NL-18-053, Response to Request for Additional Information Regarding Relief Request IP3-ISI-RR-11 for Reactor Vessel Weld Inservice Inspection Frequency Extension

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Response to Request for Additional Information Regarding Relief Request IP3-ISI-RR-11 for Reactor Vessel Weld Inservice Inspection Frequency Extension
ML18211A297
Person / Time
Site: Indian Point Entergy icon.png
Issue date: 07/18/2018
From: Halter M
Entergy Nuclear Operations
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
EPID L-2017-LLR-0127, NL-18-053
Download: ML18211A297 (17)
v • d • e


Text

    • ='~ Entergx Entergy Nuclear Operations, Inc.

1340 Echelon Parkway Jackson, MS 39213 Mandy K. Halter Director, Nuclear Licensing NL-18-053 July 18, 2018 U.S. Nuclear Regulatory Commission Document Control Desk 11545 Rockville Pike, TWFN-2 F1 Rockville, MD 20852-2738

SUBJECT:

Response to Request for Additional Information Regarding Relief Request IP3-ISI-RR-11 for Reactor Vessel Weld lnservice Inspection Frequency Extension Indian Point Nuclear Generating Unit No. 3 Docket No. 50-286 License No. DPR-64

REFERENCES:

1. Entergy Letter NL-17-131 dated October 18, 2017, "Relief Requests IP3-ISI-RR-11 and IP3-ISI-RR-12 Associated with Extension of Fourth Interval Reactor Vessel and Piping Weld Inspections," dated October 18, 2017 (ML17297A461)
2. NRC Electronic Mail dated June 4, 2018, "Indian Point, Unit 3 -

Request for Additional Information - Relief Request IP3-ISI-RR-11 (EPID: L-2017-LLR-0127)"

Dear Sir or Madam:

Pursuant to 10 CFR 50.55a(z)(1 ), Entergy Nuclear Operations, Inc. (Entergy) submitted Indian Point Unit 3 (IP3) Relief Request No. IP3-ISI-RR-11 for NRC review and approval, which was included in the Reference 1 submittal. The relief request proposed a one time extension of the current lnservice Inspection (ISi) 10 year interval reactor vessel weld inspection for IP3 from the Spring 2019 refueling outage (RFO 20) to the Spring 2021 refueling outage (RFO 21).

In Reference 2, the U.S. Nuclear Regulatory Commission (NRC) identified a need for additional information to complete its review of Relief Request No. IP3-ISI-RR-11 and issued a request for additional information (RAI). Attachment 1 to this letter provides Entergy's responses to the NRC staff's RAI. In addition, as discussed in the RAI response, Entergy has revised the previously submitted relief request (i.e., IP3-ISI-RR-11) to remove the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code,Section XI, examination category B-N-2 and B-N-3 items. The revised relief request is provided as . Entergy has created a new relief request IP3-ISI-RR-15 to extend the ISi interval for the ASME BPV Code,Section XI, examination category B-N-2 and B-N-3 items.

NL-18-053 Docket No. 50-286 Page 2 of 2 The new relief request is provided in Attachment 3, and is being submitted in accordance with 10 CFR 50.55a(z)(2). Entergy requests approval of the new relief request by October 1, 2018, which is consistent with the approval date requested in the Reference 1 submittal.

There are no new commitments being made in this submittal.

If you have any questions, or require additional information, please contact Mr. Robert Walpole at (914) 254-6710.

Sincerely, MKH/cdm Attachments: 1. Response to Request for Additional Information Regarding Relief Request IP3-ISI-RR-11

2. Indian Point Unit 3 Revised Relief Request IP3-ISI-RR-11, Reactor Vessel Weld lnservice Inspection Interval Extension
3. Indian Point Unit 3 Relief Request IP3-ISI-RR-15, Reactor Vessel lnservice Inspection Interval Extension I

cc: Mr. Richard V. Guzman, Senior Project Manager, NRG NRR DORL Mr. David Lew, Acting Regional Administrator, NRG Region I Ms. Bridget Frymire, New York State Department of Public Service Ms. Alicia Barton, President and CEO NYSERDA NRG Resident Inspector's Office

NL-18-053 ATTACHMENT 1 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING RELIEF REQUEST IP3-ISI-RR-11 ENTERGY NUCLEAR OPERATIONS, INC.

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

NL-18-053 Attachment 1 Page 1 of 3 Response to Request for Additional Information Regarding Relief Request IP3-ISI-RR-11 By letter dated October 18, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17297A461), Entergy Nuclear Operations, Inc. (Entergy, the licensee), submitted Request for Relief No. IP3-ISI-RR-11 which proposed alternatives to the inservice inspection (ISi) interval requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, Paragraph IWB-2500 and Table IWB-2500-1 for Indian Point Nuclear Generating Unit No. 3 (IP3). The ASME Code requires volumetric examination once each 10-year interval of essentially 100 percent of reactor 1

vessel pressure-retaining welds identified in Table IWB-2500-1, Examination Categories B-A, "Pressure Retaining Welds in Reactor Vessel," B-D, "Full Penetration Welded Nozzles in Vessels," B-N-2, "Welded Core Support Structures and Interior Attachments to Reactor Vessels," and B-N-3, "Removable Core Support Structures." By letter dated July 23, 2014 (ADAMS Accession No. ML14198A331), NRC staff approved a licensee request to extend the inspection interval for Examination Category B-A and B-D welds at IP3 to 20 years.

Pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Part 50.55a(z)(1 ), the licensee requested the use of the proposed alternatives to extend certain required examinations of the IP3 reactor vessel welds for the fourth ISi inspection intervals, on the basis that the alternatives provide an acceptable level of quality and safety. Specifically, the licensee requested to extend the required examinations for Examination Category B-A and B-D welds from 20 years to 22 years. The licensee also requested to extend the required examinations for Examination Category B-N-2 and B-N-3 welds from 10 years to 12 years. The current fourth 10-year ISi interval for IP3 is scheduled to end on July 20, 2020. The licensee's proposed alternative would allow for the deferral of the subject examinations for 2 years, no later than April 30, 2021.

In its October 18, 2017 submittal, the licensee stated that prior relief request IP3-ISI-RR-06 (ADAMS Accession No. ML14017A055), which was approved by the NRC staff by safety evaluation (SE) dated July 23, 2014 (ADAMS Accession No. ML14198A331) and which extended the inspection interval for Examination Category B-A and B-D welds at IP3 to 20 years, demonstrated the following:

  • The licensee "provided sufficient_information as requested in Sections 3.4 and 4.0 of the SER forWCAP-16168-NP-A" (see ADAMS Accession No. ML111600303).
  • The licensee "provided a plant-specific change in TWCF [through-wall cracking frequency] analysis to demonstrate that the proposed change in the IP3 Reactor Pressure Vessel (RPV) ISi program meets the Regulatory Guide 1.174 guidelines discussed in the SER forWCAP-16168-NP-A" (see ADAMS Accession No.

_ ML111600303).

  • The "plant specific change-in-risk analysis adequately bounds the requested 22 year extension to the Spring of 2021 since the proposed inspection deferral remains within 48 EFPY [effective full power years] and 60 calendar years."

NL-18-053 Attachment 1 Page 2 of 3 The following requests for additional information (RAls) outline the information needed for the NRC staff to complete its review:

RAl-1 In its 2014 SE which extended the inspection interval of Examination Category B-A and B-D welds to 20 years, the NRC staff stated that the licensee has, in essence, satisfactorily addressed Plant Specific Information Item 1 from WCAP-A through an approved plant-specific TWCF analysis, and that the embrittlement of the IP3 RPV was addressed appropriately in this analysis. The 2014 SE notes that the licensee's original TWCF calculation was not bounded by WCAP-A, but that the licensee provided a plant-specific TWCF analysis based on 48 EFPY which bounds the neutron fluence up to 20 years following the original license period.

NRC staff requests that the licensee (1) evaluate plant operating experience (including the occurrence of the limiting fatigue transients) since the submission *of the last alternative proposal and (2) confirm (or correct as appropriate) that the previous TWFC calculations remain valid and that the plant will not reach 48 EFPY until after the time period under consideration for this request.

Response

A review of the inputs used in WCAP-16168-NP-A indicated that fatigue crack growth was dominated primarily by plant cooldowns (CDs). The reason for this is that cooldown transients introduce tensile stresses at the inside surface of the reactor vessel which coincide with the lower material fracture toughness resulting from neutron embrittlement. Additionally, feedwater cycling and inadvertent reactor coolant system (RCS) depressurization transients during plant heatups (HUs) also contribute to fatigue crack growth, although an order of magnitude lower than plant CDs. In order to bound all affected Westinghouse designed plants up to 80 years of operation, the WCAP assumed 7 HU/CD cycles per year for a total of 560 cycles during the life of the plant. The 7 HU/CD cycles were chosen to account for up to 5 HU/CD cycles per year plus another 2 HU/CD cycles per year to account for the secondary effects of the other transients, including feedwater cycling and inadvertent RCS depressurization events.

A review of the IP3 cycle monitoring program indicated that IP3 has accumulated 65 HU/CD cycles from initial plant operation through June 1, 2018. This includes 9 HU/CD cycles which occurred between March 2009 and June 2018, resulting in approximately 1 HU/CD cycle per operating year. Applying a margin of safety of 2 and extrapolating through April 30, 2021, results in a total estimated number of 71 (65 + (2 x 3)) HU/CD cycles, which is well below the 560 HU/CD cycles evaluated in WCAP-16168-NP-A. Note that even if IP3 were to accumulate HU/CD cycles at a higher rate, the cycle monitoring program would require corrective actions to be implemented prior to exceeding the current design limit of 114 HU/CD cycles, and thereby preclude exceedance of the 560 HU/CD cycle limit established in the WCAP.

A review was also performed to ensure that the period of full power operation assumed in the WCAP-16168-NP-A evaluations supports the proposed extension of the IP3 reactor vessel inspection interval from 20 to 22 years. Based on plant operating history, IP3 accumulated 28.62 EFPY from initial plant operation through the March 31, 2017 refueling outage.

Conservatively assuming that IP3 operates at 100 percent power from March 2017 through April 30, 2021, an additional 4.1 EFPY would be accumulated. This would result in a total IP3 operating life of 32.72 EFPY, which is well below the 48 EFPY value evaluated in the WCAP.

NL-18-053 Attachment 1 Page 3 of 3 Since both the number of IP3 HU/CD cycles and its operating life (i.e., EFPY) are well within the inputs used in the WCAP-16168-NP-A evaluations, Entergy has concluded that the previous TWFC calculations performed in the WCAP remain valid, and therefore support the proposed deferral of the IP3 reactor vessel inspections for 2 years, to no later than April 30, 2021.

RAl-2 In its October 18, 2017 submittal, the licensee requests to defer the required visual examinations of Category B-N-2 and B-N-3 components to allow the inspection to be scheduled coincident with the ASME Code Case N-770-2 weld inspection and to eliminate the need to remove the core barrel in refueling outage 20 (Spring 2019). However, other than a desire to keep the personnel radiation dose as low as reasonably achievable, no technical justification was provided to support the approval of this alternative under 10 CFR 50.55a(z)(1) which applies to alternatives which provide an acceptable level of quality and safety.

NRC staff requests that the licensee submit technical justification that the request to extend the inspection interval of Category B-N-2 and B-N-3 components from 10 to 12 years will provide an acceptable level of quality and safety. Alternately, the licensee may revise the request for Category B-N-2 and B-N-3 components to demonstrate that compliance with the current ASME inspection interval requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality or safety as permitted by10 CFR 50.55a(z)(2).

Response

In response to this RAI, Entergy has revised the previously submitted relief request IP3-ISI-RR-11 to remove the ASME Code,Section XI, examination category B-N-2 and B-N-3 items. The revised relief request IP3-ISI-RR-11 is provided as Attachment 2 to this letter, and replaces the previously submitted relief request in its entirety. A new relief request IP3-ISI-RR-15 has been created to extend the ISi interval for the ASME Code,Section XI, examination category B-N-2 and B-N-3 items. The new relief request IP3-ISI-RR-15 is provided in Attachment 3 to this letter, and is being submitted on the basis that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety, as permitted by 10 CFR 50.55a(z)(2).

NL-18-053 ATTACHMENT 2 INDIAN POINT UNIT 3 REVISED RELIEF REQUEST IP3-ISI-RR-11 REACTOR VESSEL WELD INSERVICE INSPECTION INTERVAL EXTENSION ENTERGY.NUCLEAR OPERATIONS, INC.

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

NL-18-053 Attachment 2 Page 1 of 5 Indian Point Unit 3 Fourth 10 Year ISi Interval Relief Request No. IP3-1SI-RR-11 Reactor Vessel Weld lnservice Inspection Interval 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

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 inservice inspection (ISi) interval until the Spring 2019 refueling outage (RFO 20) based on approval of Relief Request IP3-ISI-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-18-053 Attachment 2 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-ISI-RR-06 (Reference 2) and the NRC SER (Reference 5) require that category B-A and 8-D vessel welds 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*1SI-RR-12) has been 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 nozzles. These proposed deferrals will extend the inspection interval from the Spring of 2019 (RFO 20) to the Spring of 2021 (RFO 21 ).

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 9), extending the third ISi interval until 2015. This original RPV weld IS I

NL-18-053 Attachment 2 Page 3 of 5 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 (JWCF) parameter. An alternative analysis to address the TWCF parameter deviation was provided in Relief Request RR-3:-43(1). The alternative TWCF analysis was later superseded by a plant specific change-in-risk analysis (Reference 10) in response to an NRC Request for Additional Information (RAI) (Reference 11).

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-ISI-RR-06 (Reference 5). Relief Request IP3-ISI-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-ISI-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-ISI-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 to the NRC in the response to Question 1 of Reference

10. 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 10). 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 criterion 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 release 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 1 0, 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 10. The response to 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 11 ), the results of the IP3 change-in-risk analysis were revised to include consideration of external events (Reference 12). This increased the bounding change in risk to 2.66E-08 events per year, which is still about a factor

NL-18-053 Attachment 2 Page 4 of 5 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 for WCAP-16168-NP-A, (b) IP3 provided a plant-specific change in TWCF analysis to demonstrate that the proposed change in the IP3 RPV IS I program meets the Regulatory Guide 1.174 guidelines discussed in the SER forWCAP-16168- NP-A, and (c) the IP3 proposed alternative provides an acceptable level of quality and safety.

As indicated in Relief Request IP3-ISI-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 defe~ral 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 10 year interval.

The duration of 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-ISI-RR-06 (ML14198A331)
2. Indian Point Unit 2 (IP2) Relief Request IP2-ISI-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-ISI-RR-06 for Reactor Vessel Weld Examination, Indian Point Unit Number 3," dated January 13, 2014 (ML14017A055)
3. NRC Letter to Entergy, "Request for Additional Information Regarding Relief Request IP3- ISI-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-ISI-RR-06, Reactor Vessel Weld Examination (TAC No. MF3345)," dated April 7, 2014 (ML14106A372)

NL-18-053 Attachment 2 Page 5 of 5

5. NRC Letter to Entergy, "Indian Point Nuclear Generating Unit No. 3 Safety Evaluation for Relief Request IP3-ISI-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. 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)

10. 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)
11. 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. MD9196 and MD9197)," dated November 20, 2008 (ML082971068)
12. 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)

- -----1 NL-18-053 ATTACHMENT 3 INDIAN POINT UNIT 3 RELIEF REQUEST IP3-ISI-RR-15 REACTOR VESSEL INSERVICE INSPECTION INTERVAL EXTENSION ENTERGY NUCLEAR OPERATIONS, INC.

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

NL-18-053 Attachment 3 Page 1 of 4 Indian Point Unit 3 Fourth 10 Year ISi Interval Relief Request No. IP3-IS1-RR-15 Reactor Vessel lnservice Inspection Interval Extension Proposed Alternative In Accordance with 10 CFR 50.55a(z)(2)

Hardship or Unusual Difficulty without a Compensating Increase in the 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-N-2 813.60 Interior Attachments Beyond Beltline Region B-N-3 813.70 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

Table IWB-2500-1, examination categories B-N-2 and B-N-3, item numbers 813.60 and 813.70, require a visual examination of the accessible interior attachment welds within and beyond the beltline region, and a visual examination of the accessible core support structure surfaces of the RPV once each ten year interval.

4. Reason for Request

In Westinghouse Topical Report WCAP-16168-NP-A, Revision 2 (Reference 2), the Pressurized

. Water Reactor Owners Group (PWROG) provided the technical and regulatory basis for decreasing the frequency of reactor vessel weld inspections by extending the ASME Code,Section XI, inservice inspection (ISi) interval from the current 10 years to 20 years for ASME Code,Section XI, examination categories B-A and B-D RPV welds. The Nuclear Regulatory Commission (NRC) approved the topical report by letter dated May 8, 2008 (Reference 3).

In revised Relief Request IP3-ISI-RR-11 (Attachment 2 to this letter), Entergy Nuclear Operations, Inc. (Entergy) requests to defer the required volumetric examination of the IP3

NL-18-053 Attachment 3 Page 2 of 4 category 8-A RPV pressure retaining welds and B-D full penetration welded nozzles from the Spring 2019 refueling outage (RFO 20) to the Spring 2021 refueling outage (RFO 21). The intent of this relief request (IP3-ISI-RR-15) is to allow deferral of the subject ASME BPV Code,Section XI, examination of the IP3 RPV category B-N-2 interior attachments and B-N-3 core support structure until the Spring 2021 refueling outage (RFO 21), so this inspection can be performed at the same time as the category B-A and 8-D inspection described in IP3-ISI-RR-11.

The B-N-2 and B-N-3 examinations were last performed in 2009 with no recordable indications.

Accordingly, Entergy is requesting to extend the B-N-2 and B-N-3 examinations for one fuel cycle, from RFO 20 to RFO 21, which will extend the inspection interval from 10 years to 12 years.

Extending the examination category B-N-2 and B-N-3 inspection to the Spring 2021 refueling outage (RFO 21) will allow the inspection to be scheduled coincident with the reactor vessel weld B-A and B-D inspection (IP3-ISI-RR-11), the ASME Code Case N-770-2 weld inspection (IP3-ISI-RR-12), and the Materials Reliability Program (MRP)-227-A reactor vessel internals inspections, and eliminate the need to remove the core barrel in 2019 (RFO 20). Since IP3 is currently scheduled to permanently cease operation no later than April 30, 2021 (Reference 4),

these inspections would only be performed if the current plan to permanently cease operation by April 2021 is changed in the future and IP3 continues to operate beyond this date.

Performing the examinations and inspections related to core barrel removal during the same refueling outage will result in significant savings in personnel radiation dose and outage duration since the same equipment and personnel used for visual and volumetric examination of the RPV shell welds and nozzle welds from the RPV interior can be used to implement the required reactor vessel internals examinations. Additionally, removing the reactor vessel internals only once to accommodate all the examinations discussed in this relief request would result in significant savings in radiation exposure.

5. Proposed Alternative and Basis for Use The proposed alternative inspection schedule would enable the subject examinations to be performed during the same refueling outage as the RPV shell, head, and nozzle weld inspections. In accordance with 10 CFR 50.55a(z)(2), this interval extension is requested on the basis that performing the examination of the RPV interior attachments and core support structure on a different schedule than the RPV shell, head, and nozzle welds would result in hardship or unusual difficulty without a compensating increase in the level of quality or safety.

The full scope examination required by ASME BPV Code,Section XI, examination categories B-N-2 and B-N-3 requires the removal of all of the fuel and the core barrel from the RPV. An unnecessary risk is created by removal of the core barrel to perform a visual examination without a compensating increase in quality or safety. Further, the personnel radiation exposure to establish the conditions for and perform the ASME Code,Section XI, B-N-2 and B-N-3 examinations would essentially double if the subject examinations were performed at a time separate from the RPV shell, lower head, and nozzle weld examinations. The visual examinations of the RPV interior attachments and the core support structure have been performed during the IP3 Third ISi Interval (2009) with no recordable indications noted.

As stated in Reference 2, " .. .it must be recognized that all reactor coolant pressure boundary failures occurring to date have been identified as a result of leakage, and were discovered by visual examination. The proposed RV ISi interval extension does not alter the visual

NL-18-053 Attachment 3 Page 3 of 4 examination interval. The reactor vessel would undergo, as a minimum, the Section XI Examination Category 8-P pressure tests and visual examinations conducted at the end of each refueling before plant start-up, as well as leak tests with visual examinations that precede each starl-up following maintenance or repair activities." The visual examinations discussed in Reference 2 are not the subject examinations (i.e., B-N-2 and B-N-3) of this relief request.

Entergy currently performs ASME examination category B-N-1 visual examinations during refueling outages, which includes the locations that are made accessible for examination by the removal of components during normal refueling outages. This examination is required once each period and provides additional assurance that the internal surfaces of the reactor vessel have not been subjected to abnormal conditions, and are not susceptible to unanticipated degradation mechanisms. These examinations have been performed each examination period, with no unacceptable conditions identified, and provide reasonable assurance of structural integrity.

As discussed further in Reference 2, defenses against human errors are preserved with the proposed extension of the inspection interval. Specifically, the increase in the inspection interval will reduce the frequency for which the reactor vessel lower internals need to be removed, thereby reducing the possibility for human error and damage to the reactor vessel or internals. Therefore, in accordance with 10 CFR 50.55a(z)(2), the proposed interval change from 10 years to 12 years for the subject ASME Code,Section XI, B-N-2 and B-N-3 examinations is requested for IP3 on the basis that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. '

6. Duration of Proposed Alternative This request is applicable to Entergy's ISi program for the IP3 fourth 10 year interval. The duration of the proposed alternative is from the Spring of 2019 (RFO 20) to the Spring of 2021 (RFO 21).
7. Precedents
1. Calvert Cliffs Nuclear Power Plant, Unit 2, Relief Request ISl-021 (ML090920077)
2. Donald C. Cook Nuclear Plant, Unit 2, Relief Request ISIR-30 (ML091320549)
3. Salem Nuclear Generating Station, Units 1 and 2, Relief Request SC-13R-95 (ML100491550)
4. Three Mile Island Nuclear Station, Unit 1, Relief Request RR-09-02 (ML102390018)

The NRC approved changes to the ASME Code,Section XI, B-N-2 and B-N-3 inspection intervals listed above have all been for extension periods that were greater than the interval change from 10 years to 12 years requested for IP3 in this relief request.

8. References
1. ASME Boiler and Pressure Vessel Code,Section XI, 2001 Edition through 2003 Addenda, American Society of Mechanical Engineers, New York

NL-18-053 Attachment 3 Page 4 of 4

2. WCAP-16168-NP-A, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval," Revision 2, June 2008 (ML082820046)
3. NRC Letter to PWROG, "Final Safety Evaluation for Pressurized Water Reactor Owners Group (PWROG) Topical Report (TR) WCAP-16168-NP, Revision 2, 'Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval (TAC No.

MC6768),"' dated May 8, 2008 (ML081060053)

4. 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)
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