NL-13-052, Reply to Request for Additional Information Regarding the License Renewal Application

Reply to Request for Additional Information Regarding the License Renewal Application

ML13142A202
Person / Time
Site:	Indian Point
Issue date:	05/07/2013
From:	Dacimo F Entergy Nuclear Northeast
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-13-052
Download: ML13142A202 (34)

Text

Enteravy Nuclear Northeast Indian Point Energy Center S'Entergy-- 450 Broadway, GSB3 P.O. Box 249 Buchanan, NY 10511-0249 Tel (914) 254-2055 Fred Dacimno Vice President Operations License Renewal NL-13-052 May 7, 2013 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Reply to Request for Additional Information Regarding the License Renewal Application Indian Point Nuclear Generating Unit Nos. 2 & 3 Docket Nos. 50-247 and 50-286 License Nos. DPR-26 and DPR-64

REFERENCE:

1. NRC letter, "Request for Additional Information for the Review of the Indian Point Nuclear Generating Unit Nos. 2 and 3, License Renewal Application, SET 2013-01" dated February 6, 2013.

2. Entergy Letter, NL-12-140, "Reply to Request for Additional Information Regarding the License Renewal Application, Indian Point Nuclear Generating Unit Nos. 2 and 3, Docket Nos. 50-247 and 50-286, License Nos. DPR-26 and DPR-64," October 17, 2012.

Dear Sir or Madam:

Entergy Nuclear Operations, Inc is providing, in Attachment 1, a reply to the additional information requested in Reference 1 pertaining to NRC review of the License Renewal Application (LRA) for Indian Point 2 and Indian Point 3.

On review of the information requested in RAI 15a, Entergy has determined that it will rely on the Fatigue Monitoring Program (FMP) to manage the effects of aging due to fatigue on the reactor internals through the PEO rather than relying on the Reactor Vessel Internals (RVI) inspection program as described in Reference 2. Therefore a revised response to RAI 15 is also included in Attachment 1.

42t

Docket Nos. 50-247 & 50-286 NL-13-052 Page 2 of 2 The revised response to RAI 15 also includes new Commitment 49 that addresses the review of reactor vessel internals for environmentally assisted fatigue. The response to RAI 11 a includes a revision to the implementation date for Commitment 47. These new and revised commitments are included in the latest list of regulatory commitments provided in Attachment 2.

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

I declare under penalty of perjury that the foregoing is true and correct. Executed on May 7, 2013.

Sincerely, FRD/rw

Attachment:

1. Reply to NRC Request for Additional Information Regarding the License Renewal Application

2. License Renewal Application IPEC List of Regulatory Commitments Revision 21.

cc: Mr. William Dean, Regional Administrator, NRC Region I Mr. Sherwin E. Turk, NRC Office of General Counsel, Special Counsel Mr. Dave Wrona, NRC Branch Chief, Engineering Review Branch I Mr. Nathaniel Ferrer, NRC Project Manager, Division of License Renewal Mr. Douglas Pickett, NRR Senior Project Manager Ms. Bridget Frymire, New York State Department of Public Service NRC Resident Inspector's Office Mr. Francis J. Murray, Jr., President and CEO NYSERDA

ATTACHMENT I TO NL-13-052 REPLY TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE LICENSE RENEWAL APPLICATION ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 & 3 DOCKET NOS. 50-247 AND 50-286

Docket Nos. 50-247 & 50-286 NL-13-052 Attachment 1 Page 1 of 10 INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3 LICENSE RENEWAL APPLICATION (LRA)

REQUESTS FOR ADDITIONAL INFORMATION (RAI)

NRC RAI Ila (RVI Program and combined effects of embrittlement)

Back-ground In request for additional information (RAI) 11, the staff requested additional information on the approach to be used for the plant-specific evaluation of the cast austenitic stainless steel (CASS) lower support column bodies. The applicant's response indicates it plans to use a screening approach using the screening criteria for thermal aging embrittlement susceptibility from the Nuclear Regulatory Commission (NRC) staff's May 19, 2000 letter (Reference 1). The applicant provided a table of the screening criteria based on chemistry, casting method, and delta ferrite content identical to Table 2 of Reference 1.

Issue However, Reference 1 also recommends that to account for a potential synergistic effect on loss of fracture toughness due to the combined effects of thermal embrittlement (TE) and neutron irradiation embrittlement (IE), a component-specific assessment should be performed for components that will experience neutron fluence of lx1 017 neutrons per square centimeter (n/cm 2) or greater. Supplemental inspections would be recommended for those components that are potentially susceptible to TE and IE, that are also subject to significant tensile loadings under any normal operating or design basis condition. Per Table 4-6 of MRP-191 (Reference 2), the screening value of the neutron fluence for the lower support column bodies for Westinghouse-designed reactor vessel internals (RVI) is 1x10 22 to 5x1 022 n/cm 2 . This is significantly greater than the lx1017 n/cm 2 threshold value provided in Reference 1.

Request Describe how the effects of neutron fluence, with respect to a potential synergistic effect of TE and IE, will be addressed in the plant-specific evaluation of the lower support column bodies.

The applicant should propose modifications of the aging management requirements for the lower support column bodies as necessary to address the concern with a potential synergistic effect.

The staff notes that for CASS, different neutron fluence thresholds, above which the synergistic effect of TE and IE must be considered, have been proposed in various industry documents. If a threshold value greater than 1x10 17 n/cm 2 is used in the applicant's evaluation of the potential synergistic effect, a technical justification should be included for the threshold value chosen.

The technical justification should include a description of the material test data used as the basis for the threshold including material type(s), thermal aging time and temperature, neutron fluence, type of reactor in which the irradiation was conducted, and relevant mechanical testing results.

Response to NRC RAI 1Ia The screening approach used to evaluate the CASS components at IP2 and IP3 is consistent with the methodology developed in MRP-227-A (Reference 1). The screening value for

Docket Nos. 50-247 & 50-286 NL-13-052 Attachment 1 Page 2 of 10 irradiation embrittlement of CASS components is based on the values cited in MRP-227-A and originally identified with supporting information in MRP-191 (Reference 2). Both MRP-227-A and MRP-191 have been made available by the industry for several years and MRP-227-A has been the subject of a safety evaluation (SE) by the NRC. On this basis, the lower support columns screen in because the most irradiated regions of these components are expected to be exposed to fluences in the range of 1 x 1022 to 5 x 1022 n/cm 2 as reported in Table 4-6 of MRP-191. Since effects of embrittlement are only significant in the presence of pre-existing flaws (e.g. from the casting process) and tensile stresses capable of propagating these flaws, the screening analysis will identify regions of individual columns where thermal and irradiation effects could give rise to embrittled materials and would also be subjected to significant tensile stresses under design and service loadings. For such regions, a functionality assessment will be conducted to determine the impact of column fracture on the lower core support plate structure. Based on the lack of any documented history of fracture in the lower core support columns, it will be assumed that only a limited number of columns could actually contain flaws of significant size. The assessment will evaluate distributions of fractured columns that can be tolerated without the loss of critical core support function.

The screening-in value of fluence for thermal embrittlement of CASS depends on the chemistry and fabrication of the CASS. As noted in the NRC document "Thermal Aging Embrittlement of Cast Austenitic Steel Components" (Reference 3), embrittlement of the composite Austenite+Ferrite structure depends on the distribution of the Ferrite constituent in the microstructure. If the Ferrite is distributed so that "it forms a continuous phase surrounding the grain boundaries" the material will be "made susceptible to low energy fracture", i.e., the fracture of the material is controlled by the low energy fracture of the embrittled Ferrite. Conversely, thermally exposed CASS microstructures in which the potentially embrittled Ferrite constituent particles are isolated and surrounded by larger volumes of Austenite structures do not exhibit thermally embrittled behavior. In these cases, Austenite, which is not susceptible to thermal embrittlement, controls the overall fracture behavior. The microstructures that exhibit such differences in behavior are produced by different alloy chemistries and casting practices. These effects on CASS susceptibility to thermal embrittlement were summarized in Table 1 of Reference 3 as shown below.

Table I Thermal Aging Screening Criteria from Table 1 of Reference 3 Molybdenum Casting Delta-Ferrite NRC Susceptibility Evaluation (Wt%) Method  %

Static >14% Potentially Susceptible to TE High 2.0-3.0 or< 14% Not Susceptible to TE Centrifugal >20% Potentially Susceptible to TE C or< 20% Not Susceptible to TE Static >20% Potentially Susceptible to TE Low 0.5 max = or< 20% Not Susceptible to TE Centrifugal All Not Susceptible to TE The microstructural distribution of Austenite and Ferrite within CASS has a similar but not identical effect on irradiation embrittlement. When irradiated, both Ferrite and Austenite can embrittle. However, the degree to which embrittlement is incurred and the onset of embrittled behavior occurs at different fluences for the two phases. Ferrite embrittles at a much lower

Docket Nos. 50-247 & 50-286 NL-13-052 Attachment 1 Page 3 of 10 fluence than Austenite. Also, even after saturation embrittlement, the remaining toughness of Austenite is still significantly higher than that of embrittled Ferrite. U. S. Nuclear Regulatory Commission reports indicate that embrittlement of pure austenite occurs in the regime of 3.5 x1021 to 5.6 x1021 n/cm 2 (References 4 and 5). Even in the presence of small amounts of Ferrite distributed within the CASS, it has been reported that the onset of this form of embrittlement only occurs at fluences above 6.7 x 1020 n/cm 2 (References 5 and 6). Since Austenite does not thermally embrittle, there cannot be synergistic effects when only small amounts of Ferrite are present.

In the case that the Ferrite forms continuous grain boundary constituents, it is expected that the CASS will embrittle at much lower fluences. The fluence that would impart irradiation embrittlement to this form of CASS is that at which the Ferrite phase undergoes irradiation embrittlement. The RAI has proposed that this level should be as low as 1 x 1017 n/cm 2 . While there are several reasons why a higher value may be more appropriate for the embrittlement of Ferrite within CASS structures (viz., the elemental composition of this Ferrite can be significantly different from that which has been shown to exhibit embrittlement at such low fluences as observed in certain low alloy pressure vessel steels and particularly their welds), it is not necessary to consider this behavior for the CASS in Indian Point 2 and 3 since the CASS component regions that will be exposed to potentially embrittling cumulative fluences will also have been sufficiently thermally exposed to render full embrittlement of the Ferrite. The materials will, therefore, have already been screened-in on the basis of susceptibility to thermal embrittlement. Thus there is no need to consider irradiation embrittlement or synergistic action.

Based on the foregoing, the screening criteria for thermal and irradiation embrittlement of the CASS in the Indian Point Plants 2 and 3 may be summarized as:

Table 2 Thermal Aging and Irradiation Embrittlement Screening Criteria for CASS Molybdenum Casting Delta- Thermal Irradiation Susceptibility (Wt%) Method Ferrite % Susceptibility Screening Fluence 2

n/cm

>14% Screen-In as Potentially Susceptible Static to TE

= or< 14% Not Susceptible to TE Screen-In if Screen-in as Fluence > 6.7 x 1020

>20%

High 2.0-3.0 Potentially Susceptible Centrifugal to TE

= or< 20% Not Susceptible to TE Screen-in if Fluence > 6.7 x 1020

>20% Screen-In as Potentially Susceptible Static to TE Low 0.5 max = or< 20% Not Susceptible to TE Screen-in if Fluence > 6.7 x 1020 Centrifugal All Not Susceptible to TE Screen-in if I I Fluence > 6.7 x1020

Docket Nos. 50-247 & 50-286 NL-13-052 Attachment 1 Page 4 of 10 Commitment 47 - revision to implementation date The implementation date for commitment 47 for IP2 is being revised from September 28, 2013 to March 1, 2015 based on the following:

In accordance with MRP-227-A, the lower support column bodies are identified as Expansion Components at IP2. They are only required to be inspected if surface-breaking indications are detected in two or more control rod guide tube (CRGT) lower flange welds, which are the Primary Components that are linked to the lower support column bodies as Expansion Components. The initial inspections for the CRGT lower flange welds are required to be performed no later than 2 refueling outages from the beginning of the license renewal period, which for IP2 is March 2016.

As indicated in Section 3.3.7 of the SER for MRP-227, the purpose of the analysis is to demonstrate that the MRP-227 recommended inspections will ensure functionality of the set of components between scheduled inspections. Changing the implementation schedule for the IP2 lower support column bodies analysis from September 28, 2013 to March 1, 2015 still ensures the analyses in Commitment 47 are completed prior to the first required MRP-227-A inspection in 2016, which is consistent with MRP-227-A and the associated NRC safety evaluation.

References

1. Materials Reliability Program : Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A) EPRI, Palo Alto, CA: 2011. TR-1 022863

2. Materials Reliability Program: Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design (MRP-1 91).

EPRI, Palo Alto, CA: 2006. TR-1013234

3. U S Nuclear Regulatory Commission "Thermal Aging Embrittlement of Cast Austentic Stainless Steel Components," License Renewal Issue N 98-0030, C.l.Grimes, U S Nuclear Regulatory Commission, May 19th 2000 - availability ADAMS Doc base ML003717179.

4. 0. K. Chopra, "Degradation of LWR Core Internal Materials due to Neutron Irradiation,"

NUREG/CR-7027, U S Nuclear Regulatory Commission (December 2010).

5. 0. K. Chopra and W. L. Shack, "Crack Growth Rates and Fracture Toughness of Irradiated Austenitic Stainless Steels in BWR Environments," NUREG/CR-6960, US Nuclear Regulatory Commission (March 2008).

6. Westinghouse Report WCAP-17141-NP, Rev. 0, "Thermal and Irradiation Embrittlement of Cast and Welded Austenitic Stainless Steel Reactor Internals," September 2009.

NRC RAI 15a (RVI program and Fatigue)

Background

In its response to RAI 15, Question 1, by letter dated October 17, 2012 (Reference 3), the applicant revised its response to RAI 12 to indicate that it intends to use the RVI Program to manage the cracking-fatigue aging effect for RVI components that have a time-limited aging analysis (TLAA) that determined a cumulative usage factor (CUF). The applicant provided a list of the RVI components that have a CUF analysis, a table cross-indexing

Docket Nos. 50-247 & 50-286 NL-13-052 Attachment 1 Page 5 of 10 these components with the equivalent component name in MRP-227-A, along with the inspection requirements, and a justification for each component with a CUF that the inspection requirements are adequate to manage the cumulative fatigue damage aging effect.

Part 5 of Action Item 8 of the staff's final safety evaluation (SE) of MRP-227-A contains two requirements that must be fulfilled by licensees that intend to use the RVI Program to manage the cracking-fatigue aging effect for components with a TLAA for fatigue:

1. For those CUF analyses that are TLAAs, the applicant may use the pressurized-water reactor (PWR) Vessel Internals Program as the basis for accepting these CUF analyses in accordance with 10 CFR 54.21 (c)(1)(iii) only if the RVI components in the CUF analyses are periodically inspected for fatigue-induced cracking in the components during the period of extended operation.

2. The periodicity of the inspections of these components shall be justified to be adequate to resolve the TLAA.

Many of the RVI components with TLAA analyses for both Indian Point (IP) Units 2 and 3 (IP2 and IP3) are either "existing programs" or "no additional measures" components under MRP-227-A, which are inspected under the ASME Section Xl, Inservice Inspection Program and are thus only subject to a VT-3 visual examination. Those components categorized as "expansion" may or may not be inspected under the RVI Program based on the findings of the RVI Inspection Program examinations of the linked "primary" component(s). Additionally, a VT-3 visual examination may not be adequate for all components for detecting fatigue cracking prior to the occurrence of structurally significant cracking, although the staff notes that VT-3 examination is used for some components that were determined to be primary components for fatigue (such as thermal shield flexures and baffle-edge bolts).

In general, a justification for the inspection periodicity was not provided in the response to RAI

15. The default inspection periodicity for most "primary" inspection category components in MRP-227-A is every ten years following the initial inspection.

All of the CUFs for RVI components provided in Tables 4.3-5 and 4.3-6 of the IP2 & IP3 license renewal application (LRA) are less than 1.0. However, these CUFs were determined without the application of an environmental correction factor (Fen) to account for the effects of the reactor coolant environment. However, it is reasonable to conclude that if the reactor coolant environment affects the fatigue usage of other components in the reactor coolant system and reactor pressure vessel, then it would affect the RVI components similarly. The Fen for the reactor pressure vessel (RPV) components and reactor coolant system given in Section 4.3.4 of the LRA range from 2.45 to 15.35. Application of Fen in this range could cause the CUF of some RVI components to exceed 1.0. This would affect the required periodicity of inspection. For a very high environmentally-adjusted CUF, even a 10-year inspection interval may not be adequate.

Issue

1. Most of the RVI components with a fatigue TLAA analysis are not "Primary" inspection category components under the RVI Program, thus may be subject to no inspection other than a VT-3 visual examination under the ASME Section XI, Inservice Inspection Program, since "Expansion" category component inspections

Docket Nos. 50-247 & 50-286 NL-13-052 Attachment 1 Page 6 of 10 are only triggered in the event of degradation of the linked "Primary" inspection category component(s).

2. The licensee did not justify the adequacy of the periodicity of the RVI Program inspections performed on RVI components that have fatigue TLAA analyses.

3. The staff considers the inspection techniques required by MRP-227-A for components in the "Primary," "Expansion," or "Existing Programs" categories, for which fatigue is a screened-in aging mechanism, adequate to detect cracking due to fatigue if the RVI Program is credited for managing a fatigue TLAA. However, those components that fall into the "No Additional Measures" category under MRP-227-A have no specified examination techniques, periodicity, coverage, and acceptance criteria in MRP-227-A.

Requested Information

1. For those RVI components having fatigue TLAA analyses for which the cumulative fatigue damage aging effect is to be managed by the RVI Inspection Program, but which are classified as "Expansion," "Existing Programs," or "No Additional Measures" inspection category components, provide a modification to the RVI Inspection Program to re-categorize these components as "Primary" inspection category components. If any such components are to remain in the "Expansion" category, provide a technical justification for potentially never inspecting these components. Discuss your plans to re-categorize these components as "Primary" inspection category components.

2. For those RVI components having fatigue TLAA analyses for which the cumulative fatigue damage aging effect is to be managed by the RVI Inspection Program, provide a quantitative justification that the periodicity of inspections for fatigue is adequate, either in terms of the calculated CUF (considering the effects of the environment on the CUF analysis), or by using a flaw tolerance approach.

3. For those RVI components having fatigue TLAA analyses for which the cumulative fatigue damage aging effect is to be managed by the RVI Inspection Program and which are classified as "No Additional Measures" components under MRP-227-A, identify the examination technique, coverage, periodicity, and acceptance criteria (i.e., provide the equivalent information to that provided in Tables 4-3 and 5-3 of MRP-227-A). In addition, provide the information requested in Parts 1 and 2 of this RAI for these components.

References

1. U.S. Nuclear Regulatory Commission Letter, "License Renewal Issue No. 98-0030, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Components," May 19, 2000 (NRC ADAMS Accession No. ML003717179)

2. MRP-191 Revision 0, "Materials Reliability Program: Screening, Categorization and Ranking of Reactor Internals of Westinghouse and Combustion Engineering PWR Designs," ADAMS Accession Number ML091910130

3. Letter from F. Dacimo to NRC dated October 17, 2012,

Subject:

Indian Point Nuclear Generating Unit Nos. 2 & 3 - Reply to Request for Additional Information

V.-

Docket Nos. 50-247 & 50-286 NL-13-052 Attachment 1 Page 7 of 10 Regarding the License Renewal Application. (ADAMS Accession No.

ML12300A391)

Response to RAI 15a In Reference 1 Entergy responded to NRC RAI 15 which requested clarification on how Entergy planned to address RVI locations with existing CUFs as described in the Fatigue Monitoring Program (FMP). In the response to RAI 15, Entergy indicated that it planned to use the RVI Inspection (MRP-227) program rather than the FMP because the inspections provided in the RVI program were sufficient to ensure that the effects of aging due to fatigue would be adequately managed.

In Reference 2 (NRC letter transmitting RAI 15a) the NRC requested additional justification to demonstrate that the inspection plan and inspection frequency provided in the IPEC RVI inspection program were sufficient to ensure that the effects of aging due to fatigue on those internals locations with existing CUFs would be adequately managed during the period of extended operation (PEO) including either consideration of the environmental effects from the reactor coolant environment or a flaw tolerance evaluation.

On review of the information requested in RAI 15a, Entergy has determined that it will rely on the FMP to manage the effects of fatigue on the reactor internals during the PEO rather than the RVI inspection program as previously indicated in Reference 1. The revised response to the original RAI 15 is provided below.

NRC RAI 15 The response to RAI 12 states that, for RVI components that are not covered by a time-limited aging analysis, Entergy will use the RVI Program to manage the effects of aging due to fatigue on the reactor vessel internals. The response also states that, as provided in Section 3.5.1 of the NRC's safety evaluation for MRP-227-A, for locations with a fatigue time-limited aging analysis, Entergy will manage the effects of aging due to fatigue through its Fatigue Monitoring Program in accordance with 10 CFR 54.21 (c)(1)(iii).

In its response, the applicant also stated that the Fatigue Monitoring Program as described in LRA Section B. 1.12 provides assurance that the cumulative usage factors (CUFs) remain below the allowable limit of 1.0 and that, consistent with Section 3.5.1 of the safety evaluation for MRP-227-A, prior to entering the period of extended operation, Entergy will review the existing RVI fatigue calculations to evaluate the effects of the reactor coolant system water environment on the CUF. Specifically, under Commitment 43, Entergy stated that it will review the units' design basis ASME Code Class 1 fatigue evaluations to determine whether the NUREG/CR-6260 locations that have been evaluated for the effects of the reactor coolant environment on fatigue usage are the limiting locations for IP2 and IP3. The applicant stated that this review will also include ASME Code Class 1 fatigue evaluations for reactor vessel internals. Based on this review, if more limiting locations are identified, Entergy will evaluate the most limiting location for the effects of the reactor coolant environment on fatigue usage. The applicant's response is not clear regarding how the "ASME Code Class 1 fatigue evaluations for reactor vessel internals" will account for the effects of the reactor coolant environment, nor what actions will be taken if CUF's for RVI components exceed 1.0.

Docket Nos. 50-247 & 50-286 NL-13-052 Attachment 1 Page 8 of 10 Requested Information

1. Clarify whether, as a result of the review described in the response to RAI 12, CUF calculations for RVI components that incorporate environmental factors (Fen) will be performed in response to Applicant/Licensee Action Item 8 of the Staff SE of MRP-227-A. If such calculations will not be performed, discuss how the effects of the reactor water environment on the existing CUF analyses for RVIs will be evaluated in response to Applicant/Licensee Action Item 8 of the Staff SE of MRP-227-A.

2. Clarify what action(s) will be taken if the consideration of environmental effects results in a CUF exceeding 1.0 for any RVI component.

3. Since ASME Code Class 1 components are designed to ASME Section III, Subsection NB (i.e., reactor coolant pressure boundary components, not reactor vessel internals),

provide necessary revisions to clarify the term "ASME Code Class 1 fatigue evaluations for reactor vessel internals" and any inconsistency in the response to RAI 12.

4. For the purposes of clarity, provide a new commitment and an associated new UFSAR Supplement to address the review of reactor vessel internals for environmentally-assisted fatigue as part of the Fatigue Monitoring Program in response to Applicant/Licensee Action Item 8 of the Staff SE of MRP-227-A, in lieu of your proposal to use Commitment 43.

Revised Response to RAI 15

1. As part of the review described in response to RAI 12, Entergy will recalculate the limiting reactor vessel internals CUFs provided in section 4.3 of the License Renewal Application (LRA) to include the effects of the reactor coolant environment. Entergy will use the environmental correction factors (Fen) provided in NUREG/CR-5704, "Effects of LWR Coolant Environments on the Fatigue Design Curves of Austenitic Stainless Steels" or NUREG/CR-6909, "Effects of LWR Coolant Environments on the Fatigue of Reactor Materials" for austenitic stainless steel components. Entergy will use the environmental correction factors (Fen) provided in NUREG/CR-6909, "Effects of LWR Coolant Environments on the Fatigue of Reactor Materials" for nickel alloy components.

2. In accordance with the corrective actions specified in the Fatigue Monitoring Program, corrective actions include further CUF re-analysis, and/or repair or replacement of the affected components prior to the CUFen reaching 1.0. Analysis methods may include finite element analyses (FEA) or other appropriate methods using actual plant operating parameters and the number of cycles expected through the end of the PEO.

3. Term "Class 1" was inadvertently included in the response to RAI 12. The phrase "ASME Code Class 1 fatigue evaluations for reactor vessel internals" is changed to read "ASME Code Subsection NG fatigue evaluations for reactor vessel internals."

4. Entergy provides the following commitment for including the reactor coolant environmental effects in the existing CUFs for the RVI.

Docket Nos. 50-247 & 50-286 NL-13-052 Attachment 1 Page 9 of 10 Commitment 49 Recalculate each of the limiting CUFs provided in section 4.3 of the LRA for the reactor vessel internals to include the reactor coolant environment effects (Fen) as provided in the IPEC Fatigue Monitoring Program using NUREG/CR-5704 or NUREG/CR-6909. In accordance with the corrective actions specified in the Fatigue Monitoring Program, corrective actions include further CUF re-analysis, and/or repair or replacement of the affected components prior to the CUFen reaching 1.0.

The following changes (identified by strikethrough) are made to LRA Section A.2.2.2 and A.3.2.2.

A.2.2.2 Metal Fatigue A.2.2.2.1 Class 1 Metal Fatigue Class 1 components evaluated for fatigue and flaw growth include the reactor pressure vessel (RPV), reactor vessel internals, pressurizer, steam generators, reactor coolant pumps, control rod drive mechanisms, regenerative letdown heat exchanger, and Class 1 piping and in-line components.

The Fatigue Monitoring Program will assure that the analyzed number of transient cycles is not exceeded. The program requires corrective action if the analyzed number of transient cycles is approached. Consequently, the effects of aging related to these TLAA (fatigue analyses) based on those transients will be managed by the Fatigue Monitoring Program in accordance with 10 CFR 54.21 (c)(1)(iii).

As indicated in EPRI MVRP 227 A, the effects of aging duo to fatigue were considered in determ~ining the necessr ipetions for reactor vessel internals components. Consistent wt MRP 22:7 A, during the perioid of extended operation, component inspectionS pe~frmed under the Reactor VesselI nternals Program and the Inserlvico Inspecltion PrOgram wiii manage the effects of aging due to fatigue of reactor vessel internals components in accordance with 10 CER 54.21 (c)(1)(iii).

A.3.2.2 Metal Fatigue A.3.2.2.1 Class 1 Metal Fatigue Class 1 components evaluated for fatigue and flaw growth include the reactor pressure vessel (RPV), reactor vessel internals, pressurizer, steam generators, reactor coolant pumps, control rod drive mechanisms, regenerative letdown heat exchanger, and Class 1 piping and in-line components.

The Fatigue Monitoring Program will assure that the analyzed number of transient cycles is not exceeded. The program requires corrective action if the analyzed number of transient cycles is approached. Consequently, the effects of aging related to these TLAA (fatigue analyses) based on those transients will be managed by the Fatigue Monitoring Program in accordance with 10 CFR 54.21(c)(1)(iii).

Docket Nos. 50-247 & 50-286 NL-13-052 Attachment 1 Page 10 of 10 As Enteg in EPRI MRP 227 A, the effectsef aging due to fatigue wefre onsidegredin determning the necea inpctionS for reaort vessel internals ogmponents. Consistentdit MRP 227 A, during the peid fcendcd operation, component inspecstions peftrmed under the Reactor Vessel Inte Pr and the Nnserice Inspection ProgDPra will manage the effects of aging due to fatigue of reacator vessel internalS components in accordance with 10 CFR 5'1.2-1(G)(1)(ifii).

References

1. Entergy Letter, NL-12-140, "Reply to Request for Additional Information Regarding the License Renewal Application, Indian Point Nuclear Generating Unit Nos. 2 and 3, Docket Nos. 50-247 and 50-286, License Nos. DPR-26 and DPR-64," October 17, 2012

2. NRC letter, "Request for Additional Information for the Review of the Indian Point Nuclear Generating Unit Nos. 2 and 3, License Renewal Application, SET 2013-01" dated February 6, 2013

ATTACHMENT 2 TO NL-13-052 LICENSE RENEWAL APPLICATION IPEC LIST OF REGULATORY COMMITMENTS Rev. 21 ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 & 3 DOCKET NOS. 50-247 AND 50-286

Docket Nos. 50-247 & 50-286 NL-13-052 Attachment 2 Page 1 of 20 List of Regulatory Commitments Rev. 21 The following table identifies those actions committed to by Entergy in this document.

Changes are shown as strikethroughs for deletiens and underlines for additions.

COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION I AUDIT ITEM Enhance the Aboveground Steel Tanks Program for P2: NL-07-039 A.2.1.1 IP2 and IP3 to perform thickness measurements of eptember 28, A.3.1.1 the bottom surfaces of the condensate storage tanks, 013 B.1.1 city water tank, and fire water tanks once during the IP&

first ten years of the period of extended operation. December 12, Enhance the Aboveground Steel Tanks Program for 2015 IP2 and IP3 to require trending of thickness measurements when material loss is detected.

2 Enhance the Bolting Integrity Program for IP2 and IP3 eP2: NL-07-039 A.2.1.2 to clarify that actual yield strength is used in selecting eptember 28, A.3.1.2 materials for low susceptibility to SCC and clarify the 013 B.1.2 prohibition on use of lubricants containing MoS 2 for IP3: NL-07-153 Audit Items bolting. ecember 12, 201,241, The Bolting Integrity Program manages loss of 2015 270 1 preload and loss of material for all external bolting. __

Docket Nos. 50-247 & 50-286 NL-13-052 Attachment 2 Page 2 of 20

1. COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION I

I /IAUDIT ITEM IP2: NL-07-039 A.2.1.5 3 Implement the Buried Piping and Tanks Inspection A.3.1.5 September 28, Program for IP2 and IP3 as described in LRA Section 2013 B.1.6 B.1.6.

NL-07-153 Audit Item This new program will be implemented consistent with IP3: 173 the corresponding program described in NUREG- December 12, 1801 Section XI.M34, Buried Piping and Tanks 2015 Inspection.

NL-09-106 Include in the Buried Piping and Tanks Inspection Program described in LRA Section B.1.6 a risk NL-09-111 assessment of in-scope buried piping and tanks that includes consideration of the impacts of buried piping or tank leakage and of conditions affecting the risk for corrosion. Classify pipe segments and tanks as having a high, medium or low impact of leakage based on the safety class, the hazard posed by fluid contained in the piping and the impact of leakage on reliable plant operation. Determine corrosion risk through consideration of piping or tank material, soil resistivity, drainage, the presence of cathodic protection and the type of coating. Establish inspection priority and frequency for periodic inspections of the in-scope piping and tanks based on the results of the risk assessment. Perform inspections using inspection techniques with demonstrated effectiveness. NL-1 1-101

Docket Nos. 50-247 & 50-286 NL-13-052 Attachment 2 Page 3 of 20 COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION I AUDIT ITEM IP2: NL-07-039 A.2.1.8 4 Enhance the Diesel Fuel Monitoring Program to September 28, A.3.1.8 include cleaning and inspection of the IP2 GT-1 gas 2013 B.1.9 turbine fuel oil storage tanks, IP2 and IP3 EDG fuel oil NL-07-153 Audit items day tanks, IP2 SBO/Appendix R diesel generator fuel IP3: 128, 129, oil day tank, and IP3 Appendix R fuel oil storage tank December 12, 132, and day tank once every ten years. NL-08-057 491,492, 2015 Enhance the Diesel Fuel Monitoring Program to 510 include quarterly sampling and analysis of the IP2 SBO/Appendix R diesel generator fuel oil day tank, IP2 security diesel fuel oil storage tank, IP2 security diesel fuel oil day tank, and IP3 Appendix R fuel oil storage tank. Particulates, water and sediment checks will be performed on the samples. Filterable solids acceptance criterion will be less than or equal to 10mg/l. Water and sediment acceptance criterion will be less than or equal to 0.05%.

Enhance the Diesel Fuel Monitoring Program to include thickness measurement of the bottom of the following tanks once every ten years. IP2: EDG fuel oil storage tanks, EDG fuel oil day tanks, SBO/Appendix R diesel generator fuel oil day tank, GT-1 gas turbine fuel oil storage tanks, and diesel fire pump fuel oil storage tank; IP3: EDG fuel oil day tanks, EDG fuel oil storage tanks, Appendix R fuel oil storage tank, and diesel fire pump fuel oil storage tank.

Enhance the Diesel Fuel Monitoring Program to change the analysis for water and particulates to a quarterly frequency for the following tanks. IP2: GT-1 gas turbine fuel oil storage tanks and diesel fire pump fuel oil storage tank; IP3: Appendix R fuel oil day tank and diesel fire pump fuel oil storage tank.

Enhance the Diesel Fuel Monitoring Program to specify acceptance criteria for thickness measurements of the fuel oil storage tanks within the scope of the program.

Enhance the Diesel Fuel Monitoring Program to direct samples be taken and include direction to remove water when detected.

Revise applicable procedures to direct sampling of the onsite portable fuel oil contents prior to transferring the contents to the storage tanks.

Enhance the Diesel Fuel Monitoring Program to direct the addition of chemicals including biocide when the presence of biological activity is confirmed.

Docket Nos. 50-247 & 50-286 NL-13-052 Attachment 2 Page 4 of 20

1. COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION I AUDIT ITEM 5 Enhance the External Surfaces Monitoring Program eP2: NL-07-039 A.2.1.10 for IP2 and IP3 to include periodic inspections of eptember 28, A.3.1.10 systems in scope and subject to aging management 013 B.1.11 review for license renewal in accordance with 10 CFR IP3:

54.4(a)(1) and (a)(3). Inspections shall include areas ecember 12, surrounding the subject systems to identify hazards to 015 those systems. Inspections of nearby systems that could impact the subject systems will include SSCs that are in scope and subject to aging management review for license renewal in accordance with 10 CFR 54.4(a)(2).

6 Enhance the Fatigue Monitoring Program for IP2 to P2: NL-07-039 A.2.1.11 Septmber28,A.3.1.11 28, monitor steady state cycles and feedwater cycles or eptember perform an evaluation to determine monitoring is not 013 B.1.12, NL-07-153 Audit 164Item required. Review the number of allowed events and resolve discrepancies between reference documents and monitoring procedures.

Enhance the Fatigue Monitoring Program for IP3 to IP3:

include all the transients identified. Assure all fatigue December 12, analysis transients are included with the lowest 2015 limiting numbers. Update the number of design transients accumulated to date.

7 Enhance the Fire Protection Program to inspect SP2: NL-07-039 A.2.1.12 external surfaces of the IP3 RCP oil collection eptember 28, A.3.1.12 systems for loss of material each refueling cycle. 013 B.1.13 Enhance the Fire Protection Program to explicitly IP3:

state that the IP2 and IP3 diesel fire pump engine December 12, sub-systems (including the fuel supply line) shall be 2015 observed while the pump is running. Acceptance criteria will be revised to verify that the diesel engine does not exhibit signs of degradation while running; such as fuel oil, lube oil, coolant, or exhaust gas leakage.

Enhance the Fire Protection Program to specify that the IP2 and IP3 diesel fire pump engine carbon steel exhaust components are inspected for evidence of corrosion and cracking at least once each operating cycle.

Enhance the Fire Protection Program for IP3 to visually inspect the cable spreading room, 480V switchgear room, and EDG room CO 2 fire suppression system for signs of degradation, such as corrosion and mechanical damage at least once every six months.

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1. COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION II I/ AUDIT ITEM IP2: NL-07-039 A.2.1.13 8 Enhance the Fire Water Program to include inspection September 28, A.3.1.13 of IP2 and IP3 hose reels for evidence of corrosion.

2013 B.1.14 Acceptance criteria will be revised to verify no unacceptable signs of degradation. NL-07-153 Audit Items P3: 105, 106 Enhance the Fire Water Program to replace all or test December 12, NL-08-014 a sample of IP2 and IP3 sprinkler heads required for 2015 10 CFR 50.48 using guidance of NFPA 25 (2002 edition), Section 5.3.1.1.1 before the end of the 50-year sprinkler head service life and at 10-year intervals thereafter during the extended period of operation to ensure that signs of degradation, such as corrosion, are detected in a timely manner.

Enhance the Fire Water Program to perform wall thickness evaluations of IP2 and IP3 fire protection piping on system components using non-intrusive techniques (e.g., volumetric testing) to identify evidence of loss of material due to corrosion. These inspections will be performed before the end of the current operating term and at intervals thereafter during the period of extended operation. Results of the initial evaluations will be used to determine the appropriate inspection interval to ensure aging effects are identified prior to loss of intended function.

Enhance the Fire Water Program to inspect the internal surface of foam based fire suppression tanks.

Acceptance criteria will be enhanced to verify no significant corrosion.

Docket Nos. 50-247 & 50-286 NL-13-052 Attachment 2 Page 6 of 20 COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION I_ I I/ AUDIT ITEM IP2: NL-07-039 A.2.1.15 9 Enhance the Flux Thimble Tube Inspection Program September 28, A.3.1.15 for IP2 and IP3 to implement comparisons to wear rates identified in WCAP-12866. Include provisions to 2013 B.1.16 compare data to the previous performances and perform evaluations regarding change to test IP3:

frequency and scope.

December 12, 2015 Enhance the Flux Thimble Tube Inspection Program for IP2 and IP3 to specify the acceptance criteria as outlined in WCAP-1 2866 or other plant-specific values based on evaluation of previous test results.

Enhance the Flux Thimble Tube Inspection Program for IP2 and IP3 to direct evaluation and performance of corrective actions based on tubes that exceed or are projected to exceed the acceptance criteria. Also stipulate that flux thimble tubes that cannot be inspected over the tube length and cannot be shown by analysis to be satisfactory for continued service, must be removed from service to ensure the integrity of the reactor coolant system pressure boundary.

Docket Nos. 50-247 & 50-286 NL-13-052 Attachment 2 Page 7 of 20 COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION I_ I / AUDIT ITEM P2: NL-07-039 A.2.1.16 10 Enhance the Heat Exchanger Monitoring Program for A.3.1.16 September 28, IP2 and IP3 to include the following heat exchangers B.1.17, 2013 in the scope of the program.

NL-07-153 Audit Item

" Safety injection pump lube oil heat exchangers ,P3: 52 December 12,

• RHR heat exchangers 2015

• RHR pump seal coolers

• Non-regenerative heat exchangers

" Charging pump seal water heat exchangers

• Charging pump fluid drive coolers

• Charging pump crankcase oil coolers

• Spent fuel pit heat exchangers

• Secondary system steam generator sample coolers

" Waste gas compressor heat exchangers

• SBO/Appendix R diesel jacket water heat exchanger (IP2 only)

Enhance the Heat Exchanger Monitoring Program for IP2 and IP3 to perform visual inspection on heat exchangers where non-destructive examination, such as eddy current inspection, is not possible due to heat exchanger design limitations.

Enhance the Heat Exchanger Monitoring Program for IP2 and IP3 to include consideration of material-environment combinations when determining sample population of heat exchangers.

Enhance the Heat Exchanger Monitoring Program for IP2 and IP3 to establish minimum tube wall thickness for the new heat exchangers identified in the scope of the program. Establish acceptance criteria for heat exchangers visually inspected to include no indication of tube erosion, vibration wear, corrosion, pitting, NL-09-018 fouling, or scaling.

11 Deleted NL-09-056 NL-1 1-101

Docket Nos. 50-247 & 50-286 NL-13-052 Attachment 2 Page 8 of 20

1. COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION

_ _ /IAUDIT ITEM 12 Enhance the Masonry Wall Program for IP2 and IP3 P2: NL-07-039 A.2.1.18 to specify that the IP1 intake structure is included in September 28, A.3.1.18 2013 B. 1.19 the program.

IP3:'

December 12, 2015 13 Enhance the Metal-Enclosed Bus Inspection Program eP2: NL-07-039 A.2.1.19 for IP2 and IP3 to visually inspect the external surface eptember 28, A.3.1.19 of material at 013 B.1.20 of MEB enclosure assemblies for loss least once every 10 years. The first inspection will NL-07-153 Audit Items IP3: 124, occur prior to the period of extended operation and the acceptance criterion will be no significant loss of December 12, NL-08-057 133, 519 material. 2015 NL-1 3-077 Enhance the Metal-Enclosed Bus Inspection Program to add acceptance criteria for MEB internal visual inspections to include the absence of indications of dust accumulation on the bus bar, on the insulators, and in the duct, in addition to the absence of indications of moisture intrusion into the duct.

Enhance the Metal-Enclosed Bus Inspection Program for IP2 and IP3 to inspect bolted connections at least once every five years if performed visually or at least once every ten years using quantitative measurements such as thermography or contact resistance measurements. The first inspection will occur prior to the period of extended operation.

The plant will process a change to applicable site procedure to remove the reference to "re-torquing" connections for phase bus maintenance and bolted connection maintenance.

14 Implement the Non-EQ Bolted Cable Connections P2: NL-07-039 A.2.1.21 Program for IP2 and IP3 as described in LRA Section September 28, A.3.1.21 B.1.22. 2013 B.1.22 IP3:

December 12,

_015

Docket Nos. 50-247 & 50-286 NL-13-052 Attachment 2 Page 9 of 20 COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION I AUDIT ITEM P2: NL-07-039 A.2.1.22 15 Implement the Non-EQ Inaccessible Medium-Voltage eptember 28, A.3.1.22 Cable Program for IP2 and IP3 as described in LRA 2013 B.1.23 Section B.1.23.

NL-07-153 Audit item This new program will be implemented consistent with :P3: 173 the corresponding program described in NUREG- December 12, NL-11-032 1801 Section XI.E3, Inaccessible Medium-Voltage 2015 Cables Not Subject To 10 CFR 50.49 Environmental NL-11-096 Qualification Requirements.

NL-1 1-101 P2: NL-07-039 A.2.1.23 16 Implement the Non-EQ Instrumentation Circuits Test September 28, A.3.1.23 Review Program for IP2 and IP3 as described in LRA 2013 B.1.24 Section B.1.24.

NL-07-153 Audit item This new program will be implemented consistent with IP3: 173 the corresponding program described in NUREG- December 12, 1801 Section XI.E2, Electrical Cables and 2015 Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Used in Instrumentation Circuits.

Implement the Non-EQ Insulated Cables and IP2: NL-07-039 A.2.1.24 17 IP3 as described in September 28, A.3.1.24 Connections Program for IP2 and 2013 B.1.25 LRA Section B.1.25.

NL-07-153 Audit item This new program will be implemented consistent with IP3: 173 the corresponding program described in NUREG- December 12, 1801 Section XI.E1, Electrical Cables and 2015 Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements.

Docket Nos. 50-247 & 50-286 NL-13-052 Attachment 2 Page 10 of 20 COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION I AUDIT ITEM eP2: NL-07-039 A.2.1.25 18 Enhance the Oil Analysis Program for IP2 to sample eptember 28, A.3.1.25 and analyze lubricating oil used in the SBO/Appendix 013 NL-11-101 B.1.26 R diesel generator consistent with the oil analysis for other site diesel generators. P3:

Enhance the Oil Analysis Program for IP2 and IP3 to December 12, sample and analyze generator seal oil and turbine 2015 hydraulic control oil.

Enhance the Oil Analysis Program for IP2 and IP3 to formalize preliminary oil screening for water and particulates and laboratory analyses including defined acceptance criteria for all components included in the scope of this program. The program will specify corrective actions in the event acceptance criteria are not met.

Enhance the Oil Analysis Program for IP2 and IP3 to formalize trending of preliminary oil screening results as well as data provided from independent laboratories.

IP2: NL-07-039 A.2.1.26 19 Implement the One-Time Inspection Program for IP2 and IP3 as described in LRA Section B.1.27. September 28, A.3.1.26 2013 B.1.27 This new program will be implemented consistent with NL-07-153 Audit item the corresponding program described in NUREG- IP3: 173 1801,Section XI.M32, One-Time Inspection. December 12, 2015 Implement the One-Time Inspection - Small Bore P2: NL-07-039 A.2.1.27 20 Piping Program for IP2 and IP3 as described in LRA September 28, A.3.1.27 2013 B.1.28 Section B.1.28.

NL-07-153 Audit item This new program will be implemented consistent with IP3: 173 the corresponding program described in NUREG- December 12, 1801,Section XI.M35, One-Time Inspection of ASME 2015 Code Class I Small-Bore Piping.

Enhance the Periodic Surveillance and Preventive P2: NL-07-039 A.2.1.28 21 Maintenance Program for IP2 and IP3 as necessary eptember 28, A.3.1.28 to assure that the effects of aging will be managed 013 8.1.29 such that applicable components will continue to perform their intended functions consistent with the 1P32 current licensing basis through the period of extended e015 operation. 015

Docket Nos. 50-247 & 50-286 NL-13-052 Attachment 2 Page 11 of 20

1. COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION I AUDIT ITEM 1P2: NL-07-039 A.2.1.31 22 Enhance the Reactor Vessel Surveillance Program for September 28, A.3.1.31 IP2 and IP3 revising the specimen capsule withdrawal 013 B.1.32 schedules to draw and test a standby capsule to cover the peak reactor vessel fluence expected IP3:

through the end of the period of extended operation. December 12, Enhance the Reactor Vessel Surveillance Program for 2015 IP2 and IP3 to require that tested and untested specimens from all capsules pulled from the reactor I vessel are maintained in storage.

Implement the Selective Leaching Program for IP2 1P2: NL-07-039 A.2.1.32 23 and IP3 as described in LRA Section B.1.33. September 28, A.3.1.32 2013 B.1.33 This new program will be implemented consistent with NL-07-153 Audit item the corresponding program described in NUREG- IP3: 173 1801,Section XI.M33 Selective Leaching of Materials. December 12, 2015 Enhance the Steam Generator Integrity Program for P2: NL-07-039 A.2.1.34 24 IP2 and IP3 to require that the results of the condition September 28, A.3.1.34 2013 B.1.35 013 monitoring assessment are compared to the operational assessment performed for the prior IP3:

operating cycle with differences evaluated. ecember 12, 2015 25 Enhance the Structures Monitoring Program to IP2: NL-07-039 A.2.1.35 explicitly specify that the following structures are September 28, A.3.1.35 included in the program. 2013 B.1.36

• Appendix R diesel generator foundation (IP3) NL-07-153

• Appendix R diesel generator fuel oil tank vault IP3: Audit items (IP3) December 12, 86, 87, 88,

• Appendix R diesel generator switchgear and 2015 NL-08-057 417 enclosure (IP3)

" city water storage tank foundation

• condensate storage tanks foundation (IP3) NL-13-077

• containment access facility and annex (IP3)

• discharge canal (IP2/3)

• emergency lighting poles and foundations (IP2/3)

• fire pumphouse (IP2)

• fire protection pumphouse (IP3)

• fire water storage tank foundations (IP2/3)

• gas turbine 1 fuel storage tank foundation

• maintenance and outage building-elevated passageway (IP2)

• new station security building (IP2) __

Docket Nos. 50-247 & 50-286 NL-13-052 Attachment 2 Page 12 of 20 COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION I IJ/ AUDIT ITEM 0 nuclear service building (IP1) 0 primary water storage tank foundation (IP3) 0 refueling water storage tank foundation (IP3) 0 security access and office building (IP3) 0 service water pipe chase (IP2/3) 0 service water valve pit (IP3) 0 superheater stack 0 transformer/switchyard support structures (IP2) 0 waste holdup tank pits (IP2/3)

Enhance the Structures Monitoring Program for IP2 and IP3 to clarify that in addition to structural steel and concrete, the following commodities (including their anchorages) are inspected for each structure as applicable.

• cable trays and supports

• concrete portion of reactor vessel supports

• conduits and supports

" cranes, rails and girders

• equipment pads and foundations

• fire proofing (pyrocrete)

• HVAC duct supports

• jib cranes

• manholes and duct banks

• manways, hatches and hatch covers

• monorails

" new fuel storage racks

• sumps NL-13-077 Enhance the Structures Monitoring Program for IP2 and IP3 to inspect inaccessible concrete areas that are exposed by excavation for any reason. IP2 and IP3 will also inspect inaccessible concrete areas in environments where observed conditions in accessible areas exposed to the same environment indicate that significant concrete degradation is occurring.

Enhance the Structures Monitoring Program for IP2 and IP3 to perform inspections of elastomers (seals, gaskets, seismic joint filler, and roof elastomers) to identify cracking and change in material properties and for inspection of aluminum vents and louvers to identify loss of material.

Docket Nos. 50-247 & 50-286 NL-13-052 Attachment 2 Page 13 of 20

1. COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION

/ AUDIT ITEM Enhance the Structures Monitoring Program for IP2 NL-08-127 Audit Item and IP3 to perform an engineering evaluation of 360 groundwater samples to assess aggressiveness of groundwater to concrete on a periodic basis (at least once every five years). IPEC will obtain samples from at least 5 wells that are representative of the ground water surrounding below-grade site structures and perform an engineering evaluation of the results from those samples for sulfates, pH and chlorides.

Additionally, to assess potential indications of spent fuel pool leakage, IPEC will sample for tritium in groundwater wells in close proximity to the IP2 spent fuel pool at least once every 3 months.

Enhance the Structures Monitoring Program for IP2 and IP3 to perform inspection of normally submerged concrete portions of the intake structures at least once every 5 years. Inspect the baffling/grating partition and support platform of the IP3 intake structure at least once every 5 years.

Enhance the Structures Monitoring Program for IP2 Audit Item and IP3 to perform inspection of the degraded areas 358 of the water control structure once per 3 years rather than the normal frequency of once per 5 years during the PEO.

Enhance the Structures Monitoring Program to include more detailed quantitative acceptance criteria NL-1 1-032 for inspections of concrete structures in accordance with ACI 349.3R, "Evaluation of Existing Nuclear Safety-Related Concrete Structures" prior to the period of extended operation. NL-1 1-101 26 Implement the Thermal Aging Embrittlement of Cast P2: NL-07-039 A.2.1.36 Austenitic Stainless Steel (CASS) Program for IP2 3eptember 28, A.3.1.36 and IP3 as described in LRA Section B.1.37. 2013 B.1.37 NL-07-153 Audit item This new program will be implemented consistent with IP3: 173 the corresponding program described in NUREG- December 12, 1801,Section XI.M12, Thermal Aging Embrittlement 2015 of Cast Austenitic Stainless Steel (CASS) Program.

Docket Nos. 50-247 & 50-286 NL-13-052 Attachment 2 Page 14 of 20

1. COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION I AUDIT ITEM Implement the Thermal Aging and Neutron Irradiation P2: NL-07-039 A.2.1.37 27 September 28, A.3.1.37 Embrittlement of Cast Austenitic Stainless Steel 2013 8.1.38 IP2 and IP3 as described in LRA 013 N. Ad ie (CASS) Program for SecionB.138.NL-07-153 Audit7 item Section B.1.38. I3 IP3: 173 This new program will be implemented consistent with December 12, the corresponding program described in NUREG- 2015 1801 Section XI.M13, Thermal Aging and Neutron Embrittlement of Cast Austenitic Stainless Steel (CASS) Program.

P2: NL-07-039 A.2.1.39 28 Enhance the Water Chemistry Control - Closed 28, A.3.1.39 Cooling Water Program to maintain water chemistry of September 2013 8. 1.40 NL-08-057 Audit item the IP2 SBO/Appendix R diesel generator cooling IP3: 509 system per EPRI guidelines.

Enhance the Water Chemistry Control - Closed December 12, Cooling Water Program to maintain the IP2 and IP3 2015 security generator and fire protection diesel cooling water pH and glycol within limits specified by EPRI guidelines.

29 Enhance the Water Chemistry Control - Primary and 1P2: NL-07-039 A.2.1.40 Secondary Program for IP2 to test sulfates Seconary monthly in Pogram013 September 28, B.1.41 the RWST with a limit of <150 ppb. 2013 30 For aging management of the reactor vessel internals, P2: NL-07-039 A.2.1.41 IPEC will (1) participate in the industry programs for September 28, A.3.1.41 011 investigating and managing aging effects on reactor internals; (2) evaluate and implement the results of IP3:

the industry programs as applicable to the reactor December 12, internals; and (3) upon completion of these programs, 2013 but not less than 24 months before entering the period of extended operation, submit an inspection plan for Complete NL-1 1-107 reactor internals to the NRC for review and approval.

31 Additional P-T curves will be submitted as required P2: NL-07-039 A.2.2.1.2 per 10 CFR 50, Appendix G prior to the period of September 28, A.3.2.1.2 2013 4.2.3 013 extended operation as part of the Reactor Vessel Surveillance Program. IP3:

December 12, 2015 32 As required by 10 CFR 50.61(b)(4), IP3 will submit a IP3: NL-07-039 A.3.2.1.4 plant-specific safety analysis for plate B2803-3 to the December 12, 4.2.5 NRC three years prior to reaching the RTPTS 2015 NL-08-127 screening criterion. Alternatively, the site may choose to implement the revised PTS rule when approved.

Docket Nos. 50-247 & 50-286 NL-13-052 Attachment 2 Page 15 of 20

1. COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION I I /IAUDIT ITEM IP2: NL-07-039 A.2.2.2.3 33 At least 2 years prior to entering the period of A.3.2.2.3 September 28, extended operation, for the locations identified in LRA 4.3.3 2011 Table 4.3-13 (IP2) and LRA Table 4.3-14 (IP3), under NL-07-153 Audit item the Fatigue Monitoring Program, IP2 and IP3 will 146 IP3:

implement one or more of the following: December 12, NL-08-021 (1) Consistent with the Fatigue Monitoring Program, 2013 Detection of Aging Effects, update the fatigue usage calculations using refined fatigue analyses to determine valid CUFs less than 1.0 when accounting Complete NL-10-082 for the effects of reactor water environment. This includes applying the appropriate Fen factors to valid CUFs determined in accordance with one of the following:

1. For locations in LRA Table 4.3-13 (IP2) and LRA Table 4.3-14 (1P3), with existing fatigue analysis valid for the period of extended operation, use the existing CUF.

2. Additional plant-specific locations with a valid CUF may be evaluated. In particular, the pressurizer lower shell will be reviewed to ensure the surge nozzle remains the limiting component.

3. Representative CUF values from other plants, adjusted to or enveloping the IPEC plant specific external loads may be used ifdemonstrated applicable to IPEC.

4. An analysis using an NRC-approved version of the ASME code or NRC-approved alternative (e.g., NRC-approved code case) may be performed to determine a valid CUF.

(2) Consistent with the Fatigue Monitoring Program, Corrective Actions, repair or replace the affected locations before exceeding a CUF of 1.0.

34 IP2 SBO / Appendix R diesel generator will be April 30, 2008 NL-07-078 2.1.1.3.5 installed and operational by April 30, 2008. This Complete NL-08-074 committed change to the facility meets the requirements of 10 CFR 50.59(c)(1) and, therefore, a NL-11-101 license amendment pursuant to 10 CFR 50.90 is not required.

Docket Nos. 50-247 & 50-286 NL-13-052 Attachment 2 Page 16 of 20

1. COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION

/ AUDIT ITEM 35 Perform a one-time inspection of representative IP2: NL-08-127 Audit Item sample area of IP2 containment liner affected by the September 28, 27 1973 event behind the insulation, prior to entering the 013 period of extended operation, to assure liner degradation is not occurring in this area. NL-11-101 Perform a one-time inspection of representative IP3:

sample area of the IP3 containment steel liner at the December 12, juncture with the concrete floor slab, prior to entering 2015 the period of extended operation, to assure liner degradation is not occurring in this area.

Any degradation will be evaluated for updating of the NL-09-018 containment liner analyses as needed.

IP2: NL-08-127 Audit 359Item 36 Perform a one-time inspection and evaluation of a ept r NL-08-101 sample of potentially affected IP2 refueling cavity 2810 concrete prior to the period of extended operation.

The sample will be obtained by core boring the refueling cavity wall in an area that is susceptible to exposure to borated water leakage. The inspection will include an assessment of embedded reinforcing steel.

Additional core bore samples will be taken, if the NL-09-056 leakage is not stopped, prior to the end of the first ten years of the period of extended operation.

A sample of leakage fluid will be analyzed to NL-09-079 determine the composition of the fluid. If additional core samples are taken prior to the end of the first ten years of the period of extended operation, a sample of leakage fluid will be analyzed.

IP2: NL-08-127 Audit Item 37 Enhance the Containment Inservice Inspection (CII- P IWL) Program to include inspections of the September 28, 361 containment using enhanced characterization of 013 degradation (i.e., quantifying the dimensions of noted indications through the use of optical aids) during the e 12 period of extended operation. The enhancement 015 includes obtaining critical dimensional data of degradation where possible through direct measurement or the use of scaling technologies for photographs, and the use of consistent vantage points for visual inspections.

Docket Nos. 50-247 & 50-286 NL-13-052 Attachment 2 Page 17 of 20

1. COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION I AUDIT ITEM 1P2: NL-08-143 4.2.1 38 For Reactor Vessel Fluence, should future core loading patterns invalidate the basis for the projected September 28, values of RTpts or CvUSE, updated calculations will 013 be provided to the NRC. IP3:

December 12, 2015 39 Deleted NL-09-079 40 Evaluate plant specific and appropriate industry P2: NL-09-106 8.1.6 operating experience and incorporate lessons learned September 28, B.1.22 in establishing appropriate monitoring and inspection 013 B.1.23 frequencies to assess aging effects for the new aging B.1.24 management programs. Documentation of the e 12 8.1.27 operating experience evaluated for each new program December 12, B.1.27 will be available on site for NRC review prior to the B.1.33 period of extended operation. B.1.37 B.1.38 IP2: NL-11-032 N/A 41 IPEC will inspect steam generators for both units to fter the assess the condition of the divider plate assembly.

The examination technique used will be capable of beginning of the detecting PWSCC in the steam generator divider plate PEO and prior to assembly. The IP2 steam generator divider plate eptember 28, 2023 NL-1 1-074 inspections will be completed within the first ten years of the period of extended operation (PEO). The IP3 IP3: NL-1 1-090 steam generator divider plate inspections will be Prior to the end completed within the first refueling outage following of the first NL-1 1-101 the beginning of the PEO. refueling outage following the beginning of the

_PEQ. I I

Docket Nos. 50-247 & 50-286 NL-13-052 Attachment 2 Page 18 of 20 COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION I I I /IAUDIT ITEM NL-1 1-032 N/A 42 IPEC will develop a plan for each unit to address the potential for cracking of the primary to secondary pressure boundary due to PWSCC of tube-to-tubesheet welds using one of the following two options.

Option 1 (Analysis)

IPEC will perform an analytical evaluation of the steam generator tube-to-tubesheet welds in order to IP2: NL-1 1-074 establish a technical basis for either determining that Prior to March the tubesheet cladding and welds are not susceptible 2024 NL-1 1-090 to PWSCC, or redefining the pressure boundary in IP3: Prior to the which the tube-to-tubesheet weld is no longer end of the first NL-1 1-096 included and, therefore, is not required for reactor refueling outage coolant pressure boundary function. The redefinition following the of the reactor coolant pressure boundary must be beginning of the approved by the NRC as a license amendment PEO.

request.

Option 2 (Inspection) IP2:

Between March IPEC will perform a one-time inspection of a 2020 and March representative number of tube-to-tubesheet welds in 2024 each steam generator to determine if PWSCC cracking is present. If weld cracking is identified: IP3: Prior to the

a. The condition will be resolved through repair end of the first or engineering evaluation to justify continued refueling outage service, as appropriate, and following the beginning of the

b. An ongoing monitoring program will be PEO.

established to perform routine tube-to-tubesheet weld inspections for the remaining life of the steam generators.

IPEC will review design basis ASME Code Class 1 IP2: NL-1 1-032 4.3.3 43 fatigue evaluations to determine whether the Prior to NUREG/CR-6260 locations that have been evaluated September 28, for the effects of the reactor coolant environment on 2013 fatigue usage are the limiting locations for the IP2 and NL-1 1-101 IP3 configurations. If more limiting locations are IP3: Prior to identified, the most limiting location will be evaluated December 12, for the effects of the reactor coolant environment on 2015 fatigue usage.

IPEC will use the NUREG/CR-6909 methodology in the evaluation of the limiting locations consisting of nickel alloy, if any.

Docket Nos. 50-247 & 50-286 NL-13-052 Attachment 2 Page 19 of 20 COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION

/ AUDIT ITEM 44 IPEC will include written explanation and justification iP2: NL-11-032 N/A of any user intervention in future evaluations using the nor to WESTEMS "Design CUF" module. 3eptember 28, NL-1 1-101 2013 1P3: Prior to December 12, 2015 45 IPEC will not use the NB-3600 option of the P2: NL-1 1-032 N/A Drior to WESTEMS program in future design calculations until ,pri tor the issues identified during the NRC review of the eptember 28, NL-1 1-101 2013 program have been resolved.

1P3: Prior to December 12, 2015 IP2: NL-11-032 N/A 46 Include in the IP2 ISI Program that IPEC will perform nior to twenty-five volumetric weld metal inspections of Ptebr 21 socket welds during each 10-year ISI interval September 28, NL-11-074 scheduled as specified by IWB-2412 of the ASME 013 Section Xl Code during the period of extended operation.

In lieu of volumetric examinations, destructive examinations may be performed, where one destructive examination may be substituted for two volumetric examinations. I IP2: to NL-12-089 N/A 47 IPEC will perform and submit analyses that nor demonstrate that the lower support column bodies will maintain their functionality during the period of epter-, NL-13-052 extended operation considering the possible loss of arch 1 2015 fracture toughness due to thermal and irradiation lP3: Prior to embrittlement. The analyses will be consistent with December 12, the IP2/IP3 licensing basis. 015

Docket Nos. 50-247 & 50-286 NL-13-052 Attachment 2 Page 20 of 20

1. COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION I AUDIT ITEM 1P2: to NL-12-174 N/A 48 Entergy will visually inspect IPEC underground piping Prior within the scope of license renewal and subject to aging management review prior to the period of September 28, of at 013 extended operation and then on a frequency least once every two years during the period of IP3: Prior to extended operation. This inspection frequency will be ecember 12, maintained unless the piping is subsequently coated in accordance with the preventive actions specified in NUREG-1801 Section XI.M41 as modified by LR-ISG-2011-03. Visual inspections will be supplemented with surface or volumetric non-destructive testing if indications of significant loss of material are observed. Consistent with revised NUREG-1801 Section XI.M41, such adverse indications will be entered into the plant corrective action program for evaluation of extent of condition and for determination of appropriate corrective actions (e.g., increased inspection frequency, repair, replacement).

IP2:NL-13-052 A.2.2.2 49 Recalculate each of the limiting CUFs provided in 1P2: to A.2.2.2 section 4.3 of the LRA for the reactor vessel internals r to include the reactor coolant environment effects e013 (Fen) as provided in the IPEC Fatigue Monitoring Program using NUREG/CR-5704 or NUREG/CR- IP3: Prior to 6909. In accordance with the corrective actions ecember 12 specified in the Fatigue Monitoring Program, 2015 corrective actions include further CUF re-analysis, and/or repair or replacement of the affected components Prior to the CUFen reaching 1.0.