Interior Attachments Beyond Beltline Region B-N-3 B13.70 Core Support Structure

2. Applicable Code Edition and Addenda

ASME Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components,"

Code 2001 Edition with Addenda through 2003.

3. Applicable Code Requirement

Paragraph IWB-2412 of ASME BPV Code,Section XI, Inspection Program B, requires volumetric examination of essentially 100 percent of the RPV pressure retaining welds identified in Table IWB-2500-1 once each ten year interval. The interval was extended for the fourth ISi interval until the Spring 2019 refueling outage (RFD 20) based on approval of Relief Request IP3-ISl-RR-06, as supplemented, (References 2, 3, and 4) and the corresponding Nuclear Regulatory Commission (NRC) safety evaluation report (SER) (Reference 5). The purpose of this relief request is to further extend the fourth ISi interval to the refueling outage scheduled for the Spring of 2021 (RFO 21).

NL-17-131 Docket No. 50-286 Attachment 1 Page 2 of 5

4. Reason for Request

Relief is being requested to extend the reactor vessel weld inspection for IP3 from the Spring 2019 refueling outage (RFO 20) to the Spring 2021 refueling outage (RFO 21). Deferring the reactor vessel weld inspection to RFO 21 will allow the inspection to be scheduled coincident with the ASME Code Case N-770-2 weld inspection (IP3-ISl-RR-12), eliminating the need to remove the core barrel in RFO

20. In addition, Reference 6 provided notification of the Entergy Nuclear Operations, Inc. (Entergy) decision to permanently cease operation of IP3 no later than April 30, 2021. As such, approval of this relief request would effectively permit the IP3 vessel weld inspection to only be performed if the current plan to permanently cease operation is changed in the future and IP3 continues to operate beyond April 30, 2021. Not having to perform this inspection will result in a substantial personnel radiation dose savings.

Relief Request IP3-ISl-RR-06 (Reference 2) and the NRC SER (Reference 5) require that category B-A, B-D, B-N-2, and B-N-3 vessel welds and supports be inspected during the upcoming Spring 2019 refueling outage (RFO 20). Inspection of the reactor vessel welds requires removal of the lower internals, including the core barrel, and storing them in the lower cavity. This inspection had previously been planned to be performed concurrent with the ASME Code Case N-770-2 reactor vessel nozzle weld Inspections during RFO 20. Consistent with this relief request, a separate IP3 Relief Request (IP3-ISl-RR-12) is being submitted to allow deferral of the Code Case N-770-2 weld inspections from the Spring 2019 refueling outage (RFO 20) to the Spring 2021 refueling outage (RFO 21).

5. Proposed Alternative and Basis for Use 10 CFR 50.55a(z) states:

"Alternatives to the requirements of paragraphs (b) through (h) of this section or portions thereof may be used when authorized by the Director, Office of Nuclear Reactor Regulation, or Director, Office of New Reactors, as appropriate. A proposed alternative must be submitted and authorized prior to implementation. The applicant or licensee must demonstrate that:

(1) The proposed alternative would provide an acceptable level of quality and safety; or (2) Compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety."

Entergy has determined that the alternative proposed in this relief request provides an acceptable level of quality and safety.

Entergy proposes to defer completion of the ASME BPV Code,Section XI, required volumetric examination of the Category B-A RPV pressure retaining welds and Category B-D full penetration welded nozzle. Entergy also proposes to defer the visual examinations of the Category B-N-2 RPV welded core support and interior attachments and Category B-N-3 removable core support structures. These proposed deferrals will extend the inspection interval from the Spring of 2019 (RFO 20) to the Spring of 2021 (RFO 21).

NL-17-131 Docket No. 50-286 Attachment 1 Page 3 of 5 In IP3 Relief Request RR-3-43(1) (Reference 7), Entergy requested a deferral of the RPV weld inspection interval based on Westinghouse Topical Report WCAP-16168-NP-A, Revision 2 (Reference 8). The NRC approved Relief Request RR-3-43(1), as documented in the March 6, 2009 SER (Reference 12), extending the third ISi interval until 2015. This original RPV weld ISi extension relief request included a plant specific evaluation of IP3 to confirm the applicability of the parameters contained in Appendix A of the WCAP. This comparison confirmed the applicability of the parameters, with the exception of the Through-Wall Cracking Frequency (TWCF) parameter. An alternative analysis to address the TWCF parameter deviation was provided in Relief Request RR 43(1). The alternative TWCF analysis was subsequently superseded by a plant specific change-in-risk analysis (Reference 9) in response to an NRC Request for Additional Information (RAI) *

(Reference 10).

The RPV weld inspection interval was subsequently extended for the fourth ISi interval until the Spring 2019 refueling outage (RFO 20) based on NRC approval of Relief Request IP3-ISl-RR-06 (Reference 5). Relief Request IP3-ISl-RR-06 used the information provided in the original Relief Request RR-3-43(1), including the results of the WCAP-16168-NP-A analysis. As discussed in Relief Request IP3-ISl-RR-06, as supplemented (References 2, 3, and 4), and summarized below, there is reasonable assurance of continued structural integrity of the affected welds during the deferral period proposed in this relief request.

As stated in Relief Request IP3-ISl-RR-06, a change-in-risk analysis was performed for IP3 using the same methodology as was used in WCAP-16168-NP-A for the Westinghouse pilot plant, Beaver Valley Unit 1. The analysis was performed using plant specific inputs for IP3, including fluence, beltline material properties, and dimensions. The IP3 change-in-risk analysis was performed using fluence values at 48 Effective Full Power Years (EFPY) to bound the potential license renewal period through 60 calendar years of operation. The results of the change-in-risk analysis were provided in the response to Question 1 of Reference 9. Consistent with the WCAP pilot plant evaluations, the change-in-risk analysis considered the effects of ISi frequency and fatigue crack growth from design basis transients. Two cases were considered in the analyses: (1) inspection performed every 1O years and (2) inspection performed after the first 10 years, but none performed thereafter (this approach is discussed in more detail in the response to Question 2 of Reference 9). The bounding change-in-risk between these two cases was determined to be 2.15E-08 events per year, which is about a factor of 5 below the 1.0E-07 events per year criteria in the risk-informed decision making guidance of Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," for an acceptably small change in large early relea.se frequency (LERF).

The inputs developed for the NRC's re-evaluation of the Pressurized Thermal Shock (PTS) Rule were used in the WCAP-16168-NP-A pilot plant analyses, and these inputs were also used in the IP3 change-in-risk analysis. The use of the Westinghouse pilot plant PTS Rule transients as input to the IP3 change-in-risk analysis was discussed in detail in the responses to Questions 3 and 4 of Reference 9, and justified through a detailed comparison of plant features that contribute to the likelihood of having a PTS event. The basis for several plant specific inputs to the IP3 change-in-risk analysis was further provided in the response to Question 6 of Reference 9. The response to

NL-17-131 Docket No. 50-286 Attachment 1 Page 4 of 5 Question 6 also demonstrated through the performance of surveillance data checks that the embrittlement trend curve correlations used in the change-in-risk analysis were appropriate for predicting the embrittlement of the IP3 reactor vessel beltline materials.

In response to an additional NRC question (Reference 10), the results of the IP3 change-in-risk analysis were revised to include consideration of external events (Reference 11 ). This increased the bounding change in risk to 2.66E-08 events per year, which is still about a factor of 4 below the criterion in Regulatory Guide 1.174 of 1.0E-07 events per year for an acceptably small change in LERF.

Relief Request IP3-ISl-RR-06, as supplemented, demonstrated that: (a) IP3 provided sufficient information as requested in Sections 3.4 and 4.0 of the SER forWCAP16168-NP-A, (b) IP3 provided a plant-specific change in TWCF analysis to demonstrate that the proposed change in the IP3 RPV ISi program meets the Regulatory Guide 1.174 guidelines discussed in the SER for WCAP 16168-NP-A, and (c) the IP3 proposed alternative provides an acceptable level of quality and safety.

As indicated in Relief Request IP3-ISl-RR-06, the IP3 change-in-risk analysis was performed for 48 EFPY, corresponding to 60 calendar years of operation. However, NRC staff approval of the relief request was limited to 2019 (Reference 5). The IP3 plant specific change-in-risk analysis adequately bounds the requested 22 year extension to the Spring of 2021 (RFO 21) since the proposed inspection deferral remains within 48 EFPY and 60 calendar years.

6. Duration of Proposed Alternative This relief request is applicable to Entergy's ISi program for the IP3 fourth 1O year interval. The proposed alternative is from the Spring of 2019 (RFO 20) to the Spring of 2021 (RFO 21).
7. Precedents
1. Indian Point Unit 3 (IP3) Relief Request IP3-ISl-RR-06(ML14198A331)
2. Indian Point Unit 2 (IP2) Relief Request IP2-ISl-RR-16(ML13228A167)
3. Arkansas Nuclear One, Unit 2, Request for Alternative AN02-ISl-004 (ML102450654, ML14150A163)
4. Waterford Steam Electric Station, Unit 3, Request for Alternative W3-ISl-006 (ML091210375)
8. References
1. ASME Boiler and Pressure Vessel Code,Section XI, 2001 Edition through 2003 Addenda, American Society of Mechanical Engineers, New York
2. Entergy Letter NL-14-004 to NRC, "Relief Request IP3-ISl-RR-06 for Reactor Vessel Weld Examination, Indian Point Unit Number 3," dated January 13, 2014(ML14017A055)

NL-17-131 Docket No. 50-286 Attachment 1 Page 5 of 5

3. NRC Letter to Entergy, "Request for Additional Information Regarding Relief Request IP3-ISl-RR-06, Reactor Vessel Weld Examinations (TAC No. MF3345)," dated March 21, 2014 (ML14064A386)
4. Entergy Letter NL-14-044 to NRC, "Response to Request for Additional Information Regarding Relief Request IP3-ISl-RR-06, Reactor Vessel Weld Examination (TAC No.

MF3345)," dated April 7, 2014(ML14106A372)

5. NRC Letter to Entergy, "Indian Point Nuclear Generating Unit No. 3 - Safety Evaluation for Relief Request IP3-ISl-RR-06 for Reactor Vessel Weld Examinations (TAC No. MF3345),"

dated July 23, 2014(ML14198A331)

6. Entergy Letter NL-17-021 to NRC, "Notification of Permanent Cessation of Power Operations, Indian Point Nuclear Generating Unit Nos. 2 and 3," dated February 8, 2017 (ML17044A004)
7. Entergy Letter NL-08-096 to NRC, "Request for Relief to Extend the Unit 2 and 3 lnservice Inspection Interval for the Reactor Vessel Weld Examination and Request for License Amendment for Submittal of ISi Information and Analyses," Enclosure 2, dated July 8, 2008 (ML081980058)
8. WCAP-16168-NP-A, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval," Revision 2, June 2008 (ML082820046)
9. Entergy Letter NL-08-177 to NRC, "Response to Request for Additional Information on Request for Relief to Extend the Unit 2 and 3 lnservice Inspection Interval for the Reactor Vessel Weld Examination and Request for License Amendment for Submittal of ISi Information and Analyses (TAC Nos. MD9196-MD9197)," dated December 23, 2008 (ML090050020)
10. NRC Letter to Entergy, "Indian Point Nuclear Generating Unit Nos. 2 and 3 - Request for Additional Information Regarding Relief Request for Vessel Weld Inspection Extension (TAC Nos. MD9096 and MD9197)," dated November 20, 2008 (ML082971068)
11. Entergy Letter NL-09-003 to NRC, "Supplemental Response to Request for Additional Information on Request for Relief to Extend the Unit 2 and 3 lnservice Inspection Interval for the Reactor Vessel Weld Examination (TAC Nos. MD9196 and MD9197)," dated January 20, 2009 (ML090400575)
12. NRC Letter to Entergy, "Indian Point Nuclear Generating Unit Nos. 2 and 3- Relief Requests on Reactor Vessel Weld Examinations (TAC Nos. MD9196 and MD9197)," dated March 6, 2009 (ML090360460)

NL-17-131 ATTACHMENT 2 INDIAN POINT UNIT 3 RELIEF REQUEST NO. IP3-ISl-RR-12 ASME CODE CASE N-770-2 WELD INSPECTION FREQUENCY EXTENSION ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 DOCKET NO. 50-286

NL-17-131 Docket No. 50-286 Attachment 2 Page 1 of 5 Indian Point Unit 3 Fourth 10 Year ISi Interval Relief Request No: IP3-ISl-RR-12 ASME Code Case N-770-2 Weld Inspection -Frequency Extension Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

Alternative Provides Acceptable Level of Quality and Safety

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1. ASME Code Component(s) Affected The affected components are the Indian Point Unit 3 (IP3) reactor vessel cold leg nozzle to safe-end welds (1-4100-16(DM), 1-4200-16(DM), 1-4300-16(DM), and 1-4400-16 (DM)). These are Alloy 600 welds, as described in American Society of Mechanical Engineers (ASME) Code Case N-770-2, Table 1, Inspection Item B.

These welds had an Alloy 600 inside diameter (ID) onlay installed during original fabrication and do not join any cast stainless steel materials.

Examination Inspection Description Category Item CC N-770-2 B Weld 1-4100-16(DM) Loop 31 cold leg nozzle to safe-end CC N-770-2 B Weld 1-4200-16(DM) Loop 32 cold leg nozzle to safe-end CC N-770-2 B Weld 1-4300-16(DM) Loop 33 cold leg nozzle to safe-end CC N-770-2 B Weld 1-4400-16(DM) Loop 34 cold leg nozzle to safe-end

2. Applicable Code Edition and Addenda

ASME Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components,"

Code 2001 Edition with Addenda through 2003.

ASME Code Case N-770-2 (Reference 1), as incorporated by reference in 10 CFR 50.55a, with the conditions in paragraph 10 CFR 50.55a(g)(6)(ii)(F).

3. ApplicableCode Requirement Table 1 of ASME Code Case N-770-2 requires volumetric examination of essentially 100 percent of the Inspection Item B pressure retaining welds once every second inspection period not to exceed 7 years.

4. Reason for Request

Relief is being requested at this time to extend the cold leg weld inspections from the Spring 2019 refueling outage (RFO 20) to the Spring 2021 refueling outage (RFO 21). This will allow the cold leg weld inspections to be scheduled coincident with the 10 year vessel shell weld inspections. A

NL-17-131 Docket No. 50-286 Attachment 2 Page 2 of 5

  • separate IP3 Relief Request (IP3-ISl-RR-11) is being submitted for approval to allow deferral of the vessel shell weld inspections from RFO 20 to RFO 21 consistent with the deferral proposed in this relief request submittal. Since IP3 is currently scheduled to permanently cease operation no later than April 30, 2021 (Reference 2), these inspections would only be performed if the current plan to permanently cease operation is changed in the future and IP3 continues to operate beyond this date.

Examination of the ASME Code Case N-770-2 Inspection Item A-2 (hot leg) and Item B (cold leg) welds are performed from the inside surface diameter of the pipe (ID) at IP3 due to limited access from the outside surface diameter (OD) of the pipe. The IP3 Item A-2 and Item B welds are located inside a "sandbox" which was installed during original plant construction after all welding was completed. As a result of physical interferences, the ultrasonic examinations of the Item B welds cannot be performed from the OD of the pipe and, therefore, require the core barrel to be removed from the vessel to allow access to the ID of the cold leg nozzles.

Baseline inspections of Code Case N-770-2 Inspection Item B welds 1-4100-16(DM), 1-4200-16(DM), 1-4300-16(DM) and 1-4400-16(DM) were performed in March 2009. The ultrasonic examinations performed in the 2009 inspections met the AS ME Section XI, Appendix VI 11 requirements, including the examination volume of essentially 100 percent. References 3 and 4 revised the schedule for re-examination of these welds to the Spring 2019 refueling outage, which will require removal of the core barrel during RFO 20 unless this relief request is granted to defer the weld examinations to RFO 21. Note that References 3, 4, and 5 refer to Code Case N-770-1, which was in effect at the time of the 2009 weld inspections.

The 2009 inspections did not identify any recordable indications within the ASME Section XI inspection volume for Welds 1-4100-16(DM), 1-4200-16(DM), and 1-4300-16(DM). Weld 1-4400-16(DM) had a recordable circumferential indication in the vicinity of the Alloy 600 and stainless steel cladding interface near the dissimilar metal weld. The indication was determined to be entirely embedded in the Alloy 600 clad material and is, therefore, not in the ASME Section XI required examination volume. In addition, supplemental eddy current examinations yielded no recordable indications, which confirmed that the indication at Weld 1-4400-16(DM) was not open to the wetted surface. Because the indication was in the Electric Power Research Institute (EPRI) Material Reliability Program ( MRP)-139 primary system piping butt weld examination volume, the indication was conservatively assessed in terms of the criteria in the ASME Code Section XI, 1989 Edition, no Addenda, Article IWB 3000, paragraph IWB-3500 (the code in effect at the time of inspection). The indication was found to be within the allowable limits specified in IWB-3500 with no further evaluation required. -I Since inspection of Jhe subject ASME Code Case N-770-2, Table 1, Inspection Item B, welds requires that the core barrel be removed from the reactor vessel, performing these inspections concurrently with the vessel shell weld inspections will result in an expected personnel radiation dose savings of 0.9 person-Roentgen equivalrnt man (Rem).

5. Proposed Alternative and Basis for Use 10 CFR 50.55a(z) states:

"Alternatives to the requirements of paragraphs (b) through (h) of this section or portions thereof

NL-17-131 Docket No. 50-286 Attachment 2 Page 3 of 5 may be used when authorized by the Director, Office of Nuclear Reactor Regulation, or Director, Office of New Reactors, as appropriate. A proposed alternative must be submitted and authorized prior to implementation. The applicant or licensee must demonstrate that:

(1) The proposed alternative would provide an acceptable level of quality and safety; or (2) Compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety."

A one time extension to the Code Case N-770-2, Table 1, Inspection Item B, volumetric examinations is requested for the period of not to exceed 7 years to a period of not to exceed 12 years. The inspections that are currently required to be performed during the Spring 2019 refueling outage (RFO 20) will be performed no later than the Spring 2021 refueling outage (RFO

21) if the current plan to permanently cease operation is changed in the future and IP3 continues to operate beyond April 30, 2021.

Operating experience on Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 82/182 welds shows that weld repairs performed during original plant construction are a significant contributor in the initiation and propagation of cracking. A review of the construction records and a weld repair search performed for the IP3 reactor vessel nozzle Alloy 82/182 welds did not identify any weld repairs performed on these welds during original plant constructicm.

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In the March 2017 refueling outage (RFO 19), ultrasonic (volumetric) and eddy current (surface) examinations were performed on the ASME Code Case N-770-2 Inspection Item A-2 (hot leg) welds, and no indications were identified. In 2019, visual examinations from the OD are scheduled to be performed on the Code Case N-770-2 Inspection Item A-2 welds (hot leg -4 welds) and Inspection Item B welds (cold leg - 2 welds). Since PWSCC is temperature dependent, and the hot legs are subjected to significantly higher operating temperatures, if cracking were to occur it would be expected to occur in the hot leg welds well before they would occur in the cold leg welds. Therefore, the absence of any indications in the Item A-2 (hot leg) welds during the 2017 inspections provides reasonable assurance that extension of the examinations of the Item B (cold leg) welds from RFO 20 to RFO 21 provides an acceptable level of quality and safety.

During the March 2009 baseline inspections of the ASME Code Case N-770-2 Inspection Item B (cold leg) welds, a surface examination utilizing an eddy current technique was performed in addition to the ultrasonic (volumetric) examination. The ultrasonic examinations performed in the 2009 inspections met the ASME Section XI, Appendix VIII requirements, including the weld examination scope of essentially 100 percent. The use of an eddy current examination in addition to the Code Case N-770-2 required ultrasonic examination provides a higher probability of detection of small, near surface, or surface connected flaws than an ultrasonic examination alone. The Code Case N-770-2 inspection frequency was chosen to ensure that new or pre-existing flaws do not grow to critical flaw sizes. The absence of any recordable indications in 2009 after 33 years of plant operation in conjunction with the results of the flaw tolerance analysis discussed below demonstrates that the proposed alternative examination schedule will provide an acceptable level of protection against PWSCC.

The Reference 5 technical justification to support an extended volumetric examination interval (LTR-PAFM-13-115-P) provides a flaw tolerance analysis performed by Westinghouse for the

NL-17-131 Docket No. 50-286 Attachment 2 Page 4 of 5 IP3 reactor vessel inlet nozzle to safe end dissimilar metal (OM) welds. The purpose of this analysis was to assess the impact of extending the inspection frequency beyond the 7 year inspection frequency required by Code Case N-770-2. This analysis calculated the length and the depth of the largest axial and circumferential flaws which, if left in service for more than 7 years, would not grow beyond the limits provided in paragraph IWB-3600 of the ASME Section XI Code.

The Reference 5 analysis also established the maximum flaw size which could have reasonably been missed during the 2009 baseline inspection considering the detection capabilities of the nondestructive examination (NOE) techniques used during the inspection. It was estimated that the volumetric examination technique (i.e., ultrasound) used during the inspection was capable of reliably detecting a 10 percent through wall, surface breaking flaw. This size flaw (i.e., 0.257 inch deep flaw) was then assumed to have been missed during the 2009 inspection even though an eddy current surface examination did not identify any surface breaking flaws. Hence, the assumption that the 2009 inspection missed a 0.275 inch deep flaw is considered conservative.

Specifically, Figures 7-1 and 7-2 of the Reference 5 analysis show the results of a flaw growth evaluation for axial and circumferential flaws which could have been missed during the 2009 inspection. From Figure 7-1, a 10 percent (0.275 inch) deep axial flaw is expected grow to approximately 23 percent (0.633 inches) after 12 years of service when subjected to operating and design basis loading conditions. This 23 percent deep flaw, when compared to the ASME Section XI allowable flaw depth size of 75 percent, results in a margin of safety of approximately 3.2 above the ASME Code requirement. Similarly, from Figure 7-2, a 10 percent (0.275 inch) deep circumferential flaw would grow to approximately 19 percent (0.523 inches) after 12 years of service when subjected to operating and design basis loading conditions. This 19 percent deep flaw, when compared to the ASME Section XI allowable flaw depth size of 75 percent, results in a margin of safety of approximately 3.9 above the ASME Code requirement. Based on these results and the details provided in the Reference 5 analysis, deferring the next cold leg nozzle OM weld volumetric examinations from the Spring 2019 refueling outage (RFO 20) to the Spring 2021 refueling outage (RFO

21) will not result in flaws which could have been missed during the last inspection to grow beyond the ASME Section XI IWB-3600 Code limits. Furthermore, since the hot leg nozzle to safe end welds were inspected during the Spring 2017 refueling outage (RFO 19) and no recordable indications were detected, it is reasonable to conclude that the proposed deferral of the cold leg nozzle to safe end weld volumetric examinations from RFO 20 to RFO 21 will provide an acceptable level of quality and safety.
6. Duration of Proposed Alternative This relief request is applicable to the fourth interval of the Entergy Nuclear Operations, Inc.

(Entergy) IP3 lnservice Inspection Program. The proposed alternative is to defer the ASME Code Case N-770~2 Inspection Item B (cold leg) weld inspections from the Spring of 2019 (RFO 20) to the Spring of 2021 (RFO 21).

7. Precedents
1. Indian Point Unit 3 (IP3) Relief Request IP3-ISl-RR-07(ML14199A444)

-1 NL-17-131 Docket No. 50-286 Attachment 2 Page 5 of 5

2. Indian Point Unit 2 (IP2) Relief Request IP2-ISl-RR-17 (ML 1331 OA575)
8. References
1. ASME Code Case N-770-2, "Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation ActivitiesSection XI, Division 1," approval date: June 9, 2011
2. Entergy letter NL-17-021 to NRC, "Notification of Permanent Cessation of Power Operations," dated February 8, 2017(ML17044A004)
3. Entergy Letter NL-14-005 to NRC, "Relief Request IP3-ISl-RR-07 for Code Case N-770-1 Weld Inspection Frequency Extension," dated January 13, 2014(ML14017A054)
4. NRC Letter, "Safety Evaluation for Relief Request No. IP3-ISl-RR-07 for Reactor Vessel Cold Leg Nozzle to Safe-End Weld Examinations (TAC No. MF3346)," dated August 4, 2014(ML14199A444)
5. Entergy Letter NL-14-027 to NRC, "Supplemental Information to Relief Request IP3-ISl-RR-07 for Code Case N-770-1 Weld Inspection Frequency Extension," dated February 4, 2014 (ML14051A166)

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