ML13277A007

Indian Point Nuclear Generating Unit Nos. 2 & 3 - Reply to Request for Additional Information Regarding the License Renewal Application

ML13277A007
Person / Time
Site:	Indian Point
Issue date:	09/27/2013
From:	Dacimo F R Entergy Nuclear Northeast, Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-13-122
Download: ML13277A007 (35)

Text

En tergyEnter-qy Nuclear Northeast Indian Point Energy Center450 Broadway, GSBP.O. Box 249Buchanan, NY 10511-0249 Tel (914) 254-2055Fred DacimoVice President Operations License RenewalNL-1 3-122September 27, 2013U.S. Nuclear Regulatory Commission Document Control Desk11545 Rockville Pike, TWFN-2 F1Rockville, MD 20852-2738

REFERENCE:

Reply to Request for Additional Information Regarding the License Renewal Application Indian Point Nuclear Generating Unit Nos. 2 & 3Docket Nos. 50-247 and 50-286License Nos. DPR-26 and DPR-641. NRC letter, "Request for Additional Information for the Review of theIndian Point Nuclear Generating Unit Nos. 2 and 3, License RenewalApplication, SET 2013-04" dated July 26, 2013.

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 RenewalApplication (LRA) for Indian Point 2 and Indian Point 3.The response to RAI 16-A includes new Commitment 50 that concerns the planned replacement of the IP2 splits pins. The response to RAI 11-B includes a revision to the implementation datefor Commitment

47. These new and revised commitments are included in the latest list ofregulatory commitments provided in Attachment

2. This list has also been updated to reflectclosure of all the IP2 commitments required to be implemented prior to the PEO and closure ofselect IP3 commitments.

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

Docket Nos. 50-247 & 50-286NL-13-122 Page 2 of 2I declare under penalty of perjury that the foregoing is true and correct.

Executed on-2--'1 , 2013.Sincerely, FRD/rw

Attachment:

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

2. License Renewal Application IPEC List of Regulatory Commitments Revision 22cc: Mr. William Dean, Regional Administrator, NRC Region IMr. Sherwin E. Turk, NRC Office of General Counsel, Special CounselMr. Dave Wrona, NRC Branch Chief, Engineering Review Branch IMs. Kimberly Green, NRC Sr. Project Manager, Division of License RenewalMr. Douglas Pickett, NRR Senior Project ManagerMs. Bridget Frymire, New York State Department of Public ServiceNRC Resident Inspector's OfficeMr. Francis J. Murray, Jr., President and CEO NYSERDA ATTACHMENT 1 TO NL-13-122 REPLY TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THELICENSE RENEWAL APPLICATION ENTERGY NUCLEAR OPERATIONS, INC.INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 & 3DOCKET NOS. 50-247 AND 50-286 NL-13-122 Attachment 1Page 1 of 11REQUEST FOR ADDITIONAL INFORMATION, SET 2013-04RELATED TO INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3LICENSE RENEWAL APPLICATION REACTOR VESSEL INTERNALS PROGRAM AND INSPECTION PLANRAI 11-BThe response to RAI 11-A, by letter dated May 7, 2013 (Ref. 1), describes the functionality analysis approach for the evaluation of the IP2 and IP3 lower support columns in supportof Applicant/Licensee Action Item 7 from MRP-227-A, "Materials Reliability Program:Pressurized Water Reactor Internals Inspection and Evaluation Guidelines."

1) The response states, in part, that based on the lack of any documented history offracture in the lower core support columns, it will be assumed that only a limitednumber of columns could actually contain flaws of significant size. Provide a moredetailed basis for the number of columns that will be assumed to contain flaws,including a description of any relevant operating experience or research supporting the assumed incidence of cracking in the columns.

The basis for the number ofcracked columns should address flaws due to any screened-in aging mechanism forthe columns, in addition to fabrication defects.2) The response states, in part, that since the effects of embrittlement are onlysignificant in the presence of pre-existing flaws (e.g. from the casting process) andtensile stresses capable of propagating these flaws, the screening analysis willidentify regions of individual columns where thermal and irradiation effects couldgive rise to embrittled materials and would also be subjected to significant tensilestresses under design and service loadings.

Define what is meant by "significant tensile stresses"

-is there a specific numerical value of stress considered to be athreshold of significance?

3) Provide a general description of the fabrication of the IP2 and IP3 lowersupport columns, including:

a. the grade of cast stainless steel used (e.g. CF-8)b. the approximate location relative to the lower core plate of the weld joiningthe upper (cast) portion of the column (the column cap) to the lower portionof the column.4) Provide a summary of the most recent American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Section Xl Inservice Inspection ofthe lower support columns at IP2 and IP3, including the dates of the inspections, coverage obtained (including a specific description of the coverage limitations onthe columns),

and the size, location and orientation of any recordable or rejectable indications.

5) MRP-227-A, Section 4.2.7, requires the plant-specific analysis forApplicant/Licensee Action Item 7 demonstrating that the lower support columnbodies (expansion components) will maintain their functionality during the period of NL-13-122 Attachment 1Page 2 of 11extended operation to be submitted along with an applicant/licensee's submittal toapply the approved version of MRP-227.

This analysis was not provided with theapplicant's submittal of the Reactor Vessel Internals (RVI) Inspection Plan for IP2and IP3. Entergy later made a commitment to submit the analyses prior to thestart of the period of extended operation (PEO) for both units.However, Entergy's May 7, 2013, letter proposed a revision to Commitment 47changing the date for the submittal of the analysis for IP2 until March 1, 2015. Adelay of this nature would jeopardize satisfactory completion of the staff's review ofthe analysis prior to the refueling outage in 2016 when the initial inspections of theMPR-227-A primary components are scheduled for IP2. The staff estimates that itwill need at least 18 months to review the analysis once it is submitted.

The staffwould expect applicants/licensees (of Westinghouse plants) to inspect the lowersupport column bodies during the initial inspections if a plant-specific analysisshowed that the expansion components could not maintain their intended functionduring the PEO, or if the staff could not review and approve the analysis prior to theinitial inspections of the primary components.

In the absence of an NRC-approved plant-specific analysis for the lower supportcolumn bodies, please explain how these components will maintain their intendedfunction during the PEO.Response to RAI 11-B1) The assumption that only a limited number of columns will contain flaws of significant size is based on the qualitative factors discussed in detail below. These factors are thelack of significant flaws in the columns at manufacturing, the lack of a credible relevantflaw enhancement mechanism during service, and the operational experience thatshows a lack of cracking and loose parts that would be expected from failed columns.The size and number of potential pre-existing flaws in the lower core support columncaps is considered to be limited because, prior to component

assembly, all of thecolumns were inspected using dye penetrant and radiography.

All columns met ASTME-71 standards.

All columns were considered defect free to this level and weredeemed to exhibit zero surface-breaking flaws. Based on this inspection, anyremaining flaws would be expected to be of small size and number. Therefore, thepotential number of flaws of sufficient size to be relevant to embrittlement-related fracture processes would be small.Flaw development due to other screened-in mechanisms occurring during service is notconsidered a viable mechanism for the production of a significant number of size-relevant flaws either by itself or from the original as-manufactured distribution of flaws.Potential mechanisms for the development of new flaws or growth of existing flaws areirradiation-assisted stress corrosion cracking (IASCC) and fatigue.

Per the following, neither mechanism is expected to be viable for significant development of new flaws orgrowth of existing flaws. Per MRP-175 [1], IASCC is a mechanism for service agingdegradation of cast austenitic stainless steel (CASS). However, under the conditions ofoperation, it is not expected that IASCC can cause sufficient additional degradation toincrease the susceptibility to embrittlement-driven fracture.

A detailed discussion of thefactors controlling IASCC of wrought stainless steel and CASS is provided in AppendixB of [1]. This discussion demonstrates that IASCC processes are only significant for NL-13-122 Attachment 1Page 3 of 11wrought stainless steel and CASS at relatively high stresses and neutron exposures.

Even at several tens of dpa, the threshold stress for the onset of IASCC is over 40 ksi,while at the lower neutron exposures expected for the column cap regions, the threshold stress for IASCC would be approximately 70 ksi or greater.

Since the nominal stressesdeveloped in the columns during normal plant operation are significantly below thesevalues, on the order of less than 20 ksi, IASCC is not expected to contribute significantly to the development of flaws. Fatigue is a potential aging mechanism that has beenevaluated for Indian Point Unit 2. The fatigue evaluation, which determined that allenvironmentally adjusted cumulative usage factors (CUFens) for the support columnsare less than 1.0, has demonstrated that the Indian Point Unit 2 lower support columnsare acceptable for fatigue through the period of extended operation.

Because of thisevaluation, we do not expect to generate or grow any structurally significant flaws as aresult of fatigue during the period of extended operation.

Operating experience also supports the view that the number of cracked columns will belimited.

Although the limited access to lower core support column cap sections hasprecluded extensive observation and inspection, no cracked columns have beenobserved to date. Furthermore, extensive column cracking would be expected toproduce loose parts, and there has been no evidence of such parts found in the reactorcoolant system. Reference

[2] summarizes a survey of operating experience ofoperating pressurized water reactor designs in the U.S. The survey included responses from similar operating plants worldwide.

The survey specifically requested reporting ofany relevant operating experience with MRP-227-A components, including failures orinspections that have not detected off-normal conditions in the components.

As a resultof the survey, there was no reported degradation or off-normal conditions noted in thelower support columns for the operating fleet. A summary of the survey, WCAP-1 7435-NP, was provided for information to the US NRC by the Pressurized Water ReactorOwners Group (PWROG.)Based on the preceding discussion, it is expected that no column failures would occurduring the period of extended operation.

As noted in the response to item 5 of this RAI,Entergy plans to provide a plant-specific functionality analysis of this component.

Aspart of this analysis, a quantitative assessment of the impact of potentially failed columnswill be performed.

2) Industry guidance (including NUREG-1801 Rev. 1 Section XI.M13) [3,4] specifies tensile stress levels to be considered as significant in performing screening evaluations.

However, in the plant-specific screening

analysis, no complete columnswill be screened out based on the stress criteria.

Therefore, a functionality analysiswill be performed as noted in the response to item 5 of this RAI.3) The grade of stainless steel used in the upper sections of the lower support columns isASTM 296 Grade CF-8. This material designation is consistent with the chemistries ofthe columns as identified in plant CMTRs. No special casting processes weredesignated; thus, it is determined that the lower core support column caps werestatically cast. After casting, surface mechanical clean up (grinding) was permitted tomeet the requirements of a 250-microinch finish. No specific surface finishing processwas designated or disallowed.

After heat treatment, the bolt holes were centerline bored and machined to allow fitting to the lower core support column forging and toallow bolting at the correct position to the lower core plate. Finally, the cast upper NL-13-122 Attachment 1Page 4 of 11section of the lower support column was welded to the wrought lower section of the coresupport column with a circumferential weld.The circumferential weld that joins the upper (cast) section of the lower support columnto the lower (wrought) section of the lower support column is approximately 18 inchesbelow the upper section to core plate interface.

4) The most recent ASME Code Section Xl inservice inspection of the core supportstructure (ASME Section Xl Category B-N-3) was performed at IP2 in May 2006. Theinspection utilized a camera attached to a remote underwater examination vehicle(submarine) and only the portion of the lower support column bodies below the domelower support plate (specifically the exterior bottom of the core barrel) were inspected.

The portion of the lower support column bodies below the dome lower support plate isthe end of the column body that extends past the lower support column nut. The portionof the lower support column bodies that was inspected was the wrought lower section.All accessible surfaces were inspected with no limitations noted; however, a specificamount of coverage was not documented on the data sheets. All inspections weresatisfactory with no recordable or rejectable indications noted.The most recent ASME Code Section Xl inservice inspection of the core supportstructure (ASME Section Xl Category B-N-3) was performed at IP3 in March 2009. Theinspection utilized a camera attached to a remote underwater examination vehicle(submarine) and only the portion of the lower support column bodies below the domelower support plate (specifically the exterior bottom of the core barrel) were inspected.

The portion of the lower support column bodies below the dome lower support plate isthe end of the column body that extends past the lower support column nut. The portionof the lower support column bodies that was inspected was the wrought lower section.All accessible surfaces were inspected;

however, a specific amount of coverage was notdocumented on the data sheets. The lower internals exterior (core barrel) bottom sectionand sides (approximately 350 degrees clockwise thru 100 degrees) were restricted fromexamination due to the core barrel location relative to the refueling cavity wall and thestand for the internals.

All inspections were satisfactory with no recordable or rejectable indications noted.5) In order to provide the NRC staff with the requested 18 month review time, Entergy isrevising Commitment 47 as follows.Commitment 47 -revision to implementation dateThe implementation date for commitment 47 for IP2 is being revised from March 1, 2015to August 15, 2014.

NL-13-122 Attachment 1Page 5 of 11RAI 15-BThe revised response to RAI 15, provided in Reference 1, states that the term "Class 1"was inadvertently included in the response to RAI 12, and that the phrase "ASME CodeClass 1 fatigue evaluations for reactor vessel internals" is changed to read "ASME CodeSubsection NG fatigue evaluation for reactor vessel internals."

However, the markups toLicense Renewal Application (LRA) Sections A.2.2.2.1 and A.3.2.2.1 containing theproposed content for the Updated Final Safety Analysis Report (UFSAR) supplement related to metal fatigue list the reactor vessel internals fatigue time-limited aging analysisunder "Class 1 Metal Fatigue."

The staff requests that Entergy correct this apparentinconsistency in LRA Sections A.2.2.2.1 and A.3.2.2.1.

The staff also requests that Entergyadd the commitment to complete the revised fatigue cumulative usage factor analysesaccounting for environmental effects (Commitment 49 from the May 7, 2013 letter) to LRASections A.2.2.2.1 and A.3.2.2.1.

Response to RAI 15-BNew sections A.2.2.2.3 and A.3.2.2.3 have been created for the discussion of the reactor vesselinternals.

"Reactor vessel internals" has been deleted from the list of Class 1 components inSections A.2.2.2.1 and A.3.2.2.1.

The commitment to complete the revised fatigue cumulative usage factor analyses accounting forenvironmental effects (Commitment

49) is discussed in new Sections A.2.2.2.3 and A.3.2.2.3.

The commitment to complete the revised fatigue cumulative usage factor analyses accounting forenvironmental effects for reactor vessel internals also affects LRA Section B.12, FatigueMonitoring.

This section is revised to include Subsection NG for reactor vessel internals.

LRA Appendix A Sections A.2.2.2 and A.3.2.2 are revised as shown below. New sections areadded for reactor vessel internals.

Changes are shown as strikethroughs for deletlens andunderlines for additions.

Section A.2.2.2.1, first paragraph, is revised as follows:A.2.2.2.1 Class 1 Metal FatigueClass 1 components evaluated for fatigue and flaw growth include the reactor pressurevessel (RPV), rcactO, vessel intoena,,

, pressurizer, steam generators, reactor coolantpumps, control rod drive mechanisms, regenerative letdown heat exchanger, and Class-1 piping and in-line components.

New Section A.2.2.2.3 is added; existing section A.2.2.2.3 is renumbered to A.2.2.2.4:

A.2.2.2.3 Subsection NG Fatigue Analysis of Reactor Pressure Vessel Internals The reactor vessel internals were designed to meet the intent of Subsection NG of theASME Boiler and Pressure Vessel Code, Section II1. Subsequent plant uprate evaluations determined CUFs for some reactor vessel internals components.

These evaluations wereperformed to the intent of Subsection NG. The Fatigue Monitoring Program manages theeffects of aging related to these TLAAs (fatigue analyses) in accordance with 10 CFR54.21 (c)(1 )(iii).

NL-13-122 Attachment 1Page 6 of 11Each of the limiting CUFs for the reactor vessel internals will be recalculated prior toSeptember 28, 2013, to include the reactor coolant environment effects (FE_) as providedin the Fatigue Monitoring Program using NUREG/CR-5704 or NUREG/CR-6909.

Corrective actions specified in the Fatique Monitoring Program include further CUF re-analysis and/or repair or replacement of the affected components prior to the CUFFenreachinq 1.0.A.2.2.2.34 Environmental Effects on FatigueSection A.3.2.2.1, first paragraph, is revised as follows:A.3.2.2.1 Class 1 Metal FatigueClass 1 components evaluated for fatigue and flaw growth include the reactor pressurevessel (RPV), r.actOr vessel int.,nal.

, pressurizer, steam generators, reactor coolantpumps, control rod drive mechanisms, regenerative letdown heat exchanger, and Class-1piping and in-line components.

New Section A.3.2.2.3 is added; existing section A.3.2.2.3 is renumbered to A.3.2.2.4:

A.3.2.2.3 Subsection NG Fatigue Analysis of Reactor Pressure Vessel Internals The reactor vessel internals were designed to meet the intent of Subsection NG of theASME Boiler and Pressure Vessel Code, Section II1. Subsequent plant uprate evaluations determined CUFs for some reactor vessel internals components.

These evaluations wereperformed to the intent of Subsection NG. The Fatigue Monitoring Program manages theeffects of aging related to these TLAAs (fatigue analyses) in accordance with 10 CFR54.21 (c)(1)(iii).

Each of the limiting CUFs for the reactor vessel internals will be recalculated prior toDecember 12, 2015, to include the reactor coolant environment effects (Fen) as providedin the Fatigue Monitoring Program using NUREG/CR-5704 or NUREG/CR-6909.

Corrective actions specified in the Fatique Monitoring Program include further CUF re-analysis and/or repair or replacement of the affected components prior to the CUFenreachingq 1.0.A.3.2.2.34 Environmental Effects on FatigueSection B.1.12, Program Description, third paragraph, is revised as follows:The analysis methods for determination of stresses and fatigue usage will be inaccordance with an NRC endorsed Edition of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section III Rules for Construction ofNuclear Power Plant Components Division 1 Subsection NB, Class 1 Components, Subarticles NB-3200 or NB-3600 and Subsection NG, Requirements for Class CSComponents, Core Support and Internal Structures as applicable to the specificcomponent.

IPEC will utilize design transients from IPEC Design Specifications to boundall operational transients.

The numbers of cycles used for evaluation will be based on thedesign number of cycles and actual IPEC cycle counts projected out to the end of thelicense renewal period (60 years).

NL-13-122 Attachment 1Page 7 of 11RAI 16-AThe response to RAI 16, by letter dated November 20, 2012 (Ref. 2), addressed theremaining life prediction for the IP2 split pins and provided the estimated replacement schedule for the split pins. Also in the response to RAI 16, Entergy stated that if the [splitpin] replacement is not implemented as currently scheduled in 2016, it will provide the NRCstaff with a detailed inspection plan, including inspection

methods, inspection

coverage, andinspection frequency, by March 2015. The staff requests that Entergy add a commitment toprovide the NRC staff with a detailed inspection plan for the IP2 split pins, including inspection

methods, inspection

coverage, and inspection frequency, by March 2015, if theplanned replacement of the IP2 split pins is not to be implemented in 2016. LRA SectionsA.2.1.41 and A.3.1.41 containing the proposed UFSAR supplement content for the IP2 andIP3 Reactor Vessel Internals Aging Management Activities should be revised to include thenew commitment.

Response to RAI 16-AEntergy provides the following commitment for providing a detailed inspection plan for the IP2split pins if the planned replacement of the IP2 split pins is not to be implemented in 2016.Commitment 50If the planned replacement of the IP2 split pins will not be accomplished in 2016, providethe NRC staff a detailed inspection plan for the IP2 split pins, including inspection

methods, inspection

coverage, and inspection frequency, by March 31, 2015.Because the new commitment only affects IP2, LRA Section A.3.1.41 does not require revision.

Changes to the RVI Program description also affect LRA Section B.1.42.The following paragraph is added to LRA Section A.2.1.41 as the fourth paragraph (additions areunderlined):

The IP2 guide tube support pins (split pins) are scheduled to be replaced during the2016 refueling outage. If the planned replacement of the IP2 split pins will not beaccomplished in 2016, Enterqy will provide the NRC staff a detailed inspection plan,including inspection

methods, inspection

coverage, and inspection freguency, no laterthan March 31, 2015.The following paragraph is added to LRA Section B.1.42, Reactor Vessel Internals

Program, under "Evaluation/l.

Scope of Program,"

new last paragraph:

The IP2 guide tube support pins (split pins) are plant-specific components as discussed in MRP-227-A, Section 4.4.3, "Westinghouse Components."

The split pins arescheduled to be replaced during the 2016 refueling outage. See letter NL-1 2-166,Entergy to NRC, response to RAI 16, dated November 20, 2012, for further discussion.

If the planned replacement of the IP2 split pins will not be accomplished in 2016, Entergywill provide the NRC staff a detailed inspection plan, including inspection methods,inspection

coverage, and inspection frequency, no later than March 31, 2015.

NL-13-122 Attachment 1Page 8 of 11RAI 17Appendix A to MRP-227-A indicates that failures of Alloy X-750 clevis insert bolts werereported by one Westinghouse-designed plant in 2010. A recent metallurgical analysis ofbolts removed from this plant confirmed that the bolts cracked due to primary water stresscorrosion cracking (PWSCC).

Appendix A to MRP-227-A indicates that most of the failuresof Alloy X-750 material have occurred in material with heat treatment condition AH1, whileAlloy X-750 given the high temperature heat treatment (HTH) has proved more resistant toPWSCC.The only aging mechanism requiring management by MRP-227-A for the clevis insert bolts iswear. The clevis insert bolts are categorized as an "Existing Programs" component underMRP-227-A, with the ASME Code, Section Xl Inservice Inspection program credited formanaging aging due to wear only. The ASME Code, Section Xl specifies a VT-3 visualinspection for the clevis insert bolts which may not be adequate to detect cracking before itresults in bolt failure.The staff requests that Entergy modify the MRP-227-A inspection requirement for theclevis insert bolts as necessary to manage the effects of PWSCC for the IP2 and IP3bolts. If the inspection requirement is not modified, the staff requests that Entergyprovide a technical justification for the adequacy of the existing inspection requirement tomanage PWSCC.Response to RAI 17Entergy provides the following technical justification for the adequacy of the existinginspection requirement to manage the effects of PWSCC.The main function of the lower radial support system (LRSS) is to prevent tangential orrotational motion of the lower internals assembly while permitting axial displacement anddifferential radial expansion.

Indian Point Units 2 and 3 have six radial supports spaced at 60degree intervals around the circumference of the vessel (see Reference

[5], Figure 1). Becauseof the small tangential clearance between the radial keys and the clevis insert, the keys arepotentially subjected to flow-induced vibration loads and wear at the key-to-keyway (clevis)interface.

These supports are designed to prevent excessive tangential displacement of thelower internals during seismic and loss of coolant accident (LOCA) conditions.

The supportsalso limit displacements and misalignments in order to avoid overstressing the core barrel andto ensure that the control rods can be freely inserted.

Therefore, providing the clevis insertsremain in place, the design function of the LRSS will be maintained during seismic and LOCAconditions.

Crack detection prior to bolt failure is not required due to inherent design redundancy.

The abilityof the LRSS to perform its intended design function under seismic and LOCA condition loadingsis unrelated to the integrity of the cap screws and pins that are used to hold the clevis insert inplace. The cap screws and the dowel pins hold the clevis inserts in place so as to minimize longterm vibration and wear of the mating parts.Should cap screws fail during operation, it could result in potential increased wear of matingsurfaces.

Any increased wear, which would occur over several operating cycles, will not impactthe function of the reactor internals components.

This is based on operating experience with NL-13-122 Attachment 1Page 9 of 11damaged bolts and one dowel pin as described in the InfoGram

[5] which showed no discernible change in the clevis insert wear surfaces after operation for two additional cycles.Complete disengagement of one of the clevis inserts is highly unlikely based on the available gaps with surrounding components (see Figure 1). Even if it were postulated that one of theclevis inserts becomes non-functional, the other lower radial supports are capable of resisting allof the internal and external asymmetric loads. Wear or some degradation of a key might occur,but the key would still be expected to maintain functionality.

Taken as a whole, the core barreland LRSS system are expected to maintain their design function with degraded clevis insertbolts. Based on the evaluations performed to date, there are no safety or operability concernswith clevis insert bolt failure.As described in the InfoGram

[5], Westinghouse performed evaluations of the potential for looseparts with failed clevis insert bolts for the plant referenced in this RAI. The loose parts evaluation concluded that the separated cap screw heads will remain captured in the clevis insertcounterbores and will not impact operation.

However, lock bars at the degraded cap screwlocations have experienced wear-related degradation; therefore, the potential for loose parts fromthe lock bars to affect other locations in the reactor vessel was also evaluated.

Westinghouse concluded that no significant degradation of mechanical components is expected as a result ofpotential loose parts from the lock bars in the primary system.The MRP-227-A categorization for wear only is based on the primary concern for clevis insertlooseness and wear of the clevis insert and radial key interfacing surfaces that could potentially lead to increased motion at the bottom end of the core barrel, rather than bolt material cracking.

The video camera visual inspections at a ten-year interval by qualified personnel that arespecified in the ASME Code Section XI and MRP-227-A are capable of identifying wear ordislodged components of the clevis insert cap screws or dowel pins at any location, if they exist.The susceptibility of Alloy X-750 to PWSCC and low-temperature crack propagation may havebeen a contributor to the observed degradation detected in 2010; however, at this time, this hasnot been confirmed by metallurgical analysis.

It was also indicated by the Staff that most failuresof Alloy X-750 material have occurred in material with heat treatment condition AH. The Alloy X-750 material used at Indian Point Units 2 and 3 for clevis insert bolts is not heat treatment condition AH.Industry operating experience, such as metallurgical test results, will continue to beevaluated for applicability as part of the operating experience program at Indian PointEnergy Center.

NL-13-122 Attachment 1Page 10 of 11Radial KeyClevisIhsertFigure 1 Lower Radial Support System Engagement NL-13-122 Attachment 1Page 11 of 11Enterqy References

1. Materials Reliability Program:

PWR Internals Material Aging Degradation Mechanism Screening and Threshold Values (MRP-175).

EPRI, Palo Alto, CA: 2005. 1012081.2. Westinghouse Report, WCAP-1 7435-NP, Rev. 1, "Results of the Reactor Internals Operating Experience Survey Conducted under PWROG Project Authorization PA-MSC-0568,"

October22, 2012.3. BWRVIP-234:

BWR Vessel and Internals

Project, Thermal Aging and Neutron Embrittlement Evaluation of Cast Austenitic Stainless Steels for BWR Internals.

EPRI, Palo Alto CA:December 2009. 1019060.4. U. S. Nuclear Regulatory Commission NUREG-1 801, "Generic Aging Lessons Learned (GALL)Report,"

July 2001.5. Westinghouse

InfoGram, IG-10-1, "Reactor Internals Lower Radial Support Clevis Insert CapScrew Degradation,"

March 31, 2010.NRC References

1. Indian Point Nuclear Generating, Units 2 & 3 -Reply to Request for Additional Information Regarding the License Renewal Application, May 7, 2013, (ADAMSAccession No. ML13142A202).

2. Indian Point, Units 2 and 3- Reply to Request for Additional Information Regarding theLicense Renewal Application, November 20, 2012, (ADAMS Accession No.ML1 2340A 154).Footnote to NRC RAI 17:1 AH = Hot rolled "equalized" at 1625 °F (885 'C) followed by20 hours at 1300 OF (704 0C)

ATTACHMENT 2 TO NL-13-122 LICENSE RENEWAL APPLICATION IPEC LIST OF REGULATORY COMMITMENTS Rev. 22ENTERGY NUCLEAR OPERATIONS, INC.INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 & 3DOCKET NOS. 50-247 AND 50-286 Docket Nos. 50-247 & 50-286NL-13-122 Attachment 2Page 1 of 20List of Regulatory Commitments Rev. 22The following table identifies those actions committed to by Entergy in this document.

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

1. COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTION/ AUDIT ITEM1 Enhance the Aboveground Steel Tanks Program for P2: NL-07-039 A.2.1.1IP2 and IP3 to perform thickness measurements of S A.3.1.1the bottom surfaces of the condensate storage tanks, 2013-Complete NL-13-122 B.1.1city water tank, and fire water tanks once during the P3:first ten years of the period of extended operation.

December 12,Enhance the Aboveground Steel Tanks Program for 2015IP2 and IP3 to require trending of thickness I measurements when material loss is detected.

IP2: NL-07-039 A.2.1.22 Enhance the Bolting Integrity Program for IP2 and IP3 2: NL-07-289 A.3.1.2to clarify that actual yield strength is used in selecting Comle.te A.1 .2materials for low susceptibility to SCC and clarify the ?013 Complete B.1.2prohibition on use of lubricants containing MoS2 for P3: NL-07-153 Audit Itemsbolting.D mb 2, 201,241,The Bolting Integrity Program manages loss of 2-1-Complete NL-13-122 2701 preload and loss of material for all external bolting.

I I _ _

Docket Nos. 50-247 & 50-286NL-13-122 Attachment 2Page 2 of 20# COMMITMENT IMPLEMENTATIONI SOURCE I RELATEDSCHEDULE LRA SECTION/ AUDIT ITEM3Implement the Buried Piping and Tanks Inspection Program for IP2 and IP3 as described in LRA SectionB.1.6.This new program will be implemented consistent withthe corresponding program described in NUREG-1801 Section XI.M34, Buried Piping and TanksInspection.

Include in the Buried Piping and Tanks Inspection Program described in LRA Section B.1.6 a riskassessment of in-scope buried piping and tanks thatincludes consideration of the impacts of buried pipingor tank leakage and of conditions affecting the risk forcorrosion.

Classify pipe segments and tanks ashaving a high, medium or low impact of leakagebased on the safety class, the hazard posed by fluidcontained in the piping and the impact of leakage onreliable plant operation.

Determine corrosion riskthrough consideration of piping or tank material, soilresistivity,

drainage, the presence of cathodicprotection and the type of coating.

Establish inspection priority and frequency for periodicinspections of the in-scope piping and tanks based onthe results of the risk assessment.

Performinspections using inspection techniques withdemonstrated effectiveness.

P2:gepteRmbeF28-,

2013-Complete IP3:December 12,2015NL-07-039 NL-13-122 NL-07-153 NL-09-106 NL-09-111 A.2.1.5A.3.1.5B.1.6Audit Item173NL-1 1-101 Docket Nos. 50-247 & 50-286NL-13-122 Attachment 2Page 3 of 20COMMITMENT IMPLEMENTATION SOURCE I RELATEDSCHEDULE LRA SECTIONI_ I / AUDIT ITEM4Enhance the Diesel Fuel Monitoring Program toinclude cleaning and inspection of the IP2 GT-1 gasturbine fuel oil storage tanks, IP2 and IP3 EDG fuel oilday tanks, IP2 SBO/Appendix R diesel generator fueloil day tank, and IP3 Appendix R fuel oil storage tankand day tank once every ten years.Enhance the Diesel Fuel Monitoring Program toinclude quarterly sampling and analysis of the IP2SBO/Appendix R diesel generator fuel oil day tank,IP2 security diesel fuel oil storage tank, IP2 securitydiesel fuel oil day tank, and IP3 Appendix R fuel oilstorage tank. Particulates, water and sedimentchecks will be performed on the samples.

Filterable solids acceptance criterion will be less than or equalto 10mg/l. Water and sediment acceptance criterion will be less than or equal to 0.05%.Enhance the Diesel Fuel Monitoring Program toinclude thickness measurement of the bottom of thefollowing tanks once every ten years. IP2: EDG fueloil 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 firepump fuel oil storage tank; IP3: EDG fuel oil daytanks, EDG fuel oil storage tanks, Appendix R fuel oilstorage tank, and diesel fire pump fuel oil storagetank.Enhance the Diesel Fuel Monitoring Program tochange the analysis for water and particulates to aquarterly frequency for the following tanks. IP2: GT-1gas turbine fuel oil storage tanks and diesel fire pumpfuel oil storage tank; IP3: Appendix R fuel oil day tankand diesel fire pump fuel oil storage tank.Enhance the Diesel Fuel Monitoring Program tospecify acceptance criteria for thickness measurements of the fuel oil storage tanks within thescope of the program.Enhance the Diesel Fuel Monitoring Program to directsamples be taken and include direction to removewater when detected.

Revise applicable procedures to direct sampling of theonsite portable fuel oil contents prior to transferring the contents to the storage tanks.Enhance the Diesel Fuel Monitoring Program to directthe addition of chemicals including biocide when thepresence of biological activity is confirmed.

P2:mbG m-28-,_0-1-3 CompleteP3:December 12,2015NL-07-039 NL-13-122 NL-07-153 NL-08-057 A.2.1.8A.3.1.8B.1.9Audit items128, 129,132,491,492,510.1. I I. I Docket Nos. 50-247 & 50-286NL-13-122 Attachment 2Page 4 of 20# COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTION/ AUDIT ITEMIP2: NL-07-039 A.2.1.105 Enhance the External Surfaces Monitoring Program S,,,.,,,

A.3.1.10for IP2 and IP3 to include periodic inspections of ?013 ...........

systems in scope and subject to aging management Complete NL-13-122 B.1.11review for license renewal in accordance with 10 CFR P354.4(a)(1) and (a)(3). Inspections shall include areas ecember 12,surrounding the subject systems to identify hazards to 015those systems.

Inspections of nearby systems thatcould impact the subject systems will include SSCsthat are in scope and subject to aging management review for license renewal in accordance with 10 CFR54.4(a)(2).

6 Enhance the Fatigue Monitoring Program for IP2 to P2: NL-07-039 A.2.1.11monitor steady state cycles and feedwater cycles or " emb. .eF"28,A2 3.1.11perform an evaluation to determine monitoring is not _04-Complete NL-13-122 B.1.12,required.

Review the number of allowed events and 164resolve discrepancies between reference documents 164and monitoring procedures.

Enhance the Fatigue Monitoring Program for IP3 to P3:include all the transients identified.

Assure all fatigue December 12,analysis transients are included with the lowest ?015limiting numbers.

Update the number of designtransients accumulated to date.7 Enhance the Fire Protection Program to inspect P2: NL-07-039 A.2.1.12external surfaces of the IP3 RCP oil collection

t e2, A.3.1.12systems for loss of material each refueling cycle. 20-1-Complete NL-13-122 B.1.13Enhance the Fire Protection Program to explicitly P3

state that the IP2 and IP3 diesel fire pump engine December 12,sub-systems (including the fuel supply line) shall be 2015observed while the pump is running.

Acceptance criteria will be revised to verify that the diesel enginedoes not exhibit signs of degradation while running;such as fuel oil, lube oil, coolant, or exhaust gasleakage.Enhance the Fire Protection Program to specify thatthe IP2 and IP3 diesel fire pump engine carbon steelexhaust components are inspected for evidence ofcorrosion and cracking at least once each operating cycle.Enhance the Fire Protection Program for IP3 tovisually inspect the cable spreading room, 480Vswitchgear room, and EDG room C02 fire suppression system for signs of degradation, such as corrosion and mechanical damage at least once every sixmonths.

Docket Nos. 50-247 & 50-286NL-13-122 Attachment 2Page 5 of 20COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTIONI_ I__I_ _I/ AUDIT ITEM8Enhance the Fire Water Program to include inspection of IP2 and IP3 hose reels for evidence of corrosion.

Acceptance criteria will be revised to verify nounacceptable signs of degradation.

Enhance the Fire Water Program to replace all or testa sample of IP2 and IP3 sprinkler heads required for10 CFR 50.48 using guidance of NFPA 25 (2002edition),

Section 5.3.1.1.1 before the end of the 50-year sprinkler head service life and at 10-yearintervals thereafter during the extended period ofoperation to ensure that signs of degradation, such ascorrosion, are detected in a timely manner.Enhance the Fire Water Program to perform wallthickness evaluations of IP2 and IP3 fire protection piping on system components using non-intrusive techniques (e.g., volumetric testing) to identifyevidence of loss of material due to corrosion.

Theseinspections will be performed before the end of thecurrent operating term and at intervals thereafter during the period of extended operation.

Results ofthe initial evaluations will be used to determine theappropriate inspection interval to ensure aging effectsare identified prior to loss of intended function.

Enhance the Fire Water Program to inspect theinternal surface of foam based fire suppression tanks.Acceptance criteria will be enhanced to verify nosianificant corrosion.

IP2:SepteRmbeF-8 20-1-3-Complete I P3:December 12,2015NL-07-039 NL-13-122 NL-07-153 NL-08-014 A.2.1.13A.3.1.13B.1.14Audit Items105,106 Docket Nos. 50-247 & 50-286NL-13-122 Attachment 2Page 6 of 20COMMITMENT IMPLEMENTATION1 SOURCE I RELATEDSCHEDULE LRA SECTIONI I I/AUDIT ITEM9Enhance the Flux Thimble Tube Inspection Programfor IP2 and IP3 to implement comparisons to wearrates identified in WCAP-12866.

Include provisions tocompare data to the previous performances andperform evaluations regarding change to testfrequency and scope.Enhance the Flux Thimble Tube Inspection Programfor IP2 and IP3 to specify the acceptance criteria asoutlined in WCAP-12866 or other plant-specific valuesbased on evaluation of previous test results.Enhance the Flux Thimble Tube Inspection Programfor IP2 and IP3 to direct evaluation and performance of corrective actions based on tubes that exceed orare projected to exceed the acceptance criteria.

Alsostipulate that flux thimble tubes that cannot beinspected over the tube length and cannot be shownby analysis to be satisfactory for continued service,must be removed from service to ensure the integrity of the reactor coolant system pressure boundary.

IP2:Septei:beF 298T IP3:December 12,2015NL-07-039 NL-13-122 A.2.1.15A.3.1.15B.1.16 Docket Nos. 50-247 & 50-286NL-13-122 Attachment 2Page 7 of 20COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTION/ AUDIT ITEM10Enhance the Heat Exchanger Monitoring Program forIP2 and IP3 to include the following heat exchangers in the scope of the program.* Safety injection pump lube oil heat exchangers

" RHR heat exchangers

• 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 samplecoolers" Waste gas compressor heat exchangers

• SBO/Appendix R diesel jacket water heatexchanger (IP2 only)Enhance the Heat Exchanger Monitoring Program forIP2 and IP3 to perform visual inspection on heatexchangers where non-destructive examination, suchas eddy current inspection, is not possible due to heatexchanger design limitations.

Enhance the Heat Exchanger Monitoring Program forIP2 and IP3 to include consideration of material-environment combinations when determining samplepopulation of heat exchangers.

Enhance the Heat Exchanger Monitoring Program forIP2 and IP3 to establish minimum tube wall thickness for the new heat exchangers identified in the scope ofthe program.

Establish acceptance criteria for heatexchangers visually inspected to include no indication of tube erosion, vibration wear, corrosion, pitting,foulina, or scalinQ.IP2:20-13Complete I P3:December 12,2015NL-07-039 NL-13-122 NL-07-153 NL-09-018 A.2.1.16A.3.1.16B.1.17,Audit Item5211 Deleted NL-09-056 NL-1 1-101 Docket Nos. 50-247 & 50-286NL-13-122 Attachment 2Page 8 of 20COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTIONI AUDIT ITEM12 Enhance the Masonry Wall Program for IP2 and IP3 -P2: NL-07-039 A.2.1.18to specify that the IP1 intake structure is included in C" m e te 28.. A.3....1.19 the program.

_9l-Complete NL-13-122 B.1.19IP3:_0_-1--Complete IP2: NL-07-039 A.2.1.1913 Enhance the Metal-Enclosed Bus Inspection Program P2:mNLr7,039 A.3.1.19for IP2 and IP3 to visually inspect the external surface m32of MEB enclosure assemblies for loss of material at NL-07-153 Audit Itemsleast once every 10 years. The first inspection will P& 124,occur prior to the period of extended operation and D r01the acceptance criterion will be no significant loss of 2015material.

NL-13-077 Enhance the Metal-Enclosed Bus Inspection Programto add acceptance criteria for MEB internal visualinspections to include the absence of indications ofdust accumulation on the bus bar, on the insulators, and in the duct, in addition to the absence ofindications of moisture intrusion into the duct.Enhance the Metal-Enclosed Bus Inspection Programfor IP2 and IP3 to inspect bolted connections at leastonce every five years if performed visually or at leastonce every ten years using quantitative measurements such as thermography or contactresistance measurements.

The first inspection willoccur prior to the period of extended operation.

The plant will process a change to applicable siteprocedure to remove the reference to "re-torquing" connections for phase bus maintenance and boltedconnection maintenance.

14 Implement the Non-EQ Bolted Cable Connections P2: NL-07-039 A.2.1.21Program for IP2 and IP3 as described in LRA Section Complete 28-, A.3.1.21B.1.22. .GI-l-Complete NL-13-122 B.1.22P3:ecember 12,__015 Docket Nos. 50-247 & 50-286NL-13-122 Attachment 2Page 9 of 20# COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTIONI AUDIT ITEMIP2: NL-07-039 A.2.1.2215 Implement the Non-EQ Inaccessible Medium-Voltage P2: N8, A.3.1.22Cable Program for IP2 and IP3 as described in LRA Complete NL-13-122 B.1.23Section B. 1.23. _;t=-o lt L1-2 .12NL-07-153 Audit itemThis new program will be implemented consistent with P3: 173the corresponding program described in NUREG- December 12, NL-1 1-0321801 Section XI.E3, Inaccessible Medium-Voltage 2015Cables Not Subject To 10 CFR 50.49 Environmental NL-1 1-096Qualification Requirements.

NL-1 1-10116 Implement the Non-EQ Instrumentation Circuits Test ,P2: NL-07-039 A.2.1.23SeptmbeF28,A.3.1

.23Review Program for IP2 and IP3 as described in LRA Coo.teN.........

Section B.1.24. 2013-Complete NL-13-122 B.1.24NL-07-153 Audit itemThis new program will be implemented consistent with P3: 173the corresponding program described in NUREG- December 12,1801 Section XI.E2, Electrical Cables and 2015Connections Not Subject to 10 CFR 50.49Environmental Qualification Requirements Used inInstrumentation Circuits.

17 Implement the Non-EQ Insulated Cables and P2: NL-07-039 A.2.1.24Connections Program for IP2 and IP3 as described in Ceot ete N2.82, A.3.1.24LRA Section B.1.25. 2-0-13-Complete NL-13-122 B.1.25NL-07-153 Audit itemThis new program will be implemented consistent with IP3: 173the corresponding program described in NUREG- December 12,1801 Section XI.E1, Electrical Cables and 2015Connections Not Subject to 10 CFR 50.49Environmental Qualification Requirements.

Docket Nos. 50-247 & 50-286NL-13-122 Attachment 2Page 10 of 20COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTION_ AUDIT ITEM18 Enhance the Oil Analysis Program for IP2 to sample P2: NL-07-039 A.2.1.25and analyze lubricating oil used in the SBO/Appendix Com e A.3.1.25R diesel generator consistent with the oil analysis for 2-43-Complete NL-13-122 B.1.26other site diesel generators.

P3:Enhance the Oil Analysis Program for IP2 and IP3 to December 12,sample and analyze generator seal oil and turbine 2015hydraulic control oil.Enhance the Oil Analysis Program for IP2 and IP3 toformalize preliminary oil screening for water andparticulates and laboratory analyses including definedacceptance criteria for all components included in thescope of this program.

The program will specifycorrective actions in the event acceptance criteria arenot met.Enhance the Oil Analysis Program for IP2 and IP3 toformalize trending of preliminary oil screening resultsas well as data provided from independent laboratories.

19 Implement the One-Time Inspection Program for IP2 P2: NL-07-039 A.2.1.26and IP3 as described in LRA Section B.1.27. Comolmbe e 2--, A.3.1.2620-l-3-Com plete NL-13-122 B. 1.27This new program will be implemented consistent with NL-07-153 Audit itemthe corresponding program described in NUREG- P3: 1731801,Section XI.M32, One-Time Inspection.

December 12,2_01520 Implement the One-Time Inspection

-Small Bore P2: NL-07-039 A.2.1.27Piping Program for IP2 and IP3 as described in LRA omplet 28-1 A.3.1.27Section B.1.28. 20l-3-Complete NL-13-122 B. 1.28NL-07-153 Audit itemThis new program will be implemented consistent with IP3: 173the corresponding program described in NUREG- December 12,1801,Section XI.M35, One-Time Inspection of ASME 2015Code Class I Small-Bore Piping.21 Enhance the Periodic Surveillance and Preventive

,P2: NL-07-039 A.2.1.28Maintenance Program for IP2 and IP3 as necessary Com le t ....... A...1.28to assure that the effects of aging will be managed _0-3-Complete NL-13-122 B.1.29such that applicable components will continue toperform their intended functions consistent with the P31current licensing basis through the period of extended e0151 operation.

015 Docket Nos. 50-247 & 50-286NL-13-122 Attachment 2Page 11 of 20# COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTION/ AUDIT ITEM22 Enhance the Reactor Vessel Surveillance Program for P2: NL-07-039 A.2.1.31 28,A.3.1

.31IP2 and IP3 revising the specimen capsule withdrawal o-,-et- 28, A...1 .32schedules to draw and test a standby capsule to 204=3-Complete NL-13-122 B.1.32cover the peak reactor vessel fluence expected P3:through the end of the period of extended operation.

December 12,Enhance the Reactor Vessel Surveillance Program for 2015IP2 and IP3 to require that tested and untestedspecimens from all capsules pulled from the reactorvessel are maintained in storage.23 Implement the Selective Leaching Program for IP2 P2: NL-07-039 A.2.1.32and IP3 as described in LRA Section B.1.33. Copteml-et3, A.3.1.3220-1-3-Com plete NL-13-122 B.1.33This new program will be implemented consistent with NL-07-153 Audit itemthe corresponding program described in NUREG- IP3: 1731801,Section XI.M33 Selective Leaching of Materials.

December 12,201524 Enhance the Steam Generator Integrity Program for P2: NL-07-039 A.2.1.34IP2 and IP3 to require that the results of the condition Complete 2832 A.3.1.34monitoring assessment are compared to the 2-l--Complete NL-13-122 B.1.35operational assessment performed for the prior IP3:operating cycle with differences evaluated.

D ,,,, 22--l6Complete 25 Enhance the Structures Monitoring Program to P2: NL-07-039 A.2.1.35explicitly specify that the following structures are , A.3.1.35included in the program.

2-43-Complete NL-13-122 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 417enclosure (1P3)* 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 (I P2)* new station security building (IP2) __

Docket Nos. 50-247 & 50-286NL-13-122 Attachment 2Page 12 of 20COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTIONI_ I / AUDIT ITEM000000000nuclear service building (IP1)primary water storage tank foundation (IP3)refueling water storage tank foundation (IP3)security access and office building (IP3)service water pipe chase (IP2/3)service water valve pit (IP3)superheater stacktransformer/switchyard support structures (IP2)waste holdup tank pits (IP2/3)Enhance the Structures Monitoring Program for IP2and IP3 to clarify that in addition to structural steeland concrete, the following commodities (including their anchorages) are inspected for each structure asapplicable.

• 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* sumpsEnhance the Structures Monitoring Program for IP2and IP3 to inspect inaccessible concrete areas thatare exposed by excavation for any reason. IP2 andIP3 will also inspect inaccessible concrete areas inenvironments where observed conditions inaccessible areas exposed to the same environment indicate that significant concrete degradation isoccurring.

Enhance the Structures Monitoring Program for IP2and IP3 to perform inspections of elastomers (seals,gaskets, seismic joint filler, and roof elastomers) toidentify cracking and change in material properties and for inspection of aluminum vents and louvers toidentify loss of material.

NL-13-077 Docket Nos. 50-247 & 50-286NL-13-122 Attachment 2Page 13 of 20# COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTION/ AUDIT ITEMEnhance the Structures Monitoring Program for IP2and IP3 to perform an engineering evaluation ofgroundwater samples to assess aggressiveness ofgroundwater to concrete on a periodic basis (at leastonce every five years). IPEC will obtain samples fromat least 5 wells that are representative of the groundwater surrounding below-grade site structures andperform an engineering evaluation of the results fromthose samples for sulfates, pH and chlorides.

Additionally, to assess potential indications of spentfuel pool leakage, IPEC will sample for tritium ingroundwater wells in close proximity to the IP2 spentfuel pool at least once every 3 months.Enhance the Structures Monitoring Program for IP2and IP3 to perform inspection of normally submerged concrete portions of the intake structures at least onceevery 5 years. Inspect the baffling/grating partition andsupport platform of the IP3 intake structure at leastonce every 5 years.Enhance the Structures Monitoring Program for IP2and IP3 to perform inspection of the degraded areasof the water control structure once per 3 years ratherthan the normal frequency of once per 5 years duringthe PEO.Enhance the Structures Monitoring Program toinclude more detailed quantitative acceptance criteriafor inspections of concrete structures in accordance with ACI 349.3R, "Evaluation of Existing NuclearSafety-Related Concrete Structures" prior to theperiod of extended operation.

NL-08-127 NL-1 1-032NL-11-101 Audit Item360Audit Item35826 Implement the Thermal Aging Embrittlement of Cast P2: NL-07-039 A.2.1.36Austenitic Stainless Steel (CASS) Program for IP2 Combee 2831 A.3.1.36and IP3 as described in LRA Section B.1.37. 2_ NL-07-153 Audit itemThis new program will be implemented consistent with IP3: 173the corresponding program described in NUREG- ecember 12,1801,Section XI.M12, Thermal Aging Embrittlement 015of Cast Austenitic Stainless Steel (CASS) Program.

Docket Nos. 50-247 & 50-286NL-13-122 Attachment 2Page 14 of 20# COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTION/ AUDIT ITEM27 Implement the Thermal Aging and Neutron Irradiation P2: NL-07-039 A.2.1.37Embrittlement of Cast Austenitic Stainless Steel "" ... .A.3.1..37 (CASS) Program for IP2 and IP3 as described in LRA 2-4-Complete NL-13-122 B.1.38Section B.1.38. NL-07-153 Audit itemIP3: 173This new program will be implemented consistent with Deccmber 126the corresponding program described in NUREG- ?016-Complete 1801 Section XI.M13, Thermal Aging and NeutronEmbrittlement of Cast Austenitic Stainless Steel(CASS) Program.28 Enhance the Water Chemistry Control -Closed P2: NL-07-039 A.2.1.39Cooling Water Program to maintain water chemistry of Com lete 2... NA....1.39 the IP2 SBO/Appendix R diesel generator cooling _0l3-Complete NL-13-122 B. 1.40system per EPRI guidelines.

P3: 509Enhance the Water Chemistry Control -Closed "e .... 12-,Cooling Water Program to maintain the IP2 and IP3 _0l6-Complete security generator and fire protection diesel coolingwater pH and glycol within limits specified by EPRIguidelines.

IP2: NL-07-039 A.2.1.4029 Enhance the Water Chemistry Control -Primary and P NSecondary Program for IP2 to test sulfates monthly in Complet N, B.11.41the RWST with a limit of <150 ppb. 20-3-Complete NL-13-122 30P2: NL-07-039 A.2.1 .4130 For aging management of the reactor vessel internals, P2: , N 7 A.3.1.41IPEC will (1) participate in the industry programs for 2 4 o" NL.......

investigating and managing aging effects on reactor Complete NL-13-122 internals; (2) evaluate and implement the results of P3:the industry programs as applicable to the reactorinternals; and (3) upon completion of these programs, but not less than 24 months before entering the periodof extended operation, submit an inspection plan forreactor internals to the NRC for review and approval.

Complete NL-11-107 31 Additional P-T curves will be submitted as required 1P2: NL-07-039 A.2.2.1.2 per 10 CFR 50, Appendix G prior to the period of S A.3.2.1.2 extended operation as part of the Reactor Vessel 2 Complete NL-13-122 4.2.3Surveillance Program.

P3:December 12,201532 As required by 10 CFR 50.61 (b)(4), IP3 will submit a P3: NL-07-039 A.3.2.1.4 plant-specific safety analysis for plate B2803-3 to the December 12, 4.2.5NRC three years prior to reaching the RTPTS 2015 NL-08-127 screening criterion.

Alternatively, the site may chooseto implement the revised PTS rule when approved.

Docket Nos. 50-247 & 50-286NL-13-122 Attachment 2Page 15 of 20COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTIONI_ I / AUDIT ITEM33At least 2 years prior to entering the period ofextended operation, for the locations identified in LRATable 4.3-13 (IP2) and LRA Table 4.3-14 (IP3), underthe Fatigue Monitoring

Program, IP2 and IP3 willimplement one or more of the following:

(1) Consistent with the Fatigue Monitoring Program,Detection of Aging Effects, update the fatigue usagecalculations using refined fatigue analyses todetermine valid CUFs less than 1.0 when accounting for the effects of reactor water environment.

Thisincludes applying the appropriate Fen factors to validCUFs determined in accordance with one of thefollowing:

1. For locations in LRA Table 4.3-13 (IP2) and LRATable 4.3-14 (IP3), with existing fatigue analysis validfor the period of extended operation, use the existingCUF.2. Additional plant-specific locations with a valid CUFmay be evaluated.

In particular, the pressurizer lowershell will be reviewed to ensure the surge nozzleremains the limiting component.

3. Representative CUF values from other plants,adjusted to or enveloping the IPEC plant specificexternal loads may be used if demonstrated applicable to IPEC.4. An analysis using an NRC-approved version of theASME code or NRC-approved alternative (e.g., NRC-approved code case) may be performed to determine avalid CUF.(2) Consistent with the Fatigue Monitoring Program,Corrective

Actions, repair or replace the affectedlocations before exceeding a CUF of 1.0.I P2:SepteRbeF 204-4--Complete I P3:DCombet 12,CompleteNL-07-039 NL-13-122 NL-07-153 NL-08-021 NL-10-082 A.2.2.2.3 A.3.2.2.3 4.3.3Audit item14634 IP2 SBO / Appendix R diesel generator will be Ai,-. 30,20 NL-13-122 2.1.1.3.5 installed and operational by April 30, 2008. This Completecommitted change to the facility meets the NL-08-074 requirements of 10 CFR 50.59(c)(1) and, therefore, alicense amendment pursuant to 10 CFR 50.90 is not NL-1 1-101required.

Docket Nos. 50-247 & 50-286NL-13-122 Attachment 2Page 16 of 20# COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTION/ AUDIT ITEMIP2: NL-08-127 Audit Item35 Perform a one-time inspection of representative 27sample area of IP2 containment liner affected by the Sef 2, 271973 event behind the insulation, prior to entering the 20-t Complete NL-13-122 period of extended operation, to assure linerdegradation is not occurring in this area. NL-1 1-101Perform a one-time inspection of representative

'P3:sample area of the IP3 containment steel liner at the December 12,juncture with the concrete floor slab, prior to entering 2015the period of extended operation, to assure linerdegradation is not occurring in this area.Any degradation will be evaluated for updating of the NL-09-018 containment liner analyses as needed.36Perform a one-time inspection and evaluation of asample of potentially affected IP2 refueling cavityconcrete prior to the period of extended operation.

The sample will be obtained by core boring therefueling cavity wall in an area that is susceptible toexposure to borated water leakage.

The inspection will include an assessment of embedded reinforcing steel.Additional core bore samples will be taken, if theleakage is not stopped, prior to the end of the first tenyears of the period of extended operation.

A sample of leakage fluid will be analyzed todetermine the composition of the fluid. If additional core samples are taken prior to the end of the first tenyears of the period of extended operation, a sample ofleakaae fluid will be analyzed.

I P2:geptenmbeF 28,20-1-3 -om[)lete NL-08-127 NL-1 1-101NL-13-122 NL-09-056 NL-09-079 Audit Item3594-+ 437Enhance the Containment Inservice Inspection (CII-IWL) Program to include inspections of thecontainment using enhanced characterization ofdegradation (i.e., quantifying the dimensions of notedindications through the use of optical aids) during theperiod of extended operation.

The enhancement includes obtaining critical dimensional data ofdegradation where possible through directmeasurement or the use of scaling technologies forphotographs, and the use of consistent vantage pointsfor visual inspections.

NL-08-127 NL-13-122 Audit Item361 Docket Nos. 50-247 & 50-286NL-13-122 Attachment 2Page 17 of 20# COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTION/ AUDIT ITEMIP2: NL-08-143 4.2.138 For R eactor V essel Fluence, should future co re I ,,,,N,,0

-,,,,422, loading patterns invalidate the basis for the projected Comb e e N.......values of RTpts or CvUSE, updated calculations will 20-1--Complete NL-13-122 be provided to the NRC. IP3:December 12,201539 Deleted NL-09-079 40 Evaluate plant specific and appropriate industry

,P2: NL-09-106 B.1..6operating experience and incorporate lessons learned " " mb e t N ........

B.1..23in establishing appropriate monitoring and inspection B.1.24frequencies to assess aging effects for the new aging P3: B.1.25management programs.

Documentation of the em 1 B.1.25operating experience evaluated for each new program 0ecember 12, B.1.27will be available on site for NRC review prior to the B.1.33period of extended operation.

B.1.37I_ B.1.38P2: NL-11-032 N/A41 IPEC will inspect steam generators for both units to fter theassess the condition of the divider plate assembly.

The examination technique used will be capable of beginning of thedetecting PWSCC in the steam generator divider plate =EO and prior toassembly.

The IP2 steam generator divider plate September 28,inspections will be completed within the first ten years 023 NL-1 1-074of the period of extended operation (PEO). The IP3 P3: NL-1 1-090steam generator divider plate inspections will be nor to the endcompleted within the first refueling outage following of the first NL-1 1-101the beginning of the PEO. refueling outagefollowing thePeginning of theI_ __PEO. II Docket Nos. 50-247 & 50-286NL-13-122 Attachment 2Page 18 of 20COMMITMENT IMPLEMENTATION1 SOURCE RELATEDSCHEDULE LRA SECTIONI / I AUDIT ITEM42IPEC will develop a plan for each unit to address thepotential for cracking of the primary to secondary pressure boundary due to PWSCC of tube-to-tubesheet welds using one of the following twooptions.Option 1 (Analysis)

IPEC will perform an analytical evaluation of thesteam generator tube-to-tubesheet welds in order toestablish a technical basis for either determining thatthe tubesheet cladding and welds are not susceptible to PWSCC, or redefining the pressure boundary inwhich the tube-to-tubesheet weld is no longerincluded and, therefore, is not required for reactorcoolant pressure boundary function.

The redefinition of the reactor coolant pressure boundary must beapproved by the NRC as a license amendment request.Option 2 (Inspection)

IPEC will perform a one-time inspection of arepresentative number of tube-to-tubesheet welds ineach steam generator to determine if PWSCCcracking is present.

If weld cracking is identified:

a. The condition will be resolved through repairor engineering evaluation to justify continued

service, as appropriate, andb. An ongoing monitoring program will beestablished to perform routine tube-to-tubesheet weld inspections for the remaining life of the steam generators.

NL-11-032 NL-11-074 NL-1 1-090NL-11-096 N/AIP2:Prior to March2024IP3: Prior to theand of the firstrefueling outagefollowing theoeginning of thePEO.IP2:Between March2020 and March2024IP3: Prior to the9nd of the firstrefueling outageFollowing theDeginning of thePEO.U -t 443IPEC will review design basis ASME Code Class 1fatigue evaluations to determine whether theNUREG/CR-6260 locations that have been evaluated for the effects of the reactor coolant environment onfatigue usage are the limiting locations for the IP2 andIP3 configurations.

If more limiting locations areidentified, the most limiting location will be evaluated for the effects of the reactor coolant environment onfatigue usage.IPEC will use the NUREG/CR-6909 methodology inthe evaluation of the limiting locations consisting ofnickel alloy, if any.P2:P;9r4ereSeptefflbeF-8

_204-3-Complete P3: Prior toDecember 12,2015NL-1 1-032NL-13-122 NL-1 1-1014.3.3-~ I Docket Nos. 50-247 & 50-286NL-13-122 Attachment 2Page 19 of 20COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTIONI AUDIT ITEM44 IPEC will include written explanation and justification P2: NL-11-032 N/Aof any user intervention in future evaluations using the NL-11-101 W ESTEMS "Design CUF" module. ;e,-. ...m.e.....

NL-13-101 2-0-1--Com pl ete NL-13-122 IP3: Prior toDecember 12,201545 IPEC will not use the NB-3600 option of the IP2: NL-1 1-032 N/AWESTEMS program in future design calculations until NL-1F1Wthe issues identified during the NRC review of the Comlete NL-11-101 program have been resolved.

20--3-Complete NL-13-122 P3: Prior toDecember 12,201546 Include in the IP2 ISI Program that IPEC will perform P2: NL--1-032 N/Atwenty-five volumetric weld metal inspections of ,F ,, , ,-socket welds during each 10-year ISI interval Ste"NL-1 1-074scheduled as specified by IWB-2412 of the ASME 20--3-Complete NL-13-122 Section Xl Code during the period of extendedoperation.

In lieu of volumetric examinations, destructive examinations may be performed, where onedestructive examination may be substituted for twovolumetric examinations.

IP2: NL- 12-089 N/A47 IPEC will perform and submit analyses that nor todemonstrate that the lower support column bodies will NL-1 3-052maintain their functionality during the period of Augu 1, 204 NL-13-12extended operation considering the possible loss of August 15, 2014 NL-13-122 fracture toughness due to thermal and irradiation P3: Prior toembrittlement.

The analyses will be consistent with ecember 12,the IP2/IP3 licensing basis. 2015 Docket Nos. 50-247 & 50-286NL-13-122 Attachment 2Page 20 of 20COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTION/ AUDIT ITEMIP2: NL-12-174 N/A48 Entergy will visually inspect IPEC underground piping P2: 2within the scope of license renewal and subject toaging management review prior to the period of ,e43-I 4 N,extended operation and then on a frequency of at Complete NL-13-122 least once every two years during the period ofextended operation.

This inspection frequency will be DecePrir tomaintained unless the piping is subsequently coated 2015in accordance with the preventive actions specified inNUREG-1801 Section XI.M41 as modified by LR-ISG-2011-03.

Visual inspections will be supplemented with surface or volumetric non-destructive testing ifindications of significant loss of material areobserved.

Consistent with revised NUREG-1801 Section XI.M41, such adverse indications will beentered into the plant corrective action program forevaluation of extent of condition and for determination of appropriate corrective actions (e.g., increased inspection frequency, repair, replacement).

49I P2: NL-13-052 A.2.2.249 Recalculate each of the limiting CUFs provided in A.3.2.2section 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 2013 Complete NL-13-122 Program using NUREG/CR-5704 or NUREG/CR-IP3: Prior to6909. In accordance with the corrective actions ecember 12,specified in the Fatigue Monitoring Program,corrective actions include further CUF re-analysis, 015and/or repair or replacement of the affectedcomponents prior to the CUFen reaching 1.0.50 If the planned replacement of the IP2 split pins IP2: NL-13-122 A.2.1.41will not be accomplished in 2016, provide the Prior to March B.1.42NRC staff a detailed inspection plan for the IP2 31,2015split pins, including inspection methods,inspection

coverage, and inspection frequency, IP3: N/Aby March 31, 2015.