ML12188A004

Reactor Vessel Internals Inspection Plan Review Request Supplement and Response to Request for Additional Information

ML12188A004
Person / Time
Site:	Kewaunee
Issue date:	06/28/2012
From:	Price J Dominion, Dominion Energy Kewaunee
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
12-295
Download: ML12188A004 (53)

Text

f 4eDominion Dominion Energy Kewaunee, Inc.

5000 Dominion Boulevard, Glen Allen, VA 23060 f

June 28, 2012 ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 Serial No.12-295 LIC/JG/RO Docket No.: 50-305 License No.: DPR-43 DOMINION ENERGY KEWAUNEE, INC.

KEWAUNEE POWER STATION REACTOR VESSEL INTERNALS INSPECTION PLAN REVIEW REQUEST SUPPLEMENT AND RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION By application dated December 12, 2011 (Reference 1), Dominion Energy Kewaunee, Inc. (DEK), requested approval, pursuant to the provisions of Renewed Operating License DPR-43, of the inspection plan for reactor vessel internal (RVI) components at Kewaunee Power Station (KPS). This inspection plan submittal was to fulfill certain requirements of Renewed Operating License DPR-43, Section 2.C(15)(b); specifically, Commitment Items 1 and 2 of Appendix A of NUREG-1958, "Safety Evaluation Report Related to the Kewaunee Power Station," dated January 2011.

Subsequently, the Nuclear Regulatory Commission (NRC) transmitted a request for additional information (RAI) regarding the inspection plan (References 2, 3, and 4). The NRC questions were discussed with NRC staff to obtain clarification during a telephone conference on April 19, 2012. The DEK responses to the NRC questions are provided in Attachment 1 to this letter. provides revised RVI components inspection plan tables.

If you have questions or require additional information, please feel free to contact Mr.

Jack Gadzala at 920-388-8604.

Very truly yours,

.an Price Vice President - Nuclear Engineering

Serial No.12-295 Page 2 of 2 Attachments:

1. Response to Request for Additional Information, KPS RVI Inspection Plan

2. KPS RVI Inspection Plan, Tables 1 and 2 (Revision 1)

References:

1. Letter from J. Alan Price (DEK) to Document Control Desk (NRC), "Reactor Vessel Internals Inspection Plan Review Request," dated December 12, 2011.

2. Email from Karl D. Feintuch (NRC) to Jack Gadzala (DEK) et al, "ME7727 -

Kewaunee - Review of RVI Inspection Plan RAI Set 1 of 2 (Rev 1)," dated March 28, 2012.

3. Email from Karl D. Feintuch (NRC) to Jack Gadzala (DEK) et al, "ME7727 (not 7277) - Kewaunee - Review of RVI Inspection Plan - (1) RAI Set 2 of 2, adding to RAI Set 1 of 2 (Rev 1); and (2) request for clarification conference call," dated April 9, 2012.

4. Email from Karl D. Feintuch (NRC) to Jack Gadzala (DEK) et al, "ME7727 Kewaunee Request for Additional Information (RAI) ME7727-RAII-EVIB-Cher-014-2012-04-27," dated April 11, 2012.

Commitments made by this letter:

1. If evaluations of the CASS materials (discussed in the response to NRC Question ME7727-RAII-EVIB-Cher-012-2012-05-09) require implementation of enhanced visual testing (EVT-1), then the results of these examinations of the CASS materials will be provided to the NRC.

cc:

Regional Administrator, Region III U. S. Nuclear Regulatory Commission 2443 Warrenville Road Suite 210 Lisle, IL 60532-4352 Mr. Karl D. Feintuch Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 08-H4A 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Kewaunee Power Station

Serial No.12-295 ATTACHMENT 1 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REACTOR VESSEL INTERNALS INSPECTION PLAN REVIEW REQUEST KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

I Serial No.12-295 Page 1 of 24 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION INSPECTION PLAN FOR THE AUGMENTED INSERVICE INSPECTION PROGRAM FOR EXAMINATION OF REACTOR VESSEL INTERNALS On March 28, April 9, and April 11, 2012, the NRC transmitted to Dominion Energy Kewaunee, (DEK) a request for additional information (RAI) (References 1, 2, and 3) concerning the inspection plan for reactor vessel internal (RVI) components at Kewaunee Power Station (KPS).

This inspection plan submittal was to fulfill certain requirements of Renewed Operating License DPR-43, Section 2.C(15)(b); specifically, Commitment Items 1 and 2 of Appendix A of NUREG-1958, "Safety Evaluation Report Related to the Kewaunee Power Station," dated January 2011.

These questions were discussed with NRC staff to obtain clarification during a telephone conference on April 19, 2012.

Each of the RAI questions are provided below, followed by the corresponding DEK response.

NRC Question ME7727-RAII-EVIB-Cher-001-2012-04-27 Applicant/Licensee Action Item 1 from the NRC staffs final SE of MRP-227-A, "Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines,"

requires that applicants/licensees submit an evaluation that demonstrates that their plant is bounded by the assumptions regarding plant design and operating history that were made in the failure modes, effects and consequences analyses (FMECA) and functionality analyses for reactors of their design.

KPS's response to Applicant/Licensee Action Item 1 in the RVI inspection plan addresses the core loading assumptions (switch to a low-leakage core) and operational (base loaded plant) aspects of design and operation that are mentioned in MRP-227-A, Section 2.4. An additional assumption listed in Section 2.4 of MRP-227-A is that there have been no design changes to the RVI beyond those identified in general industry guidance or recommended by the original vendors. Section 2.4 of MRP-227-A indicated that these assumptions are considered to represent any U.S PWR operating plant provided that these three assumptions are met, given the information on design and operation known to the MRP as of May 2007.

MRP-191, Revision 0, "Materials Reliability Program: Screening, Categorization and Ranking of Reactor Internals of Westinghouse and Combustion Engineering PWR

[pressurized water reactor] Designs" (proprietary), documents the screening for susceptibility to aging effects, the FMECA results, and the categorization and ranking of the RVI components. In addition to the assumptions listed in Section 2.4 of MRP-227-A, MRP-191 documents additional assumptions that were used. In particular, neutron

Serial No.12-295 Page 2 of 24 fluence range, temperature, and material grade for each generic component of the Westinghouse design internals were used for input to the screening process. These values were determined based on an "expert elicitation" process. Stress values were not explicitly tabulated, but were recorded as either above the stress threshold (>30 ksi) or not based on the expert interviews.

MRP-232, Revision 0, "Materials Reliability Program: Aging Management Strategies for Westinghouse and Combustion Engineering PWR Internals" (proprietary), reported more specific stress, temperature and neutron fluence values based on finite element analyses for selected high consequence of failure components identified in MRP-1 91.

The EPRI-MRP did not verify that the values of fluence, temperature, stress, and material, documented in MRP-191 and MRP-232 were bounding for all individual plants, and in fact MRP-227 states, "These evaluations were based on representative configurations and operational histories, which were generally conservative, but not necessarily bounding in every parameter."

The NRC staff expects that the licensee should have access to design information enabling verification that the material for each RVI component is bounded by the design assumptions of the MRP.

In this context, the NRC staff requests that the licensee provide the following information:

1) To provide reasonable assurance that the RVI components are bounded by assumptions in the FMECA and functionality analyses supporting the development of MRP-227-A, the licensee is requested to respond to either part a) or part b) of this RAI:

a) Provide the plant-specific values of neutron fluence (n/cm 2, E>11.0 MeV),

temperature, stress, and materials for a sample of RVI components. The components selected should represent a range of neutron fluences, and temperatures. This information should identify whether the stress is greater or less than 30 ksi. Values of neutron fluence and temperature may be estimated or analytical values. The values should be the peak values of each parameter for each component (e.g., peak end-of-life value for fluence). Provide the method used to estimate the values, or describe the analysis method. An acceptable sample of components is:

i) Lower Core Plate ii) Core Barrel Flange iii) Barrel-Former Bolts iv) Upper Core Barrel Welds v) Lower Core Barrel Welds vi) Upper Core Plate Alignment Pins

4 Serial No.12-295 Page 3 of 24 b) If the sample verification approach in Part a) is not used, describe the process used to verify that the RVI components at KPS are bounded by the assumptions regarding the neutron fluence, temperature, stress values, and materials that were made for each component in the FMECA and functionality analyses supporting the development of MRP-227-A.

2) If there are any components at KPS not bounded by assumptions regarding neutron fluence, temperature, stress or material used in the development of MRP-227-A, describe how the differences were addressed in the plant-specific RVI Inspection Plan. The NRC staff requests that the licensee, as a part of its demonstration, discuss whether there would be any changes to the screening, categorization, FMECA process and functionality analyses if the plant-specific variables (the neutron fluence, temperature, stress values, plant-specific operating experience, and materials) are used. This evaluation should address whether additional aging mechanisms would become applicable to the component.

3) For any non-bounded components, determine if any changes to the inspection requirements of MRP-227-A are needed.

Provide plant-specific inspection requirements or an alternate aging management program, as appropriate. If no changes to the inspection requirements are proposed, provide a justification for the adequacy of the existing MRP-227-A inspections for the unbounded components.

4) In its e-mail submittal dated March 19, 2012, the licensee submitted the Technical Report KLR-1309A, to the NRC staff for review. On page 13 of the KLR-1309A report, the licensee stated that as a part of design change, it installed flexure-less inserts. The NRC staff requests that the licensee provide the following information with regard to this design change.

a)

Reason for installing the flexure-less inserts, (b) Type of material used, and information regarding the material selection, (c) Operating experience with respect to any degradation (observed so far) of the flexure-less inserts, and (d)

If the flexure-less inserts were installed after May 2007, provide an assessment of the impact of this installation on the recommendations of the RVI Inspection Plan.

Provide plant-specific inspection requirements if necessary for the affected components.

Response

RAI Part 1)

DEK has confirmed information regarding plant specific internals components for KPS (in conjunction with Westinghouse Electric Company (Westinghouse)). The results of the favorable comparison of the KPS specific components to those components used in the generic FMECA and functionality analysis (MRP-191 Table 4-4) are documented in Westinghouse letter WPS-08-28, Revision 2, dated December 22, 2008 (LTR-ARIDA-08-63, Revision 3).

t Serial No.12-295 Page 4 of 24 The process used to verify that the RVI components at KPS are bounded by the assumptions regarding neutron fluence, temperature, stress values, and materials in the FMECA and functionality analyses is as follows:

1. Identification of Typical Westinghouse PWR Internals Components.

2. Identification of KPS PWR Internals Components.

3. Comparison of the Typical Westinghouse PWR Internals Components to the KPS PWR Internals Components. No atypical items were identified by this comparison.

4. The materials are identified in WPS-08-28, Revision 2 (LTR-ARIDA-08-63, Revision 3, dated December 5, 2008) and are consistent with those materials identified in MRP-191 and MRP-232.

5. KPS verifies through AMP-KLR-1309, Revision 3 that no modification to the KPS reactor internals have been made over the lifetime of the plant except for those specifically directed by the OEM.

Therefore, the assumptions regarding the neutron fluence, temperature, stress values, and materials used in performing the generic FMECA and functionality analysis are applicable for KPS.

6. KPS used new fuel at the peripheral locations during Fuel Cycles 1 and 2. Once-burned fuel was used at peripheral locations during Fuel Cycles 3 through 15. In Fuel Cycle 16, KPS switched to use of a low leakage core design. KPS continued to use a low leakage core design for all subsequent fuel cycles. KPS has operated under base load conditions over the life of the plant. The FMECA and functionality analyses for MRP-227 were based on the assumption of thirty years of operation with high leakage core loading patterns followed by thirty years of low leakage core loading patterns., Therefore, KPS satisfies the assumptions in MRP regarding fluence.

7. KPS has operated under base load conditions over the lifetime of the plant.

Therefore, KPS satisfies the assumption in MRP documents regarding operational parameters.

8. The KPS reactor vessel materials operate at temperatures between TcoId and Thot that are approximately not less than 525 °F for TcoId nor higher than 611 OF for Thot.

The design temperature for the KPS reactor vessel is 650 OF.

9. The Westinghouse design has been maintained by the owner over the lifetime of the plant and is effectively represented by the stresses assumed in MRP-191 and MRP-232.

10. Design changes to the KPS reactor vessel internals components have been limited to Westinghouse replacement of split pins, installation of flexure-less inserts, and replacement of the reactor vessel head.

11.

MRP-191 and MRP-232 was used to organize, characterize, and rank the reactor vessel internals parts into the various groups for development of aging management strategies:

Existing, Primary, Expansion, and No Additional Measures.

Serial No.12-295 Page 5 of 24 NRC requested that DEK qualitatively address how the input parameters used in the FMECA and functionality analyses are bounded by the assumptions in MRP-227-A.

Table A (below) illustrates the input parameters for the typical Westinghouse PWR internals compared to those used for the KPS internals for the six sample components identified in the NRC question above.

Serial No.12-295 Page 6 of 24 Table A Review of sample components RAI Description Parameters Item Neutron Fluence Temperature Stress Materials Typical1 KPS 2 Typical KPS3 Typical KPS Typical KPS Plant Plant Plant Plant Lower Core Reference Same as Reference Same as Reference Same as Reference Reference Plate MRP-191, MRP-191, MRP-191, MRP-191, MRP-191, MRP-191, MRP-191, LTR-ARIDA-08-63, Table 4-6, Table 4-6, Table 4-6, Table 4-6, Table A-i, Results Table A-i, Table 4-6, Rev. 3, dated Screening Input Screening Input Screening Input Screening Input of Parameter Results of Screening Input December 5, 2008.

Parameters for Parameters for Parameters for Parameters for Screening and Parameter Parameters for Westinghouse-Westinghouse-Westinghouse-Westinghouse-Interviews with Screening and Westinghouse-Designed Plants Designed Plants Designed Designed Plants.

Analysts-Interviews with Designed Plants.

Westinghouse Analysts-Plants.

Estimated Estimated Reactor Internals Westinghouse Fluence Fluence

> 608°F

> 608°F Reactor Internals 304 SS 304 SS Range Range (n/cm2, (n/cm2, Effective Stress Effective Stress E > 1 MeV)

E > 1 MeV)

> Threshold

-> Threshold 1 x 1022 to 1 x 1022 to 5 x 1022 5 x 1022 bounds KPS Core Barrel Reference Same as Reference The KPS reactor Reference Same as Reference Reference Flange MRP-161, MRP-161, MRP-161, vessel materials MRP-191, MRP-191, MRP-191, LTR-ARIDA-08-63, Table 4-6, Table 4-6, Table 4-6, operate at Table A-i, Results Table A-i, Table 4-6, Rev. 3, dated Screening Input Screening Input Screening Input temperatures of Parameter Results of Screening Input December 5, 2008.

Parameters for Parameters for Parameters for between TCOd and Screening and Parameter Parameters for Westinghouse-Westinghouse-Westinghouse-Thot that are Interviews with Screening and Westinghouse-Designed Plants Designed Plants Designed approximately Analysts-Interviews with Designed Plants.

not less than Westinghouse Analysts-Plants.

Estimated Estimated 525 *F for T.u Reactor Internals Westinghouse Fluence Fluence T-hot nor higher than Reactor Internals 304 SS 304 SS Range Range 611 *F for Tho.

(n/cm2, (n/cm2, Effective Stress Effective Stress E > 1 MeV)

E > 1 MeV)

> Threshold

> Threshold

<1020

<1020 bounds KPS

Serial No.12-295 Page 7 of 24 Table A Review of sample components RAI' Description Parameters Item Neutron Fluence Temperature Stress Materials Typical1 KPS 2 Typical KPS 3 Typical KPS Typical KPS Plant Plant Plant Plant Barrel-Reference Same as Reference Same as Reference Same as Reference Reference Former MRP-161, MRP-161, MRP-161, MRP-161, MRP-191, Table MRP-191, Table MRP-191, WCAP-13266, Bolts Table 4-6, Table 4-6, Table 4-6, Table 4-6, A-i, Results of A-i, Results of Table 4-6, Rev. 1, Proprietary Screening Input Screening Input Screening Input Screening Input Parameter Parameter Screening Input Class 2 Parameters for Parameters for Parameters for Parameters for Screening and Screening and Parameters for Westinghouse-Westinghouse-Westinghouse-Westinghouse-Interviews with Interviews with Westinghouse-Designed Plants Designed Plants Designed Designed Plants.

Analysts-Analysts-Designed Plants.

Westinghouse Westinghouse Plants.

Estimated Estimated Reactor Internals Reactor Internals Fluence Fluence

> 608°F

> 608'F 316 SS Range Range Effective Stress Effective Stress or

> Threshold

> Threshold 347 SS 347 SS (n/cm2, (n/cm2, E>1MeV)

E>1MeV) 5x1022 5x1022 bounds KPS Upper Core Reference Same as Reference The KPS reactor Reference Same as Reference Reference Barrel MRP-161, MRP-161, MRP-161, vessel materials MRP-191, Table MRP-191, Table MRP-191, LTR-ARIDA-08-63 Welds Table 4-6, Table 4-6, Table 4-6, operate at A-i, Results of A-i, Results of Table 4-6, Rev. 3, dated Screening Input Screening Input Screening Input temperatures Parameter Parameter Screening Input December 5, 2008.

Parameters for Parameters for Parameters for between T.od and Screening and Screening and Parameters for Westinghouse-Westinghouse-Westinghouse-Thot that are Interviews with Interviews with Westinghouse-Designed Plants Designed Plants Designed approximately Analysts-Analysts-Designed Plants.

not less than Westinghouse Westinghouse Plants.

Estimated Estimated 525 *F for T=Id Reactor Internals Reactor Internals Fluence Fluence T-hot nor higher than 304 SS 304 SS Range Range 611 *F for Thor.

Effective Stress Effective Stress a Threshold

> Threshold (n/cm2, E > 1 (n/cm2, E > 1 MeV)

MeV)

<1020

< 102o bounds KPS

Serial No.12-295 Page 8 of 24 Table A Review of sample components RAI Description Parameters Item Neutron Fluence Temperature Stress Materials Typical1 KPS 2 Typical KPS 3 Typical KPS Typical KPS Plant Plant Plant Plant Lower Core Reference Same as Reference The KPS reactor Reference Same as Reference Reference Barrel MRP-161, MRP-161, MRP-161, vessel materials MRP-191, Table MRP-191, Table MRP-191, LTR-ARIDA-08-63 Welds Table 4-6, Table 4-6, Table 4-6, operate at A-i, Results of A-i, Results of Table 4-6, Rev. 3, dated Screening Input Screening Input Screening Input temperatures Parameter Parameter Screening Input December 5, 2008.

Parameters for Parameters for Parameters for between T.,d and Screening and Screening and Parameters for Westinghouse-Westinghouse-Westinghouse-Thot that are Interviews with Interviews with Westinghouse-Designed Plants Designed Plants Designed approximately not Analysts-Analysts-Designed Plants.

less than 525 'F Westinghouse Westinghouse Plants.

Estimated Estimated for T_,d nor Reactor Internals Reactor Internals Fluence Fluence T-hot higher than 611 304 SS 304 SS Range Range

• F for Too.

Effective Stress Effective Stress

> Threshold

Ž Threshold (n/cm2, (n/cm2, E > 1 MeV)

E > 1 MeV)

< 1020

< 1020 bounds KPS Upper Core Reference Same as Reference The KPS reactor Reference Same as Reference Reference Plate MRP-161, MRP-161, MRP-161, vessel materials MRP-191, Table MRP-191, Table MRP-191, LTR-ARIDA-08-63 Alignment Table 4-6, Table 4-6, Table 4'-6, operate at A-i, Results of A-i, Results of Table 4-6, Rev. 3, dated Pins Screening Input Screening Input Screening Input temperatures Parameter Parameter Screening Input December 5, 2008.

Parameters for Parameters for Parameters for between Tcld and Screening and Screening and Parameters for Westinghouse-Westinghouse-Westinghouse-Thor that are Interviews with Interviews with Westinghouse-Designed Plants Designed Plants Designed approximately Analysts-Analysts-Designed Plants.

not less than Westinghouse Westinghouse Plants.

Estimated Estimated 525 °F for Tc.Id Reactor Internals Reactor Internals Fluence Fluence T-hot nor higher than 304 SS 304 SS Range Range 611 °F for Thor.

Effective Stress Effective Stress

> Threshold

> Threshold (n/cm2, (n/cm2, E > 1 MeV)

E > 1 MeV) 7 x 1020 to 7 x 1020 to 1 x 1021 1 x 1021 bounds KPS

Serial No.12-295 Page 9 of 24 Table A Review of sample components RAI Description Parameters Item Neutron Fluence Temperature Stress Materials Typical' KPS2 Typical KPS 3 Typical KPS Typical KPS Plant Plant Plant Plant Notes

1)

Assumed 30 years of high leakage core loading followed by 30 years of low leakage core loading.

2)

Fuel Cycles 1 and 2. Once-burned fuel was used at peripheral locations during Fuel Cycles 3 through 15. In Fuel Cycle 16, KPS switched to use of a low leakage core design. KPS continued to use a low leakage core design for all subsequent fuel cycles.

3)

The KPS reactor vessel materials operate at temperatures between T.1d and Thot that are approximately not less than 525 *F for T.od nor higher than 611 *F for Tho,. The design temperature for the KPS reactor vessel is 650 *F.

4)

Criteria for material, temperature, and fluence are listed in MRP-1 91, Table 4-6, Screening Input Parameters for Westinghouse-Designed Plants.

5)

Criteria for stress is depicted in MRP-191, Table 3-2, Irradiation Assisted Stress Corrosion Cracking (IASCC) Screening Criteria and MRP-191, Figure 3-1, MRP-175 Screening Criteria for IASCC.

6)

Criteria for SCC is listed in MRP-1 91, Table 3-1, Stress Corrosion Cracking (SCC) Screening Criteria for PWR Internals Materials.

Serial No.12-295 Page 10 of 24 RAI Part 2)

All of the subject RVI components at KPS are representative of assumptions regarding neutron fluence, temperature, stress or material used in the development of MRP-227-A. WPS-08-28, Revision 2 (LTR-ARIDA-08-63, Revision 3) documents the materials and compares the KPS components to those in a list of typical Westinghouse PWR internals components. The KPS components, materials, and design are consistent with those identified as Typical Westinghouse PWR Internals Components.

KLR-1309-A documents compliance to the fluence requirements and specific design changes to the reactor vessel internals implemented by the OEM over the lifetime of the plant, all of which are addressed by MRP-191, MRP-232, and MRP-227-A. The reactor vessel internals operate between Thot and Tcold which are approximately not less than 525 OF for Tcold nor higher than 611 °F for Thot. Because the design has been maintained over the lifetime of the plant and the materials are the same as those identified on MRP-191, MRP-232, and MRP-227-A the stress values are bounded by the assumptions in MRP-191, MRP-232, and MRP-227-A.

RAI Part 3)

The KPS reactor vessel internals components are bounded by the typical Westinghouse PWR internals components outlined in MRP-227-A and the applicable referenced documents, including MRP-191 and MRP-232.

The KPS reactor vessel internals inspection program was written to comply with MRP-227, Revision 0 and the seven (7) topical report conditions and eight (8) licensee action items identified in the NRC SER issued on June 22, 2011. Together, the requirements outlined by MRP-227, Revision 0, along with the seven (7) topical report conditions and eight (8) licensee action items from the NRC SER issued on June 22, 2011, satisfy the requirements specified in MRP-227-A.

No changes in the inspection requirements are being proposed at this time because the KPS inspection program complies with MRP-227-A as indicated above.

The additional notes identified in NRC Questions ME7727-RAII-EVIB-Cher-004-2012-05-09, ME7727-RAII-EVIB-Cher-009-2012-05-09, ME7727-RAII-EVIB-Cher-010-2012-05-09, and ME7727-RAII-EVIB-Cher-013-2012-05-09 have been added to the inspection tables.

RAI Part 4)

Flexure-less inserts have been installed at KPS in response to cracking issues identified in the commercial nuclear power industry. The flow restrictors on the 29 RCCA guide tubes were replaced under a design change (DCR 2030) in 1989. The flow restrictors on the four (4) plutonium recycle port guide tubes were replaced under a design change (DCR 1844) in 1986.

KPS did not experience cracking of the original flexures. The replacement flexure-less insert is fabricated from subparts:

Serial No.12-295 Page 11 of 24

" Latch housing (1) - 304 SS

" Latch retainer (1) - 304 SS Latches (3)- 316 SS Spring (1)-

718 inconel

" Dowel pins (3) - 316 SS Spring retainer (1) 304 SS KPS has not experienced any problems with the replacement flexure-less inserts. The replacement flexure-less inserts were removed and reinstalled in 2004 to facilitate replacement of the split pins. The replacement flexure-less inserts were also removed and reinstalled for the 29 RCCA active guide tubes to support the guide card inspection activity performed during the spring 2012 refueling outage.

NRC Question ME7727-RAII-EVIB-Cher-002-2012-04-27 Applicant/Licensee Action Item 2, Section 3.2.5.2, from the NRC staffs final SE of MRP-227, Rev.1, requires the following:

"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 license renewal (LR) for its facility. Applicants/licensees shall review the information in Tables 4-1 and 4-2 in MRP-1 89, Revision 1, "Materials Reliability Program: Screening, Categorization, and Ranking of B&W-Designed PWR Internals," 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 their facilities 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 identify the missing component(s) and propose any necessary modifications to the program defined in MRP-227-A, as modified by this SE, when submitting its plant-specific AMP such that the effects of aging on the missing component(s) will be managed for the period of extended operation."

On page 5, license action item-2 of its submittal dated December 12, 2011, the licensee stated that Westinghouse classified various RVI components based on their susceptibility to aging degradation. The NRC staff requests that the licensee confirm that Westinghouse complied with the aforementioned requirement in its entirety while performing scoping and screening for the license renewal for the KPS.

Response

WPS-08-28, Revision 2 (LTR-ARIDA-08-63, Revision 3) tabulates the components in the KPS reactor vessel internals and compares them to a list of the typical Westinghouse PWR internals components.

LTR-ARIDA-08-63 used MRP-191 for

Serial No.12-295 Page 12 of 24 categorization, analysis, and aging management strategy development.

In addition, scoping and screening of the reactor vessel internals was performed by DEK and documented in Technical Report KLR-1 104, License Renewal Project Integrated Plant Assessment Report Reactor Vessel, Internals, and Reactor Coolant System Kewaunee Power Station. There are no reactor vessel internal components included in Technical Report KLR-1 104 that are not covered by LTR-ARIDA-08-63.

NRC Question ME7727-RAII-EVIB-Cher-003-2012-05-09 Contrary to the requirements addressed in the MRP-227-A report, the licensee did not include the following components in its submittal dated December 12, 2011:

(a)

Lower core barrel flange shall be placed under a "Primary" inspection category as addressed in Table 4-3 of the MRP-227-A report.

(b)

Upper and lower core barrel cylinder axial welds shall be included under an "Expansion" inspection category as addressed in Table 4-6 of the MRP-227-A report.

Response

The inspection plan was revised to include the lower core barrel flange weld (i.e. lower core barrel girth weld) under a "Primary" inspection category as addressed in Table 4-3 of the MRP-227-A report. The revised Reactor Vessel Internals Inspection Plan, Table 1, showing this change is provided in Attachment 2 to the letter transmitting this response.

The inspection plan was revised to include the upper and lower core barrel cylinder axial welds under an "Expansion" inspection category as addressed in Table 4-6 of the MRP-227-A report. The revised Reactor Vessel Internals Inspection Plan, Table 2, showing these changes is provided in Attachment 2 to the letter transmitting this response.

NRC Question ME7727-RAII-EVIB-Cher-004-2012-05-09 Condition 4 of the NRC staff's SE Revision 1, dated December 16, 2011 requires that the licensee shall include inspection coverage for the RVI components. The NRC staff noted that the licensee did not include the inspection coverage for the following RVI components - (i) the control rod guide tube (CRGT) lower flange welds; (ii) For core barrel baffle-former bolts under the "Primary" inspection category; and (iii) core barrel baffle-former bolts under the "Expansion" inspection category.

Serial No.12-295 Page 13 of 24 (a)

For CRGT lower flange welds, in Table 1 on page 8 of the December 12, 2011, submittal, a footnote shall be added consistent with the note 2 in Table 4-3 of the MRP-227-A report.

(b)

For core barrel baffle-former bolts in Table 1 on page 12 of the December 12, 2011, submittal, a footnote shall be added consistent with the note 3 in Table 4-3 of the MRP-227-A report.

(c)

For core barrel baffle-former bolts in Table 2 on page 1 of the December 12, 2011, submittal, a footnote shall be added consistent with the note 2 in Table 4-6 of the MRP-227-A report.

Response

(a)

A footnote was modified in Table 1 for the CRGT lower flange welds, consistent with Note 2 in Table 4-3 of the MRP-227-A report. The footnote was modified to add the following statement:

"A minimum of 75% of the total identified sample population must be examined."

(b)

A footnote was added to Table 1 for the core barrel baffle-former bolts under the "Primary" inspection category, consistent with Note 3 in Table 4-3 of the MRP-227-A report.

The footnote states:

"A minimum of 75% of the total population (examined + unexamined), including coverage consistent with the Expansion criteria in Table 5-3, must be examined for inspection."

(c)

A footnote was added to Table 2 for the core barrel former baffle-former bolts under the "Expansion" inspection category, consistent with Note 2 in Table 4-6 of the MRP-227-A report. The footnote states: "A minimum of 75% coverage of the entire examination area or volume, or a minimum sample size of 75% of the total population of like components of the examination is required (including both the accessible and inaccessible portions)."

The revised tables are provided in Attachment 2 of the letter transmitting this response.

NRC Question ME7727-RAII-EVIB-Cher-005-2012-05-09 Condition 7 of the NRC staff's SE Revision 1, dated December 16, 2011 requires that the licensee shall include a summary of the operating experience related to the aging degradation in the RVI components. The NRC staff requests that the licensee provide information regarding the extent of aging degradation (if any) that occurred thus far in all of the RVI components specifically, include the operating history of the following components at KPS:

Serial No.12-295 Page 14 of 24 (a)

Baffle-former bolts, baffle-edge bolts, clevis insert bolts, flux thimble tubes, core barrel bolting, and thermal shields.

Response

On April 19, 2010, the PWROG conducted a reactor internals operating experience survey under project authorization PA-MSC-0568. The survey included baffle-former bolts, baffle-edge bolts, clevis insert bolts, flux thimble tubes, thermal shield flexures, and core barrel assembly barrel-former bolts. The survey did not include the thermal shield.

Information included in the RAI response is the same as provided to the PWROG for the baffle-former bolts, baffle-edge bolts, clevis insert bolts, flux thimble tubes, and core barrel assembly barrel-former bolts.

KPS has inspected portions of the reactor vessel internals 10 times (during the following years): 1976, 1980, 1985, 1988, 1995, 1996, 1998, 2004, 2006, and 2009.

VT-3 inspection has been performed on a portion of the baffle-former bolts during refueling outages KR1, KR10, KR13, KR20, KR26, KR27, and KR29.

No recordable indications were noted.

A review of the annual inservice inspection (ISI) reports was performed to ascertain whether the edge bolts have been inspected to VT-3 criteria. The data sheets did not use the term edge bolts. However, portions of the baffles were inspected, and the edge bolts are part of the baffles. Therefore, evidence exists that portions of the edge bolts were inspected to VT-3 criteria concurrent with inspection of the baffle-former bolts during refueling outages KR1, KR10, KR13, KR20, KR26, KR27, and KR29.

No recordable indications were noted.

VT-3 inspections were performed on the clevis insert bolts during refueling outages KR10, KR20, and KR26. No recordable indications were found.

Eddy current inspections were performed on the flux thimble tubes during refueling outages KR19, KR20, KR21, KR23, and KR26. The flux thimble tubes were inspected under procedure ER-AA-NDE-ET-501.

Wear was observed in seven (7) flux thimble tubes during refueling outage KR19. The thimble tubes were replaced in 1988 due to wear. A baseline inspection was performed in 1988. The thimble tubes were inspected again in 1993, 1994, 1995, 1996, 2000, 2004, and 2009. Between the 1993 and 1994 inspections, wear damage was observed on seven tubes. Tubes with the most severe damage were repositioned to move the wear scars away from the lower nozzle.

Subsequent inspections in 1995, 1996, 2000, 2004, and 2009 indicated no significant wear progression and no additional repositioning was recommended.

VT-3 inspections were performed on a portion of the core barrel assembly barrel-former bolts during refueling outages KR1, KR10, KR13, KR 20, KR26, KR27, and KR29. No recordable indications were found.

Serial No.12-295 Page 15 of 24 VT-3 inspection has been performed on portions of the thermal shield during refueling outages KR10, KR20, and KR26. No recordable indications were found.

NRC Question ME7727-RAII-EVIB-Cher-006-2012-05-09 Historically, the following materials used in the PWR RVI components were known to be susceptible to some of the aging degradation mechanisms that are identified in the MRP-227-A report. In this context, the NRC staff requests that the licensee confirm that these materials are not currently used in the RVI components at KPS.

(1)

Nickel base alloys - Inconel 600; Weld Metals - Alloy 82 and 182 and Alloy X-750 (2)

Alloy A-286 ASTM A 453 Grade 660, Condition A or B (3)

Stainless steel type 347 material (excluding baffle-former bolts)

(4) Precipitation hardened (PH) stainless steel materials 4 and 15-5 (5)

Type 431 stainless steel material

Response

Materials used in fabrication of the reactor internals are listed in WPS-08-28, Revision 2 (LTR-ARIDA-08-63, Revision 3, dated December 5, 2008). DEK reviewed the contents of this document to determine if any of the listed materials are used in RVI components.

The results are as follows.

(1) The following KPS reactor internals are fabricated from Nickel base alloys -

Inconel 600; Weld Metals - Alloy 82 and 182 and Alloy X-750: Clevis Insert Lock Keys (Alloy 600), Clevis Inserts (Alloy 600), and Clevis Insert Bolts (Alloy X-750).

(2)

No reactor internal components are listed in LTR-ARIDA-08-63, Revision 3 as being fabricated from alloy A-286, ASTM A 453, Grade 660, Condition A or B.

(3)

No reactor internal components are listed in LTR-ARIDA-08-63, Revision 3 as being fabricated from stainless steel type 347 material (excluding baffle-former bolts and barrel-former bolts). At KPS, the baffle-former bolts and barrel-former bolts are fabricated from type 347 materials.

(4)

No reactor internal components are listed in LTR-ARIDA-08-63, Revision 3 as being fabricated from precipitation hardened (PH) stainless steel materials 4 and 15-5.

(5)

No reactor internal components are listed in LTR-ARIDA-08-63, Revision 3 as being fabricated from type 431 stainless steel materials.

Serial No.12-295 Page 16 of 24 NRC Question ME7727-RAII-EVIB-Cher-007-2012-05-09 To verify that the licensee is in compliance with the implementation and control of the ten elements of the aging management program (AMP) addressed in GALL AMP XI.M16-A, "PWR Vessel Internals," the NRC staff requests that the licensee submit the KPS's AMP-KLR-1309A

report, Revision 3, "License Renewal Project, Aging Management Program, ASME Section XI, In-service Inspection, Subsection IWB, IWC, and IWD, Reactor Vessel Internals Inspection, Kewaunee Power Station," effective date September 30, 2011 as part of this review.

Response

KLR-1 309A report, Revision 3, was submitted to the NRC by letter dated April 2, 2012.

NRC Question ME7727-RAII-EVIB-Cher-008-2012-05-09 The licensee is required to inspect 20% of CRGT guide card assemblies per Table 4-3 on page 4-26 of the MRP-227-A report.

The NRC staff requests that the licensee provide an explanation how 6 out of 36 CRGT guide plate cards were selected. The explanation for the selection process should include the following aspects:

(a) most susceptible areas to experience aging degradation, (b) high stress areas, and, (c) plant-specific operating experience.

Response

There are a total of 37 control rod guide tubes at KPS (29 control rod guide tubes contain active drive rods; four (4) tubes are for plutonium recycle ports which do not contain rods; and four (4) tubes are for part length control rods which do not contain

.rods).

Because the four (4) plutonium recycle locations and the four (4) part length locations do not contain rods, and are not subject to wear, they are not part of the population of guide tubes subject to VT-3 inspection.

During the spring 2012 refueling outage (KR32), KPS performed a VT-3 inspection of the guide cards in all 29 of the control rod guide tubes with active drive rods. Various amounts of wear were observed in guide cards for all 29 of the control rod guide tubes.

Engineering analysis per the guidance in WCAP-17562-P, Revision 0, "Westinghouse Pressurized Water Reactor Internals Guide Tube Guide Card Wear Criteria," confirmed that operation of the control tubes is not impaired with wear in excess of 85% of the effective slot width opening (ESWO) for up to three (3) adjacent guide plates in a guide tube.

Engineering analysis indicates that wear is not projected to reach an unacceptable limit of 85% ESWO in three (3) adjacent guide plates in any guide tube within 20 effective full-power years (EFPY). A pattern analysis was performed, which

Serial No.12-295 Page 17 of 24 confirms that most of the observable wear occurs in guide cards located in the lower portion of the guide tube. A summary of the guide card hole locations that are projected to wear to 85% ESWO within 20 EFPY of the current inspection time is provided in Table B below.

Table B Guide Card Hole Locations - Projected wear to reach 85% within 20 EFPY Guide Tube Guide Card Hole Location EFPY to reach 85%

Constant Operation Curve Volumetric (Aligned High Flow Case)

B08 9

E5 27.6 18 D10 7

E5 30.3 19.9 D10 7

E2 27.6 18 G1l 9

B2 29.7 19.5 L06 7

B2 22.5 14.4 L06 8

B2 18.6 11.9 L06 9

B2 30.3 19.9 DEK plans one of the following options in 10 years: 1) select six (6) of the 29 active locations for inspection per MRP-227-A; or, 2) submit a deviation request to not inspect again in 10 years since no unacceptable wear was observed nor projected to occur within 20 EFPY from the date of the current inspection (2012). DEK anticipates that the next inspection would include the guide cards in guide tube assembly locations B08, D10, Gll, and L06.

NRC Question ME7727-RAII-EVIB-Cher-009-2012-05-09 Editorial corrections-The NRC staff requests that licensee include the following revisions in the December 12, 2011, submittal (a) _ Table 1, on page 9 in "Comments" column related to upper core barrel flange weld, last sentence should be revised to read-Expansion Link-"Lower Support Column Bodies." -

Reference-Table 4-3 on page 4-26 of the MRP-227-A report.

(b)

Table 1, on page 9 in "Comments" column related to core barrel girth weld, last sentence should be revised to read-Expansion Link-"Upper and Lower Core Barrel Cylinder Axial welds." --- Reference-Table 4-3 on page 4-26 of the MRP-227-A report.

(c)

Table 1, on page 10 in "Comments" column related to core barrel girth welds, last sentence should be revised to read-Expansion Link-"Upper and Lower Core

Serial No.12-295 Page 18 of 24 Barrel Cylinder Axial welds."

- Reference-Table 4-3 on page 4-26 of the MRP-227-A report.

(d)

Table 1, on page 13 in "Parts Examined" column related to core barrel baffle-former assembly should be revised to include description of the components (baffle plates, baffle-edge bolts, and former plates) which is consistent with Table 4-3 on page 4-28 in the MRP-227-A report.

Response

Table 1 on (originally numbered) pages 9, 10, and 13 has been revised as requested above. The revised Table 1 is provided in Attachment 2 to.the letter transmitting this response (the data that appeared on page 13 of the previous version of Table 1 now appears on page 14).

NRC Question ME7727-RAII-EVIB-Cher-010-2012-05-09 In Table 1, on page 10 of the December 12, 2011, submittal, the licensee stated that Type 347 stainless steel baffle-former bolts are used at KPS.

In Appendix A of the MRP-227-A report on page A-3, the MRP states that Type 347 stainless steel material is susceptible to irradiation-assisted stress corrosion cracking (IASCC). With respect to IASCC in these bolts, the NRC staff requests that the licensee address the following issues:

(a) The number of Type 347 bolts in the baffle-former bolt assembly that are classified under "Primary," and "Expansion," categories, and; (b)

The number of Type 347 bolts in the baffle-edge bolt assembly that are classified under" Primary," category, and; (c)

The percentage of Type 347 bolts in the aforementioned assemblies that were inspected thus far, at KPS and the results of the inspections.

Response

Table 1, under the column entitled Equipment No., indicates there are 728 baffle-former bolts (Primary) and 688 baffle-edge bolts (Primary).

Table 2, under the column Equipment No., indicates there are 344 barrel bolts (Expansion). These installed items are fabricated of type 347 material.

There are 688 baffle-edge bolts (Primary) fabricated of type 316 stainless steel installed at KPS. There are no type 347 stainless steel baffle-edge bolts installed at KPS.

MRP-227-A identifies ultrasonic examination (UT) as the required examination method for inspection of the baffle-former bolts (Primary) and barrel-former bolts (Expansion).

Serial No.12-295 Page 19 of 24 DEK has not performed an ultrasonic inspection of either the baffle bolts or barrel-former bolts at KPS.

Various portions of the baffle bolts or barrel-former bolts have been inspected to VT-3 criteria during refueling outages KR-1, KR-10, KR-13, KR-20, KR-26, KR-27, and KR-

29. No recordable indications were found during these VT-3 inspections.

NRC Question ME7727-RAII-EVIB-Cher-011-2012-05-09 In Table 1, on page 10 of the December 12, 2011, submittal, the licensee stated that the flux thimble tubes are examined every 5 year interval.

Provide an explanation for selecting this inspection frequency.

Response

The KPS response to NRC Bulletin 88-09, dated November 7, 1988, stated that the flux thimble tubes would be inspected at five year intervals.

The thimble tubes were replaced in 1988 due to wear.

A baseline inspection was performed in 1988. The thimble tubes were inspected again in 1993, 1994, 1995, 1996, 2000, 2004, and 2009. Between the 1993 and 1994 inspections, wear damage was observed on seven tubes. Tubes with the most severe damage were repositioned to move the wear scars away from the lower nozzle. The subsequent inspections in 1995, 1996, 2000, 2004, and 2009 indicated no significant wear progression and no additional repositioning was recommended.

Based on the inspection history, inspection of the flux thimble tubes at five year intervals remains appropriate.

NRC Question ME7727-RAII-EVIB-Cher-012-2012-05-09 On page 8 in Attachment 1 of the December 12, 2011 submittal, the licensee stated that cast austenitic stainless steel (CASS) materials that were classified under "No Additional Measures (NAM)" by the MRP will be evaluated for their susceptibility to thermal and neutron embrittlement.

RVI components under NAM classification were included in Table 4 of the December 12, 2011, submittal. After the review of Table 4, the NRC staff requests that the licensee provide following information.

(a) Provide the time frame for performing evaluations of the CASS materials per the criteria (fluence values, stress and delta ferrite) addressed on page 8 in of the December 12, 2011 submittal, and, (b) If the evaluations result in the implementation of enhanced visual testing (EVT-1) as a part supplemental examination addressed in Table 4 of the December 12,

Serial No.12-295 Page 20 of 24 2011, submittal, the licensee shall provide the results of the examinations of the CASS materials.

Response

DEK plans to complete the evaluations of the CASS materials prior to refueling outage KR34 or conduct the enhanced visual examination during refueling outage KR34. KR34 is projected to occur in spring 2015.

If evaluations of the CASS materials require implementation of enhanced visual testing (EVT-1), then the results of these examinations would be provided to the NRC.

NRC Question ME7727-RAII-EVIB-Cher-013-2012-05-09 In Tables 2 and 4 of the licensee's AMP KLR-1 309A, Revision 3, the staff noted several inconsistencies between the inspection and evaluation (I&E) guidelines that are addressed in the licensee's AMP and the MRP-227-A report.

The following table includes these inconsistencies for the various RVI components at KPS and the NRC staff requests that the licensee revise its AMP accordingly.

RVI Component -- AMP KLR-1309A, Inconsistencies with MRP-227-A Report I&E Revision 3 Guidelines CRGT Lower Flange Weld Note 2 in Table 4-3 of MRP-227-A is not included Upper Core Barrel Flange Weld Note 4 in Table 4-3 of MRP-227-A is not included Upper and Lower Core Barrel Cylinder These welds are not included (Reference Table 4-3 Girth Welds of MRP-227-A report)

Lower Core Barrel Flange Weld These welds are not included (Reference Table 4-3 of MRP-227-A report)

Baffle-Edge Bolts Note 3 in Table 4-3 of MRP-227-A is not included Baffle-Former Bolts Note 3 in Table 4-3 of the MRP-227-A is not included; Subsequent examination is required every 10 year interval (Table 4-3 of MRP-227-A)

Upper Core Plate This component is not included (Reference Table 4-6 of MRP-227-A report)

Lower Support Forging/Casting These components are not included (Reference Table 4-6 of MRP-227-A report)

Baffle-Former Bolts Note 2 in Table 4-6 of the MRP-227-A is not included; Subsequent examination is required every 10 year interval (Table 4-3 of MRP-227-A)

Lower Support Column Bolts Note 2 in Table 4-6 of the MRP-227-A is not included; Subsequent examination is required every 10 year interval (Table 4-3 of MRP-227-A)

Serial No.12-295 Page 21 of 24 RVI Component -- AMP KLR-1309A, Inconsistencies with MRP-227-A Report I&E Revision 3 Guidelines Lower Support Column Bodies (non-Note 2 in Table 4-6 of the MRP-227-A is not cast) included; Subsequent examination is required every 10 year interval (Table 4-3 of MRP-227-A)

Lower Support Column Bodies (cast)

Note 2 in Table 4-6 of the MRP-227-A is not included; Subsequent examination is required every 10 year interval (Table 4-3 of MRP-227-A)

Response

AMP KLR-1309A, Revision 3, was intended to provide guidance for developing the Reactor Vessel Internals Inspection Plan. Rather than revising AMP KLR-1309A, the corrected information was placed directly into Reactor Vessel Internals Inspection Plan, Tables 1 and 2. Table C below provides a listing of the specific changes made in response to this RAI item.

Table C Resolution of Inconsistencies RVI Component -- AMP KLR-1309A, Inconsistencies with MRP-227-A Resolution Revision 3 Report I&E Guidelines CRGT Lower Flange Weld Note 2 in Table 4-3 of MRP-227-A is not The following note has been added to included Table 1, for the lower flange welds: "A minimum of 75% of the total identified sample population must be examined."

Upper Core Barrel Flange Weld Note 4 in Table 4-3 of MRP-227-A is not The following note has been added to included Table 1 for the upper core barrel flange weld: "A minimum of 75% of the total weld length (examined + unexamined),

including coverage consistent with the Expansion criteria in Table 5-3, must be examined from either the inner or outer diameter for inspection credit."

Upper and Lower Core Barrel Cylinder These welds are not included (Reference DEK has verified that the upper core Girth Welds Table 4-3 of MRP-227-A report) barrel flange to core barrel weld is included in Table 1.

Lower Core Barrel Flange Weld These welds are not included (Reference There are 4 circumferential welds in the Table 4-3 of MRP-227-A report)

KPS core barrel: upper core barrel flange to core barrel weld, core barrel mid plane weld, core barrel lower mid plane weld and core barrel bottom weld. DEK has verified that all 4 circumferential welds are included in Table 1.

Baffle-Edge Bolts Note 3 in Table 4-3 of MRP-227-A is not The following note has been added to included Table 1 for the baffle-edge bolts: "A minimum of 75% of the total population (examined + unexamined), including coverage consistent with the Expansion criteria in Table 5-3, must be examined for inspection credit".

Serial No.12-295 Page 22 of 24 Table C Resolution of Inconsistencies RVI Component -- AMP KLR-1309A, Inconsistencies with MRP-227-A Resolution Revision 3 Report I&E Guidelines Baffle-Former Bolts Note 3 in Table 4-3 of the MRP-227-A is The following note has been added to not included; Subsequent examination is Table 1 for the baffle-former bolts: "A required every 10 year interval (Table 4-3 minimum of 75% of the total population of MRP-227-A)

(examined + unexamined), including coverage consistent with the Expansion criteria in Table 5-3, must be examined for inspection credit." KPS is adding a note that inspection of the baffle-former bolts is required at 10 year intervals.

Upper Core Plate This component is not included The upper core plate was verified to be in (Reference Table 4-6 of MRP-227-A KLR-1 309A Rev 3. The upper core plate report) is also listed in Table 2.

Lower Support Forging/Casting These components are not included The lower support forging was verified to (Reference Table 4-6 of MRP-227-A be in both LTR-ARIDA-08-63 Rev. 3 report) dated December 5, 2008 & KLR-1309A Rev 3. The lower support forging/casting is also in Table 2.

Barrel -Former Bolts Note 2 in Table 4-6 of the MRP-227-A is The following note has been added to not included; Subsequent examination is Table 1 for the baffle-former bolts:

required every 10 year interval (Table 4-3 "Subsequent examination is required of MRP-227-A) every 10 year interval."

Lower Support Column Bolts Note 2 in Table 4-6 of the MRP-227-A is The following note has been added to not included; Subsequent examination is Table 2 for the lower support column required every 10 year interval (Table 4-3 bolts: "A minimum of 75% coverage of the of MRP-227-A) entire examination area or volume, or a minimum sample size of 75% of the total population of like components of the examination is required (including both the accessible and inaccessible portions)." Re-inspection on a 10-year frequency is added to Table 2.

Lower Support Column Bodies (non-Note 2 in Table 4-6 of the MRP-227-A is The following note has been added to cast) not included; Subsequent examination is Table 2 for the Lower Support Column required every 10 year interval (Table 4-3 Bodies (non-cast): "A minimum of 75%

of MRP-227-A) coverage of the entire examination area or volume, or a minimum sample size of 75% of the total population of like components of the examination is required (including both the accessible and inaccessible portions)." Re-inspection on a 10-year frequency is added to Table 2.

Lower Support Column Bodies (cast)

Note 2 in Table 4-6 of the MRP-227-A is The following note has been added to not included; Subsequent examination is Table 2 for the Lower Support Column required every 10 year interval (Table 4-3 Bodies (cast): "A minimum of 75%

of MRP-227-A) coverage of the entire examination area or volume, or a minimum sample size of 75% of the total population of like components of the examination is required (including both the accessible and inaccessible portions)." Re-inspection on a 10-year frequency is added to Table 2.

Serial No.12-295 Page 23 of 24 NRC Question ME7727-RAII-EVIB-Cher-014-2012-04-27 Based on its evaluation of MRP-227, the NRC staff believes each plant, including Kewaunee Power Station (KPS), needs to perform a plant-specific analysis to ensure that it is bounded by the MRP-227-A assumptions unless KPS can submit an evaluation derived from MRP-227-A that: (1) is specifically relevant to KPS; and (2) is bounded for applicable parameters.

As an alternative the licensee can propose for analysis: (1) other reactor vessel internals (RVI) components that can be analyzed to satisfy this requirement; or (2) another component for which it verifies that the stress/fluence values used by MRP-227-A for that component are bounding.

Response

DEK has confirmed with Westinghouse that the input parameters used for the generic FMECA and functionality analysis summarized in MRP-191 and MRP-232 are representative for KPS. There are no components at KPS that are not bounded by assumptions regarding neutron fluence, temperature, stress, or material used in the development of MPR-227-A.

Therefore, KPS is bounded by the MRP-227-A assumptions.

Serial No.12-295 Page 24 of 24 REFERENCES

1. Email from Karl D. Feintuch (NRC) to Jack Gadzala (DEK) et al, "ME7727 -

Kewaunee - Review of RVI Inspection Plan RAI Set 1 of 2 (Rev 1)," dated March 28, 2012.

2. Email from Karl D. Feintuch (NRC) to Jack Gadzala (DEK) et al, "ME7727 (not 7277)

- Kewaunee - Review of RVI Inspection Plan - (1) RAI Set 2 of 2, adding to RAI Set 1 of 2 (Rev 1); and (2) request for clarification conference call," dated April 9, 2012.

3. Email from Karl D. Feintuch (NRC) to Jack Gadzala (DEK) et al, "ME7727 Kewaunee Request for Additional Information (RAI) ME7727-RAII-EVIB-Cher-014-2012-04-27," dated April 11, 2012.

4. Westinghouse Electric Company (Westinghouse) letter WPS-08-28 Revision 2, "Dominion Energy Kewaunee, Kewaunee Power Station, KPS Reactor Vessel Internals Fabrication and Design Information, LTR-ARIDA-08-63, Revision 3, Summary Report for the Fabrication and Design Information for KPS Reactor Vessel Internals," dated December 22, 2008.

5. WCAP-17562-P, Revision 0, "Westinghouse Pressurized Water Reactor Internals Guide Tube Guide Card Wear Criteria."

6. Materials Reliability Program:

PWR Internals Material Aging Degradation Mechanism Screening and Threshold Values (MRP-175) - EPRI Report 1012081, 2005.

7. MRP-191, Revision 0, "Materials Reliability Program: Screening, Categorization and Ranking of Reactor Internals of Westinghouse and Combustion Engineering PWR

[pressurized water reactor] Designs" (proprietary).

8. MRP-232, Revision 0, "Materials Reliability Program: Aging Management Strategies for Westinghouse and Combustion Engineering PWR Internals" (proprietary).

9. EPRI Report 1016596, "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-Rev. 0)," December 2008, Electric Power Research Institute (EPRI), Palo Alto, California.

10.Final Safety Evaluation (SE) of EPRI Report, Materials Reliability Program Report 1016596 (MRP-227), Revision 0, "Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines," dated June 22, 2011.

Serial No.12-295 ATTACHMENT 2 SUPPLEMENT TO REACTOR VESSEL INTERNALS INSPECTION PLAN REVIEW REQUEST TABLES I AND 2 (REVISION 1)

KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

Serial No.12-295 Table I (Revision 1)

Reactor Vessel Internals Inspection Plan MRP-227 Westinghouse Plants Primary Components (15 pages)

KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

Serial No.12-295 TABLE 1 (Revision I)-

KEWAUNEE POWER STATION FOURTH AND FIFTH INTERVAL ISl SCHEDULE ExaminationCategory MRP-227 Descriptioh TABLE 4-3 WESTINGHOUSE PLANTS PRIMARY COMPONENTS CONTROL ROD DRIVE TUBES (CRGT) GUIDE PLATE CARDS Examination Period Examination

xemption, Eamiatio PeiodMethods Code Case, Item No.

Parts Examined ISI Drawing No.

Equipment No.

INT.

od Case, Comments or Relief Sch 1

2 3

EOI Vol Sur Vis Request Reactor Vessel Internals CRGT Guide Plate Cards Attachment Position 2F N

X Visual (VT-3) examination no Figure 2 later than 2 refueling outages from the beginning of the license renewal period, and no earlier than two refueling outages prior to the start of the license renewal period. 20% examination of the number of CRGT assemblies, with all guide cards within each selected CRGT assembly examined. A total of 29 locations. It is suggested that the population selected for initial inspection coincide with the inlet nozzle locations.

Attachment CRGT Guide Plate Cards Figure 2 Position 2H N

X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 3E N

X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 3G N

X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 31 Y

X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 4D N

X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 4J Y

X Same as above Table 1, Page 1 of 15

Serial No.12-295 TABLE I (Revision 1)

KEWAUNEE POWER STATION FOURTH AND FIFTH INTERVAL ISI SCHEDULE Examination Category MRP-227 Description TABLE 4-3 WESTINGHOUSE PLANTS PRIMARY COMPONENTS CONTROL ROD.DRIVE TUBES (CRGT) GUIDE PLATE CARDS Examination Exemption, Examination Period Methods

• Code Case,

• Cmet Item No.

Parts Examined ISI Drawing No.

Equipment No.

INT.

or Relief Comments Sch I

2 3

E0I Vol Sur Vis Request Attachment CRGT Guide Plate Cards Figure 2 Position 5C N

X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 5E N

X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 5G N

X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 51 N

X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 5K Y

X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 66 N

X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 6F N

X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 6H N

X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 6L N

X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 7C N

X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 7E N

X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 7G N

X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 71 N

X Same as above Table 1, Page 2 of 15

Serial No.12-295 TABLE 1(Revision 1)

KEWAUNEE.POWER STATION FOURTH AND FIFTH INTERVAL ISI SCHEDULE Examination Category MRP-227 Description TABLE 4-3 WESTINGHOUSE PLANTS PRIMARY COMPONENTS CONTROL ROD DRIVE TUBES (CRGT) GUIDE PLATE CARDS E

io rExamination Exemption, Examination Period Methods Code Case, Item No.

Parts Examined ISI Drawing No.

Equipment No.

INT.

-Ref".Comments

.Schý, *i' 2

3 EOI Vol Sur Vis Request Attachment CRGT Guide Plate Cards Figure 2 Position 7K N

X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 8B N

X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 8F N

X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 8H N

X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 8L N

X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 9C Y

X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 9E N

X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 9G N

X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 91 N

X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 9K N

X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 10D Y

X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 10J N

X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 11E Y

X Same as above Table 1, Page 3 of 15

Serial No.12-295 TABLE.(Revision 1)

KEWAUNEE POWER STATION FOURTH AND FIFTH INTERVAL ISISCHEDULE Examination Category MRP-227 Description TABLE 4-3 WESTINGHOUSE PLANTS PRIMARY COMPONENTS CONTROL ROD DRIVE TUBES (CRGT) GUIDE PLATE CARDS 1<

Examination PeriodExamination Exemption, Examination Period Metho~ds.*.. *Code Case, Item No.

Parts Examined ISI Drawing No.

Equipment No.

INT or Relief.,

Comments Sch 1

2 3 ý'EOI Vol Sur Vis oRelief Attachment Request CRGT Guide Plate Cards Figure 2 Position 1G N

X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 111 N

X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 12F N

X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 12H N

X Same as above Category Notes:

1. End of Original License is December 21, 2013. The examinations may be performed during Refueling Outages KR 32 (spring 2012), KR 33 (fall 2013) or KR 34 (spring 2015).

Table 1, Page 4 of 15

Serial No.12-295 TABLE I (Revision 1)

KEWAUNEE POWER STATION FOURTH. AND FIFTH INTERVAL ISI SCHEDULE Examination Category MRP-227 Description TABLE 4-3,WESTINGHOUSE PLANTS PRIMARY cOmPONENTS CONTROL ROD DRIVE TUBES (CRG LOWER FLANGE WELDS Examination Exemption, Examination PeriodMehd Item No.

Parts Examined ISI Drawing No.

Equipment No.

INT.

Peid.edsorCode

Case, Comments or Relief Sch 1

2 3

EOI Vol Sur Vis Request Reactor Vessel Internals CRGT Lower Flange Welds Attachment Position 2F Y

X Enhanced visual (EVT-1)

Figure 4 examination to determine the presence of crack-like surface flaws in flange welds no later than 2 refueling outages from the beginning of the of the license renewal period and subsequent examination on a ten year interval. 100% of outer (accessible) CRGT lower flange weld surfaces and adjacent base metal See Figure 4-21 of MRP-227. Expansion Link - Bottom-mounted (BMI) column bodies and Lower support column bodies (cast). Expansion Link -

Upper Core Plate and Lower Support Forging per NRC SER TRC-1. A total of 37 locations.

Attachment CRGT Lower Flange Welds Figure 4 Position 2H Y

X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 3E Y

X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 3G N

X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 31 Y

X Same as above Table 1, Page 5 of 15

Serial No.12-295 TABLEl (Revision1).

KEWAUNEE POWER STATION FOURTH AND FIFTH INTERVAL ISI SCHEDULE Examination Category MRP-227 Description TABLE 4-3 WESTINGHOUSE PLANTS PRIMARY COMPONENTS CONTROL ROD DRIVE TUBES (CRGT) LOWER FLANGE WELDS Examination "

Exemption,

"."Ex nation Period Methds Code Case, Commets Item No.

Parts Examined ISI Drawing No.

Equipment No.

INT.

m men

-t-or Relief

~or Relief*,

Sch 1

2 3

EOI Vol Sur Vis Request Attachment CRGT Lower Flange Welds Figure 4 Position 4D Y

X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 4J Y

X Same as above Attachment CRGT Loker Flange Welds Figure 4 Position 5C Y

X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 5E N

X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 5G N

X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 51 N

X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 5K Y

X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 6B Y

X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 6F N

X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 6H N

X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 6L Y

X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 7C X

Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 7E N

Same as above Table 1, Page 6 of 15

Serial No.12-295 TABLE 1 (Revision 1)

KEWAUNEE POWER STATION FOURTH AND FIFTH INTERVAL ISI SCHEDULE

Examination Category MRP-227 Description TABLE 4-3 WESTINGHOUSE PLANTS PRIMARY COMPONENTS CONTROL ROD DRIVE TUBES (CRGT) LOWER FLANGE WELDS Examination Exemption, Examination Period Methods Code Case, Item No.

Parts Examined ISI Drawing No.

Equipment No.

INT.

Comments

• • , L,...,.

or R elief Sch 1

2 3

EOI:

Vol.

Sur Vis Request Attachment CRGT Lower Flange Welds Figure 4 Position 7G N

X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 71 N

X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 7K N

X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 8B Y

X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 8F N

X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 8H N

X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 8L Y

X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 9C Y

X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 9E N

X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 9G N

X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 91 N

X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 9K Y

X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 10D Y

X Same as above Table 1, Page 7 of 15

Serial No.12-295 TABLE 1 (Revision 1)

KEWAUNEE POWER STATION FOURTH AND FIFTH INTERVAL ISI SCHEDULE Examination Category *! MRP-227 Description TABLE 4-3 WESTINGHOUSE PLANTS PRIMARY COMPONENTS CONTROL ROD DRIVE TUBES (CRGT) LOWER FLANGE=WELDS Examination Exemption, Examination Period Methods Code Case, Item No.

Parts Examinedý**

ISI Drawing No.

Equipment No.

INT.

or Relief Comments Sch 1

2 3

EOI Vol Sur Vis Request Attachment CRGT Lower Flange Welds Figure 4 Position 10J Y

X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 11E Y

X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 11 G N

X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 111 V

X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 12F Y

X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 12H Y

X Same as above Category Notes:

1. End of Original License is December 21, 2013. The examinations may be performed during Refueling Outages KR 32 (spring 2012), KR 33 (fall 2013) or KR 34 (spring 2015).

2.

There are a total of 20 active CRGT's on the periphery.

3.

It is anticipated that approximately 180 degrees or half the weld length is accessible on each periphery CRGT.

4. 100% of outer (accessible) CRGT lower flange weld surfaces and adjacent base metal on the individual periphery CRGT assemblies. A minimum of 75% of the total identified sample population must be examined.

Table 1, Page 8 of 15

Serial No.12-295 TABLE I (Revision 1)

KEWAUNEE POWER STATION:..

FOURTH AND FIFTH INTERVAL ISr SCHEDULE Examination Category MRP-227 Description TABLE 4-3 WESTINGHOUSE PLANTS PRIMARY COMPONENTS CORE BARREL ASSEMBLY - UPPER CORE BARREL FLANGE WELD Examination Exemption, Examination Period Methods Code Case, Item No.

Parts Examined ISI Drawing No.

Equipment No.

INT.

-d Casor Relief Comet Sch 3

• EOI.

Vol Sur Vis Request-Reactor Vessel Internals Core Barrel Upper Core Barrel M-1 199 Upper Core Y

X3 X3 Periodic enhanced visual (EVT-1)

Assembly Flange Weld Barrel Flange examination, no later than 2 to Core Barrel refueling outages from the Weld beginning of the license renewal period and subsequent examination on a ten-year interval. 100% of one side of the accessible surfaces of the selected weld and adjacent base metal. See Figure 4-22 of MRP-227. Expansion Link - Core Barrel Outlet Nozzles. Expansion Link - Lower Support Column Bodies per NRC SER TRC-1.

Reference Table 4-3, page 4-26 of MRP-227-A Core Barrel NRC SER M-1199 Core Barrel Y

X3 x 3 Periodic enhanced visual (EVT-1)

Assembly Upper Core Barrel Mid Plane Weld examination, no later than 2 refueling outages from the Cylinder Girth Welds beginning of the license renewal period and subsequent examination on a ten-year interval. 100% of one side of the accessible surfaces of the selected weld and adjacent base metal. Expansion Link - Upper and Lower Core Barrel Cylinder Axial welds. Reference Table 4-3, page 4-26 of MRP-227-A.

Table 1, Page 9 of 15

Serial No.12-295

.TABLE 1 (Revision 1)

KEWAUNEE POWER STATION

.i.FOURTH AND FIFTH INTERVAL ISI SCHEDULE Examination Category MRP-227 Description TABLE 4-3 WESTINGHOUSE PLANTS PRIMARY COMPONENTS CORE BARREL ASSEMBLY - UPPER CORE BARREL FLANGE WELD Examination Exemption,

=*

=Examination Period Methods Code Case, Cmet Item No.

Parts Examined ISI Drawing,No.

Equipment No.

INT.

P

-or ReleComments i

....,.,.or Relief...

Sch 1

2 3

EOi Vol Sur Vis Request Core Barrel Lower Core Barrel M-1 199 Core Barrel Y

x 3 x3 Periodic enhanced visual (EVT-1)

Assembly Cylinder Girth Welds Lower Mid examination, no later than 2 Plane Weld refueling outages from the beginning of the license renewal period and subsequent examination on a ten-year interval. 100% of one side of the accessible surfaces of the selected weld and adjacent base metal. No expansion required.

Expansion Link - Upper and Lower Core Barrel Cylinder Axial welds. Reference Table 4-3, page 4-26 of MRP-227-A.

Core Barrel NRC SER M-1 199 Core Barrel N

X3.4 x 3.4 Periodic enhanced visual (EVT-1)

Lower Bottom examination, no later than 2 Assembly Lower Core Barrel Weld refueling outages from the Flange Welds beginning of the license renewal period and subsequent examination on a ten-year interval. 100% of one side of the accessible surfaces of the selected weld and adjacent base metal. Non expansion required.

Expansion Link - Upper and Lower Core Barrel Cylinder Axial welds. Reference Table 4-3, page 4-27 of MRP-227-A.

Table 1, Page 10 of 15

Serial No.12-295 TABLE 1 (Revision"l)

KEWAUNEE POWER STATION FOURTH AND FIFTH INTERVAL ISI SCHEDULE Examination Category MRP-227 Description TABLE 4-3 WESTINGHOUSE PLANTS PRIMARY COMPONENTS CORE BARREL ASSEMBLY - UPPER CORE BARREL FLANGE WELD Examination Exemption, Examination Period Methods Code Case, Comment Item No.

Parts Examined ISl Drawing: No.

Equipment No.

INT.

oroRelief

*+,*+,*

I

+1 Ior ReliefCo m n s Sch,2 I

31 EOI Vol Sur vis Request Category Notes:

1. End of Original License is December 21, 2013. The examinations may be performed during Refueling Outages KR33 (fall 2013) or KR34 (spring 2015).

2.

Reference NRC SER dated June 22, 2011, Final Safety Evaluation of EPRI Report, Materials Reliability Program Report 1016596 (MRP-227), Revision 0, "Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines".

3.

Enhanced visual may be satisfied through eddy current examination if elected in lieu of EVT-1.

4. The lower core barrel flange weld may be alternatively designated as the core barrel-to-support plate weld in some Westinghouse plant designs.

5.

A minimum of 75% of the total weld length (examined + unexamined), including coverage consistent with the Expansion criteria in Table 5-3, must be examined from either the inner or outer diameter for inspection credit.

Table 1, Page 11 of 15

Serial No.12-295 TABLE 1 (Revision 1)

KEWAUNEE POWER STATION FOURTH:AND FIFTH INTERVAL 1§l SCHEDULE Examination Category MRP-227 Description TABLE 4-3 WESTINGHOUSE PLANTS PRIMARY COMPONENTS BAFFLE FORMER ASSEMBLY - BAFFLE EDGE BOLTS Examination Exemption, Examination Period Methods

'*Codie Case, Item No.

Parts Examined IS[ Drawing Equipment INT.

P

-o,-

or Relief No.

No.

Sch 1

2 3

'EOI Vol Sur Vis Request Reactor Vessel Internals Core Baffle-Edge Bolts WCAP-688 Edge Y

X Visual (VT-3) examination, Barrel 13266R1 Bolts with baseline examination Baffle-Figures 6.1, 316ss between 20 and 40 EFPY Former 6.2, 6.3 and subsequent Assembly examinations on a ten-year interval. Bolts and locking devices on high fluence seams. 100% of components accessible from the core side.

Reference Figure 4-23 from MRP-227. No expansion required.

Category Notes:

1. End of Original License is December 21, 2013.

2.

The KPS Reactor Vessel is projected to reach 33 EFPY at End of Original License and 52.1 EFPY at End of Life Extension.

3.

A minimum of 75% of the total population (examined + unexamined), including coverage consistent with the Expansion criteria in Table 5-3, must be examined for inspection credit.

Table 1, Page 12 of 15

Serial No.12-295 TABLE 1 (Revision 1)

KEWAUNEE POWER STATION FOURTH AND FIFTH INTERVAL ISl SCHEDULE Examination Category MRP-227 Description TABLE 4-3 WESTINGHOUSE PLANTS PRIMARY COMPONENTS BAFFLE-FORMER ASSEMBLY BAFFLE-FORMER BOLTS Examination Exemption, Examination Period Methods Code Case Comments Item No.

Parts Examined

.SI Drawing No.

Equipment No.

INT.

-od

"-e

=

I*

• ,*:or Reliefi*

Sch 1

2 3

.EOI Vol Sur Vis Request Reactor Vessel Internals Core Barrel Baffle-Former Bolts WCAP-13266R1 728 Baffle Y

X Baseline volumetric (UT)

Baffle-Figures 6.1,6.2, Former Bolts examination between 25 and Former 6.3 347 ss 35 EFPY, with subsequent Assembly examination on a 10-year inspection frequency per NRC SER TRC-5. 100% of accessible bolts or as supported by plant-specific justification. Heads accessible from the core side. UT accessibility may be affected by complexity of head and locking device designs.

Reference Figures 4-23 and 4-24 of MRP-227. Expansion Link - Lower support column bolts and barrel-former bolts.

Cate-gory Notes:

1. End of Original License is December 21, 2013. The KPS Reactor Vessel is projected to reach 33 EFPY at End of Original License, 34.5 EFPY at Refueling Outage KR-34 (spring 2015), and 52.1 EFPY at End of Life Extension.

2.

Reference WCAP-15425 for KPS Minimum Baffle-Former Bolt Pattern.

3.

Reference NRC SER dated June 22, 2011, Final Safety Evaluation of EPRI Report, Materials Reliability Program Report 1016596 (MRP-227), Revision 0, "Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines".

4. A minimum of 75% of the total population (examined + unexamined), including coverage consistent with the Expansion criteria in Table 5-3, must be examined for inspection credit. Subsequent inspection of the baffle-former bolts is required every 10 year interval.

Table 1, Page 13 of 15

Serial No.12-295 TABLE 1*(Revision 1,)

KEWAUNEE POWER STATION, FOURTH AND FIFTH INTERVAL ISI SCHEDULE Examination Category MRP-227.

Description TABLE 4-3 WESTINGHOUSE PLANTS PRIMARY COMPONENTS BAFFLE - FORMER ASSEMBLY - ASSEMBLY Examination Exemption, Examination. Period Methods Code Case, Comments Item No.

Parts Examined ISI Drawing No.

Equipment INT.

or Relief No.

Sch 1

2 3,.

EOI Vol Sur Vis Request Reactor Vessel Internals Core Barrel Assembly WCAP-13266R1 Baffle Y

X Visual (VT-3) examination Baffle-(baffle plates, Figures 6.1, 6.2, Former to check for evidence of Former baffle-edge 6.3 distortion, with baseline Assembly bolts, and cexamination between 20 former plates)

Attachment and 40 EFPY and

) Figure 4 subsequent examinations on a ten-year interval.

Inspections are performed on the core side surface.

Reference Figures 4-24, 4-25, 4-26, and 4-27 of MRP-227. No expansion required.

Category Notes:

1. End of Original License is December 21, 2013.

2.

The KPS Reactor Vessel is projected to reach 33 EFPY at End of Original License, 34.5 EFPY at Refueling Outage KR-34 (spring 2015), and 52.1 EFPY at End of Life Extension.

Table 1, Page 14 of 15

Serial No.12-295,,

TABLE 1 (Revision fil).

KEWAUNEE POWER STATION FOURTH AND FIFTH INTERVAL ISI SCHEDULE Examination Category'.,MRW-227' Description TABLE 4-3 WESTINGHOUSE PLANTS PRIMARY COMPONENTSTHERMAL SHIELD ASSEMBLY - THERMAL SHIELD FLEXURES Examination Period Examination Exemption,

• Methods Code Case, Item No.

PartsExamined IS[ Drawing No.

Equipment No.

INT.

Mehds__

or Relief Comments Sch 1

2 3

EOI Vol Sur Vis Request Reactor Vessel Internals Core Barrel Thermal Shield Flexures Attachment Thermal Shield Y

X Visual (VT-3) examination no Thermal Figure 4 Flexure 00 later than 2 refueling outages Shield from the beginning of the license Assembly renewal period. Subsequent examinations on a ten-year interval. 100% of thermal shield flexures. Reference Figures 4-29 and 4-36 of MRP-227. No expansion required.

Core Barrel Thermal Shield Flexures Attachment Thermal Shield Y

X Same as above Thermal Figure 4 Flexure 90° Shield Assembly Core Barrel Thermal Shield Flexures Attachment Thermal Shield Y

X Same as above Thermal Figure 4 Flexure 1800 Shield Assembly Core Barrel Thermal Shield Flexures Attachment Thermal Shield Y

X Same as above Thermal Figure 4 Flexure 2700 Shield Assembly Category Notes:

1. End of Original License is December 21, 2013. The examinations may be performed during Refueling Outages KR33 (fall 2013) or KR34 (spring 2015).

2. VT-3 examination no later than 2 refueling outages from the beginning of the license renewal period. Subsequent examinations on a ten-year interval.

Table 1, Page 15 of 15

Serial No.12-295 Table 2 (Revision 1)

Reactor Vessel Internals Inspection Plan MRP-227 Westinghouse Plants Expansion Components (8 pages)

KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

Serial No.12-295 TABLE 2 (Revision 1)

KEWAUNEE POWER STATION FOURTH AND FIFTH INTERVAL ISI SCHEDULE Examination Category MRP-227 Description TABLE 4-6 WESTINGHOUSE PLANTS EXPANSION COMPONENTS CORE BARREL ASSEMBLY - BARREL-FORMER BOLTS Examination Exemption, Examination Period Methods Code Case, Item: No.

Parts Examined ISI Drawing No.

Equipment No.

INT'..

or Relief Comments

• :,;t or 116if*

Sch 1

2 3

E0I Vol Sur Vis Request Reactor Vessel Internals Core Barrel-Barrel-Former Bolts WCAP-13266R1 344 Barrel-Former N

X Primary Link-Baffle-Former Former Figures 6.1, 6.2, Bolts Bolts. Volumetric (UT)

Assembly 6.3 347ss examination, with initial examination dependent on results of baffle-former bolt examinations. Re-inspection is on a 10-year frequency. 100%

of accessible bolts. Accessibility may be limited by presence of thermal shields or neutron pads.

Reference Figure 4-23 of MRP-227. Expansion Link - Lower support column bolts and Barrel-former bolts.

Category Notes:

1. End of Original License is December 21, 2013. The KPS RV is projected to reach 33 EFPY at End of Original License, 34.5 EFPY at KR-34 (spring 2015), and 52.1 EFPY at End of Life Extension.

2.

Reference WCAP-13266 Revision 1 for details. The barrel-former bolts are fabricated from type 347 stainless steel.

3.

Examinations are scheduled per the Corrective Action Process if the number of indications on the Baffle-Former Bolts exceeds the threshold.

4. Confirmation that more than 5% of the baffle-former bolts actually examined on the four baffle plates at the largest distance from the core (presumed to be the lowest dose locations) contain unacceptable indications shall require UT examination of the lower support column bolts within the next three fuel cycles.

5. Confirmation that more than 5% of the lower support column bolts actually examined contain unacceptable indications shall require UT examination of the barrel-former bolts.

6.

If expansion is needed/invoked then subsequent examination is required every 10 year interval.

7. A minimum of 75% coverage of the entire examination area or volume, or a minimum sample size of 75% of the total population of like components of the examination is required (including both the accessible and inaccessible portions).

Table 2, Page 1 of 8

Serial No.12-295

..................."...TABLE 2 (Revision 1)

KEWAUNEE POWER STATION FOURTH AND FIFTH INTERVAL ISI SCHEDULE Examination Category MRP-227 Description TABLE 4-6 WESTINGHOUSE PLANTS EXPANSION COMPONENTS CORE BARREL ASSEMBLY - UPPER AND LOWER CORE BARREL CYLINDER AXIAL WELDS Examination Exemption, Examination Period Methods Code Case, Item No.

Parts Examined ISI Drawing No.

Equipment No.

INT.

sor ReeComments or Relief Sch 1

2 3

EOI V01 Sur Vis Request Reactor Vessel Internals Core Barrel Upper and lower N

X Primary Link-Upper and lower Assembly core barrel cylinder core barrel cylinder girth welds.

axial welds Re-inspection is on a 10-year frequency. Enhanced visual (EVT-1) examination, with initial examination dependent on the examination results for upper and lower core barrel cylinder girth welds. Re-inspection on a 10-year frequency. 100% of one side of the accessible surfaces of the selected weld and adjacent base metal. Reference Figure 4-34 of MRP-227.

Category Notes:

1. End of Original License is December 21, 2013. The KPS RV is projected to reach 33 EFPY at End of Original License, 34.5 EFPY at KR-34 (spring 2015), and 52.1 EFPY at End of Life Extension.

2.

Examinations are scheduled per the Corrective Action Process if a surface breaking indication is observed in the upper core barrel flange weld.

3. The confirmed detection and sizing of a surface-breaking indication with a length greater than two inches in the upper core barrel cylinder girth welds shall require that the EVT-1 examination be expanded to include the upper core barrel cylinder axial welds by the completion of the next refueling outage.

4. The confirmed detection and sizing of a surface-breaking indication with a length greater than two inches in the lower core barrel cylinder girth welds shall require that the EVT-1 examination be expanded to include the lower core barrel cylinder axial welds by the completion of the next refueling outage.

5.

If expansion is neededlinvoked then subsequent examination is required every 10 year interval.

6.

A minimum of 75% coverage of the entire examination area or volume, or a minimum sample size of 75% of the total population of like components of the examination is required (including both the accessible and inaccessible portions).

Table 2, Page 2 of 8

Serial No.12-295 TABLE 2 (Revision 1)

KEWAUNEE POWER STATION FOURTH AND FIFTH INTERVAL ISI SCHEDULE Examination Category, MRP-227 Description TABLE 4-6 WESTINGHOUSE PLANTS EXPANSION COMPONENTS LOWER SUPPORT ASSEMBLY-LOWER SUPPORT COLUMN BOLTS Examination Exemption, Examination Period Methods Code Case Item No.

Parts Examined ISI Drawing No.

Equipment No.

INT.

m I

P

-e-o-

- 'Comme or Relief Sch 1

2 3

EOI Vol Sur Vis Request Reactor Vessel Internals Lower Lower Support W Drawing N

X Primary Link-Baffle-Former Support Column Bolts 882D685 Bolts. Volumetric (UT)

Assembly examination, with initial examinations dependent on results of baffle-former bolt examinations. Re-inspection is on a 10-year frequency. 100% of accessible bolts or as supported by plant-specific justification.

Reference Figures 4-32 and 4-33 of MRP-227.

Category Notes:

1. End of Original License is December 21, 2013. The KPS RV is projected to reach 33 EFPY at End of Original License, 34.5 EFPY at KR-34 (spring 2015), and 52.1 EFPY at End of Life Extension.

2.

Examinations are scheduled per the Corrective Action Process if the number of indications on the Baffle-former bolts exceeds the threshold.

3.

Confirmation that more than 5% of the lower support column bolts actually examined contain unacceptable indications shall require UT examination of the barrel-former bolts.

4.

If expansion is needed/invoked then subsequent examination is required every 10 year interval.

5.

A minimum of 75% coverage of the entire examination area or volume, or a minimum sample size of 75% of the total population of like components of the examination is required (including both the accessible and inaccessible portions).

Table 2, Page 3 of 8

Serial No.12-295 TABLE 2 (Revision 1)

KEWAUNEE POWER STATION FOURTH AND FIFTH INTERVAL ISI SCHEDULE Examination Category MRP-227 Description TABLE 4-6 WESTINGHOUSE PLANTS EXPANSION COMPONENTS CORE BARREL ASSEMBLY - CORE BARREL FLANGE, CORE BARREL: OUTLET NOZZLES, CORE BARREL SAFETY INJECTION NOZZLES AND LOWER CORE BARREL FLANGE WELD Examination Exemption, ExainaionPerod=

Methods Code Case, Item No.

Parts Examined ISI Drawing No.

Equipment No.

INT.

Ei od Ce, Comments or Relief Sch 1

2 3

EOI Vol Sur Vis Request Reactor Vessel Internals Core Barrel Core Barrel Flange (1),

N X

Primary Link-Upper Core Barrel Assembly Core Barrel Outlet Flange Weld. Enhanced visual Nozzles(2), Safety (EVT-1) examination, with initial Injection Nozzles (2) examination dependent on the examination results for upper core barrel flange. Re-inspection on a 10-year frequency. 100% of one side of the accessible surfaces of the selected weld and adjacent base metal. Reference Figure 4-34 of MRP-227.

Category Notes:

1.

End of Original License is December 21, 2013. The KPS RV is projected to reach 33 EFPY at End of Original License, 34.5 EFPY at KR-34 (spring 2015), and 52.1 EFPY at End of Life Extension.

2.

Examinations are scheduled per the Corrective Action Process if a surface breaking indication is observed in the upper core barrel flange weld.

3. The confirmed detection and sizing of a surface-breaking indication with a length greater than two inches in the upper core barrel flange weld shall require that the EVT-1 examination be expanded to include the core barrel outlet nozzle welds by the completion of the next refueling outage.

4.

If extensive cracking in the core barrel outlet nozzle welds is detected, EVT-1 examination shall be expanded to include the upper six inches of the accessible surfaces of the non-cast lower support column bodies within three fuel cycles following the initial observation.

5.

If expansion is needed/invoked then subsequent examination is required every 10 year interval.

6.

A minimum of 75% coverage of the entire examination area or volume, or a minimum sample size of 75% of the total population of like components of the examination is required (including both the accessible and inaccessible portions).

Table 2, Page 4 of 8

Serial No.12-295 TABLE 2 (Revision 1)

KEWAUNEE POWER STATION FOURTH AND FIFTH INTERVAL ISl SCHEDULE

,Examination Category MRP-227 Description TABLE 4-6 WESTINGHOUSE PLANTS EXPANSION COMPONENTS LOWER SUPPORT ASSEMBLY - LOWER SUPPORT COLUMN BODIES (NON CAST)

Examination Exemption, Item No.

Parts Examined ISI Drawing No.

Equipment No.

INT.

Examination Period Methods Code Case, Comments

~~~~or Relief/*':

Cmet Sch 1

2 3

EOI Vol Sur Vis Request Reactor Vessel Internals Lower Lower Support W685J896 N

X Primary Link-Upper Core Support Column Bodies Barrel Flange Weld. Enhanced Assembly (Non Cast) visual (EVT-1) examination, with initial examination dependent on the examination results for upper core barrel flange. Re-inspection on a 10-year frequency. 100% of accessible surfaces.

Reference Figure 4-34 of MRP-227.

Category Notes:

1. End of Original License is December 21, 2013.

2.

Examinations are scheduled per the Corrective Action Process if a surface breaking indication is observed in the upper core barrel flange weld.

3.

The confirmed detection and sizing of a surface-breaking indication with a length greater than two inches in the upper core barrel flange weld shall require that the EVT-1 examination be expanded to include the core barrel outlet-nozzle welds by the completion of the next refueling outage.

4.

If extensive cracking in the core barrel outlet nozzle welds is detected, EVT-1 examination shall be expanded to include the upper six inches of the accessible surfaces of the non-cast lower support column bodies within three fuel cycles following the initial observation.

5.

If expansion is needed/invoked then subsequent examination is required every 10 year interval.

6.

A minimum of 75% coverage of the entire examination area or volume, or a minimum sample size of 75% of the total population of like components of the examination is required (including both the accessible and inaccessible portions).

Table 2, Page 5 of 8

Serial No.12-295 Sr TABLE 2 (Revision 1)

KEWAUNEE POWER STATION FOURTH AND FIFTH INTERVAL ISl SCHEDULE Examination Category MRP-227 Description TABLE 4-6 WESTINGHOUSE PLANTS EXPANSION COMPONENTS LOWER SUPPORT ASSEMBLY - LOWER SUPPORT COLUMN BODIES (CAST)

Examination Period Examination Exemption, Methods Code Case, Item No.

Parts Examined ISI Drawing No.,'ý Equipment No.

INT.

-Comments or Relief Cmet Sch 1

2 3

EOI Vol Sur Vis Request Reactor Vessel Internals Lower Lower Support W685J896 N

X Lower Support Column Bodies Support Column Bodies are not cast at Kewaunee Assembly (Cast)

Power Station. This expansion item [from the CRGT lower flange welds] is Not Applicable to KPS. Primary Link-Control Rod Guide Tube Lower Flanges. Visual (EVT-1) examination. 100% of accessible support columns.

Reference Figure 4-34 of MRP-227.

Category Notes:

1. End of Original License is December 21, 2013.

2.

Examinations are scheduled per the Corrective Action Process if a crack-like surface indication is observed.

3. Lower support column bodies (cast) are not applicable to Kewaunee Power Station. Confirmation of surface breaking indications in CRGT lower flange welds shall require EVT-1 examination of cast lower support column bodies within three fuel cycles following the initial observation. For cast lower support column bodies, the specific relevant condition is a detectable crack-like surface indication.

4.

If expansion is needed/invoked then subsequent examination is required every 10 year interval.

5.

A minimum of 75% coverage of the entire examination area or volume, or a minimum sample size of 75% of the total population of like components of the examination is required (including both the accessible and inaccessible portions).

Table 2, Page 6 of 8

Serial No.12-295 11...

TABLE 2 (Revision 1)

KEWAUNEE POWER STATION FOURTH AND FIFTH INTERVAL ISl SCHEDULE Examination Category MRP-227 Description TABLE 4-6 WESTINGHOUSE PLANTS EXPANSION COMPONENTS BOTTOM MOUNTED INSTRUMENTATION SYSTEM - BMI COLUMN BODIES Examination Exemption, Examination Period Methods Code Case, Cmet Item No.

Parts Examined ISI Drawing No.

Equipment No.

INT.

E or Relief Comments Sch 1

2 3

EOI Vol Sur Vis Request Reactor Vessel Internals Reactor Vessel Bottom-Mounted 685J896 N

X Primary Link-Control rod guide Bottom Instrumentation tube lower flanges. Visual (Vr-Mounted (BMI) Column

3) examination of BMI column Instrumentation Bodies (36) bodies as indicated by difficulty System of insertion/withdrawal of flux thimbles. Flux thimble insertion/withdrawal to be monitored at each inspection interval. 100% of BMI column bodies for which difficulty is detected during flux thimble insertion/withdrawal. Re-inspection on a 10-year frequency. Reference Figure 4-35 of MRP-227.

Cate-gory Notes:

1. End of Original License is December 21, 2013.

2.

Examinations are scheduled per the Corrective Action Process if a detectable crack-like surface indication is detected in the CRGT lower flange welds.

3.

Bottom-Mounted Instrumentation (BMI) column bodies. For BMI column bodies, the specific relevant condition for the VT-3 examination is completely fractured column bodies. Confirmation of surface breaking indications in two or more CRGT lower flange welds, combined with flux thimble insertion/withdrawal difficulty, shall require visual (VT-3) examination of BMI column bodies by the completion of the next refueling outage.

4.

If expansion is neededlinvoked then subsequent examination is required every 10 year interval.

5.

A minimum of 75% coverage of the entire examination area or volume, or a minimum sample size of 75% of the total population of like components of the examination is required (including both the accessible and inaccessible portions).

Table 2, Page 7 of 8

Serial No.12-295 TABLE 2 (Revision 1)

KEWAUNEE POWER STATION FOURTH AND FIFTH INTERVAL ISl SCHEDULE Examination Category - MRP-227 Description TABLE 4-6 WESTINGHOUSE PLANTS EXPANSION COMPONENTS Upper Internals & Lower Internals Assembly Examination Exemption,

• i..*i. *: "

.~~~~~~~~Examination Period'

.ehd~

i,**oeCs.C m et Item No.

Partrs Examined IS[ Drawing No.

'Eqlm~ent No.

INT.

Comm.en.t....osRlef, or Relief Sch 1

2 3

EOI Vol Sur Vis Request Reactor Vessel Internals Lower Lower Support M-1199 N

X Primary Link - Control Rod Support Forging Guide Tube Lower Flange Forging Welds. Enhanced visual (EVT-1) examination. Re-inspection on a 10-year frequency. 100% of accessible surfaces (Note 5).

Ref. MRP-227, Figure 4-33.

Upper Core Upper Core Plate M-1199 N

X Primary Link - Control Rod Plate Guide Tube Lower Flange Welds. Enhanced visual (EVT-1) examination. Re-inspection on a 10-year frequency. 100% of accessible surfaces (Note 5).

Category Notes:

1. End of Original License is December 21, 2013.

2.

Examinations are scheduled per the Corrective Action Process if a crack-like surface indication is observed.

3.

Control Rod Guide Tube Flange Welds. Confirmation of surface breaking indications shall require EVT-1 examination of the lower support forging & upper core support plate within three fuel cycles following the initial observation. For the upper core plate and lower support forging/castings, the specific relevant condition is a detectable crack-like surface indication.

4.

If expansion is needed/invoked then subsequent examination is required every 10 year interval.

5.

A minimum of 75% coverage of the entire examination area or volume, or a minimum sample size of 75% of the total population of like components of the examination is required (including both the accessible and inaccessible portions).

Table 2, Page 8 of 8