AEP-NRC-2014-60, Final Response to Request for Additional Information Concerning the Reactor Vessel Internals Aging Management Program

Final Response to Request for Additional Information Concerning the Reactor Vessel Internals Aging Management Program

ML14316A449
Person / Time
Site:	Cook
Issue date:	10/22/2014
From:	Gebbie J Indiana Michigan Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML14316A453	List: AEP-NRC-2014-60, Final Response to Request for Additional Information Concerning the Reactor Vessel Internals Aging Management Program
References
AEP-NRC-2014-60
Download: ML14316A449 (50)

Text

INDIANA MICHIGAN Ihdiana Michigan Power Cook Nuclear Plant POWER ° One Cook Piece Widgman,Ml 49106 A unit ofAmerican Electric Power IndianaMichiganPowercom October 22, 2014 AEP-NRC-2014-60 10 CFR 50.4 Docket Nos.: 50-315 50-316 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Donald C. Cook Nuclear Plant Units I and 2 FINAL RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION CONCERNING THE REACTOR VESSEL INTERNALS AGING MANAGEMENT PROGRAM

References:

1) Letter from J. P. Gebbie, Indiana Michigan Power Company (I&M), to U.S. Nuclear Regulatory Commission (NRC), "Donald C. Cook Nuclear Plant Units 1 and 2, Transmittal of Reactor Vessel Internals Aging Management Program," dated October 1, 2012, Agencywide Documents Access and Management System (ADAMS) Accession No. ML12284A320.

2) Letter from T. J. Wengert, NRC, to L. J. Weber, I&M, "Donald C. Cook Nuclear Plant, Units 1 and 2 - Request for Additional Information Concerning the Reactor Vessel Internals Aging Management Program Submittal (TAC Nos. MF0050 and MF0051),"

dated June 6, 2014, ADAMS Accession No. ML14135A320.

3) Letter from J. P. Gebbie, I&M, to NRC, "Donald C. Cook Nuclear Plant Units 1 and 2

- First Response to Request for Additional Information Concerning the Reactor Vessel Internals Aging Management Program," dated July 30, 2014, ADAMS Accession No. ML14216A497.

4) Letter from J. P. Gebbie, I&M, to NRC, "Donald C. Cook Nuclear Plant Units 1 and 2

- Second Response to Request for Additional Information Concerning the Reactor Vessel Internals Aging Management Program, dated September 4, 2014, ADAMS Accession No. ML14253A316.

This letter provides Indiana Michigan Power Company's (I&M), licensee for Donald C. Cook Nuclear Plant (CNP) Units 1 and 2, final response to Requests for Additional Information (RAI) by the U. S. Nuclear Regulatory Commission (NRC) regarding CNP's Reactor Vessel Internals (RVI) Aging Management Program (AMP).

PROPRIETARY INFORMATION Enclosure 4 to this Letter contains proprietary information.

Withhold from public disclosure under 10 CFR 2.390.

Upon removal of Enclosure 4, this Letter is decontrolled. AOQ(

U.S. Nuclear Regulatory Commission AEP-NRC-2014-60 Page 2 By Reference 1, I&M submitted the CNP RVI AMP. By Reference 2, the NRC transmitted RAIs regarding the program. References 3 and 4 provided I&M's response to Reference 2, RAI-1, RAI-5, RAI-7, and RAI-8. By E-Mail dated October 6, 2014, I&M requested a due date extension from October 10, 2014, to October 25, 2014, for the final response to Reference 2. By E-mail dated October 7, 2014, the NRC approved the due date extension to October 25, 2014. to this letter provides response to Reference 2, RAI-2, RAI-3, RAI-4, and RAI-6. contains additional information regarding the detailed component review and cold work assessments performed for CNP Units 1 and 2. Enclosures 3, 4, and 5 contain additional information regarding the fuel design and fuel management assessments performed for CNP Units 1 and 2. contains additional information regarding results of cast austenitic stainless steel evaluations for CNP Units 1 and 2. Enclosure 7 provides a status of the commitments made by Reference 1 to this letter. Enclosure 8 contains a new regulatory commitment.

Should you have any questions, please contact Mr. Michael K. Scarpello, Regulatory Affairs Manager, at (269) 466-2649.

Sincerely, Joel P. Gebbie Site Vice President DMB/amp

Enclosures:

1. Donald C. Cook Nuclear Plant Response to Request for Additional Information Regarding the Reactor Vessel Internals Aging Management Program

2. Appendix A from PWROG-14044-P, Revision 0, "D.C. Cook Units 1 and 2 Summary Report for the Cold Work Assessment in Support of RAIs Regarding A/LAI 1 for MRP-227"

[Non-Proprietary]

3. PWROG-14049-NP, Revision 0, "D. C. Cook Units 1 and 2 Summary Report for the Fuel Design / Fuel Management Assessments" [Non-Proprietary]

4. PWROG-14049-P, Revision 0, "D. C. Cook Units 1 and 2 Summary Report for the Fuel Design / Fuel Management Assessments" [Proprietary]

5. CAW-14-4041, "Application for Withholding Proprietary Information from Public Disclosure"

6. LTR-RIAM-1 4-24, Revision 1, "Reports for D.C. Cook Units 1 and 2 for PWROG PA-MSC-0983 Cafeteria Tasks 3, 4, and 5 Deliverables" [Non-Proprietary]

7. Status of Regulatory Commitments

8. Regulatory Commitment PROPRIETARY INFORMATION Enclosure 4 to this Letter contains proprietary information.

Withhold from public disclosure under 10 CFR 2.390.

Upon removal of Enclosure 4, this Letter is decontrolled.

U.S. Nuclear Regulatory Commission AEP-NRC-2014-60 Page 3 c: M. L. Chawla, NRC Washington, D.C.

J. T. King - MPSC MDEQ- RMD/RPS NRC Resident Inspector C. D. Pederson, NRC Region III A. J. Williamson - AEP Ft. Wayne PROPRIETARY INFORMATION Enclosure 4 to this Letter contains proprietary information.

Withhold from public disclosure under 10 CFR 2.390.

Upon removal of Enclosure 4, this Letter is decontrolled.

ENCLOSURE 1 TO AEP-NRC-2014-60 DONALD C. COOK NUCLEAR PLANT REACTOR VESSEL INTERNALS AGING MANAGEMENT PROGRAM RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING THE REACTOR VESSEL INTERNALS AGING MANAGEMENT PROGRAM List of Acronyms:

A/LAI Applicant / Licensee Action Item from the Safety Evaluation for MRPO-227 ADAMS Agencywide Documents Access and Management System AMP Aging Management Program ASME American Society of Mechanical Engineers CASS Cast Austenitic Stainless Steel CNP Donald C. Cook Nuclear Power Plant EPRI Electric Power Research Institute I&M Indiana Michigan Power LRSS Lower Radial Support System LSC Lower Support Column MRP Materials Reliability Program MSC Materials Subcommittee NRC U. S. Nuclear Regulatory Commission PA Project Authorization PWROG Pressurized Water Reactor Owners Group RAI Request for Additional Information RVI Reactor Vessel Internals By letter dated October 1, 2012 (ADAMS Accession No. ML12284A320), I&M, the licensee for CNP, submitted an AMP for CNP, Units 1 and 2, RVI to the NRC. By letter dated June 6, 2014 (ADAMS Accession No. ML14135A320), the NRC staff reviewed the submittal and requested additional information to complete its review. By letter dated July 30, 2014, the responses to RAI-1, RAI-5, and RAI-7 were provided to the NRC (ADAMS Accession No. ML14216A497). By letter dated September 4, 2014, the response to RAI-8 was provided to the NRC (ADAMS Accession Nos.

ML14253A316, ML14253A317, and ML14253A318). The response to RAI-2, RAI-3, RAI-4, and RAI-6 is provided in this enclosure.

RAI-2

In Section 4.4.2.1 of the submittal, I&M states that a response to Action Item 1 will be submitted priorto the period of extended operationfor each unit. Justify that the Pressurized WaterReactor Owners Group (PWROG)-providedresults of project PA-MSC-0938 are fully bounding for CNP or identify any gaps between these results and those for CNP. The response to RAI Question 3 should be incorporatedinto this response.

Response to RAI-2 I&M has assessed the design and operating history of CNP Units 1 and 2 to demonstrate that MRP-227-A is applicable to both units. The evaluations considered the assumptions regarding plant design and operating history made in the failure mode, effects, and criticality analysis and functionality analysis for Westinghouse reactors which support MRP-227-A.

to AEP-NRC-2014-60 Page 2 No changes are required to the reactor internals aging management strategy for CNP Units 1 or 2 as a result of these evaluations.

Details regarding the design evaluations for CNP Units 1 and 2 can be found in the response to RAI-3(a), RAI-4, and RAI-6. Details regarding the operating history evaluations for CNP Units 1 and 2 can be found in the response to RAI-3(b) and RAI-6.

RAI-3

As discussedin References I and 2, provide the following informationrelated to verification of the applicabilityof MRP-227-A to CNP, Units 1 and 2:

(a) Do the CNP, Units I and 2 RVI have non-weld or bolting austeniticstainless steel components with 20 percent cold work or greater,and subject to operating stresses greaterthan 30 kilopounds per square inch? If so, perform a plant specific evaluation to determine the aging management requirementsfor the affected components.

(b) Have CNP, Units I and 2 ever utilized a typical design or fuel management that could make the assumptions of MRP-227-A regarding core loading/core design non-representativefor the plant,includingpower changes/uprates? If so, describe how the differences were reconciled with the assumptions of MRP-227-A, or provide a plant-specific AMP for the affected components as appropriate.

Response to RAI-3 Evaluations were performed through the PWROG program PA-MSC-0983, "Support for Applicant Action Items 1, 2, and 7 from the Final Safety Evaluation on MRP-227, Revision 0," using guidance provided by the EPRI Materials Reliability Program.

(a) I&M has evaluated CNP Units 1 and 2 reactor internals components according to the MRP-191 industry generic component listings and screening criteria, including consideration of cold work as defined in MRP-175, noting the requirements of MRP-175 Section 3.2.3. In addition to consideration of the material fabrication, forming, and finishing process, a general screening definition of a resulting reduction in wall thickness of 20 percent (%) was applied as an evaluation limit. It was confirmed that all of CNP Units 1 and 2 components, as applicable for the design, are included directly in Table 4-4 of MRP-191, except for the components identified in Table 3-1 and 3-2 of this enclosure, respectively.

Differences in material type from wrought austenitic stainless steel to CASS, or vice versa, were resolved by expert panel elicitation. Consensus was used to resolve differences in material type consistent with the process outlined in MRP-191. The remaining material differences were within the wrought austenitic stainless steel family, for which the MRP-1 91 Section 3 screening criteria was the same. These evaluations do not affect any component inspection categories.

No changes are required to the reactor internals aging management strategy for CNP Units 1 or 2 as a result of these evaluations.

Additional information regarding the detailed component review and cold work assessments performed for CNP Units 1 and 2 can be found in the following enclosure to this letter:

Enclosure 1 to AEP-NRC-2014-60 Page 3 Enclosure 2: Appendix A from PWROG-14044-P, Revision 0, "D.C. Cook Units 1 and 2 Summary Report for the Cold Work Assessment in Support of RAIs Regarding A/LAI 1 for MRP-227" [Non-Proprietary].

Table 3-1: CNP Unit I Deviations from Generic MRP-191 Comrponent List Cold worked Assembly Subassembly Component Material Category (CW) 20%

Assessment Upper Control Rod Guide plates/cards 1'2 CF8 1 No Internals Assemblies and Flow A351 Assembly Downcomers Housing plates1 '" CF8 1 No A351 Mixing Devices Mixing devices" 304SS 3 No A240 Upper Core Plate and Fuel alignment pins 304 SS 4 No Fuel Alignment Pin A276 Upper Bolting 304 SS Instrumentation A213 4 No Conduit and Supports A276 4 No Brackets, clamps, terminal CF8 No blocks, and conduit straps 1 A351 Locking caps 304L SS A276 4 No Upper Support Plate Lock keys 304L SS 4 No Assembly A276 Lower Bottom-Mounted BMI column cruciform 304 SS Internals Instrumentation (BMI) A276 3 No Assembly Column Assemblies A240 3 No Lower Core Plate and Fuel alignment pins 304 SS 4 No Fuel Alignment Pins A276 Lower Support Lower support column bolts 316 SS 4 No Column Assemblies 70041 EA (AIS1316)

Radial Support Keys Radial support key bolts 316 SS 4 No 70041 EA YS4 (AISI316)

Interfacing Interfacing Clevis insert lock keys 304L 4 Yes Components Components SA-479 (The Notes and Category Key apply to Table 3-1 and Table 3-2)

Notes:

1 The plant-specific component MRP-227-A category was confirmed by an expert panel as remaining consistent with the generic MRP-232 classification.

2 Drawing lists CF8 as an alternate material 3 This component is a mounting ring for the mixing device 4 This component is a stainless steel crimp cup locking device on replacement LRSS clevis insert bolts.

5 The drawing lists 304 SS vanes / pads that are welded to the BMI column bodies which serve the same purpose as a BMI column cruciform.

Category Key:

1 CASS 2 Hot formed austenitic stainless steel 3 Annealed austenitic stainless steel 4 Fasteners austenitic stainless steel 5 Cold formed austenitic stainless steel (without subsequent solution annealing)

Enclosure 1 to AEP-NRC-2014-60 Page 4 Table 3-2: CNP Unit 2 Deviations from Generic MRP-191 Component List CW 20%

Assembly Subassembly Component Material Category Assessment Upper Upper Core Plate and Fuel alignment pins 304 SS 4 No Internals Fuel Alignment Pin A276 Assembly Upper Bolting 304 SS Instrumentation A276 4 No Conduit and Supports A213 4 No Brackets, clamps, terminal CF8 blocks, and conduit straps 1 A351 1 No Locking caps 304L SS 4 No A276 Upper Support Plate Flange' CF8 1 No Assembly A351 Lock keys 304 SS 4 No A276 Upper support plate1 CF8 1 No A351 Upper support ring or skirt1 CF8 1 No A351 Lower BMI Column BMI column cruciforms" = 304 SS Internals Assemblies A276 3 No Assembly A240 3 No Lower Core Plate and Fuel alignment pins 304 SS 4 No Fuel Alignment Pins A276 Lower Support Lower support column bolts 316 SS 4 No Column Assemblies 70041 EA (AISI 316)

Neutron Panels / Thermal shield dowels 304 SS Thermal Shield A276 4 No AIS1302 or 4 Yes AISI 304 308L 1 No AISI 302 or 4 Yes AISI 304 Radial Support Keys Radial support key bolts 316 SS 4 No 70041 EA (AISI 316)

(b) I&M has evaluated the CNP Units 1 and 2 fuel design and fuel management and compared them with the assumptions of MRP-227-A. Neutron fluence and heat generation rates are concluded to be acceptable based on an assessment comparing plant specific values to the limiting MRP threshold values and guidance. However, a projection of future operation for CNP Unit 1 with the current fuel management strategy shows that the MRP guidelines will be

.exceeded in the future.

to AEP-NRC-2014-60 Page 5 The MRP guidelines indicate that the heat generation rate figure of merit (HGR-FOM) may not exceed 68 W/cm 3 after the first 30 years of operation. Short periods of operation outside guidance limits of less than two years do not invalidate applicability of assumptions made in development of MRP-227-A for a plant. The MRP guidance indicates that plants which operate outside the limits for more than two years of operation need to perform further evaluations to demonstrate that fuel design and fuel management are consistent with the assumptions of MRP-227-A.

CNP Units 1 and 2 both converted from an "out-in" fuel management strategy to a "low-leakage" fuel management strategy in the first 30 years of operation. The Unit 1 HGR-FOM exceeded 68W/cm 3 during the U1C25 fuel cycle which started in the3 spring of 2013 and completed in the fall of 2014. The HGR-FOM will exceed 68 W/cm during the U1C26 fuel cycle scheduled to start in the fall of 2014 and complete in the spring of 2016. Therefore, CNP Unit 1 will exceed 2 years of operation outside the HGR-FOM guidance provided by the EPRI MRP during U1C26. No other guidelines have exceeded, or are projected to exceed, MRP guidelines for CNP Units 1 or 2.

No changes are required to the reactor internals aging management strategy for CNP Unit 1 as a result of these evaluations; however, I&M continues to evaluate the interaction of the current fuel management strategy and the reactor internals aging management strategy. No changes are required to the reactor internals aging management strategy for CNP Unit 2 as a result of these evaluations.

Additional information regarding the fuel design and fuel management assessments performed for CNP Units 1 and 2 can be found in the following enclosures to this letter:

Enclosure 3: PWROG-14049-NP, Revision 0, "D. C. Cook Units 1 and 2 Summary Report for the Fuel Design / Fuel Management Assessments" [Non-Proprietary];

Enclosure 4: PWROG-1 4049-P, Revision 0, "D. C. Cook Units 1 and 2 Summary Report for the Fuel Design I Fuel Management Assessments" [Proprietary]; and Enclosure 5: CAW-14-4041, "Application for Withholding Proprietary Information from Public Disclosure."

RAI-4

In Section 4.4.2.2 of the submittal, I&M states that Action Item 2 will be satisfied through action taken by the PWROG. It is unclearfrom the licensee's response if it was understood, that this Action Item requiresan applicant/licenseeto evaluate whether any plant-specific components that should be included in the SNP AMP were overlooked in MRP-227-A. The NRC staff requests that I&M perform the plant-specific actions described by Action Item 2 and provide the results.

Response to RAI-4 I&M has compared the plant specific RVI components at CNP Units 1 and 2 against those listed in MRP-191, Table 4-4. All plant specific components were found for each plant in MRP-191, but some material differences exist. A list of all differences between plant specific components and components identified in Table 4-4 of MRP-191 are in Tables 3-1 and 3-2 of this enclosure, for CNP to AEP-NRC-2014-60 Page 6 Units 1 and 2, respectively. The RAI-3(a) response text describes how these material differences are reconciled.

No changes are required to the reactor internals aging management strategy for CNP Units 1 or 2 as a result of these evaluations.

RAI-6

In Section 4.4.2.7 of its submittal, I&M states that responses to Action Item 7 will be submitted prior to the period of extended operation for each unit.

(a) The NRC staff requests that the licensee identify all components for which analyses will be provided underAction Item 7.

The cast austenitic stainlesssteel (CASS) lower supportcolumn bodies are prone to thermal embrittlement, neutron embrittlement, and irradiation-assistedstress corrosion cracking.

Since CASS materials with delta ferrite content greaterthan 20 percent are susceptible to thermal embrittlement, establishing the delta ferrite content in each column is essential to assess the extent of aging degradationdue to thermal embrittlement in these columns. In addition, the casting method (i.e., static or centrifugal)will affect the occurrenceof thermal embrittlement. The value of the delta ferrite content can be obtainedfrom a certifiedmaterial test report (CMTR) for each column.

(b) The NRC staff requests that the licensee provide the delta ferrite content for each lower support column from CMTRs, and provide the casting method for each column.

Response to RAI-6 I&M has reviewed and evaluated the RVI components and their materials for CNP Units 1 and 2 in accordance with A/LAI 7 from the Safety Evaluation for MRP-227.

(a) The CNP Unit 1 and 2 RVI CASS components have been reviewed and the evaluations of their susceptibility to thermal and irradiation embrittlement have been completed. A number of components in both units are found to be potentially susceptible to thermal and/or irradiation embrittlement. However, all potentially susceptible components have either been considered in MRP-1 91, or in a plant specific expert panel, as being potentially susceptible.

No martensitic stainless steel or martensitic precipitation-hardened stainless steel materials were identified at CNP Units 1 or 2.

No further analysis will be provided for CASS components under A/LAI 7, with the exception of lower support columns as described in part (b).

Therefore, lower support columns at CNP Units 1 and 2 are the only components for which further analysis is pursued under A/LAI 7.

Additional information regarding results of CASS evaluations for CNP Units 1 and 2 can be found in the following enclosure to this letter:

Enclosure 6: LTR-RIAM-14-24, Revision 1, "Reports for D.C. Cook Units 1 and 2 for to AEP-NRC-2014-60 Page 7 PWROG PA-MSC-0983 Cafeteria Tasks 3, 4, and 5 Deliverables" [Non-Proprietary].

(b) The CNP Unit 1 and 2 LSC casting method and ferrite content are listed in Table 3.1-1 and 4.1-1 in Enclosure 6 to this letter. All LSCs in both units were manufactured with static casting methods. No CNP Unit 1 LSC Certified Material Test Reports (CMTR) or specific construction records were located; therefore, the delta ferrite content cannot be calculated.

The CNP Unit 1 LSCs are conservatively assumed to have delta ferrite content greater than 20%, and they are considered potentially susceptible to thermal embrittlement. CNP Unit 2 LSC CMTRs were located and used to calculate a delta ferrite content less than or equal to 20%. Since they are less than 20%, they are considered not susceptible to thermal embrittlement.

I&M is pursuing demonstration of LSC functionality in the absence of component specific CMTRs. There is no established methodology for assessing the impact of non-functional LSCs at this time. I&M is participating in a PWROG project to develop a LSC functionality methodology to demonstrate sufficient redundancy within the lower support structure to tolerate a number of LSC failures. Once the methodology is developed, then it may be applied to representative plants, to specific plants, or otherwise used as required to satisfy NRC Staff questions and concerns.

I&M will continue to participate in the PWROG project for LSCs. I&M will provide a supplemental response to the NRC on this RAI when an acceptable methodology is developed by the PWROG project as discussed above. Resolution is anticipated by December 31, 2014. If resolution is not achieved by the end of 2014, I&M will update the supplemental response date as appropriate.

ENCLOSURE 2 TO AEP-NRC-2014-60 Appendix A from PWROG-14044-P, Revision 0, "D.C. Cook Units 1 and 2 Summary Report for the Cold Work Assessment in Support of RAIs Regarding A/LAI 1 for MRP-227"

[Non-Proprietary]

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-1 APPENDIX A

©2014 Westinghouse Electric Company LLC All Rights Reserved PWROG-14044-P Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-2 The scope of this effort was authorized by American Electric Power through PA-MSC-0983, Revision 1, Task 6.

SE Applicant/Licensee Action Item 1: Applicability of FMECA and Functionality Analysis Assumptions "As addressedin Section 3.2.5.1 of this SE, each applicant/licensee is responsible for assessing its plant's design and operating history and demonstrating that the approved version of MRP-227 is applicable to the facility. Each applicant/licensee shall refer, in particular,to the assumptions regardingplant design and operating history made in the FMECA and functionality analyses for reactors of their design (i.e., Westinghouse, CE, or B&W) which support MRP-227 and describe the process used for determining plant-specific differences in the design of their RVI components or plant operating conditions, which result in different component inspection categories. The applicant/licensee shall submit this evaluation for NRC review and approval as part of its application to implement the approved version of MRP-227. This is Applicant/Licensee Action Item 1"

[5].

In the Request for Additional Information (RAI) 3 Part A [4], the U. S. Nuclear Regulatory Commission (NRC) requested the following:

RAI 3(a)

"Do the CNP, Units I and 2 RVI have non-weld or bolting austenitic stainless steel components with 20 percent cold work or greater, and subject to operating stresses greaterthan 30 kilopounds per square inch? If so, perform a plant-specific evaluation to determine the aging management requirementsfor the affected components."

The evaluation included a review of all plant modifications affecting reactor internals and the plant operating history recorded in Tasks 3, 4, and 5 of the PA-MSC-0983 effort to confirm MRP-227 applicability. The components were procured according to ASTM or ASME material specifications through applicable quality controlled protocols. D.C. Cook, Units 1 and 2 (CNP 1 and CNP 2) components were binned according to the following categories for material fabrication and cold work potential:

I. Cold work categories include the following:

• Cast austenitic stainless steel (CASS) (Category 1)

• Hot-formed austenitic stainless steel (Category 2)

• Annealed austenitic stainless steel (Category 3)

• Fasteners austenitic stainless steel (Category 4)

" Cold-formed austenitic stainless steel without subsequent solution annealing (Category 5)

Revision 0 PWROG-14044-P PWROG-14044-P Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-3

2. Cold work potential was based on MRP-227-A generic criteria:

• No (N) typically applies to cold work Categories 1, 2, and 3.

• Yes (Y) typically applies to cold work Categories 4 and 5.

Where multiple options existed for a component or assembly, the bounding condition, taken as including cold work, was selected for the purpose of the assessment. In some instances cascading fabrication (sequential processing such as forming, working, and heat treatment) would appear to mitigate any potential for cold work; however, since the historical record was not detailed, the potential is noted and a conservative approach was selected for this assessment.

D.C. Cook Unit I Westinghouse has evaluated the D.C. Cook Unit 1 (CNP 1) reactor internals components according to industry guideline MRP 2013-025 [1], as well as the MRP-191

[2] industry generic component listings and screening criteria (including consideration of cold work as defined in MRP-175 [31, noting the requirements of Section 3.2.3). The cold work assessment for CNP 1 was based on reactor vessel internals design drawings, which include materials specifications containing process information. In addition to consideration of the material fabrication, forming, and finishing process, a general screening definition of a resulting reduction in wall thickness of 20% was applied as an evaluation limit. It was confirmed that all CNP 1 components, as applicable for the design, are included directly in the MRP-191 component lists.

The evaluation, performed consistent with the industry guidelines, concluded that the reactor internals Category 1, 2, and 3 (non-bolting) components at CNP 1 contain no cold work greater than 20% as a result of material specification and controlled fabrication construction. Category 4 components were already assumed to have the potential for cold work in the MRP-191 generic assessments. Material fabrication specifications used for CNP 1 suggest that processes (e.g., heat treatment and strength characteristics) were limiting, which precluded the introduction of significant cold work in some of the Category 4 and 5 components. In these cases, the components were conservatively considered to be cold worked for the purposes of this assessment. There are no Category 5 components with significant cold work identified for CNP 1. Because no Category 5 components with significant cold work were identified, no stress evaluation was conducted, and therefore, components were not evaluated for operating stress that may exceed 30 ksi. The detailed evaluation for the A/LAI for the CNP 1 cold work assessments concluded that the plant-specific material fabrication and design were consistent with the MRP-191 basis and that the MRP-227-A sampling inspection aging management requirements as related to cold work are directly applicable to CNP 1.

PWROG-14044-P Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-4 D.C. Cook Unit 2 Westinghouse has evaluated the D.C. Cook Unit 2 (CNP 2) reactor internals components according to industry guideline MRP 2013-025 [1], as well as the MRP-191

[2] industry generic component listings and screening criteria (including consideration of cold work as defined in MRP-175 [3], noting the requirements of Section 3.2.3). The cold work assessment for CNP 2 was based on reactor vessel internals design drawings, which include materials specifications containing process information. In addition to consideration of the material fabrication, forming, and finishing process, a general screening definition of a resulting reduction in wall thickness of 20% was applied as an evaluation limit. It was confirmed that all CNP 2 components, as applicable for design, are included directly in the MRP-191 component lists.

The evaluation, performed consistent with the industry guidelines, concluded that the reactor internals Category 1, 2, and 3 (non-bolting) components at CNP 2 contain no cold work greater than 20% as a result of material specification and controlled fabrication construction. Category 4 components were already assumed to have the potential for cold work in the MRP-191 generic assessments. Material fabrication specifications used for CNP 2 would suggest that processes (e.g., heat treatment and strength characteristics) were limiting, which precluded the introduction of significant cold work in some of the Category 4 and 5 components. In these cases, the components were conservatively considered to be cold worked for the purposes of this assessment. There are no Category 5 components with significant cold work identified for CNP 2. Because no Category 5 components with significant cold work were identified, no stress evaluation was conducted, and therefore, components were not evaluated for operating stress that may exceed 30 ksi. The detailed evaluation for the A/LAI for the CNP 2 cold work assessments concluded that the plant-specific material fabrication and design were consistent with the MRP-191 basis and that the MRP-227-A sampling inspection aging management requirements as related to cold work are directly applicable to CNP 2.

Revision 0 PWROG-1 4044-P PWROG-14044-P Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-5 References

1. EPRI Letter, MRP 2013-025, "MRP-227-A Applicability Template Guideline," October 14, 2013.

2. Materials Reliability Program: Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design (MRP-191). EPRI, Palo Alto, CA: 2006. 1013234.

3. MaterialsReliability Program:PWR Internals MaterialAging DegradationMechanism Screening and Threshold Values (MRP-175). EPRI, Palo Alto, CA: 2005. 1012081.

4. U.S. NRC Letter, "Request for Additional Information (RAI) Regarding Aging Management Program for Reactor Vessel Internals, Donald C. Cook Nuclear plant, Units 1 and 2" October 1, 2012 (NRC ADAMS Accession Number 12284A320).

5. Safety Evaluation for MRP-227, "Safety Evaluation by the Office of Nuclear Reactor Regulation Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227, Revision 0)," March 28, 2011 (NRC ADAMS Accession Number ML110820773).

PWROG-14044-P Revision 0

ENCLOSURE 5 TO AEP-NRC-2014-60 CAW-14-4041, "Application for Withholding Proprietary Information from Public Disclosure"

Westinghouse Electric Company Engineering, Equipment and Major Projects 1000 Westinghouse Drive, Building 3 Cranberry Township, Pennsylvania 16066 USA U.S. Nuclear Regulatory Commission Direct tel: (412) 374-4643 Document Control Desk Direct fax: (724) 940-8560 11555 Rockville Pike e-mail: greshaja@westinghouse.com Rockville, MD 20852 Proj. letter CAW-14-4041 October 2, 2014 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

PWROG-14049-P, Rev. 0, "D. C. Cook Units 1 and 2 Summary Report for the Fuel Design /

Fuel Management Assessments for Reactor Internals Aging Management MRP-227-A Applicability" (Proprietary)

The proprietary information for which withholding is being requested in the above-referenced report is further identified in Affidavit CAW-14-4041 signed by the owner of the proprietary information, Westinghouse Electric Company LLC. The Affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Commission's regulations.

Accordingly, this letter authorizes the utilization of the accompanying Affidavit by Pressurized Water Reactor Owners Group (PWROG).

Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse Affidavit should reference CAW-] 4-4041 and should be addressed to James A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company, 1000 Westinghouse Drive, Building 3 Suite 3 10, Cranberry Township, Pennsylvania 16066.

Very truly yours, Jmes A..Gresham, Manager Regulatory Compliance Enclosures

CAW-14-4041 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:

ss COUNTY OF BUTLER:

Before me, the undersigned authority, personally appeared James A. Gresham, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse), and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief:

himes A. Gresham, Manager Regulatory Compliance Sworn to and subscribed before me this iJD day of 0.of 2014 Notary Pubhlc COMMONWEALTH OF PENNSYLVANIA C " NOTARJAL SEAL

[ Anne M. Stagrnan, Notary public utigo Cor m Twp., Westmoreland County y mssinExpIres Aug. 7, 2016 MEMAER. PENNSYLVANIA ASSOCIAVON OF NOTARIE

2 CAW-14-4041 (1) 1 am Manager, Regulatory Compliance, Westinghouse Electric Company LLC (Westinghouse),

and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of Westinghouse.

(2) I am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the Commission's regulations and in conjunction with the Westinghouse Application for Withholding Proprietary Information from Public Disclosure accompanying this Affidavit.

(3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged or as confidential commercial or financial information.

(4) Pursuant to the provisions of paragraph (b)(4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosure is owned and has been held

-in confidence by Westinghouse.

(ii) The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitute Westinghouse policy and provide the rational basis required.

Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as -follows:

(a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of

3 CAW-14-4041 Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.

(c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

(e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f) It contains patentable ideas, for which patent protection may be desirable.

(iii) There are sound policy reasons behind the Westinghouse system which include the following:

(a) The use of such information by Westinghouse givesWestinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position.

(b) It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.

(c) Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.

4 CAW-14-4041 (d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.

(e) Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries.

(f) The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.

(iv) The information is being transmitted to the Commission in confidence and, under -the provisions of 10 CFR Section 2.390, it is to be received in confidence by the Commission.

(v) The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief.

(vi) The proprietary information sought to be withheld in this submittal is that which is appropriately marked in PWROG-14049-P, Rev. 0, "D. C. Cook Units 1 and 2 Summary Report for the Fuel Design / Fuel Management Assessments for Reactor Internals Aging Management MRP-227-A Applicability" (Proprietary), for submittal to the Commission, being transmitted by PWROG letter OG-14-345 and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk. The proprietary information as submitted by Westinghouse is that associated with the NRC letter, "Donald C. Cook Nuclear Plant, Units 1 and 2 -'Requests for Additional Information Concerning the Reactor Vessel Internals Aging Management Program Submittal (TAC NOS. MFOO50 and MF0051)," ML14135A320, June 6, 2014, and may be used only for that purpose.

5 CAW-14-4041 (a) This information is part of that which will enable Westinghouse to:

(i) Support reactor vessel internals aging management.

(b) Further this information has substantial commercial value as follows:

(i) Westinghouse plans to sell the use of similar information to its customers for the purpose of supporting reactor internals aging management.

(iii) The information requested to be withheld reveals the distinguishing aspects of a methodology which was developed by Westinghouse.

Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar technical evaluation justifications and licensing defense services for commercial power reactors without commensurate expenses, Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and.

the expenditure of a considerable sum of money.

In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended.

Further the deponent sayeth not.

Proprietary Information Notice Transmitted herewith are proprietary and/or non-proprietary versions of documents furnished to the NRC in connection with requests associated with the NRC letter, "Donald C. Cook Nuclear Plant, Units 1 and 2

- Requests for Additional Information Concerning the Reactor Vessel Internals Aging Management Program Submittal (TAC NOS. MF0050 and MF0051)," ML14135A320, June 6, 2014, and may be used only for that purpose.

In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower~case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the Affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(1).

Copyright Notice The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.

Indiana Michigan Power Letter for Transmittal to the NRC The following paragraphs should be included in your letter to the NRC Document Control Desk:

Enclosed are:

1. One (1) copy of PWROG-14049-P, Rev. 0, "D. C. Cook Units 1 and 2 Summary Report for the Fuel Design / Fuel Management Assessments for Reactor Internals Aging Management MRP-227-A Applicability" (Proprietary)

2. One (1) copy of PWROG-14049-NP, Rev. 0, "D. C. Cook Units 1 and 2 Summary Report for the Fuel Design / Fuel Management Assessments for Reactor Internals Aging Management MRP-227-A Applicability" (Non-Proprietary)

Also enclosed is the Westinghouse Application for Withholding Proprietary Information from Public Disclosure CAW- 14-4041, accompanying Affidavit, Proprietary Information Notice, and Copyright Notice.

As Item 1 contains information proprietary to Westinghouse Electric Company LLC, it is supported by an Affidavit signed by Westinghouse, the owner of the information. The Affidavit sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of Section 2.390 of the Commission's regulations.

Accordingly, it is respectfully requested that the information which is proprietary to Westinghouse be withheld from public disclosure in accordance with 10 CFR Section 2.390 of the Commission's regulations.

Correspondence with respect to the copyright or proprietary aspects of the items listed above or the supporting Westinghouse Affidavit should reference CAW- 14-4041 and should be addressed to James A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric .Company, 1.000 Westinghouse Drive, Building 3 Suite 310, Cranberry Township, Pennsylvania 16066.

ENCLOSURE 6 TO AEP-NRC-2014-60 LTR-RIAM-14-24, Revision 1, "Reports for D.C. Cook Units 1 and 2 for PWROG PA-MSC-0983 Cafeteria Tasks 3, 4, and 5 Deliverables" [Non-Proprietary]

Westinghouse Non-Proprietary Class 3 O Westinghouse To: James P. Molkenthin Date: October 3, 2014 cc: Cheryl L. Boggess, Thomas S. Hamm From: Reactor Internals Aging Management Your ref: N/A Ext: 860-731-6607 Our ref: LTR-RIAM-14-24, Rev. I Fax: 860-731-2480

Subject:

Reports for D.C. Cook Units 1 and 2 for PWROG PA-MSC-0983 Cafeteria Tasks 3, 4, and 5 Deliverables

1. PWROG Project Authorization, PA-MSC-0983, Rev. 0, "Support for Applicant Action Items 1,

Reference:

2, and 7 fromr the Final Safety Evaluation on MRP-227, Revision 0," June 2012.

Attachment 1: D.C. Cook Unit I Applicant/Licensee Action Items I and 2 Report Attachment 2: D.C. Cook Unit 2 Applicant/Licensee Action Items I and 2 Report Attachment 3: D.C. Cook Unit 1 Applicant/Licensee Action Item 7 Report Attachment 4: D.C. Cook Unit 2 Applicant/Licensee Action Item 7 Report The purpose of this letter revision is to change the proprietary classification to Non-Proprietary Class 3.

Proprietary Class 2 information relating to customer comments has been removed in this revision of the letter.

D.C. Cook Units 1 and 2 (CNP I and CNP 2) are participating in Cafeteria Tasks 3, 4, and 5 of PA-MSC-0983 [1]. This letter provides the letter report deliverable, at the request of American Electric Power (AEP). Attachments I and 2 discuss applicant/licensee action items I and 2; Attachments 3 and 4 discuss Applicant/Licensee Action Item 7. Please transmit the attachments of this letter to Mr. Kevin Kalchik and Mr. Bruce Mickatavage.

If you have any questions, please contact Thomas S. Hamm at (412) 374-2142 or Jessica L. Tatarczuk at (860) 731-6607.

Authored by: ELECTRONICALLY APPROVED' Jessica L. Tatarczuk Transient & LOCA Analysis Authored by: ELECTRONICALLY APPROVED' Joshua K. McKinley Materials Center of Excellence Verified by: ELECTRONICALLY APPROVED' Steven G. Toney Aging Management, Radiation, and Nuclear Operations Services Approved by: ELECTRONICALLY APPROVED 1 Patricia C. Paesano, Manager Reactor Internals Aging Management Electronicallyapproved records are authenticatedin the electronic document management system.

© 2014 Westinghouse Electric Company LLC All Rights Reserved

Westinghouse Non-Proprietary Class 3 Page 2 of 2 Our ref: LTR-RIAM-14-24, Rev. I October 3, 2014 LEGAL NOTICE This report was prepared as an account of work performed by Westinghouse Electric Company LLC.

Neither Westinghouse Electric Company LLC, nor any person acting on its behalf:

A. Makes any warranty or representation, express or implied including the warranties of fitness for a particular purpose or merchantability, with respect to the accuracy, completeness, or usefulness of the information contained in this report, or that the use of any information, apparatus, method, or process disclosed in this report may not infringe privately owned rights; or B. Assumes any liabilities with respect to the use of, or for damages resulting from the use of, any information, apparatus, method, or process disclosed in this report.

COPYRIGHT NOTICE This report has been prepared by Westinghouse Electric Company LLC and bears a Westinghouse Electric Company LLC copyright notice. As a member of the PWR Owners Group, you are permitted to copy and redistribute all or portions of the report within your organization; however all copies made by you must include the copyright notice in all instances DISTRIBUTION NOTICE This report was prepared for the PWR Owners Group. This Distribution Notice is intended to establish guidance for access to this information. This report (including proprietary and non-proprietary versions) is not to be provided to any individual or organization outside of the PWR Owners Group program participants without prior written approval of the PWR Owners Group Program Management Office.

However, prior written approval is not required for program participants to provide copies of Class 3 Non Proprietary reports to third parties that are supporting implementation at their plant, and for submittals to the NRC.

Westinghouse Non-Proprietary Class 3 Attachment 1, Page 1 of 5 Our ref: LTR-RIAM-14-24, Rev. 1 October 3, 2014 : D.C. Cook Unit 1 Applicant/Licensee Action Items 1 and 2 Report 1.1 SE Applicant/Licensee Action Item 1: Applicability of FMECA and Functionality Analysis Assumptions The action item text from MRP-227-A states:

"As addressed in Section 3.2.5.1 of this SE, each applicant/licenseeis responsible for assessing its plant's design and operatinghistory and demonstratingthat the approved version of A1RP-22 7 is applicable to the facility. Each applicant/licenseeshall refer, in particular,to the assumptions regardingplant design and operatinghistory made in the FMECA andfunctionality analyses for reactors of their design (i.e., Westinghouse, CE, or B& W) which support MRP-227 and describe the process used for determining plant-specific differences in the design of their RVI components or plant operating conditions, which result in different component inspection categories. The applicant/licensee shall submit this evaluation for NRC review and approval as part of its application to implement the approved version of MARP-227. This is Applicant/Licensee Action Item V" [I].

D.C. Cook Unit 1 (CNP 1)

The process used to verify that CNP I is reasonably represented by the generic industry program assumptions with regard to neutron fluence, temperature, materials, and stress values used in the development of MRP-227-A is:

1. Identify typical Westinghouse pressurized water reactor (PWR) internals components (MRP-191, Table 4-4 [2]).

2. Identify CNP I PWR internals components.

3. Compare the typical Westinghouse PWR internals components to the CNP I PWR internals components identified in [7].

a. Confirm that no additional items were identified by this comparison (primarily supports Applicant/Licensee Action Item 2).

b. Confirm that the materials identified for CNP I are consistent with those materials identified in MRP-191, Table 4-4.

c. Confirm that the CNP I internals are the same as, or equivalent to, the typical Westinghouse PWR internals regarding design and fabrication.

4. Confirm that the CNP I operating history is consistent with the assumptions in MRP-227-A regarding core loading patterns.

5. Confirm that CNP I operates at base load.

6. Confirm that the CNP I reactor vessel internals (RVI) materials operated at temperatures within the original design basis parameters.

7. Determine stress values based on design basis documents.

8. Confirm that any changes to the CNP 1 RVI components do not impact the application of the MRP-227-A generic aging management strategy.

Westinghouse Non-Proprietary Class 3 Attachment 1, Page 2 of 5 Our ref: LTR-RIAM-14-24, Rev. I October 3, 2014 CNP 1 reactor internals components are reasonably represented by the design and operating history assumptions regarding neutron fluence, temperature, materials, and stress values in the MRP-191 generic failure modes, effects, and criticality analysis (FMECA) and the MRP-232 functionality analyses based on the following:

1. CNP 1 operating history is consistent with the assumptions in MRP-227-A with regard to neutron fluence.

a. The FMECA and functionality analyses for MRP-227-A were based on the assumption of 30 years of operation with high-leakage core loading patterns followed by 30 years of low-leakage core fuel management strategy. CNP 1 switched to use of a low-leakage core design prior to 30 years of operation [5]. Therefore, CNP I meets the fluence and fuel management timing assumptions in MRP-191. CNP 1 is evaluating operating history to confirm compliance with the numerical limits regarding core loading/core design in [II].

b. CNP I operates as a base load unit [5]. Therefore, CNP I satisfies the assumptions in MRP documents regarding operational parameters affecting fluence.

2. The CNP I reactor vessel materials operate at temperatures between Th., and TcoId that have been not less than 514.9°F for Totd and not higher than 607.5'F for Th., [3].

3. CNP 1 internals components and materials are comparable to the typical Westinghouse PWR internals components, as summarized in MRP-191, Table 4-4.

a. No additional components are identified for CNP I by this comparison to MRP-191.

b. Most of the materials identified for components at CNP I are consistent with, or equivalent to, those materials identified in MRP-191, Table 4-4 for Westinghouse-designed plants [2]. Where differences exist, there is no impact on the CNP I RVI program or the component is already credited as being managed under an alternate CNP I aging management program.

As stated above, CNP 1 RVI component materials are consistent with, or equivalent to, those materials identified in [2]. The components that contain materials that differ from

[2] are control rod guide tube assemblies - guide cards/plates; control rod guide tube assemblies - housing plates; mixing devices; upper core plate and fuel alignment pins -

alignment pins, upper instrumentation conduit and supports - bolting; upper instrumentation conduit and supports- brackets, clamps, terminal blocks, and conduit straps; upper instrumentation conduit and supports - locking caps; upper support plate assembly - lock keys; bottom-mounted instrumentation (BMI) column assemblies -

BMI cruciforms; lower core plate and fuel alignment pins - fuel alignment pins, lower support column assemblies - lower support column bolts; and radial support keys -

radial support key bolts.

Four of these components required further evaluation to determine the applicability of MIRP-227-A to CNP 1. These were the control rod guide tube assemblies - guide cards/plates, the control rod guide tube assemblies - housing plates, the upper instrumentation conduit and supports - brackets, clamps, terminal blocks, and conduit straps, and the BMI column assemblies - BMI cruciform. These were evaluated for the

Westinghouse Non-Proprietary Class 3 Attachment 1, Page 3 of 5 Our ref: LTR-RIAM-14-24, Rev. I October 3, 2014 potential impacts of a significant change in material: from wrought Type 304SS to cast austenitic stainless steel (CASS) for the first three components and from CASS to wrought material for the BMI cruciform. A FMECA expert panel review applying the same methodology as used in the development of MRP-191 was conducted for these four components [4]. The FMECA concluded that the aging management strategies of MRP-227-A were still applicable based on a consideration of the likelihood of failure and the likelihood of damage and the resulting classification of the components. There is no change to the CNP I MRP-227-A inspection requirements as a result of the differences in the material of fabrication for these components.

c. Design and fabrication of CNP 1 RVI components are the same as, or equivalent to, the typical Westinghouse-designed PWR RVI components.

4. CNP I has made modifications over the lifetime of the plant. They are identified in general industry guidance or specifically directed by the original equipment manufacturer (OEM).

These modifications include a control rod guide tube and split pin replacement in 1985, Cycle 9 [5], a barrel-former bolt replacement of three bolts in 1997, Cycle 16 [6], and a clevis insert bolt replacement in 2013 [8].

CNP 1 had degraded bolts and a degraded dowelpin discovered in the lower radial support system clevis inserts in the reactor vessel in 2010, Cycle 23 [5]. During the Cycle 25 refueling outage, 28 clevis insert bolts were replaced. The replacement bolt design and materials are similar to the original bolt design considered in MRP-227-A [1]. The unit continues to be bounded following repair. CNP 1 has not made any additional modifications to the reactor internals components, with the exception of the above-mentioned degraded clevis insert bolts, since May 2007 [5]. Operational parameters with regard to fluence and temperature are compliant with MRP-227-A requirements, and the tomponents and materials are the same as those considered in MRP-191. Therefore, the CNP I components have remained within the original structural design configuration, and stress values are represented by the assumptions in MRP-227-A, MRP-191, and MRP-232.

Conclusion The assumptions regarding plant design and operating history made in the FMECA and functionality analyses for the Westinghouse design apply to CNP 1. There are no components at CNP 1 not contained in the FMECA and functionality analysis. There are components with materials different than those assumed in the FMECA; however, evaluations have been completed to verify that these differences do not affect the current aging management strategy.

CNP I meets the requirements for application of MRP-227-A as a strategy for managing age-related material degradation in reactor internals components. CNP I will implement and apply the approved version of MRP-227 (MRP-227-A) as a strategy for managing age-related material degradation in reactor internals components.

Westinghouse Non-Proprietary Class 3 Attachment 1, Page 4 of 5 Our ref: LTR-RIAM-14-24, Rev. I October 3, 2014 1.2 SE Applicant/Licensee Action Item 2: PWR Vessel Internal Components within the Scope of License Renewal The action item text from MRP-227-A states:

"As discussed in Section 3.2.5.2 of this SE, consistent with the requirements addressed in 10 CFR 54.4, each applicant/licensee is responsible for identifying which RVI components are within the scope of LR for its facility. Applicants/licenseesshall review the information in Tables 4-1 and 4-2 in AMRP-189, Revision 1, and Tables 4-4 and 4-5 in MRP-191 and identify whether these tables contain all of the RI'I components that are within the scope of LR for theirfacilities in accordance with 10 CFR 54.4. If the tables do not identify all the R V1 components that are within the scope of LR for its facility, the applicant or licensee shall identift the missing component(s) andpropose any necessary modifications to the program defined in MRP-227, as modifed by this SE, when submitting its plant-specific Aging Management Program Plan. The Aging Management Program Plan shall provide assurance that the effects of aging on the missing component(s) will be managed for the period of extended operation. This issue is Applicant/Licensee Action Item 2" [1].

CNP 1 Compliance This A/LAI requires comparison of the RVI components that are within the scope of license renewal for CNP 1 [7] to those components contained in MRP-191, Table 4-4. A detailed tabulation of the CNP I RVI components was completed, and it was compared to the typical Westinghouse PWR components in MRP- 191.

Several components have different materials than those specified in the MRP-191 assessment, but the differences have no effect on the recommended MRP aging strategy or aging is already managed by an alternate CNP I program. Therefore, no modifications to the program details in MRP-227-A need to be proposed.

This supports the requirement that the NRC-Aging Management Program Plan shall provide assurance that the effects of aging on the CNP I RVI components within the scope of license renewal, but not included in the generic Westinghouse-designed PWR RVI components from Table 4-4 ofMRP-191, will be managed for the period of extended operation.

The generic scoping and screening of the RVI, as summarized in MRP-191 and MRP-232 [9] to support the inspection sampling approach for aging management of RVI specified in MRP-227-A, are applicable to CNP I with no modifications.

Conclusion CNP 1 complies with A/LAI 2 of the NRC SE on MRP-227, revision 0; therefore, it meets the requirement for application of MRP-227-A as a strategy for managing age-related material degradation in reactor internal components.

Westinghouse Non-Proprietary Class 3 Attachment 1, Page 5 of 5 Our ref: LTR-RIAM-14-24, Rev. 1 October 3, 2014 1.3 References

1. Materials Reliability Program: Pressurized Water Reactor Internals Inspection and EvaluationGuidelines (MRP-227-A). EPRI, Palo Alto, CA: 2011. 1022863.

2. Materials Reliability Program. Screening. Categorization,and Ranking of Reactor Internals Componentsfor Westinghouse and Combustion EngineeringPWR Design (MRP-191). EPRI, Palo Alto, CA: 2006. 1013234.

3. D.C. Cook UFSAR, Rev. 25, "Indiana and Michigan Power D. C. Cook Nuclear Plant Updated Final Safety Analysis Report," September 9, 2013.

4. Westinghouse Letter, LTR-RIAM-14-87, Rev. 1, "Summary of DC Cook Units I and 2 Expert Elicitation Panel Meeting Minutes for Reactor Internals Components and Materials,"

September 26, 2014. (Westinghouse Proprietary Class 2)

5. D.C. Cook Letter, AEP-NRC-2012-82, "Donald C. Cook Nuclear Plants Units 1 and 2 Transmittal of Reactor Vessel Internals Aging Management Program," ADAMS #

ML12284A320.

6. Westinghouse Report, WCAP-14823, Rev. 0, "Donald C. Cook Unit 1 - Barrel Former Bolt Issue - Replacement Bolt Evaluation," March 1997. (Westinghouse Proprietary Class 2)

7. License Renewal Application D.C. Cook Nuclear Generating Station, "License Renewal Application," October 2003.

8. Indiana Michigan Power Technical Bulletin, "Root Cause Evaluation of Unit I Rx Vessel Core Support Lug Bolting Anomalies," March 21, 2010 (AR 2010-1804-10).

9. Materials Reliability Program: Aging Management Strategies for Westinghouse and Combustion EngineeringPWR Internals (MRP-232). EPRI, Palo Alto, CA: 2008. 1016593.

10. AEP Design Information Transmittal, DIT-B-03618-00, "Reactor Vessel Internals Aging Management Assessment," September 16, 2014.

11. EPRI Letter, MRP 2013-025, "MRP-227-A Applicability Template Guideline," October 14, 2013.

Westinghouse Non-Proprietary Class 3 Attachment 2, Page 1 of 5 Our ref: LTR-RIAM-14-24, Rev. I October 3, 2014 : D.C. Cook Unit 2 Applicant/Licensee Action Items 1 and 2 Report 2.1 SE Applicant/Licensee Action Item 1: Applicability of FMECA and Functionality Analysis Assumptions The action item text from MRP-227-A states:

"As addressedin Section 3.2.5.1 of this SE, each applicant/licenseeis responsiblefor assessing its plant's design and operatinghistory and demonstratingthat the approved version of ARP-22 7 is applicable to the facility. Each applicant/licenseeshall refer, in particular,to the assumptions regardingplant design and operating history made in the FMECA andfinctionality analyses for reactors of their design (i.e., Westinghouse, CE, or B&W) which support AMRP-227 and describe the process used for determining plant-specific difierences in the design of their RVI components or plant operating conditions, which result in different component inspection categories. The applicant/licensee shall submit this evaluation for NRC review and approval as part of its application to implement the approved version of MRP-227. This is Applicant/Licensee Action Item 1" [1].

D.C. Cook Unit 2 (CNP 2)

The process used to verify that CNP 2 is reasonably represented by the generic industry program assumptions with regard to neutron fluence, temperature, materials, and stress values used in the development of MIRP-227-A is:

I. Identify typical Westinghouse pressurized water reactor (PWR) internals components (MRP-191, Table 4-4 [2]).

2. Identify CNP 2 PWR internals components.

3. Compare the typical Westinghouse PWR internals components to the CNP 2 PWR internals components identified in [5].

a. Confirm that no additional items were identified by this comparison (primarily supports Applicant/Licensee Action Item 2).

b. Confirm that the materials identified for CNP 2 are consistent with those materials identified in MRP-191, Table 4-4.

c. Confirm that the CNP 2 internals are the same as, or equivalent to, the typical Westinghouse PWR internals regarding design and fabrication.

4. Confirm that the CNP 2 operating history is consistent with the assumptions in MRP-227-A regarding core loading patterns.

5. Confirm that CNP 2 operates at base load.

6. Confirm that the CNP 2 RVI materials operated at temperatures within the original design basis parameters.

7. Determine stress values based on design basis documents.

Westinghouse Non-Proprietary Class 3 Attachment 2, Page 2 of 5 Our ref: LTR-RIAM-14-24, Rev. I October 3, 2014

8. Confirm that any changes to the CNP 2 reactor vessel internals (RVI) components do not impact the application of the MRP-227-A generic aging management strategy.

CNP 2 reactor internals components are reasonably represented by the design and operating history assumptions regarding neutron fluence, temperature, materials, and stress values in the MRP-191 generic failure modes, effects, and criticality analysis (FMECA) and the MRP-232 functionality analyses based on the following:

1. CNP 2 operating history is consistent with the assumptions in MRP-227-A with regard to neutron fluence.

a. The FMECA and functionality analyses for MRP-227-A were based on the assumption of 30 years of operation with high-leakage core loading patterns followed by 30 years of low-leakage core fuel management strategy. D.C. Cook Unit 2 fuel management program changed from high- to low-leakage core loading pattern prior to 30 years of operation [4]. Therefore, CNP 2 meets the fluence and fuel management assumptions in MRP-191 and requirements for MRP-227-A application.

b. CNP 2 operates as a base load unit [4]. Therefore, CNP 2 satisfies the assumptions in MRP documents regarding operational parameters affecting fluence.

2. The CNP 2 reactor vessel materials operate at temperatures between Thot and TCOId that have nominally been not less than 541.27°F for TcoId and not higher than 606.357F for Th., [6].

3. CNP 2 internals components and materials are comparable to the typical Westinghouse PWR internals components, as summarized in MRP-191, Table 4-4.

a. No additional components are identified for CNP 2 by this comparison to MRP-191.

b. Most of the materials identified for components at CNP 2 are consistent with, or equivalent to, those materials identified in MRP-191, Table 4-4 for Westinghouse-designed plants [2]. Where differences exist, there is no impact on the CNP 2 RVI program or the component is already credited as being managed under an alternate CNP 2 aging management program.

As stated above, CNP 2 RVI component materials are consistent with, or equivalent to, those materials identified in [2]. The components that contain materials that differ from

[2] are upper core plate and fuel alignment pins - alignment pins, upper instrumentation conduit and supports - bolting; upper instrumentation conduit and supports - brackets, clamps, terminal blocks, and conduit straps; upper instrumentation conduit and supports -

locking caps; upper support plate assembly - flange; tipper support plate assembly - lock keys; upper support plate assembly - upper support plate; upper support plate assembly -

tipper support ring or skirt; lower internals assembly - bottom-mounted instrumentation (BMI) column cruciforrns; lower core plate and fuel alignment pins - fuel alignment pins, lower support column assemblies - lower support column bolts; neutron panels/thermal shield - thermal shield dowels; and radial support keys - radial support key bolts.

Five of these components required further evaluation to determine the applicability of MRP-227-A to CNP 2. These were the upper instrumentation conduit and supports -

brackets, clamps, terminal blocks, and conduit straps, upper support plate assembly -

Westinghouse Non-Proprietary Class 3 Attachment 2, Page 3 of 5 Our ref: LTR-RIAM-14-24, Rev. 1 October 3, 2014 flange, upper support plate assembly - upper support plate, upper support plate assembly

- upper support ring or skirt, and BMI column assemblies - BMI column cruciforns.

These were evaluated for the potential impacts of a significant change in material: from wrought Type 304SS to cast austenitic stainless steel (CASS) for the first four components and from CASS to wrought material for the BMI column cruciforms. A FMECA expert panel review applying the same methodology as used in the development of MRP-191 was conducted for these four components [3]. The FMECA concluded that the aging management strategies of MRP-227-A were still applicable based on a consideration of the likelihood of failure and the likelihood of damage and the resulting classification of the components. There is no change to the CNP 2 MRP-227-A inspection requirements as a result of the differences in the material of fabrication for these components.

c. Design and fabrication of CNP 2 RVI components are the same as, or equivalent to, the typical Westinghouse designed PWR RVI components

4. CNP 2 has made modifications to the RVI. These modifications include a control rod guide tube split pin replacement by Babcock & Wilcox in 1986, Cycle 6; a control rod guide tube cap screw modification in 1986, Cycle 6; and a baffle-former bolt replacement of 52 bolts in 2010, Cycle 19 [4]. MRP-227-A states that the recommendations are applicable to all U.S.

PWR operating plants as of May 2007 for the three designs considered. Baffle-former bolt degradation was observed and repaired in 2010, and repair was performed by the original equipment manufacturer (OEM). No other modifications have been made to the CNP 2 RVIs; therefore, CNP 2 is bounded [4]. Operational parameters with regard to fluence and temperature are compliant with MRP-227-A requirements, and the components and materials are the same as those considered in MRP-191. Therefore, the CNP 2 components have remained within the original structural design configuration, and stress values are represented by the assumptions in MRP-227-A, MRP-191, and MRP-232.

Conclusion The assumptions regarding plant design and operating history made in the FMECA and functionality analyses for the Westinghouse design apply to CNP 2. There are no components at CNP 2 not contained in the FMECA and functionality analysis. There are components with materials different than those assumed in the FMECA; however, evaluations have been completed to verify that these differences do not affect the current aging management strategy.

CNP 2 meets the requirements for application of MRP-227-A as a strategy for managing age-related material degradation in reactor internals components. CNP 2 will implement and apply the approved version of MRP-227 (MRP-227-A) as a strategy for managing age-related material degradation in reactor internals components.

Westinghouse Non-Proprietary Class 3 Attachment 2, Page 4 of 5 Our ref: LTR-RIAM-14-24, Rev. 1 October 3, 2014 2.2 SE Applicant/Licensee Action Item 2: PWR Vessel Internal Components within the Scope of License Renewal The action item text from MRP-227-A states:

"As discussed in Section 3.2.5.2 of this SE, consistent with the requirements addressed in 10 CFR 54.4, each applicant/licensee is responsiblefor identfjing which RVI components are within the scope of LR for its facility. Applicants/licensees shall review the information in Tables 4-1 and 4-2 in M.IRP-189, Revision 1, and Tables 4-4 and 4-5 in MRP-191 and identify whether these tables contain all of the RVI components that are within the scope of LR for theirfacilities in accordance with 10 CFR 54.4. If the tables do not identify all the RVI components that are within the scope of LR for its facility, the applicant or licensee shall identifj, the missing component(s) and propose any necessary modifications to the program defined in AIRP-22 7, as modified by this SE, when submitting its plant-specific Aging Management Program Plan. The Aging Management Program Plan shall provide assurance that the effects of aging on the missing component(s) will be managed for the period of extended operation. This issue is Applicant/LicenseeAction Item 2" [1].

CNP 2 Compliance This Applicant/Licensee Action Item requires comparison of the RVI components that are within the scope of license renewal for CNP 2 [5] to those components contained in MRP-191, Table 4-

4. A detailed tabulation of the CNP 2 RVI components [5] was completed, and it was compared to the typical Westinghouse PWR components in MiRP-191.

Several components have different materials than those specified in the MRP-191 assessment, but the differences have no effect on the recommended MRP aging strategy or aging is already managed by an alternate CNP 2 program. Therefore, no modifications to the program details in MRP-227-A need to be proposed.

This supports the requirement that the NRC-Aging Management Program Plan shall provide assurance that the effects of aging on the CNP 2 RVI components within the scope of license renewal, but not included in the generic Westinghouse-designed PWR RVI components from Table 4-4 of MRP-191, will be managed for the period of extended operation.

The generic scoping and screening of the RVI, as summarized in MRP-191 and MRP-232 [7] to support the inspection sampling approach for aging management of RVI specified in MRP-227-A, are applicable to CNP 2 with no modifications.

Conclusion CNP 2 complies with A/LAI 2 of the NRC SE on MRP-227, revision 0; therefore, it meets the requirement for application of MRP-227-A as a strategy for managing age-related material degradation in reactor internal components.

Westinghouse Non-Proprietary Class 3 Attachment 2, Page 5 of 5 Our ref: LTR-RIAM-14-24, Rev. 1 October 3, 2014 2.3 References

1. Materials ReliabilitY Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A). EPRI, Palo Alto, CA: 2011. 1022863.

2. Materials Reliabilit, Program:Screening. Categorization,and Ranking of Reactor Internals Components for Westinghouse and Combustion EngineeringPWR Design (MRP-191). EPRI, Palo Alto, CA: 2006. 1013234.

3. Westinghouse Letter LTR-RIAM-14-87, Rev. 1, "Summary of DC Cook Units I and 2 Expert Elicitation Panel Meeting Minutes for Reactor Internals Components and Materials,"

September 26, 2014. (Westinghouse Proprietary Class 2)

4. D.C. Cook Letter AEP-NRC-2012-82, "Donald C. Cook Nuclear Plants Units I and 2 Transmittal of Reactor Vessel Internals Aging Management Program," ADAMS #

ML12284A320.

5. License Renewal Application D.C. Cook Nuclear Generating Station, "License Renewal Application," October 2003.

6. D.C. Cook UFSAR, Rev. 25, "Indiana and Michigan Power D. C. Cook Nuclear Plant Updated Final Safety Analysis Report," September 9, 2013.

7. Materials Reliability Program: Aging Management Strategies for Westinghouse and Combustion EngineeringPWR Internals (MRP-232). EPRI, Palo Alto, CA: 2008. 1016593.

Westinghouse Non-Proprietary Class 3 Attachment 3, Page 1 of 4 Our ref: LTR-RIAM-14-24, Rev. I October 3, 2014 : D.C. Cook Unit 1 Applicant/Licensee Action Item 7 Report 3.1 SE Applicant/Licensee Action Item 7: Plant-specific Evaluation of CASS Materials "As discussed in Section 3.3.7 of this SE, the applicants/licensees of B& W, CE, and Westinghouse reactors are required to develop plant-specific analyses to be applied for their facilities to demonstrate that B& W IMI guide tube assembly spiders and CRGT spacer castings, CE lower support columns, and Westinghouse lower support column bodies will maintain their functionality during the period of extended operationor for additionalRVI components that may be fabricated from CASS, mnartensitic stainless steel or precipitation hardened stainless steel materials. These analyses shall also consider the possible loss offracture toughness in these components due to thermal and irradiation embrittlement, and may also need to consider limitations on accessibility for inspection and the resolution/sensitivity of the inspection techniques. The requirementmay not apply to components that were previously evaluated as not requiring aging management during development of MRP-227. That is, the requirement would apply to components fabricatedfiom susceptible materials for which an individual licensee has determined aging management is required, for example during their review performed in accordance with Applicant/Licensee Action Item 2. The plant-specific analysis shall be consistent with the plant's licensing basis and the need to maintain the functionality of the components being evaluated under all licensing basis conditions of operation. The applicant/licenseeshall include the plant-specific analysis as part of their submittal to apply the approved version of MRP-227. This is Applicant/Licensee Action Item 7" [1].

D.C. Cook Unit 1 (CNP 1)

Applicant/Licensee Action Item 7 from the staff's final SE on MRP-227-A [1] states:

"For CASS, if the application of applicable screening criteriafor the component's material demonstrates that the components are not susceptible to either thermal embrittlement or irradiation embrittlement, or the synergistic effects of thermal embrittlement and irradiation embrittlement combined, then no other evaluation would be necessary. For assessment of CASS materials,the licensee or applicantfor license renewal may apply the criteriain the NRC letter of May 19, 2000, "License Renewal Issue No. 98-0030, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Components" (NRC ADAMS Accession No. MLO03 717179) as the basis fior determining whether the CASS materials are susceptible to the thermal aging mechanism."

The CNP 1 reactor vessel internals (RVI) cast austenitic stainless steel (CASS) components and the assessment of their susceptibility to thermal embrittlement (TE) are summarized in Table 3.1-1. Susceptibility to TE is based on the NRC acceptance criteria [3]. The ferrite content was calculated based on Hull's formula [4].

In the CNP I RVI, 28 of 29 upper support column assemblies column bases (orifice base style),

18 of 19 upper support column assemblies column bases (mixing style), and the lower internals assembly lower support casting, are screened as not susceptible to TE.

Westinghouse Non-Proprietary Class 3 Attachment 3, Page 2 of 4 Our ref: LTR-RIAM-14-24, Rev. I October 3, 2014 The upper internals assembly mixing devices (of the stand-alone style) are potentially susceptible to TE. One support column base (orifice style) and one support column base (mixing style) are also potentially susceptible to TE. The susceptibility of the mixing devices and column bases to TE was considered in the development of MRP-227-A [1].

The CNP 1 brackets, clamps, terminal blocks, and conduit straps on the upper instrumentation columns are CASS, and are assumed to be susceptible to TE. In MRP-191, the Lipper instrumentation conduit and support - brackets, clamps, terminal blocks, and straps were screened as 304 stainless steel (SS), not as CF8. These parts are evaluated as CASS under the guidelines of the MRP-191 failure modes, effects, and critically analysis (FMECA) in support of A/LAIs I and 2.

Documentation confirming the materials used to construct the control rod guide tube assembly housing plates, intermediate and lower flanges, and guide plates/cards for CNP I were not located. Therefore, for A/LAI 7, it is conservatively assumed that they are CASS and that they are potentially susceptible to TE. The susceptibility of the guide tube intermediate and lower flanges to TE was considered in the development of MRP-227-A [1]. However, in MRP-191 [2],

the control rod guide tube housing plates and guide plates/cards were screened as 304 SS, not as CF8.

The lower support column bodies (column caps) are considered potentially susceptible to TE.

The susceptibility of the lower support column bodies to TE was considered in the development of MRP-227-A.

Irradiation may also cause a material to become embrittled. The upper support column bases, control rod guide tube lower flanges, upper internals assembly mixing devices, and lower support column bodies screened-in at the MRP-191 irradiation screening level [2]; thus, for these components, susceptibility to irradiation embrittlement (IE) was considered in the development of MRP-227-A [1]. The control rod guide tube housing plates, intermediate flanges and guide plates/cards, and the upper instrumentation conduit and support brackets, clamps, terminal blocks screened below the MRP-191 irradiation screening level [2]; thus, IE of these components was not considered in the development of MRP-227-A [1].

The material difference for the control rod guide tube housing plates, intermediate flanges and guide plates/cards is evaluated tinder the guidelines of the MRP-191 Failure Modes, Effects, and Critically Analysis in support of A/LAIs I and 2.

No martensitic stainless steel or martensitic precipitation-hardened stainless steel materials were identified in the CNP 1 RVI.

Conclusion The results of this CASS evaluation do not conflict with the MRP-227-A strategy for aging management of RVI. It is concluded that continued application of the strategy of MRP-227-A will meet the requirement for managing age-related degradation of the CNP I CASS RVI components.

Westinghouse Non-Proprietary Class 3 Attachment 3, Page 3 of 4 Our ref: LTR-RIAM-14-24, Rev. 1 October 3, 2014 Table 3.1-1: Summary of CNP I CASS Components and Their Susceptibility to TE Molybdenum Casting Ferrite Susceptibility to TE CASS Component Content Method Content (Based on NRC Criteria (Wt %) o(Wt %) 131)

Upper Internals Assembly Control Rod Assemblies - Guide Low 0.5 Max. Static > 20(2) Potentially Susceptiblet 5 )

t1 Plates/Cards Control Rod Assemblies -Flanges -()5 5

Potentially Susceptiblet 1 Static >> 20 Potentially LowtesCrds Control Rod Assemblies- l Flanges - Low 0.5 Max.

Low 0.5 Max. Static 20(2) Susceptible(5 )

Lower"')

Control Rod Assemblies- Flanges - > 0(2)(5 Low 0.5 Max. Static >Potentially Susceptible Intermediatel Control Rod Assemblies - Housing Low 0.5 Max. Static > 20(2 Potentially Susceptible 5 Plates")y Upper Instrumentation Conduit and 5 Supports - Brackets, Clamps, Low 0.5 Max. Static > 20(2) Potentially Susceptible( )

Terminal Blocks, and Conduit Straps 5

Upper Internals Assembly - Mixing Low 0.5 Max. Static > 20(2) Potentially Susceptible(s Devices (Standalone Style)

< 20(3 18 Not Susceptible"4 )

I Potentially Low 0.5 Max. Static Upper Support Column Assemblies -

Column Bases (Mixing Style) > 20o2) I Potentially Susceptible (5)

< 20(3 28 Not Susceptible14 )

I Potentially Low 0.5 Max. Static Upper Support Column Assemblies -

Column Bases > 20() Potentially(5)

Susceptible I >2 Lower Internals Assembly Lower Internals Assembly - Lower Low 0.5 Max. Static < 20(3) Not Susceptible 4)

Support Casting Lower Support Column Assemblies Low 0.5 Max. Static > 20t2) Potentially Susceptible(5)

- Lower Support Column Bodies Notes:

1. Per engineering drawing, component may be 304 SS or alternate material CASS.

2. Where component-specific certified material test report data is not available, the ferrite content is calculated (per guidance of [4]) based on permitted variations in ASTM A351, Grade CF8 chemistry requirements. Allowable variants of Grade CF8 chemistry requirements may result in ferrite content greater than 20%; thus, the ferrite content is identified as potentially exceeding 20%.

3. The elemental percentages from retrieved certified material test reports (CMTRs) are input into Hull's formula (per guidance of[4]) to calculate the ferrite content of the CASS material.

4. Conclusion is based on CMTR chemistry data.

5. Conclusion is based on material specification chemistry requirements.

Westinghouse Non-Proprietary Class 3 Attachment 3, Page 4 of 4 Our ref: LTR-RIAM-14-24, Rev. I October 3, 2014 3.2 References

1. Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A). EPRI, Palo Alto, CA: 2011. 1022863.

2. Materials Reliability Program:Screening, Categorization,and Ranking of Reactor hIternals Componentsfor Westinghouse and Combustion EngineeringPWR Design (MRP-191). EPRI, Palo Alto, CA: 2006. 1013234.

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

4. U.S. Nuclear Regulatory Commission, NUREG/CR-4513, Rev. 1, "Estimation of Fracture Toughness of Cast Stainless Steels during Thermal Aging in LWR Systems," May 1994 (NRC ADAMS Accession No. ML052360554).

Westinghouse Non-Proprietary Class 3 Attachment 4, Page 1 of 4 Our ref: LTR-RIAM-14-24, Rev. I October 3, 2014 Attachment 4: D.C. Cook Unit 2 Applicant/Licensee Action Item 7 Report 4.1 SE Applicant/Licensee Action Item 7: Plant-specific Evaluation of CASS Materials "As discussed in Section 3.3.7 of this SE, the applicants/licensees of B&W, CE, and Westinghouse reactors are required to develop plant-specific analyses to be applied for their facilities to demonstrate that B& W IMI guide tube assembly spiders and CRGT spacer castings, CE lower support columns, and Westinghouse loiwer support column bodies will maintain their functionality during the period of extended operation orfor additionalRVI components that may be fabricatedfrom CASS, martensitic stainless steel or precipitation hardened stainless steel materials. These analyses shall also consider the possible loss of fracture toughness in these components due to thermal and irradiation embrittlement, and may also need to consider limitations on accessibility for inspection and the resolution/sensitivity of the inspection techniques. The requirementmay not apply to components that were previously evaluated as not requiring aging management during development of MRP-227. That is, the requirement would apply to components fabricatedfirom susceptible materialsfor which an individual licensee has determined aging management is required, for example during their review performed in accordancewith Applicant/Licensee Action Item 2. The plant-specific analysis shall be consistent with the plant's licensing basis and the need to maintain the fiunctionality of the components being evaluated under all licensing basis conditions of operation. The applicant/licensee shall include the plant-specific analysis as part of their submittal to apply the approved version of MRP-22 7. This is Applicant/Licensee Action Item 7" [1].

D.C. Cook Unit 2 (CNP 2)

Applicant/Licensee Action Item 7 from the staff's final SE on MRP-227-A [I ] states:

"For CASS, if the application of applicable screening criterlafor the component's material derronstrates that the components are not susceptible to either thermal embrittlement or irradiationembrittlement, or the synergistic effects of thermal embrittlement and irradiation embrittlement combined, then no other evaluation would be necessary. For assessment of CASS materials, the licensee or applicantfor license renewal may apply the criteriain the NRC letter of May 19, 2000, "License Renewal Issue No. 98-0030, Thermal Aging Embrittlerrent of Cast Austenitic Stainless Steel Components" (NRC ADAMS Accession No. ML003717179) as the basis for determiningwhether the CASS materialsare susceptible to the thermal agingmechanism. "

The CNP 2 reactor vessel internals (RVI) cast austenitic stainless steel (CASS) components and the assessment of their susceptibility to TE are summarized in Table 4.1-1. Susceptibility to TE is based on the NRC acceptance criteria [3]. The ferrite content was calculated based on Hull's formula [4].

The CNP 2 upper support plate assembly upper support plate (includes flange, ring or skirt), 20 of 21 upper support column assemblies column bases (orifice style), upper support column assemblies column bases (mixing style), lower internals assembly lower support casting, and the

Westinghouse Non-Proprietary Class 3 Attachment 4, Page 2 of 4 Our ref: LTR-RIAM-14-24, Rev. 1 October 3, 2014 lower support column assemblies lower support column bodies (column.caps) are screened as not susceptible to TE. In MRP-191, the upper support plate, flange, and ring were screened as 304 SS, not as CF8.

The upper internals assembly mixing devices (of the stand-alone style), and one of the 21 upper support column bases (orifice style), are potentially susceptible to TE. The susceptibility of the mixing devices and upper support column bases to TE was considered in the development of MRP-227-A.

The CNP 2 brackets, clamps, terminal blocks, and conduit straps on the upper instrumentation columns are CASS, and are assumed to be susceptible to TE. In MRP-191, the upper instrumentation conduit and supports, brackets, clamps, terminal blocks, and straps were screened as 304 SS, not as CF8. These parts are evaluated as CASS tinder the guidelines of the MRP-191 Failure Modes, Effects, and Critically Analysis (FIVIECA) in support of A/LAls 1 and 2.

Irradiation may also cause a material to become embrittled. The Lipper support column bases, upper internals assembly mixing devices, and lower support column bodies screened-in at the MRP-191 irradiation screening level [2]; thus, for these components, susceptibility to irradiation embrittlement (IE) was considered in the development of MRP-227-A [1]. The upper support plate, flange, ring or skirt and the tipper instrumentation conduit and support brackets, clamps, terninal blocks and conduit straps screened below the MRP-191 irradiation screening level [2];

thus, IE of these components was not considered in the development of MRP-227-A [1].

The material difference for the upper support casting, which includes the upper support plate, flange, and ring, is evaluated under the guidelines of the MRP-191 Failure Modes, Effects, and Critically Analysis in support of AILAls 1 and 2.

No martensitic stainless steel or martensitic precipitation-hardened stainless steel materials were identified in the CNP 2 RVI.

Conclusion The results of this CASS evaluation do not conflict with the MRP-227-A strategy for aging management of RVI. It is concluded that continued application of the strategy of MRP-227-A will meet the requirement for managing age-related degradation of the CNP 2 CASS RVI components.

Westinghouse Non-Proprietary Class 3 Attachment 4, Page 3 of 4 Our ref: LTR-RIAM-14-24, Rev. 1 October 3, 2014 Table 4.1-1: Summary of CNP 2 CASS Components and Their Susceptibility to TE Susceptibility to TE CASS Component (Based on the NRC Criteria 131)

Upper Internals Assembly Upper Support Plate Assembly Low 0.5 Max. Static < 20(2) Not Susceptible( 3)

- Upper Support Plate Upper Instrumentation Conduit and Supports -Brackets, Low 0.5 Max. Static > 20") Potentially Susceptible 4)

Clamps, Terminal Blocks, and Conduit Straps Upper Internals Assembly -

Mixing Devices (Standalone Low 0.5 Max. Static > 20"1 Potentially Susceptible4

• Style)

Upper Support Column Assemblies - Column Bases Low 0.5 Max. Static < 202) Not Susceptible'3 *

(Mixing Style) 3 Low 0.5 Max. StaticSusceptible Upper Support Column Assemblies - Column Bases L 05 x StI Potentially

> 20(l) Susceptible, 4 )

Lower Internals Assembly Lower Internals Assembly - Low 0.5 Max. Static < 2012, Not Susceptible 3 )

Lower Support Casting Lower Support Column 3

Assemblies - Lower Support Low 0.5 Max. Static < 202) Not Susceptible° Column Bodies Notes:

1. Where component-specific certified material test report data is not available, the ferrite content is calculated (per guidance of [4]) based on permitted variations in ASTM A351, Grade CF8 chemistry requirements. Allowable variants of Grade CF8 chemistry requirements may result in ferrite content greater than 20%; thus, the ferrite content is identified as potentially exceeding 20%.

2. The elemental percentages from the chemical data retrieved from certified material test reports (CMTRs) are input into Hull's formula (per guidance of [4]) to calculate the delta ferrite content of the CASS material.

3. Conclusion is based on CMTR chemistry data.

4. Conclusion is based on material specification chemistry requirements.

Westinghouse Non-Proprietary Class 3 Attachment 4, Page 4 of 4 Our ref: LTR-RIAM-14-24, Rev. 1 October 3, 2014 4.2 References

1. Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A). EPRI, Palo Alto, CA: 2011. 1022863.

2. Materials Reliability,Program."Screening, Categorization,and Ranking of Reactor Internals Componentsfor Westinghouse and Combustion EngineeringPWVR Design (MRP-191). EPRI, Palo Alto, CA: 2006. 1013234.

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

4. U.S. Nuclear Regulatory Commission, NUREG/CR-4513, Rev. 1, "Estimation of Fracture Toughness of Cast Stainless Steels during Thermal Aging in LWR Systems," May 1994 (NRC ADAMS Accession No. ML052360554).

ENCLOSURE 7 TO AEP-NRC-2014-60 Status of Regulatory Commitments The following table identifies the status of actions committed to by Indiana Michigan Power Company by Letter from J. P. Gebbie, Indiana Michigan Power Company (I&M), to U. S. Nuclear Regulatory Commission (NRC), "Donald C. Cook Nuclear Plant Units 1 and 2, Transmittal of Reactor Vessel Internals Aging Management Program," dated October 1, 2012, Agencywide Documents Access and Management System (ADAMS) Accession No. ML12284A320.

COMMITMENT SCHEDULED COMPLETION STATUS DATE This Action Item addresses the applicability of the FMECA Satisfied by this letter.

and functionality analysis assumptions made in the development of MRP-227-A to individual facilities. I&M is participating in Pressurized Water Reactor Owners Group Unit 1: October 25, 2014 (PWROG) project PA-MSC-0938, "Support for Applicant Action Items 1, 2, and 7 from the Final Safety Evaluation on MRP-227, Revision 0" to address this item. Unit 2: December 23, 2017 The results of the evaluation will be provided to the NRC prior to the period of extended operation for each unit.

This Action Item requires the licensee to verify that all the Satisfied by this letter.

reactor vessel internals (RVI) components withi6 the scope for license renewal at that facility have been considered in applicable documents in development of MRP-227-A. I&M is participating in PWROG project PA-MSC-0938, "Support for Applicant Action Items 1, 2, and 7 from the Final Safety Evaluation on MRP-227, Revision 0" to address this item. Unit 2: December 23, 2017 The results of the evaluation will be provided to the NRC prior to the period of extended operation for each unit.

to AEP-NRC-2014-60 Page 2 COMMITMENT SCHEDULED COMPLETION STATUS DATE COMMITMENT SCHEDULED COMPLETION Status DATE CNP Unit 1 and Unit 2 both have X-750 split pins. Project Satisfied by Letter from J. P. Gebbie, requests have been initiated to investigate split pin I&M, to NRC, "Donald C. Cook replacement for each unit. Nuclear Plant Units 1 and 2 - First Response to Request for Additional I&M will provide the NRC with the strategy for managing Information Concerning the Reactor split pins prior to the period of extended operation for each Vessel Internals Aging Management unit. Unit 2: December 23, 2017 Program," dated July 30, 2014, ADAMS Accession No.

MIL14216A497 CNP Unit 1 and Unit 2 both have 304 SS hold down Remains Open springs. MRP-227-A guidance includes physical Unit 1: Prior to the first measurement of 304 SS hold down springs. This action required physical item requires acceptance criteria to be provided to the measurement.

NRC. CNP plant specific acceptance criteria will be developed and submitted to the NRC prior to the first required physical measurement. The hold down springs Unit 2: Prior to the first will be replaced if acceptance criteria are not developed in required physical lieu of performing the first required physical measurement. measurement.

to AEP-NRC-2014-60 Page 3 COMMITMENT SCHEDULED COMPLETION STATUS DATE A plant specific evaluation of RVI cast austenitic stainless Partially satisfied by this letter.

steel (CASS) materials is required in this Action Item. I&M Remaining portion will be addressed is participating in PWROG project PA-MSC-0938, "Support Unit 1: October 25, 2014 by the commitment provided in for Applicant Action Items 1, 2, and 7 from the Final Safety Enclosure 8 of this letter.

Evaluation on MRP-227, Revision 0" to address this item The results of the evaluation will be provided to the NRC Unit 2: December 23, 2017 prior to the period of extended operation for each unit.

Enclosure 8 to AEP-NRC-2014-60 REGULATORY COMMITMENT The following table identifies an action committed to by Indiana Michigan Power Company (I&M) in this document. Any other actions discussed in this submittal represent intended or planned actions by I&M. They are described to the U. S. Nuclear Regulatory Commission (NRC) for the NRC's information and are not regulatory commitments.

Commitment Date I&M will continue to participate in the Pressurized Water Reactor March 31, 2015 Owners Group (PWROG) project for lower support columns.

I&M will provide a supplemental response to the NRC on this request for additional information when an acceptable methodology is developed by the PWROG project.