ML15331A181

Official Exhibit - NYS000498-00-BD01 - Dacimo, Fred, Entergy, Letter to Document Control Desk, Usnrc, License Renewal Application - Revised Reactor Vessel Internals Program and Inspection Plan, NL-12-134 (September 28, 2012) (ML12285A232)

ML15331A181
Person / Time
Site:	Indian Point
Issue date:	06/09/2015
From:	State of NY, Office of the Attorney General
To:	Atomic Safety and Licensing Board Panel
SECY RAS
References
RAS 27908, ASLBP 07-858-03-LR-BD01, 50-247-LR, 50-286-LR
Download: ML15331A181 (16)

Text

EntoW0Entergy Nuclear Northeast Indian Point Energy Center 450 Broadway, GSBP.O. Box 249Buchanan, NY 10511-0249 Tel (914) 254-2055Fred DacimnoVice President Operations License RenewalNL-12-134 September 28, 2012U.S. Nuclear Regulatory Commission ATTN: Document Control -DeskWashington, DC 20555-0001

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,"

dated May 15, 20122. Entergy letter (NL-12-089)," Reply to Request for Additional Information Regarding the License Renewal Application,"

dated June14, 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 RenewalApplication (LRA) for Indian Point 2 and Indian Point 3. The reply provided in this transmittal addresses RAIs 6, 7, 9, 10 and 11 on the Reactor Vessel Internals (RVI) Program.

Responses to RAIs 1-5, 8, and 12 were provided in Reference 2.If you have any questions, or require additional information, please contact Mr. Robert Walpoleat 914-254-6710.

I declare under penalty of perjury that the foregoing is true and correct.

Executed onSincerely, FRD/rw/

Docket Nos. 50-247 & 50-286NL-12-134 Page 2 of 2

Attachment:

1. Reply to NRC Request for Additional Information Regarding the LicenseRenewal Application cc: 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 IMr. Robert F. Kuntz, 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-12-134 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-12-134 Attachment 1Page 1 of 13INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3LICENSE RENEWAL APPLICATION (LRA)REQUESTS FOR ADDITIONAL INFORMATION (RAI)RAI's Related to Reactor Vessel Internals Inspection Plan fNRC Ref. 51RAI 6Applicant/Licensee Action Item 1 from the staff's final SE on MRP-227, Revision 0 requires thatapplicants/licensees submit an evaluation that demonstrates that their plant is bounded by theassumptions regarding plant design and operating history that were made in the failure modes,effects and consequences analyses (FMECA) and functionality analyses for reactors of theirdesign.The applicant's response to Applicant/Licensee Action Item 1 in the RVI inspection planaddresses the core loading assumptions (switch to a low-leakage core) and operational (baseloaded plant) aspects of design and operation that are mentioned in MRP-227-A, Section 2.4.An additional assumption listed in Section 2.4 of MRP-227.-A is that there have been no designchanges to the RVI beyond those identified in general industry guidance or recommended bythe original vendors.

Section 2.4 of MRP-227-A indicated that these assumptions areconsidered to conservatively represent any U.S. Pressurized Water Reactor operating plantprovided that these three assumptions are met, given the information on design and operation known to the MRP as of May 2007.MRP-191, Revision 0, "Materials Reliability Program:

Screening, Categorization and Ranking ofReactor Internals of Westinghouse and Combustion Engineering PWR Designs,"

documents thescreening for susceptibility to aging effects, the FMECA results, and the categorization andranking of the RVI components.

In addition to the assumptions listed in Section 2.4 of MRP-227-A, MRP-191 documents additional assumptions that were used. In particular, neutron fluencerange, temperature, and material grade for each generic component of the Westinghouse design internals were used for input to the screening process.

These values were determined based on an "expert elicitation" process.

Stress values were not explicitly tabulated, but wererecorded as either above the stress threshold

(>30 ksi) or not based on the expert interviews.

MRP-232, Revision 0, "Materials Reliability Program:

Aging Management Strategies forWestinghouse and Combustion Engineering PWR Internals,"

reported more specific stress,temperature and neutron fluence values based on finite element analyses for selected highconsequence of failure components identified in MRP-191.MRP-227 -A did not verify that the values of fluence, temperature, stress, and material, documented in MRP-1 91 and MRP-232 were bounding for all individual plants, and in fact MRP-227-A states, "These evaluations were based on representative configurations and operational histories, which were generally conservative, but not necessarily bounding in every parameter."

Each plant should have access to design information enabling verification that the material foreach RVI component is bounded by the design assumptions of the MRP. In this context, thestaff requests the following information:

NL-12-134 Attachment 1Page 2 of 131 ) To provide reasonable assurance that the RVI components are bounded by assumptions inthe FMECA and functionality analyses supporting the development of MRP-227-A, theapplicant is requested to respond to either 2.a or 2.b of this RAI:2.a)Provide the plant-specific values of neutron fluence (n/cm2, E>1.0 MeV), temperature, stress, and materials for a sample of RVI components.

The components selected shouldrepresent a range of neutron fluences, and temperatures.

This information should identifywhether the stress is greater or less than 30 ksi. Values of neutron fluence andtemperature may be estimated or analytical values. The values should be the peak valuesof each parameter for each component (e.g., peak end-of-life value for fluence).

Providethe method used to estimate the values, or describe the analysis method. An acceptable sample of components is:i) Lower Core Plateii) Core Barrel Flangeiii) Barrel-Former Boltsiv) Upper Core Barrel Weldsv) Lower Core Barrel Weldsvi) Upper Core Plate Alignment Pins2.b) If the sample verification approach in Part (a) is not used, describe the process used toverify that the RVI components at IP2 and IP3 are bounded by the assumptions regarding the neutron fluence, temperature, stress values, and materials that were made for eachcomponent in the FMECA and functionality analyses supporting the development of MRP-227-A.3) If there are any components at IP2 or IP3 not bounded by assumptions regarding neutronfluence, temperature, stress or material used in the development of MRP-227-A, describehow the differences were addressed in the plant-specific RVI Inspection Plan. The staffrequests that the applicant, as a part of its demonstration, discuss whether there would beany changes to the screening, categorization, FMECA process and functionality analyses ifthe plant-specific variables (the neutron fluence, temperature, stress values, plant-specific operating experience, and materials) are used. This evaluation should address whetheradditional aging mechanisms would become applicable to the component.

4) For any non-bounded components, determine if any changes to the inspection requirements of MRP-227-A are needed. Provide plant-specific inspection requirements or an alternate aging management

program, as appropriate.

If no changes to the inspection requirementsare proposed, provide a justification for the adequacy of the existing MRP-227-A inspections for the unbounded components.

5) Identify all design changes to the IP2 and IP3 RVI, and describe (1) if any of these arebeyond those identified in general industry guidance or recommended by the originalvendors, and (2) if any of the design changes were implemented after May 2007. Assess theimpact of these design changes on the recommendations of the RVI Inspection Plan.Provide plant-specific inspection requirements if necessary for the affected components.

NL-12-134 Attachment 1Page 3 of 13Response to RAI 6Indian Point Unit 21) To provide reasonable assurance that the RVI components are bounded byassumptions in the generic industry failure modes, effects and consequences analyses(FMECA),

MRP-191, and functionality

analyses, MRP-232, supporting the development of MRP-227-A, a response to request 2.b is provided.

2.a) Based on the response in item 1 above, no response is necessary.2.b) The process used to verify that Indian Point Unit 2 (IP2) is reasonably represented bythe generic industry program assumptions with regard to neutron fluence, temperature,materials, and stress values used in the development of MRP-227-A is as follows:1. Identification of typical Westinghouse pressurized water reactor (PWR) internalcomponents in MRP-191, Table 4-4.2. Identification of IP2 PWR internal components.

3. Comparison of the typical Westinghouse PWR internal components to the IP2PWR internal components.

a. Confirm that no additional items were identified by this comparison

b. Confirm that the materials identified for IP2 are consistent with thosematerials identified in MRP-191, Table 4-4.c. Confirm that the IP2 internals are the same as, or equivalent to, thetypical Westinghouse PWR internals regarding design and fabrication.

4. Confirmation that the IP2 operating history is consistent with the assumptions inMRP-227-A regarding core loading patterns.

5. Confirmation that the IP2 RVI materials operated at temperatures within theoriginal design basis parameters.

6. Determination of stress values based on design basis documents.

7. Confirmation that any changes to the IP2 RVI components do not impact theapplication of the MRP-227-A generic aging management strategy.

The IP2 RVI components are reasonably represented by assumptions regarding neutronfluence, temperature, materials, and stress values in the generic industry FMECAand functionality analyses based on the following:

1. IP2 operating history is consistent with the assumptions used to develop theMRP-227-A aging management strategy with regard to neutron fluence.a. The FMECA and functionality analyses for MRP-227-A were based onthe assumption of 30 years of operation with high-leakage core loading NL-12-134 Attachment 1Page 4 of 13patterns followed by 30 years of low-leakage core fuel management strategy.

IP2 started to go from high-leakage to low-leakage loadingpattern in Fuel Cycle 6. In Fuel Cycle 6 (12/29/82) at 9 years of operation, IP2 switched to use of a low-low leakage loading pattern (L3p). Then, inFuel Cycle 12 (4/20/93) at 19 years of operation, IP2 switched to use of alow-low-low leakage loading pattern (L4p). Therefore, IP2 meets thefluence and fuel management assumptions in MRP-191, MRP-232,and the requirements for application of the MRP-227-A agingmanagement strategy.

b. IP2 has operated under base load conditions over the life of the plant.Therefore, IP2 satisfies the assumptions in MRP-227-A and supporting documents regarding operational parameters affecting fluence.2. The IP2 RVI operate between Thot and TcoId, which are not less thanapproximately 514°F (515.50F prior to Stretch Power Uprate [SPU]) for ToId andnot higher than 605.8°F (611.7°F prior to SPU) for Thot. The design temperature for the vessel is 650'F. Therefore, IP2 historical operation is within originaldesign basis parameters and is consistent with the assumptions used to developthe aging management strategy in MPR-227-A with regard to temperature operational parameters.

3. IP2 RVI components and materials are covered by the list of genericWestinghouse-designed PWR RVI components in MRP-191, Table 4-4.Components and materials are summarized in [Entergy reference 1].a. As summarized in [Entergy reference 1], no additional components wereidentified for IP2 by this comparison.

b. Most of the IP2 RVI component materials are consistent or nearlyequivalent with those materials identified in MRP-191, Table 4-4 forWestinghouse-designed plants, except for a few components that werefabricated from CF8 cast austenitic stainless steel (CASS) material ratherthan the Type 304 stainless steel (SS) called out in MRP-191.

Theseitems along with the items that were fabricated from different butessentially equivalent materials are summarized in a Westinghouse proprietary document

[Entergy reference 1]. Plant-specific details areproprietary and not typically released as part of RAI responses.

If theNRC requires additional

details, the calculation would be made available for the NRC to review the supporting technical basis. This typically occursat either a plant location or at the Westinghouse Rockville office withWestinghouse personnel supporting the discussion.

c. The design and fabrication of IP2 RVI components are the same as, orequivalent to, the typical Westinghouse-designed PWR RVI components.

4. Modifications to the IP2 RVI made over the lifetime of the plant are thoseidentified in general industry guidance or specifically directed by Westinghouse.

IP2 performed a SPU in 2004 where analyses were performed on RVIcomponents and it was determined that the structural integrity of the reactorinternals was maintained at the SPU conditions.

The design has been maintained NL-12-134 Attachment 1Page 5 of 13over the lifetime of the plant as specified by Westinghouse, and IP2 has notmade any modifications since May 2007, which meets requirements from MRP-227-A. Operational parameters with regard to fluence and temperature arecompliant with requirements from MRP-227-A, and the components andmaterials are the same as those considered in MRP-1 91. Therefore, the IP2stress values are reasonably represented by the assumptions in MRP-1 91, MPR-227- A, and MRP-232.3) Per the response to item 2.b of this RAI, IP2 complied with Applicant/Licensee ActionItem 1 of the NRC Safety Evaluation (SE) regarding MRP-227, Revision 0, except for theCF8 components that were called out as 304 SS in MRP-1 91. Those materialdifferences were evaluated and determined, as summarized in (Entergy reference 1],to require no modifications to the generic MRP-227-A inspection strategy.

Therefore, the requirement is met for application of MRP-227-A as a strategy for managing age-related material degradation in the IP2 components, and no additional agingmechanisms are applicable to components at IP2.4) See response to item 3 of this RAI. A FMECA was done addressing the CF8components that were called out as 304 SS in MRP-191 and determined that thecomponents could be classified as "No Additional Measures" components based on aconsideration of the likelihood of failure and the likelihood of damage. Documentation forthis FMECA is summarized in [Entergy reference 1].5) No design changes were made to IP2 that were beyond those identified in generalindustry guidance or as were recommended by the original vendors.

There were nodesign changes implemented after May 2007 [Entergy reference 5].Indian Point Unit 31) To provide reasonable assurance that the RVI components are bounded byassumptions in the generic industry FMECA, MRP-191, and functionality

analyses, MRP-232, supporting the development of MRP-227-A, a response to request 2.b isprovided.

2.a) Based on the response in item 1 above, no response is necessary.

2.b) The process used to verify that Indian Point Unit 3 (IP3) is reasonably represented bythe generic industry program assumptions with regard to neutron fluence, temperature, materials, and stress values used in the development of MRP-227-A is as follows:1. Identification of typical Westinghouse PWR internal components MRP-191, Table4-4.2. Identification of IP3 PWR internal components.

3. Comparison of the typical Westinghouse PWR internal components to the IP3PWR internal components.

a. Confirm that no additional items were identified by this comparison.

NL-12-134 Attachment 1Page 6 of 13b. Confirm that the materials identified for IP3 are consistent with thosematerials identified in MRP-191, Table 4-4.c. Confirm that the IP3 internals are the same as, or equivalent to, thetypical Westinghouse PWR internals regarding design and fabrication.

4. Confirmation that the IP3 operating history is consistent with the assumptions inMRP-227-A regarding core loading patterns.

5. Confirmation that the IP3 RVI materials operated at temperatures within theoriginal design basis parameters.

6. Determination of stress values based on design basis documents.

7. Confirmation that any changes to the IP3 RVI components do not impact theapplication of the MRP-227-A generic aging management strategy.

The IP3 RVI components are reasonably represented by assumptions regarding neutronfluence, temperature, materials, and stress values in the generic industry FMECAand functionality analyses based on the following:

1. IP3 operating history is consistent with the assumptions used to develop theMRP-227-A with regard to neutron fluence.a. The FMECA and functionality analyses for MRP-227-A were based onthe assumption of 30 years of operation with high-leakage core loadingpatterns followed by 30 years of low-leakage core fuel management strategy.

IP3 started to go from high-leakage to low-leakage loadingpattern in Fuel Cycle 4. In Fuel Cycle 7 (6/24/84) at 9 years of operation, IP3 switched to use of a low-low-leakage loading pattern (L3p). Then, inFuel Cycle 14 (4/7/05) at 30 years of operation, IP3 switched to use of alow-low-low-leakage loading pattern (L4p). Therefore, IP3 meets thefluence and fuel management assumptions in MRP-191, MRP-232,and the requirements for application of the MRP-227-A agingmanagement strategy.

b. IP3 has operated under base load conditions over the life of the plant.Therefore, IP3 satisfies the assumptions in MRP-227-A and supporting documentation regarding operational parameters affecting fluence.2. The IP3 RVI operate between Thot and T,0od, which are not less thanapproximately 517°F (541.90F prior to SPU) for Tood and not higher than 603'F(600.80F prior to SPU) for Thot. The design temperature for the vessel is6500F. Therefore, IP3 historical operation is within original design basisparameters and is consistent with the assumptions used to develop the agingmanagement strategy in MRP-227-A with regard to temperature operational parameters.

3. IP3 RVI components and materials are covered by the list of genericWestinghouse-designed PWR RVI components in MRP-191, Table 4-4.Components and materials are summarized in [Entergy reference 1].

NL-12-134 Attachment 1Page 7 of 13a. As summarized in [Entergy reference 1], no additional components wereidentified for IP3 by this comparison.

b. Most of the IP3 RVI component materials are consistent or nearlyequivalent with those materials identified in MRP-191, Table 4-4 forWestinghouse-designed plants, except for a few components that werefabricated from CF8 CASS material rather than the Type 304 SS calledout in MRP-1 91. These items along with the items that were fabricated from different but essentially equivalent materials are summarized in aWestinghouse proprietary document

[Entergy reference 1]. Plant-specific details are proprietary and not typically released as part of RAIresponses.

If the NRC requires additional

details, the calculation wouldbe made available for the NRC to review the supporting technical basis.This typically occurs at either a plant location or at the WestinghouseRockville office with Westinghouse personnel supporting the discussion.

c. The design and fabrication of IP3 RVI components are the same as, orequivalent to, the typical Westinghouse-designed PWR RVI components.

4. Modifications to the IP3 RVI made over the lifetime of the plant are thoseidentified in general industry guidance or specifically directed by Westinghouse.

IP3 performed a SPU in 2005 where analyses were performed on RVIcomponents and it was determined that the structural integrity of the reactorinternals was maintained at the SPU conditions.

The design has been maintained over the lifetime of the plant as specified by Westinghouse, and IP3 has notmade any modifications since May 2007, which meets requirements from MRP-227-A. Operational parameters with regard to fluence and temperature arecompliant with requirements from MRP-227-A, and the components andmaterials are the same as those considered in MRP-1 91. Therefore, the IP3stress values are reasonably represented by the assumptions in MRP-1 91, MPR-227-A, and MRP-232.3) Per the response to item 2.b of this RAI, IP3 complied with Applicant/Licensee ActionItem 1 of the NRC SE regarding MRP-227, Revision 0, except for the CF8 components that were called out as 304 SS in MRP-191.

Those material differences were evaluatedand determined, as summarized in [Entergy reference 1], to require no modifications tothe generic MRP-227-A inspection strategy.

Therefore, the requirement is met forapplication of MRP-227-A as a strategy for managing age-related material degradation in the IP3 components, and no additional aging mechanisms are applicable tocomponents at IP3.4) See response to item 3 of this RAI. A FMECA was done addressing the CF8components that were called out as 304 SS in MRP-1 91 and determined that thecomponents could be classified as "No Additional Measures" components based on aconsideration of the likelihood of failure and the likelihood of damage. Documentation forthis FMECA is summarized in [Entergy reference 1].5) No design changes were made to IP3 that were beyond those identified in generalindustry guidance or as were recommended by the original vendors.

There were nodesign changes implemented after May 2007 [Entergy reference 5].

NL-12-134 Attachment 1Page 8 of 13RAI 7The staff reviewed the applicant's response to Applicant/Licensee Action Item 2 from the NRCstaff's final SE on MRP-227, Revision

0. In Section 3.6 of the RVI Inspection Plan (Ref. 5), theapplicant stated that it reviewed the information in Table 4-4 of MRP-191 and determined thatthis table contains all the RVI components that are within the scope of license renewal and thatthis is shown in Table 5-7. The staff notes that Table 5-1 contains a cross-index between thecomponent designations in Entergy Letter NL-10-063 (Amendment 9 to the LRA, Ref. 1) and thecomponent names as designated in MRP-191, Table 4-4 (Ref. 6). All the IPEC component designations correlate with an equivalent component designation in MRP-191 (Ref. 7), Table 44with the exception of the Lower Internals Assembly

-Column Cap.The staff therefore requests that the applicant verify that the Lower Internals Assembly

-ColumnCap would be subject to the same inspection requirements that are applied to the lower supportassembly, lower support column bodies (cast) in MRP-227-A, Table 4-6. If not, provide plant-specific aging management requirements for the Lower Internals Assembly

-Column Cap.Response to RAI 7Indian Point Unit 2In RAI 7, the NRC notes that all IP2 components within the scope of license renewal comparewith an equivalent component designation in MRP-191 with the exception of the lower internals assembly column cap. The lower internals assembly column cap is a CASS piece welded onto the top of the core support column shaft. These two pieces together constitute the lowerinternals assembly

-column body.

The development of the MRP-227-A aging managementstrategy considered the lower support column as one complete unit denoted as the "lowersupport column assemblies

-lower support column bodies."

Under the lower support columnbodies in MRP-191, both Type 304 SS and CF8 CASS material were considered, assummarized in [Entergy reference 1]. Since the lower internals assembly column cap is part ofthe lower support column body, it was addressed by the generic industry FMECA andfunctionality analysis; therefore, it is subject to the same inspection requirements as the lowersupport column bodies (cast). Since the column cap is a part of the lower support column body,it would be subject to the same inspection requirements as the lower support column body. Theinspection of the lower support column body includes the column cap.Indian Point Unit 3In RAI 7, the NRC notes that all IP3 components within the scope of license renewal comparewith an equivalent component designation in MRP-1 91 with the exception of the lower internals assembly column cap. The lower internals assembly column cap is a CASS piece welded ontothe top of the core support column shaft. These two pieces together constitute the lowerinternals assembly

-column body. The development of the MRP-227-A aging management strategy considered the lower support column as one complete unit denoted as the "lowersupport column assemblies

-lower support column bodies."

Under the lower support columnbodies in MRP-191, both Type 304 SS and CF8 CASS material were considered, assummarized in [Entergy reference 1]. Since the lower internals assembly column cap is part ofthe lower support column body, it was addressed by the generic industry FMECA andfunctionality analysis; therefore, it is subject to the same inspection requirements as the lower NL-12-134 Attachment 1Page 9 of 13support column bodies (cast).

Since the column cap is a part of the lower support column body,it would be subject to the same inspection requirements as the lower support column body. Theinspection of the lower support column body includes the column cap.RAI 9The applicant's response to Applicant/Licensee Action Item 5 from Revision 1 of the staff's finalSE on MRP-227, states in part that the acceptance criteria will ensure the remaining compressible height of the spring shall provide hold down forces within the IPEC designtolerance.

If a plant specific acceptance criterion is not developed for the hold down spring,IPEC will replace the spring in lieu of performing the first required physical measurement.

MRP-227-A, Table 4-3, calls for direct measurement of the hold-down spring height within threecycles of the beginning of the license renewal period. If the first set of measurements is notsufficient to determine life, spring height measurements must be taken during the next twooutages, in order to extrapolate the expected spring height to 60 years.The staff requires clarification of how the applicant will determine whether the first set ofmeasurements could be extrapolated to demonstrate acceptable spring functionality through 60years. Therefore, the staff requests the following information:

1. Provide the specific acceptance criteria for spring height and/or hold down force from theIP2/IP3 licensing basis.2. Describe the procedure by which the remaining hold down forces will be projected toend-of-life based on one measurement.

Address whether the decrease in spring heightor hold-down force is assumed to occur linearly over time or via some other function oftime.3. What results of the first spring measurements would indicate a need for successive measurements?

Response to RAI 9Indian Point Unit 2 and Indian Point Unit 31. The acceptable criteria are based on the measured height of the spring as a function oftime relative to the required hold-down force. Applicable plant loading conditions wereevaluated.

Time-dependent details for the hold down spring (HDS) height measurements are summarized in [Entergy reference 1]. Confirmatory actions are included if applicable after completion of the evaluation.

Plant-specific details are proprietary and not typically released as part of RAI responses.

If the NRC requires additional details, the calculation would be made available for the NRC to review the supporting technical basis. Thistypically occurs at either a plant location or at the Westinghouse Rockville office withWestinghouse personnel supporting the discussion.

2. The decrease in HDS height is assumed to occur linearly over time. The approach usedto develop the HDS height acceptance criteria is to consider the actual HDS height atplant start-up and the required HDS heights at the end of 60 years. A linear interpolationat the time of the HDS height measurement determines the required minimum HDSheight.

NL-12-134 Attachment 1Page 10 of 133. The HDS height measures less than the required minimum HDS heights, assummarized in [Entergy reference 1], indicating a need for re-evaluation and successive measurements or a replacement HDS.RAI 10The applicant's response to Applicant/Licensee Action Item 7 indicates that the plant-specific analysis to demonstrate functionality of the lower support column bodies during the period ofextended operation will be submitted to the NRC prior to the PEO. In the aging management review tables submitted in LRA Amendment 9, the applicant credits the "Thermal Aging andNeutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS) Program" formanaging loss of fracture toughness of the lower core support column bodies, as well asseveral other CASS components.

NUREG-1 930 indicates that the staff determined this programwas consistent with the Generic Aging Lessons Learned Report, Revision 1, AgingManagement Program (AMP) XI.M13, "Thermal Aging and Neutron Irradiation Embrittlement ofCast Austenitic Stainless Steel (CASS) Program."

Per GALL, Rev. 1,Section XI.M13, the"Thermal Aging and Neutron Irradiation Embrittlement of CASS Program" generally requiressupplemental visual inspections (equivalent to an EVT-1) for CASS RVI components that areeither susceptible to thermal aging based on chemistry and other manufacturing parameters, orreceive a neutron fluence > lx1017 n/cm2, unless it can be demonstrated that the stresses onthe component are either compressive or low in magnitude if tensile.

The RVI Program iscredited with managing cracking of the core support column bodies and other CASScomponents.

Under the RVI Program, the core support column bodies are expansion components that would be subject to an EVT-1 visual examination for cracking due to irradiation assisted stress corrosion cracking if cracking were found in the associated primary component.

The staff requests the following information:

Since both the plant-specific analysis and Thermal Aging and Neutron Irradiation Embrittlement of CASS Program could both potentially involve screening for thermal or neutron irradiation embrittlement, stress analyses, and flaw tolerance evaluations, and both the RVI Program andThermal Aging and Neutron Irradiation Embrittlement of CASS Program could potentially requireinspections, discuss the relationship of the two programs and the plant-specific analysis.

Response to RAI 10Indian Point Unit 2 and Indian Point Unit 3The RVI Program is the base program that addresses the RVI components that require agingmanagement and specifies inspection requirements.

The Thermal Aging and Neutron Irradiation Embrittlement of CASS Program at Indian Point (IPEC LRA) augments the RVI program byevaluating the potential susceptibility of plant-specific CASS components based on component specific chemistry.

RAI 11In response to Applicant/Licensee Action Item 7, the applicant stated that the plant-specific analyses to demonstrate the lower support column bodies will maintain their functionality duringthe period of extended operation will consider the possible loss of fracture toughness in thesecomponents due to thermal and irradiation embrittlement.

The analyses will be consistent with NL-12-134 Attachment 1Page 11 of 13the IP2/1 P3 licensing basis and the need to maintain the functionality of the lower supportcolumn bodies under all licensing basis conditions of operations.

The staff requests the following additional information:

1) Section 3.3.7 of Revision 1 of the staff's final SE on MRP-227, Revision 0 lists threepossible options for the type of plant-specific analysis used to fulfill the requirements ofthis action item. The three approaches are 1) functionality analyses of the set of likecomponents,

2) component-specific flaw tolerance evaluations, or 3) a screening approach demonstrating that the CASS Components are not susceptible to thermalembrittlement, neutron embrittlement, or the combined effects of both. Discuss which ofthese approaches will be used and why.2) Describe the acceptance criteria for the plant-specific analysis results that are derivedfrom the IP2/IP3 licensing basis.3) Since the applicant stated that the analysis of the core support columns will be submitted prior to the period of extended operation for IP2 and IP3, the staff requests the applicant submit a letter documenting this as a formal licensing commitment.

Response to RAI 11Indian Point Unit 2 and Indian Point Unit 31) The approach used to evaluate the susceptibility of the CASS components at IP2 andIP3 to thermal and neutron embrittlement is a screening approach.

A screening approach was used as this is consistent with the MRP-227-A development methodology and a methodology and screening values have been endorsed by the NRC [Entergyreference 3] for application with CASS materials.

2) The acceptance criteria for the plant-specific CASS component susceptibility evaluations based on fabrication and chemistry, taken from [Entergy references 2 and 3], aresummarized in the following table.Molybdenum Casting NRC Susceptibility (wt %) Method Delta-Ferrite

% Evaluation High 2.0-3.0 Static >14% Potentially susceptible to TE<14% Not susceptible to TECentrifugal

>20% Potentially susceptible to TE<20% Not susceptible to TELow 0.5 max Static >20% Potentially susceptible to TE<20% Not susceptible to TECentrifugal all Not susceptible to TE NL-12-134 Attachment 1Page 12 of 133) Commitment 47 in [Entergy reference 4] states that IPEC will perform and submitanalyses that demonstrate that the lower support column bodies will maintain theirfunctionality during the period of extended operation.

These analyses will consider thepossible loss of fracture toughness due to thermal and irradiation embrittlement.

Theanalyses will be consistent with the IP2/IP3 licensing basis.NRC References

1. Letter from Fred Dacimo, Entergy, to NRC dated July 14, 2010,

Subject:

Amendment 9 toLicense Renewal Application (LRA) -Reactor Vessel Internals Program Indian Point NuclearGenerating Unit Nos. 2 & 3, Docket Nos.

50-247 and 50-286 License Nos. DPR-26 andDPR-64 (ADAMS Accession No. ML102010102)

2. Letter from Robert Nelson, NRC, to Neil Wilmshurst, EPRI dated December 16, 2011;

Subject:

Revision 1 of the Final Safety Evaluation of EPRI Report, Materials Reliability Program Report 1016596 (MRP-227),

Revision 0, "Pressurized Water Reactor (PWR)Internals Inspection and Evaluation Guidelines" (TAC No. ME0680) (ADAMS Accession No.ML11308A770)

3. Reactor Internals Acceptance Criteria Methodology and Data Requirements, WCAP-17096-NP, Rev. 2, Westinghouse Non-Proprietary Class 3 Report, December 2009, ADAMSAccession No. ML1014601570

4. Materials Reliability Program:

Inspection Standard for PWR Internals (MRP-228) 1016609Final Report, July 2009 Electric Power Research Institute, Palo Alto, CA (EPRI Product No.1016609)

(ADAMS Accession No. ML092120573)

5. Indian Point Energy Center Revised Reactor Vessel Internals Inspection Plan Compliant with MRP-227-A.

Attachment 2 to Entergy Letter NL-12-037, Letter from Fred Dacimo toNRC dated February 17, 2012,

Subject:

License Renewal Application

-Revised ReactorVessel Internals Program and Inspection Plan Compliant with MRP-227-A, Indian PointNuclear Generating Unit Nos. 2 and 3, Docket Nos. 50-247 and 50-286-License Nos. DPR-26 and DPR-64 (ADAMS Accession No. ML1206A312)

6. MRP-191 Revision 0, "Materials Reliability Program:

Screening, Categorization and Rankingof Reactor Internals of Westinghouse and Combustion Engineering PWR Designs,"

ADAMSAccession No. ML091910130

7. NUREG-1930, Volume 2, "Safety Evaluation Report Related to The License Renewal ofIndian Point Nuclear Generating Unit Nos. 2 and 3, Dockets No. 50-247 and 50-286,November 30, 2009 (ADAMS Accession No. ML093170671)

Enterqy References

1. Westinghouse Letter, LTR-RIAM-12-112, Rev. 0, "Indian Point Units 2 and 3 Support for ReactorVessel Internals Aging Management Request for Additional Information (RAI) Project -Request forTransmittal of Final Comment Resolutions on RAI Responses,"

September 25, 2012.

NL-12-134 Attachment 1Page 13 of 132. U.S. Nuclear Regulatory Commission, NUREG/CR-4513, Rev. 1, "Estimation of FractureToughness of Cast Stainless Steels During Thermal Aging in LWR Systems,"

May 1994 (NRCADAMS Accession No. ML052360554).

3. U.S. Nuclear Regulatory Commission Letter, "License Renewal Issue No. 98-0030, Thermal AgingEmbrittlement of Cast Austenitic Stainless Steel Components,"

May 19, 2000 (NRC ADAMSAccession No. ML003717179).

4. Entergy Letter, NL-12-089, "Reply to Request for Additional Information Regarding the LicenseRenewal Application Indian Point Nuclear Generating Unit Nos. 2 & 3 Docket Nos. 50-247 and 50-286 License Nos. DPR-26 and DPR-64,"

June 14, 2012.5. Entergy Letter, NL-10-063, "Amendment 9 to License Renewal Application (LRA) -Reactor VesselInternals Program Indian Point Nuclear Generating Unit Nos. 2 & 3 Docket Nos. 50-247 and 50-286License Nos. DPR-26 and DPR-64,"

July 14, 2010.

EntoW0Entergy Nuclear Northeast Indian Point Energy Center 450 Broadway, GSBP.O. Box 249Buchanan, NY 10511-0249 Tel (914) 254-2055Fred DacimnoVice President Operations License RenewalNL-12-134 September 28, 2012U.S. Nuclear Regulatory Commission ATTN: Document Control -DeskWashington, DC 20555-0001

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,"

dated May 15, 20122. Entergy letter (NL-12-089)," Reply to Request for Additional Information Regarding the License Renewal Application,"

dated June14, 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 RenewalApplication (LRA) for Indian Point 2 and Indian Point 3. The reply provided in this transmittal addresses RAIs 6, 7, 9, 10 and 11 on the Reactor Vessel Internals (RVI) Program.

Responses to RAIs 1-5, 8, and 12 were provided in Reference 2.If you have any questions, or require additional information, please contact Mr. Robert Walpoleat 914-254-6710.

I declare under penalty of perjury that the foregoing is true and correct.

Executed onSincerely, FRD/rw/

Docket Nos. 50-247 & 50-286NL-12-134 Page 2 of 2

Attachment:

1. Reply to NRC Request for Additional Information Regarding the LicenseRenewal Application cc: 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 IMr. Robert F. Kuntz, 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-12-134 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-12-134 Attachment 1Page 1 of 13INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3LICENSE RENEWAL APPLICATION (LRA)REQUESTS FOR ADDITIONAL INFORMATION (RAI)RAI's Related to Reactor Vessel Internals Inspection Plan fNRC Ref. 51RAI 6Applicant/Licensee Action Item 1 from the staff's final SE on MRP-227, Revision 0 requires thatapplicants/licensees submit an evaluation that demonstrates that their plant is bounded by theassumptions regarding plant design and operating history that were made in the failure modes,effects and consequences analyses (FMECA) and functionality analyses for reactors of theirdesign.The applicant's response to Applicant/Licensee Action Item 1 in the RVI inspection planaddresses the core loading assumptions (switch to a low-leakage core) and operational (baseloaded plant) aspects of design and operation that are mentioned in MRP-227-A, Section 2.4.An additional assumption listed in Section 2.4 of MRP-227.-A is that there have been no designchanges to the RVI beyond those identified in general industry guidance or recommended bythe original vendors.

Section 2.4 of MRP-227-A indicated that these assumptions areconsidered to conservatively represent any U.S. Pressurized Water Reactor operating plantprovided that these three assumptions are met, given the information on design and operation known to the MRP as of May 2007.MRP-191, Revision 0, "Materials Reliability Program:

Screening, Categorization and Ranking ofReactor Internals of Westinghouse and Combustion Engineering PWR Designs,"

documents thescreening for susceptibility to aging effects, the FMECA results, and the categorization andranking of the RVI components.

In addition to the assumptions listed in Section 2.4 of MRP-227-A, MRP-191 documents additional assumptions that were used. In particular, neutron fluencerange, temperature, and material grade for each generic component of the Westinghouse design internals were used for input to the screening process.

These values were determined based on an "expert elicitation" process.

Stress values were not explicitly tabulated, but wererecorded as either above the stress threshold

(>30 ksi) or not based on the expert interviews.

MRP-232, Revision 0, "Materials Reliability Program:

Aging Management Strategies forWestinghouse and Combustion Engineering PWR Internals,"

reported more specific stress,temperature and neutron fluence values based on finite element analyses for selected highconsequence of failure components identified in MRP-191.MRP-227 -A did not verify that the values of fluence, temperature, stress, and material, documented in MRP-1 91 and MRP-232 were bounding for all individual plants, and in fact MRP-227-A states, "These evaluations were based on representative configurations and operational histories, which were generally conservative, but not necessarily bounding in every parameter."

Each plant should have access to design information enabling verification that the material foreach RVI component is bounded by the design assumptions of the MRP. In this context, thestaff requests the following information:

NL-12-134 Attachment 1Page 2 of 131 ) To provide reasonable assurance that the RVI components are bounded by assumptions inthe FMECA and functionality analyses supporting the development of MRP-227-A, theapplicant is requested to respond to either 2.a or 2.b of this RAI:2.a)Provide the plant-specific values of neutron fluence (n/cm2, E>1.0 MeV), temperature, stress, and materials for a sample of RVI components.

The components selected shouldrepresent a range of neutron fluences, and temperatures.

This information should identifywhether the stress is greater or less than 30 ksi. Values of neutron fluence andtemperature may be estimated or analytical values. The values should be the peak valuesof each parameter for each component (e.g., peak end-of-life value for fluence).

Providethe method used to estimate the values, or describe the analysis method. An acceptable sample of components is:i) Lower Core Plateii) Core Barrel Flangeiii) Barrel-Former Boltsiv) Upper Core Barrel Weldsv) Lower Core Barrel Weldsvi) Upper Core Plate Alignment Pins2.b) If the sample verification approach in Part (a) is not used, describe the process used toverify that the RVI components at IP2 and IP3 are bounded by the assumptions regarding the neutron fluence, temperature, stress values, and materials that were made for eachcomponent in the FMECA and functionality analyses supporting the development of MRP-227-A.3) If there are any components at IP2 or IP3 not bounded by assumptions regarding neutronfluence, temperature, stress or material used in the development of MRP-227-A, describehow the differences were addressed in the plant-specific RVI Inspection Plan. The staffrequests that the applicant, as a part of its demonstration, discuss whether there would beany changes to the screening, categorization, FMECA process and functionality analyses ifthe plant-specific variables (the neutron fluence, temperature, stress values, plant-specific operating experience, and materials) are used. This evaluation should address whetheradditional aging mechanisms would become applicable to the component.

4) For any non-bounded components, determine if any changes to the inspection requirements of MRP-227-A are needed. Provide plant-specific inspection requirements or an alternate aging management

program, as appropriate.

If no changes to the inspection requirementsare proposed, provide a justification for the adequacy of the existing MRP-227-A inspections for the unbounded components.

5) Identify all design changes to the IP2 and IP3 RVI, and describe (1) if any of these arebeyond those identified in general industry guidance or recommended by the originalvendors, and (2) if any of the design changes were implemented after May 2007. Assess theimpact of these design changes on the recommendations of the RVI Inspection Plan.Provide plant-specific inspection requirements if necessary for the affected components.

NL-12-134 Attachment 1Page 3 of 13Response to RAI 6Indian Point Unit 21) To provide reasonable assurance that the RVI components are bounded byassumptions in the generic industry failure modes, effects and consequences analyses(FMECA),

MRP-191, and functionality

analyses, MRP-232, supporting the development of MRP-227-A, a response to request 2.b is provided.

2.a) Based on the response in item 1 above, no response is necessary.2.b) The process used to verify that Indian Point Unit 2 (IP2) is reasonably represented bythe generic industry program assumptions with regard to neutron fluence, temperature,materials, and stress values used in the development of MRP-227-A is as follows:1. Identification of typical Westinghouse pressurized water reactor (PWR) internalcomponents in MRP-191, Table 4-4.2. Identification of IP2 PWR internal components.

3. Comparison of the typical Westinghouse PWR internal components to the IP2PWR internal components.

a. Confirm that no additional items were identified by this comparison

b. Confirm that the materials identified for IP2 are consistent with thosematerials identified in MRP-191, Table 4-4.c. Confirm that the IP2 internals are the same as, or equivalent to, thetypical Westinghouse PWR internals regarding design and fabrication.

4. Confirmation that the IP2 operating history is consistent with the assumptions inMRP-227-A regarding core loading patterns.

5. Confirmation that the IP2 RVI materials operated at temperatures within theoriginal design basis parameters.

6. Determination of stress values based on design basis documents.

7. Confirmation that any changes to the IP2 RVI components do not impact theapplication of the MRP-227-A generic aging management strategy.

The IP2 RVI components are reasonably represented by assumptions regarding neutronfluence, temperature, materials, and stress values in the generic industry FMECAand functionality analyses based on the following:

1. IP2 operating history is consistent with the assumptions used to develop theMRP-227-A aging management strategy with regard to neutron fluence.a. The FMECA and functionality analyses for MRP-227-A were based onthe assumption of 30 years of operation with high-leakage core loading NL-12-134 Attachment 1Page 4 of 13patterns followed by 30 years of low-leakage core fuel management strategy.

IP2 started to go from high-leakage to low-leakage loadingpattern in Fuel Cycle 6. In Fuel Cycle 6 (12/29/82) at 9 years of operation, IP2 switched to use of a low-low leakage loading pattern (L3p). Then, inFuel Cycle 12 (4/20/93) at 19 years of operation, IP2 switched to use of alow-low-low leakage loading pattern (L4p). Therefore, IP2 meets thefluence and fuel management assumptions in MRP-191, MRP-232,and the requirements for application of the MRP-227-A agingmanagement strategy.

b. IP2 has operated under base load conditions over the life of the plant.Therefore, IP2 satisfies the assumptions in MRP-227-A and supporting documents regarding operational parameters affecting fluence.2. The IP2 RVI operate between Thot and TcoId, which are not less thanapproximately 514°F (515.50F prior to Stretch Power Uprate [SPU]) for ToId andnot higher than 605.8°F (611.7°F prior to SPU) for Thot. The design temperature for the vessel is 650'F. Therefore, IP2 historical operation is within originaldesign basis parameters and is consistent with the assumptions used to developthe aging management strategy in MPR-227-A with regard to temperature operational parameters.

3. IP2 RVI components and materials are covered by the list of genericWestinghouse-designed PWR RVI components in MRP-191, Table 4-4.Components and materials are summarized in [Entergy reference 1].a. As summarized in [Entergy reference 1], no additional components wereidentified for IP2 by this comparison.

b. Most of the IP2 RVI component materials are consistent or nearlyequivalent with those materials identified in MRP-191, Table 4-4 forWestinghouse-designed plants, except for a few components that werefabricated from CF8 cast austenitic stainless steel (CASS) material ratherthan the Type 304 stainless steel (SS) called out in MRP-191.

Theseitems along with the items that were fabricated from different butessentially equivalent materials are summarized in a Westinghouse proprietary document

[Entergy reference 1]. Plant-specific details areproprietary and not typically released as part of RAI responses.

If theNRC requires additional

details, the calculation would be made available for the NRC to review the supporting technical basis. This typically occursat either a plant location or at the Westinghouse Rockville office withWestinghouse personnel supporting the discussion.

c. The design and fabrication of IP2 RVI components are the same as, orequivalent to, the typical Westinghouse-designed PWR RVI components.

4. Modifications to the IP2 RVI made over the lifetime of the plant are thoseidentified in general industry guidance or specifically directed by Westinghouse.

IP2 performed a SPU in 2004 where analyses were performed on RVIcomponents and it was determined that the structural integrity of the reactorinternals was maintained at the SPU conditions.

The design has been maintained NL-12-134 Attachment 1Page 5 of 13over the lifetime of the plant as specified by Westinghouse, and IP2 has notmade any modifications since May 2007, which meets requirements from MRP-227-A. Operational parameters with regard to fluence and temperature arecompliant with requirements from MRP-227-A, and the components andmaterials are the same as those considered in MRP-1 91. Therefore, the IP2stress values are reasonably represented by the assumptions in MRP-1 91, MPR-227- A, and MRP-232.3) Per the response to item 2.b of this RAI, IP2 complied with Applicant/Licensee ActionItem 1 of the NRC Safety Evaluation (SE) regarding MRP-227, Revision 0, except for theCF8 components that were called out as 304 SS in MRP-1 91. Those materialdifferences were evaluated and determined, as summarized in (Entergy reference 1],to require no modifications to the generic MRP-227-A inspection strategy.

Therefore, the requirement is met for application of MRP-227-A as a strategy for managing age-related material degradation in the IP2 components, and no additional agingmechanisms are applicable to components at IP2.4) See response to item 3 of this RAI. A FMECA was done addressing the CF8components that were called out as 304 SS in MRP-191 and determined that thecomponents could be classified as "No Additional Measures" components based on aconsideration of the likelihood of failure and the likelihood of damage. Documentation forthis FMECA is summarized in [Entergy reference 1].5) No design changes were made to IP2 that were beyond those identified in generalindustry guidance or as were recommended by the original vendors.

There were nodesign changes implemented after May 2007 [Entergy reference 5].Indian Point Unit 31) To provide reasonable assurance that the RVI components are bounded byassumptions in the generic industry FMECA, MRP-191, and functionality

analyses, MRP-232, supporting the development of MRP-227-A, a response to request 2.b isprovided.

2.a) Based on the response in item 1 above, no response is necessary.

2.b) The process used to verify that Indian Point Unit 3 (IP3) is reasonably represented bythe generic industry program assumptions with regard to neutron fluence, temperature, materials, and stress values used in the development of MRP-227-A is as follows:1. Identification of typical Westinghouse PWR internal components MRP-191, Table4-4.2. Identification of IP3 PWR internal components.

3. Comparison of the typical Westinghouse PWR internal components to the IP3PWR internal components.

a. Confirm that no additional items were identified by this comparison.

NL-12-134 Attachment 1Page 6 of 13b. Confirm that the materials identified for IP3 are consistent with thosematerials identified in MRP-191, Table 4-4.c. Confirm that the IP3 internals are the same as, or equivalent to, thetypical Westinghouse PWR internals regarding design and fabrication.

4. Confirmation that the IP3 operating history is consistent with the assumptions inMRP-227-A regarding core loading patterns.

5. Confirmation that the IP3 RVI materials operated at temperatures within theoriginal design basis parameters.

6. Determination of stress values based on design basis documents.

7. Confirmation that any changes to the IP3 RVI components do not impact theapplication of the MRP-227-A generic aging management strategy.

The IP3 RVI components are reasonably represented by assumptions regarding neutronfluence, temperature, materials, and stress values in the generic industry FMECAand functionality analyses based on the following:

1. IP3 operating history is consistent with the assumptions used to develop theMRP-227-A with regard to neutron fluence.a. The FMECA and functionality analyses for MRP-227-A were based onthe assumption of 30 years of operation with high-leakage core loadingpatterns followed by 30 years of low-leakage core fuel management strategy.

IP3 started to go from high-leakage to low-leakage loadingpattern in Fuel Cycle 4. In Fuel Cycle 7 (6/24/84) at 9 years of operation, IP3 switched to use of a low-low-leakage loading pattern (L3p). Then, inFuel Cycle 14 (4/7/05) at 30 years of operation, IP3 switched to use of alow-low-low-leakage loading pattern (L4p). Therefore, IP3 meets thefluence and fuel management assumptions in MRP-191, MRP-232,and the requirements for application of the MRP-227-A agingmanagement strategy.

b. IP3 has operated under base load conditions over the life of the plant.Therefore, IP3 satisfies the assumptions in MRP-227-A and supporting documentation regarding operational parameters affecting fluence.2. The IP3 RVI operate between Thot and T,0od, which are not less thanapproximately 517°F (541.90F prior to SPU) for Tood and not higher than 603'F(600.80F prior to SPU) for Thot. The design temperature for the vessel is6500F. Therefore, IP3 historical operation is within original design basisparameters and is consistent with the assumptions used to develop the agingmanagement strategy in MRP-227-A with regard to temperature operational parameters.

3. IP3 RVI components and materials are covered by the list of genericWestinghouse-designed PWR RVI components in MRP-191, Table 4-4.Components and materials are summarized in [Entergy reference 1].

NL-12-134 Attachment 1Page 7 of 13a. As summarized in [Entergy reference 1], no additional components wereidentified for IP3 by this comparison.

b. Most of the IP3 RVI component materials are consistent or nearlyequivalent with those materials identified in MRP-191, Table 4-4 forWestinghouse-designed plants, except for a few components that werefabricated from CF8 CASS material rather than the Type 304 SS calledout in MRP-1 91. These items along with the items that were fabricated from different but essentially equivalent materials are summarized in aWestinghouse proprietary document

[Entergy reference 1]. Plant-specific details are proprietary and not typically released as part of RAIresponses.

If the NRC requires additional

details, the calculation wouldbe made available for the NRC to review the supporting technical basis.This typically occurs at either a plant location or at the WestinghouseRockville office with Westinghouse personnel supporting the discussion.

c. The design and fabrication of IP3 RVI components are the same as, orequivalent to, the typical Westinghouse-designed PWR RVI components.

4. Modifications to the IP3 RVI made over the lifetime of the plant are thoseidentified in general industry guidance or specifically directed by Westinghouse.

IP3 performed a SPU in 2005 where analyses were performed on RVIcomponents and it was determined that the structural integrity of the reactorinternals was maintained at the SPU conditions.

The design has been maintained over the lifetime of the plant as specified by Westinghouse, and IP3 has notmade any modifications since May 2007, which meets requirements from MRP-227-A. Operational parameters with regard to fluence and temperature arecompliant with requirements from MRP-227-A, and the components andmaterials are the same as those considered in MRP-1 91. Therefore, the IP3stress values are reasonably represented by the assumptions in MRP-1 91, MPR-227-A, and MRP-232.3) Per the response to item 2.b of this RAI, IP3 complied with Applicant/Licensee ActionItem 1 of the NRC SE regarding MRP-227, Revision 0, except for the CF8 components that were called out as 304 SS in MRP-191.

Those material differences were evaluatedand determined, as summarized in [Entergy reference 1], to require no modifications tothe generic MRP-227-A inspection strategy.

Therefore, the requirement is met forapplication of MRP-227-A as a strategy for managing age-related material degradation in the IP3 components, and no additional aging mechanisms are applicable tocomponents at IP3.4) See response to item 3 of this RAI. A FMECA was done addressing the CF8components that were called out as 304 SS in MRP-1 91 and determined that thecomponents could be classified as "No Additional Measures" components based on aconsideration of the likelihood of failure and the likelihood of damage. Documentation forthis FMECA is summarized in [Entergy reference 1].5) No design changes were made to IP3 that were beyond those identified in generalindustry guidance or as were recommended by the original vendors.

There were nodesign changes implemented after May 2007 [Entergy reference 5].

NL-12-134 Attachment 1Page 8 of 13RAI 7The staff reviewed the applicant's response to Applicant/Licensee Action Item 2 from the NRCstaff's final SE on MRP-227, Revision

0. In Section 3.6 of the RVI Inspection Plan (Ref. 5), theapplicant stated that it reviewed the information in Table 4-4 of MRP-191 and determined thatthis table contains all the RVI components that are within the scope of license renewal and thatthis is shown in Table 5-7. The staff notes that Table 5-1 contains a cross-index between thecomponent designations in Entergy Letter NL-10-063 (Amendment 9 to the LRA, Ref. 1) and thecomponent names as designated in MRP-191, Table 4-4 (Ref. 6). All the IPEC component designations correlate with an equivalent component designation in MRP-191 (Ref. 7), Table 44with the exception of the Lower Internals Assembly

-Column Cap.The staff therefore requests that the applicant verify that the Lower Internals Assembly

-ColumnCap would be subject to the same inspection requirements that are applied to the lower supportassembly, lower support column bodies (cast) in MRP-227-A, Table 4-6. If not, provide plant-specific aging management requirements for the Lower Internals Assembly

-Column Cap.Response to RAI 7Indian Point Unit 2In RAI 7, the NRC notes that all IP2 components within the scope of license renewal comparewith an equivalent component designation in MRP-191 with the exception of the lower internals assembly column cap. The lower internals assembly column cap is a CASS piece welded onto the top of the core support column shaft. These two pieces together constitute the lowerinternals assembly

-column body.

The development of the MRP-227-A aging managementstrategy considered the lower support column as one complete unit denoted as the "lowersupport column assemblies

-lower support column bodies."

Under the lower support columnbodies in MRP-191, both Type 304 SS and CF8 CASS material were considered, assummarized in [Entergy reference 1]. Since the lower internals assembly column cap is part ofthe lower support column body, it was addressed by the generic industry FMECA andfunctionality analysis; therefore, it is subject to the same inspection requirements as the lowersupport column bodies (cast). Since the column cap is a part of the lower support column body,it would be subject to the same inspection requirements as the lower support column body. Theinspection of the lower support column body includes the column cap.Indian Point Unit 3In RAI 7, the NRC notes that all IP3 components within the scope of license renewal comparewith an equivalent component designation in MRP-1 91 with the exception of the lower internals assembly column cap. The lower internals assembly column cap is a CASS piece welded ontothe top of the core support column shaft. These two pieces together constitute the lowerinternals assembly

-column body. The development of the MRP-227-A aging management strategy considered the lower support column as one complete unit denoted as the "lowersupport column assemblies

-lower support column bodies."

Under the lower support columnbodies in MRP-191, both Type 304 SS and CF8 CASS material were considered, assummarized in [Entergy reference 1]. Since the lower internals assembly column cap is part ofthe lower support column body, it was addressed by the generic industry FMECA andfunctionality analysis; therefore, it is subject to the same inspection requirements as the lower NL-12-134 Attachment 1Page 9 of 13support column bodies (cast).

Since the column cap is a part of the lower support column body,it would be subject to the same inspection requirements as the lower support column body. Theinspection of the lower support column body includes the column cap.RAI 9The applicant's response to Applicant/Licensee Action Item 5 from Revision 1 of the staff's finalSE on MRP-227, states in part that the acceptance criteria will ensure the remaining compressible height of the spring shall provide hold down forces within the IPEC designtolerance.

If a plant specific acceptance criterion is not developed for the hold down spring,IPEC will replace the spring in lieu of performing the first required physical measurement.

MRP-227-A, Table 4-3, calls for direct measurement of the hold-down spring height within threecycles of the beginning of the license renewal period. If the first set of measurements is notsufficient to determine life, spring height measurements must be taken during the next twooutages, in order to extrapolate the expected spring height to 60 years.The staff requires clarification of how the applicant will determine whether the first set ofmeasurements could be extrapolated to demonstrate acceptable spring functionality through 60years. Therefore, the staff requests the following information:

1. Provide the specific acceptance criteria for spring height and/or hold down force from theIP2/IP3 licensing basis.2. Describe the procedure by which the remaining hold down forces will be projected toend-of-life based on one measurement.

Address whether the decrease in spring heightor hold-down force is assumed to occur linearly over time or via some other function oftime.3. What results of the first spring measurements would indicate a need for successive measurements?

Response to RAI 9Indian Point Unit 2 and Indian Point Unit 31. The acceptable criteria are based on the measured height of the spring as a function oftime relative to the required hold-down force. Applicable plant loading conditions wereevaluated.

Time-dependent details for the hold down spring (HDS) height measurements are summarized in [Entergy reference 1]. Confirmatory actions are included if applicable after completion of the evaluation.

Plant-specific details are proprietary and not typically released as part of RAI responses.

If the NRC requires additional details, the calculation would be made available for the NRC to review the supporting technical basis. Thistypically occurs at either a plant location or at the Westinghouse Rockville office withWestinghouse personnel supporting the discussion.

2. The decrease in HDS height is assumed to occur linearly over time. The approach usedto develop the HDS height acceptance criteria is to consider the actual HDS height atplant start-up and the required HDS heights at the end of 60 years. A linear interpolationat the time of the HDS height measurement determines the required minimum HDSheight.

NL-12-134 Attachment 1Page 10 of 133. The HDS height measures less than the required minimum HDS heights, assummarized in [Entergy reference 1], indicating a need for re-evaluation and successive measurements or a replacement HDS.RAI 10The applicant's response to Applicant/Licensee Action Item 7 indicates that the plant-specific analysis to demonstrate functionality of the lower support column bodies during the period ofextended operation will be submitted to the NRC prior to the PEO. In the aging management review tables submitted in LRA Amendment 9, the applicant credits the "Thermal Aging andNeutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS) Program" formanaging loss of fracture toughness of the lower core support column bodies, as well asseveral other CASS components.

NUREG-1 930 indicates that the staff determined this programwas consistent with the Generic Aging Lessons Learned Report, Revision 1, AgingManagement Program (AMP) XI.M13, "Thermal Aging and Neutron Irradiation Embrittlement ofCast Austenitic Stainless Steel (CASS) Program."

Per GALL, Rev. 1,Section XI.M13, the"Thermal Aging and Neutron Irradiation Embrittlement of CASS Program" generally requiressupplemental visual inspections (equivalent to an EVT-1) for CASS RVI components that areeither susceptible to thermal aging based on chemistry and other manufacturing parameters, orreceive a neutron fluence > lx1017 n/cm2, unless it can be demonstrated that the stresses onthe component are either compressive or low in magnitude if tensile.

The RVI Program iscredited with managing cracking of the core support column bodies and other CASScomponents.

Under the RVI Program, the core support column bodies are expansion components that would be subject to an EVT-1 visual examination for cracking due to irradiation assisted stress corrosion cracking if cracking were found in the associated primary component.

The staff requests the following information:

Since both the plant-specific analysis and Thermal Aging and Neutron Irradiation Embrittlement of CASS Program could both potentially involve screening for thermal or neutron irradiation embrittlement, stress analyses, and flaw tolerance evaluations, and both the RVI Program andThermal Aging and Neutron Irradiation Embrittlement of CASS Program could potentially requireinspections, discuss the relationship of the two programs and the plant-specific analysis.

Response to RAI 10Indian Point Unit 2 and Indian Point Unit 3The RVI Program is the base program that addresses the RVI components that require agingmanagement and specifies inspection requirements.

The Thermal Aging and Neutron Irradiation Embrittlement of CASS Program at Indian Point (IPEC LRA) augments the RVI program byevaluating the potential susceptibility of plant-specific CASS components based on component specific chemistry.

RAI 11In response to Applicant/Licensee Action Item 7, the applicant stated that the plant-specific analyses to demonstrate the lower support column bodies will maintain their functionality duringthe period of extended operation will consider the possible loss of fracture toughness in thesecomponents due to thermal and irradiation embrittlement.

The analyses will be consistent with NL-12-134 Attachment 1Page 11 of 13the IP2/1 P3 licensing basis and the need to maintain the functionality of the lower supportcolumn bodies under all licensing basis conditions of operations.

The staff requests the following additional information:

1) Section 3.3.7 of Revision 1 of the staff's final SE on MRP-227, Revision 0 lists threepossible options for the type of plant-specific analysis used to fulfill the requirements ofthis action item. The three approaches are 1) functionality analyses of the set of likecomponents,

2) component-specific flaw tolerance evaluations, or 3) a screening approach demonstrating that the CASS Components are not susceptible to thermalembrittlement, neutron embrittlement, or the combined effects of both. Discuss which ofthese approaches will be used and why.2) Describe the acceptance criteria for the plant-specific analysis results that are derivedfrom the IP2/IP3 licensing basis.3) Since the applicant stated that the analysis of the core support columns will be submitted prior to the period of extended operation for IP2 and IP3, the staff requests the applicant submit a letter documenting this as a formal licensing commitment.

Response to RAI 11Indian Point Unit 2 and Indian Point Unit 31) The approach used to evaluate the susceptibility of the CASS components at IP2 andIP3 to thermal and neutron embrittlement is a screening approach.

A screening approach was used as this is consistent with the MRP-227-A development methodology and a methodology and screening values have been endorsed by the NRC [Entergyreference 3] for application with CASS materials.

2) The acceptance criteria for the plant-specific CASS component susceptibility evaluations based on fabrication and chemistry, taken from [Entergy references 2 and 3], aresummarized in the following table.Molybdenum Casting NRC Susceptibility (wt %) Method Delta-Ferrite

% Evaluation High 2.0-3.0 Static >14% Potentially susceptible to TE<14% Not susceptible to TECentrifugal

>20% Potentially susceptible to TE<20% Not susceptible to TELow 0.5 max Static >20% Potentially susceptible to TE<20% Not susceptible to TECentrifugal all Not susceptible to TE NL-12-134 Attachment 1Page 12 of 133) Commitment 47 in [Entergy reference 4] states that IPEC will perform and submitanalyses that demonstrate that the lower support column bodies will maintain theirfunctionality during the period of extended operation.

These analyses will consider thepossible loss of fracture toughness due to thermal and irradiation embrittlement.

Theanalyses will be consistent with the IP2/IP3 licensing basis.NRC References

1. Letter from Fred Dacimo, Entergy, to NRC dated July 14, 2010,

Subject:

Amendment 9 toLicense Renewal Application (LRA) -Reactor Vessel Internals Program Indian Point NuclearGenerating Unit Nos. 2 & 3, Docket Nos.

50-247 and 50-286 License Nos. DPR-26 andDPR-64 (ADAMS Accession No. ML102010102)

2. Letter from Robert Nelson, NRC, to Neil Wilmshurst, EPRI dated December 16, 2011;

Subject:

Revision 1 of the Final Safety Evaluation of EPRI Report, Materials Reliability Program Report 1016596 (MRP-227),

Revision 0, "Pressurized Water Reactor (PWR)Internals Inspection and Evaluation Guidelines" (TAC No. ME0680) (ADAMS Accession No.ML11308A770)

3. Reactor Internals Acceptance Criteria Methodology and Data Requirements, WCAP-17096-NP, Rev. 2, Westinghouse Non-Proprietary Class 3 Report, December 2009, ADAMSAccession No. ML1014601570

4. Materials Reliability Program:

Inspection Standard for PWR Internals (MRP-228) 1016609Final Report, July 2009 Electric Power Research Institute, Palo Alto, CA (EPRI Product No.1016609)

(ADAMS Accession No. ML092120573)

5. Indian Point Energy Center Revised Reactor Vessel Internals Inspection Plan Compliant with MRP-227-A.

Attachment 2 to Entergy Letter NL-12-037, Letter from Fred Dacimo toNRC dated February 17, 2012,

Subject:

License Renewal Application

-Revised ReactorVessel Internals Program and Inspection Plan Compliant with MRP-227-A, Indian PointNuclear Generating Unit Nos. 2 and 3, Docket Nos. 50-247 and 50-286-License Nos. DPR-26 and DPR-64 (ADAMS Accession No. ML1206A312)

6. MRP-191 Revision 0, "Materials Reliability Program:

Screening, Categorization and Rankingof Reactor Internals of Westinghouse and Combustion Engineering PWR Designs,"

ADAMSAccession No. ML091910130

7. NUREG-1930, Volume 2, "Safety Evaluation Report Related to The License Renewal ofIndian Point Nuclear Generating Unit Nos. 2 and 3, Dockets No. 50-247 and 50-286,November 30, 2009 (ADAMS Accession No. ML093170671)

Enterqy References

1. Westinghouse Letter, LTR-RIAM-12-112, Rev. 0, "Indian Point Units 2 and 3 Support for ReactorVessel Internals Aging Management Request for Additional Information (RAI) Project -Request forTransmittal of Final Comment Resolutions on RAI Responses,"

September 25, 2012.

NL-12-134 Attachment 1Page 13 of 132. U.S. Nuclear Regulatory Commission, NUREG/CR-4513, Rev. 1, "Estimation of FractureToughness of Cast Stainless Steels During Thermal Aging in LWR Systems,"

May 1994 (NRCADAMS Accession No. ML052360554).

3. U.S. Nuclear Regulatory Commission Letter, "License Renewal Issue No. 98-0030, Thermal AgingEmbrittlement of Cast Austenitic Stainless Steel Components,"

May 19, 2000 (NRC ADAMSAccession No. ML003717179).

4. Entergy Letter, NL-12-089, "Reply to Request for Additional Information Regarding the LicenseRenewal Application Indian Point Nuclear Generating Unit Nos. 2 & 3 Docket Nos. 50-247 and 50-286 License Nos. DPR-26 and DPR-64,"

June 14, 2012.5. Entergy Letter, NL-10-063, "Amendment 9 to License Renewal Application (LRA) -Reactor VesselInternals Program Indian Point Nuclear Generating Unit Nos. 2 & 3 Docket Nos. 50-247 and 50-286License Nos. DPR-26 and DPR-64,"

July 14, 2010.