Text

Response to Request for Additional Information, Reactor Vessel Internals Inspection Plan Review Request
	ML12101A231
Person / Time
Site:	Kewaunee
Issue date:	04/02/2012
From:	Price J; Dominion Energy Kewaunee
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	12-151, TAC ME7727
	Download: ML12101A231 (65)
	v • d • e

Dominion Energy Kewaunee, Inc.

5000 Dominion Boulevard, Glen Allen, VA 23060 P Dominion April 2, 2012 ATTN: Document Control Desk Serial No.12-151 U. S. Nuclear Regulatory Commission LIC/JG/RO Washington, DC 20555-0001 Docket No.: 50-305 License No.: DPR-43 DOMINION ENERGY KEWAUNEE, INC.

KEWAUNEE POWER STATION RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REACTOR VESSEL INTERNALS INSPECTION PLAN REVIEW REQUEST (ME7727)

By application dated December 12, 2011 (reference 1), Dominion Energy Kewaunee, Inc. (DEK), requested NRC approval of an inspection plan for reactor vessel internal (RVI) components at Kewaunee Power Station (KPS). The proposed inspection plan fulfils two commitments required by Renewed Operating License DPR-43, Section 2.C(15)(b), to submit an inspection plan for reactor internals to the NRC staff for review and approval to augment the current inspections.

Subsequently, the Nuclear Regulatory Commission (NRC) staff transmitted a request for additional information (RAI) regarding the proposed inspection plan (reference 2).

Specifically, the staff requested a copy of Aging Management Program Technical Report AMP-KLR-1309A, "ASME Section XA Inservice Inspection, Subsections IWB, IWC, and IWD Reactor Vessel Internals Inspections, Kewaunee Power Station." The requested report is enclosed with this letter.

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

Jack Gadzala at 920-388-8604.

Very truly yours, J. l i V(4ce stidert - Nuclear Engineering

Enclosure:

Aging Management Program Technical Report AMP-KLR-1309A, "ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Reactor Vessel Internals Inspections, Kewaunee Power Station," Revision 3, dated September 30, 2011. (ýI- - /

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Serial No.12-151 RV Internals Inspection Plan Page 2 of 2

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), Craig Sly (DEK), and Ganesh Cheruvenki (NRC), "ME7727 - Kewaunee - Reactor Vessel Internals Inspection Plan review - an Aging Management Program report is needed to prepare RAI items," dated March 2, 2012.

Commitments made by this letter: None 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.12-151 ENCLOSURE RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REACTOR VESSEL INTERNALS INSPECTION PLAN REVIEW REQUEST Aging Management Program Technical Report AMP-KLR-1309A, "ASME Section Xl Inservice Inspection, Subsections IWB, IWC, and IWD Reactor Vessel Internals Inspections, Kewaunee Power Station,"

Revision 3, dated September 30, 2011 (62 pages)

KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

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-"TECHNICAL REPORT: KLR-1309A LICENSE RENEWAL PROJECT AGING MANAGEMENT PROGRAM ASME SECTION XI INSERVICE INSPECTION, SUBSECTIONS IWB, IWC, AND: IWD REACTOR VESSEL INTERNALS INSPECTIONS KEWAUNEE POWER STATION Prepared by Date:

RICHARDA, REMER Reviewed by 04 Date: __-- _____

CHARLES Ak TOMES Reviewed b Py LIP E. BUKrSz Approved by. Date 1(

Approved by Date: l -.

PAUL C; AiTKEN Approved by _/-/,, 4,-, Date: '- 2 )

TIMoTHY P. OLSON Effective Date: 0913012011.

Revision Number: 3

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION XI ISI - RVI INSPECTIONS REvISION RECORD Page(s) Revision Revision Change Description Revised Number Date All 0 3/24/2010 Initial Issue Various 1 10/29/2010 DMRs 2758, 2816 Various 2 06/3012011 Incorporation of MRP-227 Requirements per Charter ER-31 14 3 09/30/2011 Alignment with NRC Letter for Commitments 1 & 2 Page 2 of 62 KLR-1309A, Rev. 3 KLR-1 309A, Rev. 3 Page 2 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION XI ISI - RVI INSPECTIONS TABLE OF CONTENTS 1.0 PUR PO SE ....................................................................................................... 5 2.0 PROGRAM DESCRIPTION ................................................................................. 5 3.0

SUMMARY

PROGRAM INFORMATION ........................................................... 7 4.0 EVALUATION OF NUREG-1800 AND MRP-227 PROGRAM ELEMENTS ...... 7 4.1 PROGRAM ELEMENT 1 -SCOPE OF PROGRAM .................................................. 8 4.2 PROGRAM ELEMENT 2-- PREVENTIVE ACTIONS .................................................... 10 4.3 PROGRAM ELEMENT 3- PARAMETERS MONITORED/INSPECTED ........................ 11 4.4 PROGRAM ELEMENT 4- DETECTION OF AGING EFFECTS ................................... 11 4.5 PROGRAM ELEMENT 5-- MONITORING AND TRENDING ...................................... 15 4.6 PROGRAM ELEMENT 6- ACCEPTANCE CRITERIA ............................................... 16 4.7 PROGRAM ELEMENT 7-- CORRECTIVE ACTIONS ................................................ 17 4.8 PROGRAM ELEMENT 8-- CONFIRMATION PROCESS ........................................... 18 4.9 PROGRAM ELEMENT 9- ADMINISTRATIVE CONTROLS ....................................... 18 4.10 PROGRAM ELEMENT 10 - OPERATING EXPERIENCE ........................................ 19 5.0 PROGRAM ENHANCEMENTS ......................................................................... 19 6.0 PROGRA M EXCEPTIONS ........... ;.................................................................. 20 7.0 C O NC LUSIO N .................................................................................................. 20 8.0 R EFER ENCES .................................................................................................. 20 9.0 ATTA CHM ENTS ................................................................................................ 22 Rv. Pag 3 f 6 KLR- 30A, KLR-1309A, Rev. 3 Page 3 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION Xl ISI - RVI INSPECTIONS LIST OF TABLES Table Title Page Table I - Comparison to MRP-227 Assumptions ...................................... 13 Table 2 - Reactor Vessel Internals Evaluation ............................................ 23 Table 3 - Aging Management Program Enhancement and Inspection Implementation Schedule ................................................... 41 Table 4 - Detailed MRP-227 Reactor Vessel Internals - Component Inspection Requirem ents ................................................................... 42 LIST OF FIGURES Figure Title Page None ae o6 KL-3IA e.

KLR-1309A, Rev. 3 Page 4 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION XI ISI - RVI INSPECTIONS 1.0 PURPOSE The purpose of the aging management program (AMP) report is to update the basis for the aging management program information summarized in the Kewaunee Power Station (KPS) renewed operating license granted under 10 CFR Part 54, "Requirements for Renewal of Operating Licenses for Nuclear Power Plants" (Reference 8.1). The AMP information is applicable to systems, structures, and components (SSCs) that are in-scope, long-lived, passive, and perform a license renewal intended function.

This document implements commitments to the NRC to enhance the ASME Section Xl Inservice Inspection, Subsections IWB, IWC, and IWD program for managing aging effects on reactor internals and on limiting susceptible cast austenitic stainless steel reactor vessel internals components (Reference 8.7, Section 5.1 and 5.2).

In order to ensure the aging of reactor vessel internal components are adequately managed, the Nuclear Energy Institute initiated a commercial nuclear utility industry Materials Initiative. The result of this effort was NEI 03-08, which was issued through the Nuclear Strategic Issue Advisory Committee (NSIAC) for management of primary pressure boundary materials. NEI 03-08 is classified as a NSIAC initiative, which is a formal agreement by 80% vote of Chief Nuclear Officers and is binding for all US nuclear utilities. To implement the NEI 03-08 requirements, the U.S. industry, through the efforts of the EPRI Materials Reliability Program (MRP) and PWR Owners Group, developed and issued MRP-227 and MRP-228 to identify and document the components and subcomponents that require aging management to support continued reliable operation during the period of extended plant operation.

In addition, a secondary purpose of this AMP is to develop and document a PWR reactor internals aging management program (AMP) as outlined in Appendix A of MRP-227 Rev. 0 and as required by Condition Report 322266 (Reference 8.2).

This document, in conjunction with the integrated plant assessment reports, demonstrates that the effects of aging on identified SSCs will be adequately managed so that the intended function(s) will be maintained consistent with the plant-specific current licensing basis for the period of extended operation.

2.0 PROGRAM DESCRIPTION The Reactor Vessel Internal Inspections program is a new plant-specific program that consists of the applicable ten elements as described in Fleet GARD ER-AA-AMP-1003, "Aging Management Programs for License Renewal Structures, Systems, and Components (Reference 8.3). These applicable ten elements are age ofI Re.

KLR-309A 3 KLR-1309A, Rev. 3 Page 5 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION Xl ISI - RVI INSPECTIONS the same as the ten elements included in MRP-227 Appendix A (Reference 8.4) and Appendix A of NUREG-1800, "Standard Review Plan for Review of License Renewal applications for Nuclear Power Plants" (Reference 8.5).

The program manages the aging effects of changes in dimensions due to void swelling: cracking due to stress corrosion cracking (SCC), primary water stress corrosion cracking (PWSCC), irradiation-assisted stress corrosion cracking (IASCC), and fatigue; loss of fracture toughness due to neutron irradiation embrittlement and thermal aging embrittlement; loss of material due to wear; and loss of preload due to stress relaxation.

Loss of material by pitting corrosion and crevice corrosion is controlled by plant chemistry by limiting the oxygen concentration in the coolant system. Additional management of such mechanisms is not required by the reactor internals program.

The Reactor Vessel Internals Inspections program performs one-time, periodic, and conditional examinations using visual, surface, and ultrasonic examination techniques in accordance with the ASME Code,Section XI and EPRI Report 1016596 (MRP-227) and NEI 03-08 (References 8.4, 8.6). The ASME Section Xl Inservice Inspection, Subsections IWB, IWC, and IWD program performs visual examinations of the reactor vessel removable core support structures under Table IWB-2500-1, Examination Category B-N-3. MRP-227 has developed additional inspection requirements for reactor vessel internals that are susceptible to the aging effects managed by the program, including the examination techniques to be used, the examination schedule, the examination acceptance criteria, and the inspection expansion criteria.

The MRP-227 requirements are administered by Dominion using Fleet Reactor Internals Inspection Program Description, ER-AA-RII-10 (Reference 8.10) and Fleet Reactor Internals Inspection Administrative Procedure ER-AA-RII-101 (Reference 8.11). The Reactor Vessel Internals Inspections program is implemented in accordance with the KPS ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD program that include the inspection procedures, the acceptance criteria for each examination technique, and the process for evaluating unsatisfactory inspection results. For the ASME Code Section XI inspections, indications and relevant conditions detected during examination will be evaluated in accordance with ASME Section XI, Article IWB-3500. For the additional MRP-227 required inspections, the evaluations will be performed in accordance with the examination acceptance criteria and the expansion criteria provided in MRP-227.

The program inspections will provide support for the ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD program for managing the reactor vessel internal components in accordance with NEI 03-08 management of material issues.

Flaw indications detected during the required examinations are dispositioned in accordance with the Corrective Action Program.

Rev l6 3 1ILR I0A IIf PageI 62 KLR-1309A, Rev. 3 Page 6 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION XI ISI - RVi INSPECTIONS 3.0

SUMMARY

PROGRAM INFORMATION Attachment 1 provides the Kewaunee and Nuclear Business Unit procedures that implement this aging management program.

Attachment 2 provides a listing of the aging management programs that are credited by this program as performing activities required by the program and a listing of the aging management programs that credited this program to perform activities required by that program.

Attachment 3 identifies the materials and aging effects managed by this aging management program.

Attachment 4 identifies the Integrated Plant Assessment Reports that credits this aging management program and identifies the systems or commodities managed by the program.

Attachment 5 is a detailed assessment of the ten program elements at the implementation level.

4.0 EVALUATION OF NUREG-1800 AND MRP-227 PROGRAM ELEMENTS This section provides a summary comparison of the Kewaunee aging management program to the ten program elements described in Fleet GARD ER-AA-AMP-1003, "Aging Management Programs for License Renewal Structures, Systems, and Components." These applicable ten elements are the same as the ten elements included in MRP-227 Appendix A and Appendix A of NUREG-1 800, "Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants." Age-related degradation in the reactor internals is managed through an integrated program. Specific features of the integrated program are listed in the following ten program elements. The comparison addresses the aging management program expectations for each element and provides a description of station implementation of these expectations.

Page 7 of 62 KLR-1309A, Rev. 33 KLR-1 309A, Rev. Page 7 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION Xl ISI - RVI INSPECTIONS Attachment 5 is a detailed assessment of the program elements at the implementation level. Attachment 5 provides the following information:

Column Content Attribute Indicates the "attribute" or detailed element recommendation Number being addressed. A program element that contains multiple recommendations is divided into more than one attribute.

Attribute Contains the detailed recommendation directly from NUREG-Description 1800 and MRP-227, Appendix A.

Attribute Evaluates the Kewaunee program against the NUREG-1 800 and Response MRP-227 recommendations.

Document Provides the procedure(s), preventive maintenance activities, or Number other documents, which implement the program.

Document Indicates if all or only a portion of the document is required to Section implement the program.

Purpose Indicates the specific activities performed by the document, which are required to address the NUREG-1800 and MRP-227 recommendations.

Document Indicates the specific document changes to properly implement Change the program.

Note: Some document changes that clarify or provide additional detail for existing activities have not been identified as a program enhancement since the change is not required to meet the NUREG-1800 and MRP-227 recommendations.

Follow-Up Indicates the internal tracking number used to track proposed Action Item document changes and other activities (FAI)

CAP/CR Indicates the number of the Corrective Action Program Action Request/Condition Report that will implement the proposed document changes and other activities.

4.1 PROGRAM ELEMENT 1 - SCOPE OF PROGRAM The scope of the Reactor Vessel Internals Inspections program is defined in Sections 5.1 and 5.2 of the ASME Section XI Inservice Inspection, Subsections KLR-1309A, Rev. 3 Page 8 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION XI ISI - RVI INSPECTIONS IWB, IWC, and IWD program (Reference 8.7). These two sections identify two program enhancements to incorporate applicable industry reactor vessel internals inspection initiatives. The two enhancements state:

"5.1 Enhancement 1: Aging Management of Reactor Vessel Internals The ASME Section Xl Inservice Inspection Subsections IWB, IWC and IWD program will be enhanced to (1) participate in the industry programs for investigating and managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC` for review and approval to augment the current inspections.

5.2 Enhancement 2: Aging Management of Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel The ASME Section XI Inservice Inspection Subsections IWB, IWC and IWD program will be enhanced to include identification of the limiting susceptible cast austenitic stainless steel reactor vessel internals components from the standpoint of thermal aging susceptibility, neutron fluence, and cracking. For each identified component, a plan will be developed, which accomplishes aging management through either a supplemental examination or a component-specific evaluation. The plan will be submitted for NRC review and approval not less than 24 months before entering the period of extended operation."

The scope of the ASME Section XA ISI - Reactor Vessel Internals Inspections program is to implement the inspections required by the two above program enhancements.

The reactor vessel internals included in the scope of the program consist of two basic assemblies, the upper internals assembly that is removed during each refueling operation to obtain access to the reactor core, and the lower internals assembly that can be removed, if desired, following a complete core off-load.

Reactor Vessel Internals Inspections program implements the NEI 03-08 (Reference 8.6) "Mandatory" requirement to use MRP-227 and MRP-228 to develop and implement a program for reactor vessel internals no later than three years after the initial industry issuance of MRP-227, which was issued in December 2008.

The reactor vessel internals components included in the program are based on EPRI Report 1016596 (MRP-227) and Westinghouse letter LTR-ARIDA-08-63 Rev. 3, (References 8.4, 8.8).

The scope of the program does not include consumable items, such as fuel assemblies, reactivity control assemblies, and nuclear instrumentation, because these components are not typically within the scope of the components that are 309, Re. age of6 3

KLR-KLR-1309A, Rev. 3 Page 9 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION Xl ISI - RVI INSPECTIONS required to be subject to an aging management review, as defined by the criteria set in 10 CFR 54.21(a)(1).

MRP-227 provides the results of the industry initiative via the Materials Reliability Program (MRP) to develop the inspection and evaluation guidelines for managing the long-term aging of pressurized water reactor (PWR) reactor vessel internals. The report provides a generic list of the reactor vessel internals components for Westinghouse PWRs. Westinghouse letter LTR-ARIDA-08-63 Rev. 3 evaluated the MRP-227 generic list of the reactor vessel internals components for applicability to KPS.

Fleet procedures provide the administrative and technical direction for the Reactor Vessel Internals Inspections program (References 8.9 - 8.11). The reactor vessel internals components, including the cast austenitic stainless steel components, in the scope of the program are identified in Table 2 and Table 4.

MRP-227 has been submitted to the NRC for review. Following NRC review and approval, MRP-227 will be revised to incorporate any necessary changes to the guidelines and reissued as MRP-227-A. The Reactor Vessel Internals Inspections program will be revised, as necessary, to incorporate the final recommendations and requirements as published in MRP-227-A .

Dominion will maintain cognizance of industry activities related to PWR internals inspection and aging management and will address/implement industry guidance stemming from those activities, as appropriate under NEI 03-08 practices.

The scope of program will be consistent with the corresponding program elements as described in Fleet GARD ER-AA-AMP-1003, "Aging Management Programs for License Renewal Structures, Systems, and Components." These program elements are the same as the program elements included in MRP-227 Appendix A and Appendix A of NUREG-1800, "Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants."

4.2 PROGRAM ELEMENT 2 - PREVENTIVE ACTIONS The Reactor Vessel Internals Inspections program is a condition monitoring program and does not include any preventive or mitigative actions.

MRP-227 does not specify any preventive actions other than their applicability limitations to base-loaded plants. However, the guidelines in MRP-227 do rely on PWR water chemistry control to manage SCC and reduce the impact of IASCC.

Therefore, an important adjunct to the aging management methodologies described by guidelines in MRP-227 is PWR water chemistry control. The water chemistry program for PWRs relies on monitoring and control of reactor water

'Any revisions to MRP-227 resufting from the issuance of MRP-227-A will not be incorporated for any inspections that are completed before MRP-227-A is issued.

KL-13IIIRvI3Pae10of6 KLR-1309A, Rev. 3 Page 10 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION Xl ISI - RVI INSPECTIONS chemistry as presented in Chapter XI.M2, "Water Chemistry," of NUREG-1801, Volume 2.

It is noted that KPS Primary Water Chemistry program manages the aging effects of cracking, loss of material, and reduction of heat transfer for nickel alloys, stainless steel and steel components due to stress corrosion cracking (SCC), including primary water stress corrosion cracking (PWSCC), and irradiation-assisted stress corrosion cracking (IASCC) pitting and crevice corrosion, and fouling (Reference 8.12).

The preventive actions will be consistent with the corresponding element as described in Fleet GARD ER-AA-AMP-1003, "Aging Management Programs for License Renewal Structures, Systems, and Components." These program elements are the same as the program elements included in MRP-227 Appendix A and Appendix A of NUREG-1800, "Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants."

4.3 PROGRAM ELEMENT 3- PARAMETERS MONITOREDIINSPECTED The Reactor Vessel Internals Inspections program manages the aging effects of Changes in dimensions due to void swelling: cracking due to stress corrosion cracking (SCC), primary water stress corrosion cracking (PWSCC), irradiation-assisted stress corrosion cracking (IASCC), and fatigue; loss of fracture toughness due to neutron irradiation embrittlement and thermal aging embrittlement; loss of material due to wear; and loss of preload due-to stress relaxation.

The aging effects related to reactor vessel internals components in the program are identified in Table 2 and Table 4.

The parameters monitored/inspected will be consistent with the corresponding element as described in Fleet GARD ER-AA-AMP-1003, "Aging Management Programs for License Renewal Structures, Systems, and Components." These program elements are the same as the program elements included in MRP-227 Appendix A and Appendix A of NUREG-1 800, "Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants."

4.4 PROGRAM ELEMENT 4- DETECTION OF AGING EFFECTS The Reactor Vessel Internals Inspections program performs one-time, periodic, and conditional examinations using visual, surface, and ultrasonic examination techniques in accordance with the ASME Code, Section Xl and MRP-227.

The ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD program conducts a visual VT-3 examination of the reactor vessel removable core support structures under Table IWB-2500-1, Examination Category B-N-3, once per Inservice Inspection interval (Reference 8.13). The visual VT-3 examination determines the general mechanical and structural condition of the components by inspecting for structural distortion or displacement of parts, 1 f6 KL-i 09,Rv 3Pg KLR-1309A, Rev. 3 Page 11 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION XI ISI - RVI INSPECTIONS loose/missing, cracked or fractured parts, loose/missing bolting or fasteners, debris, corrosion or erosion, and wear of mating surfaces. The inspections are performed by qualified personnel following procedures consistent with the ASME Code and 10 CFR Part 50, Appendix B.

Additional reactor vessel internals inspection requirements have been promulgated by MRP-227. MRP-227 developed three important precursor elements as a basis for the reactor vessel internals inspection requirements:

screening criteria, categorization of PWR vessel internals, and functionality assessment of components and assemblies of components. The evaluation of the three precursor elements resulted in the assignment of the reactor vessel internals components into four inspection categories:

Primary - those reactor vessel internals that are highly susceptible to the effects of at least one of the eight aging mechanisms.

Expansion - those reactor vessel internals that are highly or moderately susceptible to the effects of at least one of the eight aging mechanisms, but for which functionality assessment has shown a degree of tolerance to those effects.

" Existing Programs - those reactor vessel internals that are susceptible to the effects of at least one of the eight aging mechanisms and for which existing generic and plant-specific aging management programs are capable of managing those effects.

No Additional Measures - those reactor vessel internals for which the effects of all eight aging mechanisms are below the screening criteria.

MRP-227 provided appropriate recommendations for aging management for each category (Reference 8.17). The categorization and resulting inspection requirements described above do not supersede the ASME Section XI, Inservice Inspection requirements for reactor vessel internals components.

The MRP-227 guidelines were based on a broad set of assumptions about plant operation, which encompass the range of current plant conditions for the U.S.

domestic fleet of PWRs. The functionality analyses and supporting aging management strategies in MRP-231 and MRP-232 provided the basis for the MRP-227 guidelines (References 8.14, 8.15, 8.17). The following table compares KPS to the general assumptions used in the analysis.

ae1 f6 KLI 0A e.

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LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION XI ISI - RVI INSPECTIONS Table I - Comparison to MRP-227 Assumptions MRP Assumption KPS Comparison Assumption Met 30 years of operation with high KPS used new fuel at the Yes leakage core loading patterns peripheral locations during then Fuel Cycles I and 2. Once remaining 30 years of burned fuel was used at operation with a low-leakage peripheral locations during fuel management strategy. Fuel Cycles 3 through 15. In Fuel Cycle 16 KPS switched to use of a low leakage core design. KPS continues to use a Low Leakage core design for all subsequent fuel cycles.

Base load operation KPS has implemented Base Yes load operation over the life of the plant.

No design changes affecting Design changes to the KPS Yes Reactor Vessel Internals Reactor Vessel Internals components beyond those components have been limited identified in general industry to Westinghouse replacement guidance or recommended by of split pins, installation of the original vendors flexure less inserts, and replacement of the reactor vessel head.

The Reactor Vessel Internals Inspections program will perform the inspections of the reactor vessel internals in the Primary and Expansion categories consistent with the requirements of MRP-227. The aging management methodologies described in MRP-227 are based on either existing inservice examinations required by the ASME Code,Section XI or on well-documented and well-demonstrated examination methods with which the industry has considerable experience. The Reactor Vessel Internals Inspections program will perform visual examinations, surface examinations, volumetric examinations, and physical measurements. Visual examinations (VT-3) will be used to detect the general degradation conditions. Visual and enhanced visual examinations (VT-1 and EVT-1, respectively) will be conducted to detect discontinuities and imperfections on the surface of components. The surface examinations further characterize discontinuities on the surface of components, and the volumetric inspections 13 fI6 309, KLR- Rv. 3Pag KLR-1309A, Rev. 3 Page 13 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION XI ISI - RVI INSPECTIONS indicate the presence of discontinuities or flaws throughout the volume of material. Some aging effects may involve changes in clearances, settings, and physical displacements that can be monitored by visual means, supplemented by physical measurements.

The specific inspection methodologies and requirements for the NDE examinations of the components in the Primary and Expansion categories will be consistent with the guidance with EPRI Report 1016609 (MRP-228) (Reference 8.16). New procedures will be developed to perform the NDE examinations of the components in the Primary and Expansion categories as identified in Table 2 and Table 4.

The schedule for inspection of the reactor vessel internals components in the Primary and Expansion categories will be consistent with the requirements of MRP-227, as shown in Table 3.

Table 2 provides a breakdown of how each component maps to the four categories: Primary, Expansion, Existing, No Additional Measures. Those reactor vessel internals components included in the Existing Programs or No Additional Measures categories will require no additional measures for future inspections other than the ASME Code inspections per Section XI, Examination Category B-N-3 for removable internal structures, which were previously discussed.

The CASS items are identified in Table 2 as Expansion, Existing, or No Additional Measures. Per guidance from MRP-227, the CASS item in the Expansion category will be inspected if and when the Primary criteria for the RCCA guide tubes lower flange weld is exceeded. The CASS item in the Existing category will be inspected under the Section XI program on a frequency of approximately once every 10 years. Per MRP-227, the CASS items in the No Additional Measures category do not require scheduled inspections.

For the CASS items in the Expansion, Existing, or No Additional Measures category they will be inspected per the guidance in MRP-227 or evaluated per the following criteria.

For reactor vessel internal CASS components, in the Expansion and Existing category, that meet one of the following criteria: 1) have a neutron fluence of greater than [1017 n/cm 2] 2 (E>1 MeV); OR 2) are determined to be susceptible to thermal embrittlement; the Reactor Vessel Internals Inspections program will:

Perform a component-specific evaluation, including a mechanical loading assessment to determine the maximum tensile loading on the component during ASME Code Level A, B, C, and D conditions. If the loading is compressive or low enough (<5 ksi) to preclude fracture, then supplemental inspection of the component is not required.

2 Per MRP-175, Materials Reliability Program: PWR Internals Material Degradation Mechanism 2

Screening and Threshold Values - EPRI Report 1012081, 2005, a threshold value of 1020 n/cm (E>1 MeV) will be used for screening of neutron fluence for susceptibility to thermal aging embrittlement pending agreement with NRC.

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LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION XI ISI - RVI INSPECTIONS If the above criterion is not met, a supplemental inspection will be performed for the portions of the susceptible components determined to be limiting from the standpoint of thermal aging susceptibility (i.e., ferrite and molybdenum contents, casting process, and operating temperature),

neutron fluence, and cracking susceptibility (i.e., applied stress, operating temperature, and environmental conditions). The inspection technique will be capable of detecting the critical flaw size with adequate margin. The critical flaw size is determined based on the service loading condition and service-degraded material properties.

For CASS components that meet neither of the two criteria specified above, the existing ASME Section XI inspection requirements are adequate in accordance with NUREG-1 801, XI.M13.

Table 2 and Table 4 identify the reactor vessel internals components in the program, the aging effects applicable to each component, the examination techniques used to detect the aging effect(s), and the examination schedule for the Primary and Expansion category components.

The Reactor Vessel Internals Inspections program inspections are performed in accordance with the ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD program, which includes the procedures for performing the inspections, the acceptance criteria for each examination technique, and the review and disposition of inspection results.

The personnel performing the inspections will be trained in accordance with the guidance in MRP-228.

The detection of aging effects will be consistent with the corresponding element as described in Fleet GARD ER-AA-AMP-1003, "Aging Management Programs for License Renewal Structures, Systems, and Components." These program elements are the same as the program elements included in MRP-227 Appendix A and Appendix A of NUREG-1800, "Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants."

4.5 PROGRAM ELEMENT 5 - MONITORING AND TRENDING The Reactor Vessel Internals Inspections program performs one-time, periodic, and conditional examinations as scheduled and performed in accordance with the ASME Code,Section XI as well as MRP-227 to provide timely detection of aging effects.

In addition to the MRP-227 Primary components, Expansion components have been defined by MRP-227 should the scope of examination and re-examination need to be expanded beyond the Primary group due to detection of significant aging effects found in Primary components with the same material and potential aging mechanism.

Table 2 and Table 4 identify the reactor vessel internals components in the program, the aging effects applicable to each component, the examination KLR-1309A, Rev. 3 Page 15 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION X1 ISI - RVI INSPECTIONS techniques used to detect the aging effect(s), and the examination schedule for the Primary and Expansion category components.

Those reactor vessel internals components included in the Existing Programs or No Additional Measures categories will be inspected on a schedule consistent with the requirements of the ASME Code for Section X1, Examination Category B-N-3, for removable internal structures, as outlined in the ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD program. This schedule is consistent with the recommendations of MRP-227.

The EPRI MRP Reactor Internals Inspection and Evaluation Guidelines Core Group will evaluate the results of the first and subsequent rounds of augmented examinations as documented in Section 7 of MRP-227. This evaluation may result in recommendations for changes in the examination periodicity, which would be evaluated for applicability to KPS (Reference 8.17).

Flaw indications detected during the required examinations are dispositioned in accordance with the Corrective Action Program.

The monitoring and trending will be consistent with the corresponding element as described in Fleet GARD ER-AA-AMP-1003, "Aging Management Programs for License Renewal Structures, Systems, and Components." These program elements are the same as the program elements included in MRP-227 Appendix A and Appendix A of NUREG-1800, "Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants."

4.6 PROGRAM ELEMENT 6 - ACCEPTANCE CRITERIA As discussed in Section 4.4, the Reactor Vessel Internals Inspections program is implemented in accordance with the ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD program that include the inspection procedures, the acceptance criteria for each examination technique, and the process for evaluating unsatisfactory inspection results. For ASME Code Section XI inspections, indications and relevant conditions detected during examination are required to be evaluated in accordance with ASME Section XI, Article IWB-3500.

In addition to the ASME Code Section XI requirements, MRP-227, MRP-228, WCAP-17096, and WCAP-17020-P provide the examination acceptance criteria for the Primary and Expansion components and the expansion criteria for expanding the examinations beyond the Primary components to include the Expansion components (References 8.18, 8.19).

The guidance in MRP-227 contains three types of examination acceptance criteria:

For visual examination (and surface examination as an alternative to visual examination), the examination acceptance criterion is the absence of any of the specific, descriptive relevant conditions; in addition, there are requirements to record and disposition surface breaking indications that detected and sized for length by VT-1/EVT-1 examinations; ii-09,Rv.3Pge1o6 KLR-1309A, Rev. 3 Page 16 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION Xl ISI - RVI INSPECTIONS

" For volumetric examination, the examination acceptance criterion is the capability for reliable detection of indications in bolting, as demonstrated in the examination Technical Justification; in addition, there are requirements for system-level assessment of bolted or pinned assemblies with unacceptable volumetric (UT) examination indications that exceed specified limits; and

For physical measurements, the required examination acceptance criterion is related to the hold down spring. At KPS the hold down spring is fabricated from alloy 403SS and therefore does require inspection.

For the examination of the baffle-former bolts and the baffle-former edge bolts, a plant specific evaluation to determine the minimum acceptable bolting pattern in accordance with the NRC reviewed and approved methodology in WCAP-15029-P-A has been completed and is documented in WCAP-15425, (References 8.20, 8.21).

Table 2 and Table 4 identify the specific reactor vessel internals components in the program, the aging effects applicable to each component, the examination techniques used to detect the aging effect(s), and the additional examination acceptance criteria and inspection expansion criteria for the Primary and Expansion category components.

The acceptance criteria will be consistent with the corresponding element as described in Fleet GARD ER-AA-AMP-1003, "Aging Management Programs for License Renewal Structures, Systems, and Components." These program elements are the same as the program elements included in MRP-227 Appendix A and Appendix A of NUREG-1800, "Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants."

4.7 PROGRAM ELEMENT 7 - CORRECTIVE ACTIONS As discussed in Section 4.4, the Reactor Vessel Internals Inspections program is implemented for applicable components in accordance with the ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD program that include the inspection procedures, the acceptance criteria for each examination technique, and the process for evaluating unsatisfactory inspection results.

For the additional MRP-227-required inspection of the Primary and Expansion category components, MRP-227 provides information on methodology that can be used for the evaluation of detected conditions that exceed the examination acceptance criteria. The flaw evaluation methodology accounts for the accumulated neutron exposure and the resulting loss of fracture toughness due to radiation embrittlement in assessing the suitability of the component for continued service. Justification for flaw evaluation fracture toughness limits is provided in Section 6 of MRP-227.

Any detected conditions that do not satisfy the examination acceptance criteria are required to be dispositioned through the plant corrective action program, which may require repair, replacement, or analytical evaluation for continued f6 KLR-1309A Re.3Pge1 KLR-1309A, Rev. 3 Page 17 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION XI ISI - RVI INSPECTIONS service until the next inspection. The disposition will ensure that design functions of the reactor internals components will continue to be fulfilled for all licensing basis loads and events. Other alternative corrective action bases may be used to disposition relevant conditions if they have been previously approved or endorsed by the NRC. For example, previous NRC-endorsed alternative corrective actions bases include the corrective actions bases for Westinghouse-design RVI components that are defined in Tables 4-1, 4-2, 4-3, 4-4, 4-5, 4-6, 4-7, and 4-8 of Westinghouse Report No. WCAP-14577 Rev 1-A (Reference 8.22).

The Quality Assurance Program and sub-tier procedures fulfill and implement the requirements of Criterion XVI of 10 CFR Part 50, Appendix B. The process includes provisions for timely evaluation of adverse conditions and implementation of corrective actions, including root cause determinations and prevention of recurrence, where appropriate. Provisions are in place for tracking, monitoring, reviewing, and approving corrective actions to ensure effective corrective actions are taken, along with monitoring the process for adverse trends (References 8.23, 8.24).

The corrective actions will be consistent with the corresponding element as described in Fleet GARD ER-AA-AMP-1003, "Aging Management Programs for License Renewal Structures, Systems, and Components." These program elements are the same as the program elements included in MRP-227 Appendix A and Appendix A of NUREG-1800, "Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants."

4.8 PROGRAM ELEMENT 8 - CONFIRMATION PROCESS As discussed in Section 4.4, the Reactor Vessel Internals Inspections program is implemented in accordance with the ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD program that include the inspection procedures, the acceptance criteria for each examination technique, and the process for evaluating unsatisfactory inspection results.

The corrective action program, as presented above, includes provisions to ensure that all significant conditions adverse to quality receive a cause determination and that corrective actions are implemented to preclude reoccurrence (References 8.23, 8.24).

The confirmation process will be consistent with the corresponding element as described in Fleet GARD ER-AA-AMP-1003, "Aging Management Programs for License Renewal Structures, Systems, and Components." These program elements are the same as the program elements included in MRP-227 Appendix A and Appendix A of NUREG-1800, "Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants."

4.9 PROGRAM ELEMENT 9 -ADMINISTRATIVE CONTROLS The Quality Assurance Program and sub-tier procedures fulfill and implement the requirements of 10 CFR Part 50, Appendix B. Administrative controls include ag 1 ot6 ev 3 09, KLR1 KLR-1309A, Rev. 3 Page 18 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION Xi ISI - RVI INSPECTIONS management of document issue, revision, review and approval (References 8.23 and 8.25 - 8.28).

The MRP-227 requirements are administered by Dominion using Fleet Reactor Internals Inspection Program Description, ER-AA-RII-10 and Fleet Reactor Internals Inspection Administrative Procedure ER-AA-RII-1 01.

The administrative controls will be consistent with the corresponding element as described in Fleet GARD ER-AA-AMP-1003, "Aging Management Programs for License Renewal Structures, Systems, and Components." These program elements are the same as the program elements included in MRP-227 Appendix A and Appendix A of NUREG-1800, "Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants."

4.10 PROGRAM ELEMENT 10 - OPERATING EXPERIENCE The Reactor Vessel Internals Inspections program is a new program. Therefore, no internal programmatic experience is available. As operating experience is obtained, lessons learned will be used to adjust this program as needed.

Relatively few incidents of PWR vessel internals aging degradation have been reported in operating U.S. commercial PWR plants. However, a considerable amount of PWR vessel internals aging degradation has been observed in European PWRs, with emphasis on cracking of baffle-former bolting. Reactor internals failures (both domestic and international), research data, and vendor evaluations have been considered in the development of MRP-227, which forms the basis for the Reactor Vessel Internals Inspections program.

The PWROG has performed a survey under PA-MSC-0568 to gather reactor internals inspection results to date. Reactor vessel internals inspection results performed under MRP-227 in the future will be reported to the EPRI MRP Issue Program within 120 days following completionof the inspections. The EPRI MRP program plans to publish future inspection results under MRP-219 (Reference 8.29).

The Operating Experience Program ensures that additional operating experience is factored into the aging management programs to ensure program effectiveness (Reference 8.30).

5.0 PROGRAM ENHANCEMENTS The Reactor Vessel Internals Inspection program is a new plant-specific program that will be consistent with the applicable ten elements as described in Fleet GARD ER-AA-AMP-1003, "Aging Management Programs for License Renewal Structures, Systems, and Components." These program elements are the same as the program elements included in MRP-227 Appendix A and Appendix A of KLR-1309A, Rev. 3 Page 19 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION Xl ISI - RVI INSPECTIONS NUREG-1800, "Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants."

6.0 PROGRAM EXCEPTIONS The Reactor Vessel Internals Inspection program is a new plant-specific aging management program.

7.0 CONCLUSION

The Reactor Vessel Internals Inspections program ensures that the effects of aging associated with the in-scope components will be adequately managed so that there is reasonable assurance that their intended functions will be maintained consistent with the current licensing basis throughout the period of extended operation.

8.0 REFERENCES

8.1 10 CFR Part 54, "Requirements for Renewal of Operating Licenses for Nuclear Power Plants," Code of Federal Regulations, U.S. Nuclear Regulatory Commission, Washington, D.C.

8.2 Central Reporting System, Condition Report 322266, "Mandatory and Needed Requirements for Reactor Vessel Internals," February 3, 2009.

8.3 Nuclear Fleet Guidance and Reference Document ER-AA-AMP-1003, "Aging Management Programs for License Renewal Structures, Systems, and Components."

8.4 EPRI Report 1016596, "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-Rev. 0)," December 2008, Electric Power Research Institute, Palo Alto, California 8.5 NUREG-1800, Revision 1, "Standard Review Plan for Review of JLicense Renewal Applications for Nuclear Power Plants," U. S. Nuclear Regulatory Commission, September 2005.

8.6 NEI 03-08, "Guideline for the Management of Materials Issues," April 2007.

8.7 Technical Report KLR-1309, "ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD."

8.8 Memo from Westinghouse Advanced Reactor Internals Design & Analysis to Donna Rogosky, "Summary Report for the Fabrication and Design Information for KPS Reactor Vessel Internals," dated December 5, 2008, [LTR-ARIDA-08-63 Rev. 3].

8.9 Dominion Equipment Reliability (ER) Peer Group Project Charter ER-31, "Fleet Reactor Internals Inspection Program."

f6 Rev 3iae2 KLR-1309A,~ii KLR-1309A, Rev. 3 Page 20 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION Xl ISI - RVI INSPECTIONS 8.10 Nuclear Fleet Program Description ER-AA-RII-10, "Fleet Reactor Internals Inspection."

8.11 Nuclear Fleet Administrative Procedure ER-AA-RII-101, "Fleet Reactor Internals Inspection Program."

8.12 Technical Report KLR-1 310, "Primary Water Chemistry."

8.13 Nuclear Fleet Nondestructive Examination Procedure ER-AA-NDE-VT-608, 'VT-3 Visual Examination Procedure for Examination Category B-N-i, Interior of Reactor Vessel."

8.14 EPRI Report 1016592, "Materials Reliability Program: Aging Management Strategies for B&W PWR Internals (MRP-231)," Electric Power Research Institute, Palo Alto, CA.

8.15 EPRI Report 1016593, "Materials Reliability Program: Aging Management Strategies for Westinghouse and Combustion Engineering PWR Internals (MRP-232)," Electric Power Research Institute, Palo Alto, CA.

8.16 EPRI Report 1016609, "Materials Reliability Program: Inspection Standard for PWR Internals (MRP-228)," July 2009, Electric Power Research Institute, Palo Alto, CA.

8.17 Letter from Terry McAlister (EPRI) to Tanya M. Mensah (NRC), "EPRI MRP Responses to: Request for Additional Information Re: Electric Power Research Institute Topical Report 1016596, 'Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-REV.

0)' (TAC NO. ME0680), August 24, 2009," dated February 1, 2010 [MRP 2010-004].

8.18 WCAP-17096, Revision 0, "Reactor Internals Acceptance Criteria Methodology and Data Requirements".

8.19 WCAP-17020-P, Revision 0, "Point Beach Unit I Upper Internals Guide Tube -

Guide Card Wear Evaluation".

8.20 WCAP-15029-P-A, 'Westinghouse Methodology for Evaluating the Acceptability of Baffle-Former-Barrel Bolting Distributions under Faulted Loaded Conditions".

8.21 WCAP-15425, "Determination of Acceptable Baffle-Barrel-Bolting for Kewaunee and Prairie Island Plants."

8.22 WCAP-14577, Revision 1-A, "License Renewal Evaluation: Aging Management for Reactor Internals."

8.23 Topical Report DOM-QA-1 "Dominion Nuclear Facility Quality Assurance Program Description."

8.24 Nuclear Fleet Administrative Procedure PI-AA-200, "Corrective Action" 8.25 KPS Nuclear Administrative Directive NAD-03.01, "Directive, Implementing Document, and Procedure Control."

8.26 KPS General Nuclear Procedure GNP-03.01.01, "Directive, Implementing Document, and Procedure Administrative Controls."

KLR-1309A, Rev. 3 Page 21 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION Xl ISi - RVI INSPECTIONS 8.27 DNAP-0501, "Dominion Nuclear Procedure Administrative Control Program."

8.28 Nuclear Fleet Program Description AD-AA-10, "Administrative Controls Program."

8.29 MRP-219, "Inspection Data Survey Report."

8.30 Nuclear Fleet Guidance and Reference Document PI-AA-100-1007, "Operating Experience Program."

9.0 ATTACHMENTS Attachment 1 - Implementing Procedures Attachment 2 - Associated Aging Management Programs Attachment 3 - Materials and Aging Effects Managed Attachment 4 - Integrated Plant Assessment Reports / Systems or Commodities Attachment 5- Detailed NUREG-1800 Program Evaluation Page 22 of 62 KLR-1309A, Rev. 33 KLR-1 309A, Rev. Page 22 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION Xl ISI - RVI INSPECTIONS Table 2 - Reactor Vessel Internals Evaluation RVI Component LR Aging LR Aging Existing MRP-227 Information CASS Effect Mechanism Inspections Components(t1 Cte Required Inspections Ispec tion Schedule Procedure Requirements Changes in void swelling Baffle Former New dimensions Assembly procedure8 Baffle Former

Baseline required to Assembly examination perform the Primary SVT_3(6) VT-3 between 20 and stress corrosion

Baffle Former inspections.

Core side 40 EFPY.

cracking, irradiation- Assembly Cracking assisted stress surface as

Subsequent corrosion cracking Indicated examinations on a ten-year interval.

Baffle Former Assembly Loss of neutron irradiation Baffle-edge bolts Baffle-edge bolts Including: fracture embrittlement, void and locking devices and locking devices toughness swelling e Baffle Assembly Primary VT-3(6) Baseline bafflelformer Bolts and locking examination Baffle-edge plates ER-AA-NDE- bolts and devices on high between 20 and a Baffle-edge VT-608 locking fluence seams. 40 EFPY.

Loss of pitting and crevice Subsequent bolts including material corrosion devices 100% of

bolting lock bars components examinations on accessible from a ten-year

Baffle-former core side. interval.

bolts including bolting lock bars Baffle-former bolts New NAM( 31 Baffle-former bolts procedure8

" Baseline

Baffle bolting " UT required to examination lock bars

100% of between 25 and perform the Primary accessible bolts 35 EFPY. UT Loss of or as supported inspections.

preload stress relaxation

Baffle-former " Subsequent bolts by plant specific examination after justification.

Expansion 10 to 15 Heads

Baffle-former additional EFPY accessible from bolts the core side. Re-examination for high leakage core Page 23 of 62 KLR-1309A, Rev. Rev. 33 Page 23 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION XI ISI - RVI INSPECTIONS Table 2 - Reactor Vessel Internals Evaluation RVI Component LR Aging LR Aging Existing MRP-227 Informatlon CASS Effect Mechanism Inspections Components(2)

Category() [ Required Inspection Schedule Procedure designs requires continuing examinations on a ten-year interval.

KL-39,Rv Page24If 6 KLR-1309A, Rev. 3 Page 24 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION XI ISI - RVI INSPECTIONS Table 2 - Reactor Vessel Internals Evaluation RVI Component LR Aging LR Aging Existing MRP-227 Information CASS Effect Mechanism Inspections Components(21 1

Category( ) Required Inspection Schedule Procedure Inspections I i Requirements Bottom-Mounted Changes in void swelling Instrumentation dimensions columns NAM Including: stress corrosion Except as listed

BMI column Cracking cracking, irradiation- below.

assisted stress bodies corrosion cracking

" BMI column bolts Loss of thermal aging and

" BMI column fracture neutron irradiation collars toughness embdttlement

" BMI column Loss of pitting and crevice cruciforms material corrosion

" BMI column extension bars ER-AA-NDE- BMI column bodies BMI column bodies CASS - BMI

" BMI column VT-608 o VT.3(8)

As indicated by column extension tubes difficulty of New cruciforms insertion/withdra procedure 8

" BMI column lock Expansion column bodies wal of flux required to caps* BMI column for which thimbles. Flux perform the

" BMI column bodies difficulty detected isduring thimble withdrawalinsertion/

to be VT-3 inspections.

nuts flux thimble monitored at Insertion/ each inspection withdrawal. Interval.

KLR1i09ievr3olgl2 oI6 KLR-1309A, Rev. 3 Page 25 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION XI ISI - RVI INSPECTIONS Table 2 - Reactor Vessel Internals Evaluation LR Aging LR Aging Existing MRP-227 Information CASS RVI Component Effect Mechanism Inspections Componentst 2 )

Category(l) Required Inspection Schedul Procedure Inspections(7 n Requirements NAM Changes in void swelling Except as listed dimensions below.

Core barrel flange New Core barrel flange weld procedure8 weld Initial required to 9 EVT-1 examination and perform the Expansion-

100% of one side reexamination EVT-1I Core barrel of the accessible frequency inspections.

Core Barrel flange weld surfaces of the dependent on the selected weld examination Including: and adjacent results for upper

" Core Barrel stress corrosion base metal. core barrel flange cracking, irradiation- flange.

Cracking

" Core Barrel assisted stress corrosion cracking ER-AA-NDE- Core Barrel outlet New outlet nozzles Core Barrel outlet VT-608 nozzles procedure8

Lower Core nozzles

EV/T-1 Initial required to Barrel perform the Expansion-

100% oexamination and

Upper Core - 100% of one side reexamination EVT-1 Barrel

Core barrel of the accessible frequency inspections.

outlet nozzles surfaces of the dependent on the

Ring Sector selected weld examination and adjacent results for upper base metal. core barrel flange Lower core barrel Lower core barrel New flange weld flange weld procedure 8 Expansion-Loss of neutron irradiation Iinitial required to fracture

Lower core

EVT-1 perform the embrittlement, void examination and toughness swelling barrel flange

100% of one side EVT-1 weld reexamination of the accessible frequency inspections.

surfaces of the dependent on the Page 26 of 62 KLR-1 309A, Rev.

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LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION Xl ISl - RVI INSPECTIONS Table 2 - Reactor Vessel Internals Evaluation RVI Component LR Aging LR Aging Existing MRP-227 Information CASS Effect Mechanism Inspections * - Components(2)

(1) ReuiredProcedure Ctgr(

Category nsRequiredn( 7 )

Inspections Inspection Schedule Requirements Rqurmet selected weld examination and adjacent results for upper base metal, core barrel Upper core barrel New Upper core barrel flange weld procedure8 flange weld

No later than 2 required to

EVT-1 refueling outages perform the Primary from the EVT-1 Upper core

100% of one side beginning of the inspections.

barrel flange of the accessible surfaces of the license renewal weld period.

selected weld and adjacent Subsequent base metal. examination on a ten-year Interval.

Page 27 of 62 KLR-1309A, Rev. 3 KLR-1 309A, Rev. 3 Page 27 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION XI ISI - RVI INSPECTIONS Table 2 - Reactor Vessel Internals Evaluation RVI Component LR Aging LR Aging Existing MRP-227 Information CASS Effect Mechanism Inspections Components(2 )

(1) ReuiredProcedure 7 Requirements Category( ) Requiredo )

Inspections Inspection Schedule Rqurmet Interfacing NAM Components Changes in void swelling

Except as dimensions Including: noted below.

Clevis inserts 9 CJevis Insert stress corrosion Existing(4) bolts Cracking cracking, assistedirradiation-stress Existing ert

Clevis insert s Clevis insert corrosion cracking bolts lock keys

Head and Loss of neutron irradiation ER-AA-NDE-vessel fracture embrittlement, void VT-608 alignment pins toughness swelling

" Head and vessel Primary alignment pin NA - KPS does not bolts e Hold-down spring (30488 have down 304SS hold-springsý5).

" Head and vessel Loss of only) alignment pin preload stress relaxation lock cups

Hold-down spring f6 KLR-I 30A e.3Pae2 KLR-1309A, Rev. 3 Page 28 of 62

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RequiredProcedure Category( 1 ) Required npcin* Inspection Schedule Requirements Irradiation specimen guide Including:

Irradiation specimen guide ER-AA-NDE-bolts VT-608 NAM Irradiation specimen lock caps Irradlation specimen plugs Changes digensions in dimensions void swelling Lower core plate Including: stress corrosion

" Fuel alignment cracking, primary pins water stress Cracking corrosion cracking, ER-AA-NDE- NAM

" Fuel alignment Irradiation-assisted VT-608 Existing pin bolts stress corrosion

Fuel alignment cracking pin lock caps, Loss of neutron Irradiation fracture embrittlement, void toughness swelling IL-139 Rv 3rPage 29 ofI62 KLR-1309A, Rev. 3 Page 29 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION XI ISI - RVI INSPECTIONS Table 2 - Reactor Vessel Internals Evaluation RV] Component LR Aging LR Aging Existing MRP-227 Information CASS Effect Mechanism Inspections Components(21 (g) 1 Required Procedure Inspectionsr 7 ) Inspection Schedule Requirements NAM Changes in dimensions void swelling

Except as noted below.

Column bodies New 8

- Initial procedure Lower core plate Column bodies examination and required to support columns stress corrosion reexamination perform the cracking, irradiation- Expansion EVT-1 frequency EVT-1 assisted stress Ing Column 100% of dependent onthe inspections

" Column bodies corrosion cracking bodies accessible examination

Column bolts surfaces. results for upper

" Lug core barrel flange weld.

" Nuts (Bearing Nut) Loss of neutron irradiation Column bolts Column bolts New 8

" Sleeves fracture embrittlement, void

Initial and Procedure requIred to 0 UT subsequent toughness swelling Expansion

100% of examinations perform the Column bolts accessible bolts dependent on UT inspections Loss of stress relaxation or as supported results of baffle-preload by plant-specific former bolt justification, examinations.

KLR-139A, Rv. 3 age 3 of 6 KLR-1309A, Rev. 3 Page 30 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION X! ISI - RVI INSPECTIONS Table 2 - Reactor Vessel Internals Evaluation RVI Component LR Aging LR Aging Existing MRP-227 Information CASS Effect Mechanism Inspections Components( 21 i) Required (7Procedure Categorynlnspections Schedule Requirements Lower Internals void swelling Including: Changes in stress corrosion

" Diffuser plate dimensions cracking, Irradiation-assisted stress

" Flux thimble corrosion cracking tubes

" Flux thimble tube plugs stress corrosion

Head cooling Craking cracking, irradiation-spray nozzles assisted stress ER-AA-NDE- NAM

Lower support corrosion cracking VT-608 Existing forging (13 inch tk. plate)

Radial support keys Loss of neutron Irradiation

Radial support fracture embrittlement, void key bolts toughness swelling

Radial support key lock keys KLR-309, Re. 3Page31 f 6 KLR-1309A, Rev. 3 Page 31 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION Xl ISI - RVI INSPECTIONS Table 2 - Reactor Vessel Internals Evaluation RVI Component LR Aging LR Aging Existing CASS Effect Mechanism Inspections MRP-227 Information Components(2)

Category(l) Required ( Inspection Schedule redurem RCCA guide Changes in NAM tubes dimensions void swelling Except as Includes: noted below.

" Anti-rotation Flanges-lower New studs and nuts

otswelds Flanges-lower welds wlspoeue procedure8 Bolts t welds No later than 2 required to

" C tubes 9 E'T-1 refueling outages perform the Primary - 100% of outer from the EVT-1

" Enclosure pins Flanges-lower (accessible) beginning of the inspections

Upper guide welds CRGT lower license renewal tube flange weld period.

enclosures surfaces and Flanges-gnemdaes adjacent base meaexamination eSubsequent ona intermediate metal. ten-year interval.

" Flanges-lower stress corrosion Guide plates/cards New p

Guides/cak Cracking cracking, assistedIrradiation-stress Guide plates/cards uNNo later than 2 lateran 2 procedurew required to plates/cards corrosion cracking e refueling outages rform th

Housing plates - VT-3(') from the perform the

Inserts

20% examination beginning of the VT3 of the number of license renewal inspections Lock bars Primary CRGT period, and no

Sheaths

Guide assemblies, with earlier than two

Support pins plates/cards all guide cards refueling outages within each prior to the start Support pin selected CRGT of the license cover plate assembly renewal period.

Support pin examined. Subsequent cover plate exams on a ten-locking caps year basis.

KLR-1~~~~ e.3Pae3 I0A f6 KLR-1309A, Rev. 3 Page 32 of 62

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Effect Mechanism Required( Inspection Schedule Procedure Category(l Inspectionsequirements and tie straps

" Support pin nuts

" Water flow slot ligaments

" Flexures NA Loss of thermal aging and Primary KPS does not have

" Flexureless fracture neutron irradiation P ria KPS de thae Inserts toughness embrittlement

Flexures RCCA tfiexu res(guide 5

]. tubes Page 33 of 62 KLR-1 309A, Rev.

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Category(t) inspections(7)

Required Inspection Schedule IsetoScdue Procedure Requirements Secondary Core Changes in void swelling Support (SCS) dimensions Assembly Including: stress corrosion cracking, irradiation-

SCS base plate Cracking assisted stress "SCS bolts corrosion cracking ER-AA-NDE- NAM

" SCS energy VT-608 absorber

SCS guide post Loss of neutron Irradiation

" SCS housing fracture embrittlement, void

" Tie Plate toughness swelling

" SCS lock keys Rv.3I KLR139A agI3IoI6 KLR-1309A, Rev. 3 Page 34 of 62

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Category Required (7) Inspection Schedule Procedure Inspections7 p Requirements Thermal shield Changes In void swelling Including: dimensions NAM

Thermal shield stress corrosion

Except as bolts cracking, irradiation- noted below.

Thermal shield Cracking assisted stress flexures corrosion cracking

Thermal shield dowels Cracking fatigue Thermal shield New ER-AA-NDE- flexurei procedure8

" Thermal shield Loss of neutron Irradiation VT-608

No later than 2 requiredto bumper bar fracture embrittlement, void Thermal shield refueling outages perform the

" Locking Bar toughness swelling Primary flexures from the VT-3

" Adjustment Plug Thermal shield a VT_3(6) beginning of the inspections T hermal shiel license renewal

Neutron panel flexures

100% of thermal period.

" Neutron panel Loss of shield flexures. Subsequent bolts material wear examinations on

" Neutron panel a ten-year lock caps interval.

II 109,Re. Pg 35of6 KLR-1309A, Rev. 3 Page 35 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION Xl ISI - RVI INSPECTIONS Table 2 - Reactor Vessel Internals Evaluation RVI Component LR Aging LR Aging Existing MRP-227 Information CASS Effect Mechanism Inspections Components(2 )

Required Inspection Schedule Procedure Category~l) Requirements inspections(7)

Upper core plate Changes In void swelling Including : dimensions

" Upper core stress corrosion plate alignment cracking, irradiation- Upco keyways and Cracking assisted stress pins corrosion cracking ER-AA-NDE- NAM Upper core

" Upper core VT-608 Existing plate mixing plate fuel guide devices pins Loss of neutron Irradiation

" Upper core fracture embrittlement, void plate mixing toughness swelling devices f6 KL-39, e.3Pge3 KLR-1309A, Rev. 3 Page 36 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION XI ISI - RVI INSPECTIONS Table 2 - Reactor Vessel Internals Evaluation RVI Component LR Aging LR Aging Existing MRP-227 Information CASS Effect Mechanism Inspections Components Category111 Required Inspection Schedule Procedure Inspections Requirements Upper core plate Changes in void swelling instrumentation dimensions columns Including: stress corrosion

" Bolting Cracking cracking, irradiation-assisted stress

" Brackets, corrosion cracking clamps, terminal blocks, and CASS conduit straps ER-AA-NDE- N conduit

" Conduit seal VT-608 NAM support assembly-body,

clamp tubesheets Loss of neutron irradiation

Conduit seal fracture embrittlement, void assembly-tubes toughness swelling

" Conduits

" Flange base

" Locking caps

" Support tubes f6 KL-39, e.3Pge3 KLR-1309A, Rev. 3 Page 37 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION XI ISI - RVI INSPECTIONS Table 2 - Reactor Vessel Internals Evaluation RVI Component LR Aging LR Aging Existing MRP-227 Information CASS Effect Mechanism Inspections Componentsl2 )

() Required InProcedure ynspections()ur Requirements Upper core plate Changes In void swelling support columns dimensions Including stress corrosion Adapters C cracking, Irradiation-0 Bolts assisted stress

Column bases corrosion cracking e Column bodies; Loss of neutron Irradiation CASS Upper Support fracture embrittlement, void ER-AA-NDE- NAM* support Column - Type 11 toughness swelling VT-608 9 Extension tubes

thermocouple

Flanges stop e Lock keys

Thermocouple Loss of stop - located at preload stress relaxation flow mixing device

Nuts 309, Rv. 3Pag 38 f 6 KLR-KLR-1309A, Rev. 3 Page 38 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION XI ISI - RVI INSPECTIONS Table 2 - Reactor Vessel Internals Evaluation RVI Component LR Aging LR Aging Existing MRP-227 Information CASS Effect Mechanism Inspections Components Category l 1 Required Procedure Inspections(7 [nspection Schedule Requirements Upper core NAM support plate Changes In void swelling Except as assembly dimensions noted below.

Including

" Bolts stress corrosion

Deep beam ribs Cracking cracking, Irradiation-assisted stress

" Deep beam corrosion cracking stiffeners ER-AA-NDE-

Upper core VT-608 Existing support plate -

Upper support welded to deep Loss of neutron irradiation ring or skirt beam assy.

fracture embrittlement, void

Look keys toughness swelling

" Upper core support ring or skirt Table Notes:

1. MRP-227 inspection categories are discussed in Section 4.4.

2. The CASS (Cast Austenitic Stainless Steel) components identified in Reference 8.8 for evaluation in accordance with KLR-1309, Section 5.2. A component specific evaluation must be performed determine the susceptibility of each CASS component to thermal embrittlement or an inspection is required to detect cracking as discussed in Section 4.4.

3. NAM is the MRP-227 No Additional Measures category.

4. Existing is the MRP-227 Existing Programs category.

5. Reference 8.8, Table 3: Kewaunee Components.

KLR-1309A, Rev. 3 Page 39 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION XI ISI - RVI INSPECTIONS

6. The MRP-227 inspection recommendations for VT-3 to monitor for the effects of cracking are appropriate because they are limited to cases where the Intent of the examination is to monitor the general condition of the component. The VT-3 examination is specified where the objective of the examination is detection of broken or missing bolt locking devices and welds; protruding bolts; or broken or missing pieces (e.g., supports, spider arms, dowels). VT-3 is not specified where the examination objective the detection of the onset of cracking with accompanying tight crack opening displacements (CODs). V-T-3 as currently defined in Section XI of the ASME Code is capable of this level of detection. The MRP-227 recommendations are consistent with the approach used in the ASME Section XI examinations, which require VT-3 inspections for accessible core support structures (Reference 8.17).

7. The coverage requirements for the required inspections based on MRP-227 are adequate to ensure timely detection of aging effects In the reactor vessel components. Coverage for a visual examination of a single component is defined as the percentage of the target surface area observed in the examination. Consideration of accessibility in terms of target surface is included in the coverage requirements for a number of the components. Additionally, accessibility was a factor in determining coverage for inspection of a variety of bolts, locking devices and lock welds. In these cases, coverage referred to the fraction of components inspected rather than the coverage associated with any particular examination (Reference 8.17).

8. Additional NDE procedures (either fleet, plant-specific, or vendor) will be needed to implement the revised ISI Plan for Reactor Vessel Internals Inspections.

evIIPg 0 f6 KL-30A KI-R-1309A, Rev. 3 Page 40 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION Xi ISI - RVI INSPECTIONS Table 3 - Aging Management Program Enhancement and Inspection Implementation Schedule1 Refueling Target Estimated AMPInspection MethodComment Outage Month/Year EFPY Criteria KR-31 Spring 2011 30.16 Not Applicable Not Applicable MRP-227 Inspections In These items are planned for KR-KR-32 Spring 2012 31.15 RCCA guide tubes Guide Plate/Cards accordance with MRP- 32. If not performed in KR-32 ASME Section XI they will be inspected in KR-33.

MRP-227 inspections in Original license expires KR-33 Fall 2013 32.64 Baffle Former Bolts accordance with MRP- December 21, 2013. Expansion 228 specifications and scope will be implemented on an ASME Section Xi as needed basis.

Baffle Former Assembly, Baffle-edge bolts and locking MRP-227 Inspections in Core barrel removed this outage.

KR-34 Spring 2015 34.03 devices, Baffle-former bolts, Upper core barrel flange weld, accordance with MRP- Expansion scope will be RCCA guide tubes flanges lower welds, RCCA guide tubes 228 specifications and implemented on an as needed guide plates/ cards, Thermal shield flexures, CASS items ASME Section Xl basis.

KR-35 Fall 2016 35.52 Not Applicable Not Applicable KR-36 Spring 2018 36.91 Not Applicable Not Applicable KR-37 Fall 2019 38.40 Not Applicable Not Applicable KR-38 Spring 2021 39.79 Not Applicable Not Applicable KR-39 Fall 2022 41.28 Not Applicable Not Applicable Baffle Former Assembly, Baffle-edge bolts and locking MRP-227 Inspections in Core barrel may be removed this KR-40 Spring 2024 42.67 devices, Baffle-former bolts, Upper core barrel flange weld, accordance with MRP- outage. Expansion scope will be RCCA guide tubes flanges lower welds, RCCA guide tubes 228 specifications and Implemented on an as needed guide plates/ cards, Thermal shield flexures, CASS items ASME Section X] basis.

KR-41 Fall 2025 44.16 Not Applicable Not Applicable KR-42 Spring 2027 45.55 Not Applicable Not Applicable KR-43 Fall 2028 47.04 Not Applicable Not Applicable KR-44 Spring 2030 48.43 Not Applicable Not Applicable KR-45 Fall 2031 49.92 Not Applicable Not Applicable KR-46 Spring 2033 51.31 Not Applicable Not Applicable Footnote 1 - Table 3 is for illustration, refer to the KPS (Si Program and IDDEAL for the implementation Inspection schedule.

Page 41 of 62 KLR-1309A, Rev. Rev. 33 Page 41 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION Xl IS! - RVI INSPECTIONS Table 4 - Detailed MRP-227 Reactor Vessel Internals - Component Inspection Requirements 1 RVI LR Effect LR Aging Examination MRP-227 Information Component Mechanism Coverage Category Required Inspection Schedule Re-inspection Expansion Inspections Schedule Control Rod Loss of material Wear 20% Primary VT No later than 2 RFO Subsequent exams None Guide Tube examination of measurements from beginning of required on a ten Assembly the number of license renewal period, year interval per Guideplates CRGT and no earlier than two MRP-227 Rev. 0 (cards) assemblies, refueling outages prior with all guide to the start of the license cards within renewal period.

each selected Subsequent CRGT examinations are assembly required on a ten-year examined. Interval.

Control Rod Cracking Stress corrosion 100% of outer Primary EVT-1 No later than 2 refueling Subsequent exams Bottom-Guide Tube cracking, Fatigue (accessible) outages from the required on a ten mounted Assembly CRGT lower beginning of the license year interval per instrumentation Lower flange flange weld renewal period and MRP-227 Rev. 0 (BMI) column welds surfaces and subsequent examination bodies, Lower adjacent base on a ten-year Interval, support column metal. bodies (cast)

Core Barrel Cracking Stress corrosion 100% of one Primary EVT-1 No later than 2 refueling Subsequent Remaining core Assembly cracking side of the outages from the examinations on a barrel welds, Upper core accessible beginning of the license ten-year interval Lower support barrelflange surfaces of the renewal period and column bodies weld selected weld subsequent examination (non cast) and adjacent on a ten-year Interval.

base metal.

Page 42 of 62 KLR-1 309A, Rev.Rev. 33 Page 42 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION X1 ISI - RVI INSPECTIONS Table 4- Detailed MRP-227 Reactor Vessel Internals - Component Inspection Requirements' RVI LR Effect LR Agjna Examination MRP-227 Information Component Mechanism Coverage Category Required Insoectior Schedule Re-inspection .Expansion Inspections Schedule Baffle-Former Cracking that results Irradiation-assisted Bolts and Primary VT-3 Baseline examination Subsequent exams None Assembly in stress corrosion locking devices between 20 and 40 required on a ten Baffle-edge cracking, Fatigue on high fluence EFPY'and subsequent year interval per bolts -Lost or broken seams. 100% examinations on a ten- MRP-227 Rev. 0 locking devices of components year interval.

accessible from

-Failed or missing core side.

bolts

-Protrusion of bolt heads)

Baffle-Former Cracking Irradiation-assisted 100% of Primary UT Baseline volumetric (UT) Subsequent exams Lower support Assembly stress corrosion accessible bolts examination between 25 required after 10 to column bolts, Baffle-former cracking, Fatigue or as supported and 35 EFPY, with 15 additional EFPY Barrel-former bolts by plant specific subsequent examination bolts justification after 10 to 15 additional EFPY to confirm stability Heads of bolting pattern. Re-accessible from examination for high the core slde. leakage core designs UT accessibility requires continuing may be affected examinations on a ten-by complexity of year interval.

head and locking device designs.

Page 43 of 62 KLR-1 309A, Rev.

KLR-1309A, Rev. 3 3 Page 43 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION XI ISI - RVI INSPECTIONS Table 4 - Detailed MRP-227 Reactor Vessel Internals - Component Inspection Requirements 1 RVI LR Effect LR Aqing Examination MRP-227 Information

_Component Mechanism Coverage Category Required Inspection Schedule Re-inspection Expansion Inspections Schedule Baffle-Former Changes in Void swelling or Core side Primary VT-3 Baseline examination Subsequent exams None Assembly dimensions and Irradiation-assisted surface as between 20 and 40 required on a ten .

Assembly cracking that results stress corrosion indicated. EFPY and subsequent year interval per In cracking examinations on a ten- MRP - 227 Rev. 0 year interval.

- Abnormal interaction with fuel assemblies

Gaps along high fluence baffle joint

- Vertical displacement of baffle plates near high fluence joint

- Broken or damaged edge bolt locking systems along high fluence baffle joint Thermal Shield Cracking or loss of Fatigue or wear 100% of thermal Primary VT-3 No later than 2 refueling Subsequent exams None Assembly material that results shield flexures outages from the required on a ten Thermal shield inthermal shield beginning of the license year interval per flexures flexures excessive renewal period. MRP - 227 Rev. 0 wear, fracture, or Subsequent complete separation examinations on a ten-year interval.

IIIII9, 3 Page 4 of 6 ev KLR-1309A, Rev. 3 Page 44 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION Xl ISI - RVI INSPECTIONS Table 4 - Detailed MRP-227 Reactor Vessel Internals - Component Inspection Requirements' RVI LR Effect LR Aging Examination MRP-227 Information Component Mechanism Coverace Cateqory Required Inspection Schedule Re-insgection Expansion Inspections Schedule Core Barrel Cracking Irradiation-assisted 100% of Expansion UT Initial and subsequent Subsequent None Assembly stress corrosion accessible examinations dependent examinations Barrel-former cracking, Fatigue bolts, on results of baffle- dependent on bolts Accessibility former bolt examinations, results of baffle-may be limited former bolt by presence of examinations.

thermal shields or neutron pads.

Lower Support Cracking Irradiation-assisted 100% of Expansion UT Initial and subsequent Subsequent None Assembly stress corrosion accessible bolts examinations dependent examinations Lower support cracking, Fatigue or as supported on results of baffle- dependent on column bolts by plant-specific former bolt examinations, results of baffle-justification. former bolt examinations.

Core Barrel Cracking Stress corrosion 100% of one Expansion EVT-1 Initial examination and Reexamination None Assembly cracking, fatigue side of the reexamination frequency frequency accessible dependent on the dependent on the Core barrel surfaces of the examination results for examination results flange, Core selected weld upper core barrel flange. for upper core barreloutlet and adjacent barrel flange.

nozzles, Lower base metal.

core barrel flange weld Lower Support Cracking Irradiation-assisted 100% of Expansion EVT-1 Initial examination and Reexamination None Assembly stress corrosion accessible reexamination frequency frequency Lower support cracking surfaces. dependent on the dependent on the column bodies examination results for examination results (non cast) upper core barrel flange for upper core weld. barrel flange weld.

Page 45 of 62 KLR-1 309A, Rev.

KLR-1309A, Rev. 33 Page 45 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION XI ISI - RVI INSPECTIONS 1

Table 4 - Detailed MRP-227 Reactor Vessel Internals - Component Inst~ection Reouirements RVI LR Effect LR Aginq Examination MRP-227 Information Component Mechanism Coveraqe Catego[y Required ,InspectionSchedule Re-inspection Expansion Inspections Schedule Lower Support Cracking including Irradiation-assisted 100% of Expansion EVT-1 Initial examination and Reexamination None Assembly the detection of stress corrosion accessible reexamination frequency frequency Lower support fractured support cracking support dependent on the dependent on the column bodies columns columns. examination results for examination results (cast) control rod guide tube for control rod (CRGT) lower flanges. guide tube (CRGT) lower flanges.

Bottom Cracking Including Fatigue 100% of BMI Expansion VT-3 Examination of BMI Flux thimble None Mounted the detection of column bodies column bodies as insertlon/withdrawa Instrumentation completely fractured for which indicated by difficulty of I to be monitored at System column bodies difficulty Is Insertion/withdrawal of each Inspection Bottom- detected during flux thimbles. Flux Interval.

mounted flux thimble thimble instrumentation Insertion/ insertlon/withdrawal to be (BMI) column withdrawal. monitored at each bodies inspection interval.

Core Barrel Loss of material Wear All accessible Existing VT-3 Each 10 Year Interval Per Section Xl None Assembly Core surfaces at Programs Per Section XI Category Barrel Flange specified B-N-3.

frequency.

Upper Internals Cracking SCC, Fatigue All accessible Existing VT-3 Each 10 Year Interval Per Section XI None Assembly surfaces at Programs Per Section Xl Category Upper support specified B-N-3.

ring frequency.

Lower core Loss of material Wear All accessible Existing VT-3 Each 10 Year Interval Per Section Xl None plate surfaces at Programs Per Section Xl Category specified B-N-3.

frequency.

Page 46 of 62 KLR-1309A, Rev.Rev. 33 Page 46 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION Xl ISI - RVI INSPECTIONS Table 4 - Detailed MRP-227 Reactor Vessel Internals - Component Inspection Requirements 1 RVI LR Effect LR Aging Examination MRP-227 Information Component Mechanism Coverag Categor Required Inspection Schedule Re-inspection Expansion Inspections Schedule Clevis insert Loss of material and Wear and stress All accessible Existing VT-3 Each 10 Year Interval Per Section XI None bolts cracking relaxation surfaces at Programs Per Section Xl Category specified B-N-3.

frequency.

Upper core Loss of material Wear Eddy current Existing VT-3 Each 10 Year Interval Per Section XI None plate alignment surface Programs Per Section Xl Category pins examination as B-N-3.

defined in plant response to IEB 88-09.

Flux thimble Loss of material Wear All accessible Existing Eddy Current ER-AA-NDE-ET-501 Every five years None tubes surfaces at Programs (ET) WDI-ISI-088 specified frequency.

Footnotes:

1 This table is based on the guidelines of MRP-227 for Westinghouse-designed plants. The following are not listed In this table:

"No Additional Measures" components from MRP-227 for Westinghouse Plants (no additional aging management is necessary for this group).

A comprehensive list of Section Xl B-N-3 components and inspections (see KPS "ASME Section XI Inservice Inspection, Subsection IWB, IWC, IWD" program).

Page 47 of 62 KLR-1 KLR-1309A,309A, Rev.Rev. 33 Page 47 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION XI ISI - RVI INSPECTIONS ATTACHMENT 1 IMPLEMENTING PROCEDURES (1) Nuclear Fleet Nondestructive Examination Procedure ER-AA-NDE-VT-608', "VT-3 Visual Examination Procedure for Examination Category B-N-I, Interior of Reactor Vessel."

(2) Topical Report DOM-QA-1 "Dominion Nuclear Facility Quality Assurance Program Description."

(3) Nuclear Fleet Administrative Procedure PI-AA-200, "Corrective Action."

(4) KPS Nuclear Administrative Directive NAD-03.01, "Directive, Implementing Document, and Procedure Control."

(5) KPS General Nuclear Procedure GNP-03.01.01, "Directive, Implementing Document, and Procedure Administrative Controls."

(6) DNAP-0501, "Dominion Nuclear Procedure Administrative Control Program.

(7) Nuclear Fleet Program Description AD-AA-10, "Administrative Controls Program."

(8) Nuclear Fleet Guidance and Reference Document PI-AA-100-1007, "Operating Experience Program."

(9) Nuclear Fleet Program Description ER-AA-RII-10, "Fleet Reactor Internals Inspection."

(10) Nuclear Fleet Administrative Procedure ER-AA-RII-1 01, "Fleet Reactor Internals Inspection Program."

Additional NDE procedures (either fleet, plant-specific, or vendor) will be needed to implement the revised ISI Plan for Reactor Vessel Internals Inspections.

309, Rv. 3Pag 48 f 6 KLR-KI-R-1309A, Rev. 3 Page 48 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION XI ISI - RVI INSPECTIONS ATTACHMENT 2 ASSOCIATED AGING MANAGEMENT PROGRAMS Aging Management Programs Supporting KLR-1309A.

KLR-1310, "Primary Water Chemistry" Aging Management Program Supported by KLR-1 309A.

KLR-1309, "ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD" Aging Management Programs Verifying the Effectiveness of KLR-1309A.

None Aging Management Programs - Effectiveness Validated by KLR-1309A.

None 309, Rv. 3Pag 49 f 6 KLR-KLR-1309A, Rev. 3 Page 49 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION Xl ISI - RVI INSPECTIONS ATTACHMENT 3 MATERIALS AND AGING EFFECTS MANAGED Materials Aging Effects Nickel Alloys Cracking Nickel Alloys, Stainless Steel, Cracking (SCC, PWSCC, and CASS (PWR Reactor IASCC, Fatigue)

Internals Under MRP-227) Loss of material (Wear)

Loss of Fracture Toughness (Neutron Irradiation or Thermal Aging Embrittlement)

Loss of Preload (Thermal and Irradiation-Enhanced Stress relaxation or Creep)

Dimensional Changes (Void Swelling) f6 I5 KL- 09,Rv 3Pg KLR-1309A, Rev. 3 Page 50 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION Xl ISI - RVI INSPECTIONS ATTACHMENT 4 INTEGRATED PLANT ASSESSMENT REPORTS SYSTEMS OR COMMODITIES KLR-1 104, "Reactor Vessel, Internals, and Reactor Coolant System" Section 4.2, "Reactor Vessel" Section 4.3, "Reactor Vessel Internals" Section 4.4, "Steam Generator" Page 51 of 62 Rev. 3 KLR-1309A, Rev. 3 Page 51 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION XI ISI - RVI INSPECTIONS Attachment 5 I

Detailed NUREG-1800 Program Evaluation Attribute Attribute Description AttributeResponse D tm Dso" Pupose Docent Change FNb CecoCR Number Doumbent Seurn Pip. DculeaChn- Ft AP PI KLR.1309, "ASMESection XAInseri/ce Inspection, "I-s 0a~

Subseclione IWB, IWC, and IWO,. addresses the Inservice Inspecttons for theComponents Included Ir Section XI, Subsections IWB-1100, IWC-1100, and IWD-1100 WtASM5 Class 1. 2, and 3 components, respectively. The aging of the Close 1 Components, Tabte IWB.2500-1, Examination cualguY B-NS3 Items Is menaged In acourdance 1,1ththe AMSE Code, S*ction XI requirements, KLR-130D Identiles two program er-runnenmente in Sections 5.1 and 5.2 to Incorwoela applicable bIdustry reactor vessel nlernaJs inspection Iiti*tllve. Tne two enhancements state:

"5.1 Enhancemant 1: Aging Management of Reactor Veuoet tntemels The ASME Seclion XI Inloevice Inspection Subsection IWO.IWC and WD program will be enhanced to (1) particpate In the Industry programs for invesigpatig and managing aging effects 0n reactor Interlats; (2) evaluate end Implement the resutts of the indust*y programs en applicabte 10the reactor ltmerolnand (3) upon completion of theue programs. bud not less than 24 months beour entering the period of ext*nded operation, submit au Inspection pln for reactor Intemals to the NRC for reotew and approval to augment the current The IdentifledFAlend CAPECR numbera were The specieo program necessanry for License renewal should be Identofed. The scope o" the program Inspections. Initiatedin KLR-1309and am Included here to 621 LA001042 I should Include the specific structures and '5.2 Enhancement 2. Aging Management of None None provide a tie to the boros forthis program. I LADOIO43 components of which the pmgram menages the Thermro Aging And Neutron trrtation No documents to Implement the scope of INS aging. Embnttlrment of Cast Auslalenl Staruiee program were tdentited.

Steel "The ASME oectlonXI Inservice Inspection Subsections IWB, NYC end IWD program Alt be enhanced to Include identilitetaon a1 the limiting susceptlble east austeniso estanless steet reactor vessel tntemalscomponents from the standpoint of thermal aging suoscptibIily, neutron fluenca. and cradding. For each Identified oumponent a plan All be deveioped, whIch accomplestes aging mrranagement through aithere supplemental examlnatlon ore component-spedfic evaluation. The plan Alt be Submltted for NRC review ancd appeovel not losa than 24 months baoor= entedng the period of extended operallor.The scope of the KIR-13OeA, "ASME SXI lSt - Reactor Vessel Ineremals Inspoctlions" program Is to Implement the Inspections reqired by the toe program enhanocement=.

The reactor vessel lntemats(RVI) tnduded In the scope of the program consist of two bastc aesemblies, the upper Intemnals usemoby that is romrOed during each refueling operation to obtain aocess to the reactor core, and the lowerthteroolt easembly that can be removed, If desired, fotltohng a comrrplete cor ottload.

EPRI Report 1016596 {MRP-2271 Provides the Page 52 of 62 Rev. 33 KLR-1 309A, Rev.

KLR-1309A, Page 52 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION XI ISI - RVI INSPECTIONS Attachment 5 Detailed NUREG-t800 Program Evaluation Attribute Document Document Purpose Document Change FAPICR Number Atribute eri Abbut Respo Number section results of the Industr Inlifi,tivu via the Materials Retiability Program (MRP) to develop the Inspection and evaluston guidelines for managing the long.

term aging of pressurized water mattor (PWR) RVI.

The report provides a generic lsstof the RVI setrctures for Westinghouse pWRs.

Westinghowse letter LTR-ARIOA-08-63 Rev. 3 (Ref

) evaluated the generic MRP.227 scope with respect to KPS.

The RVf components Included In the program am Identified In Table 2 and Table 4. These tables ame based on the MRP-227 denltied RVl components an modified by the Westinghouse evaluation documented in LTR-ARIDA-0t-63 Rev. 3.

Note: MRP-227 has been submitted to the NRC for review. Foilowing NRC review and approval, MRP-227 will be revised to Incorporate any necessary changes to the guidelines and reissued as MRP.

227-A. The Reactor Vessel Internats Inspedlons program will be revised, ts necessary.to Incorporate the onarrecommendations and requlrmeots as published In MRP-227-A.

Howlever, any.revisions to MRP-227 resulting from the Issuance of MRP-227-A will not be Incarporated for any Inspections that are completed before MRP-227-A Is Issued.

Not Included In the program are monsumabla Items such as fuel assemblies, masllvitycontrol assemblies, and nuclear instnuorentatlon and welded attachmenot to the reactor vessel.

___........'~V~.t'~V1/2'.

~ .~.... V 1.1 ~-

Tite adviltee forpreventlon and mitigation programs The ReacerVessel Internals Inspections program s 1 should be desctibed. These actions should mitigate a ondlitionmonIto.rngprogram enddoss net Incbide or provont aging degradation. sny preventhveor mitigative actions.

An noted In Element 2, Atihbufe1, the Reactor her cendonelr or perforeavnce actionsiang programs, Vessel Intemals Inspections program does not they do net rely on preventive aetions and thus, this perfror any preventive actions.

2 typnfrmatin Implemented management be provded.rgramm be maensure that agIng effectsteay rTheReactor Vessel Internats program provides are verification trat The Primary Water Chemistry managed, program has been effective iv preventing primary water stress corrosIon crhcking.

The program manages the follow aging tffect for the RVI by monitoring for the parameters Identified:

Stress Corrosion Cractking - SCC refsira to local, non-ductle Is cd.lig of a material due to e combination of tensile stross, environment, and The parameter to be monitored or Inspected should metllturgicel properties. The observed aging affect be Identified and linked to the degradation of the of SCC on component Integrity Is cacking.

panticularstructureand component Intended Irrradiaon-Asslsled Stress Corrosion Cmctdng -

function(s). IASCC Is a unique form of SCC that occurs only in highly irradiated components. The observed aging effect of IAMCCon component Integrity Is cracking.

Wear - Wear Is caused by the relative motion between adjacent surfaces. with the extent determined by the relative properties of the adjacent Page 63 of 62 KLR-1309A, Rev. Rev. 3 3 Page 53 of 62

LtCENsE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION XI ISI - RVI INSPECTIONS Attachment 5 Attribute AttrIbule Dscription Attribute Response Umbert Donumant purpose oaumantChange FAI CAPICR materals and thei surface condition. The observed aging effect of wear on compont Intogdityis Ide5 of material.

Fatigue - Fatigue Is de.fied as the st*uctural deteriouraton that can occur as the resultl of repeated stess/stinr c~cles caused byfluctuating loeds and temperatures. The observed aging effeot offatigue on component Integrity Is an¢trng.

Thermal Aging Embrttlement - Thenral embrittlement Is the exposure of delft forrte within cast austenttlistainless steel (CASS) and pmrepltation-hardsned (PH) stainless steel to high ln.-eence temperatures, which can result In an Incease In tensile strength, a decrease In ductilty, and a loss of francturetoughness. The observed aging effect of thermal embdlttlement on raonponsnt Integrity is cracking.

irradiation Ecbrttlrnment -tWhen exposed to high energy neutrons. the mechanical properties of steel and nltckel-base alloys can be

,s"intess Changed. Such changes In mechanictl properties Include Increasing yield sierngth. Increasing ulttmate strength, decreasing ductilty, and a loss of fracture toughness. The observed aging effect of Irradiation embrittelment on component integrity Is cracking.

Void Swelting and Irradititon Growth - Votd sweling Isdefined as a gradual Increase In the volume of a conpornent caused by the omislatn of microscogic cavities In the matertat, While the Initial aging effect Is dimensional change end distortion, severe void samling may result In crackdng under stress.

Thermal And Irrradittloanhanced Stresa Relaxation Or trradisto*o-Enhanced Creep -The loss of preload aging effect can be caused by the aglng meclantsrms oa stress tefaxatsan or creep.

The aging effect t5a loss of m chanttcalclosure integriy (or, preload) that can lead to unantiipated loading which, In turn. mayeventually cause subsequent degradatlon by fatigue or wear and result in cracking.

The aging effects relted to RVI cantponents are Identifed in Element 4. Attribute 1.

For a condition monitoring program, the porameter monitored or Inspected should detect the presence The aging effects managed by the RIA Inspecotion 2 and extent of aging effects. Some examples are program and associated parameters mnitored am None None None None None None measurements of wal thtckness and detection end discussed In Element 3, Attribute 1.

sizing of cracks.

For a performance monteodng program, a lnk should be established between the degr.adaton of the psrticular structure or component intended function(s) and the parameter(s) being monitores.

An example of ltinixng thedegradation of a passive component Intended function with the performance The Reactor Vessel Interals Inspoctions program ta None None None None None None being monitored Is tinking the fouling of beat not s performlnce muxitortoigprogram.

exchanger tubes with the reat transfer Intended funttn. Thts courdbe monitored by paeodlO heat balances. Since this example deals only with one Intended function of the tubes, heat transfer, addiutonal programs may be necessaryto manage I Page 54 of 62 KLR-1309A, KLR-1309A, Rev. Rev. 3 3 Page 54 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION XI ISI - RVI INSPECTIONS Attachment 5 Detailed NUREG-1800 Program Evaluation Attribute Attribute Description Attribute Response Nume n Purpose DocumetChage FI CAPCR Number Dunta Section other intended function(s) of the tubes, such as pressure boundary.

A performance montoridng program may not ensure the structure and component Intended function(s) without linking the degradalton of pussive Itended functions with the performance being monitored. For example, a pedodli diesel generator test atone would not provide assurance that the diesel will start end run properly under et applicabie design conditions. While the test vetifles that the diesel will perform ifai the support systems ftnction, it provides little Information related to the material conditlsr of the suppoort components and their ability to withstand DOE toads. Thus. a DBE, such as a teismito event, could cause the diesel supports. such as the diesel nmbo*,ent plate anchors or the fuel oil tank, to fell If the eftects of aging on these components are not managed during the period of extended Operation.I For prevention and mittgation programs, the parameters mreornred should bo (he specitf parameters being controlled to achieve prevention The Reactor Vessel hriernalsInspection program Is None None None None None None or mlitgation of aging effects. An example Is the not a prevention and mitigation program.

coolant oxygen level that is being controlled In a water chemistry prograrm tomlgtoI rdpecracdng.

.The aging management metihodotogie descdbed in MRP-227 am based on either existing Insernlvo uxaminations required by the ASiMECode, Section

)I or on well-documented and well-demonstrated examination methods with which the edoustiy has conshderable experience. The RVI Inspections program wil use the follow methods to detect the parameters monitored identified n Elernent 3.

Attrbute 1:

Oeleclton of aging effects siould occur before there Visual (VT-3) Eoarrination - VT-3 examination has tsa loss of the structure and component Intended been determined to be an appropriate NDE method function(s). The parameters to be moeitored or for the detection of general degradation conditions In Inspected should be appropriate to eneure that the many of the susceptible components. A VT-3 can structure end component Intended function(s) will be detect sturctunt distortion or displacement of parts adequately maintained for Iloense renewal under all to the extent that component function may be CLB design conditions. This Includes "pact$s such impaired; loose, mdssing. craeked, or fractured parts, None None None None None None as method or technique (e.g., vitual. vo1umeldr. boaling, or fasteners; corrons or erosion that surface Inspection), frequency, sample size, data reduces the nominal section thickness by more than collection and timing of newlone-Ime inspections to 6%; wear of mating surfaces that may lead to loonsof ensture timely delection of egIng effects&Provide funclion; end structural degradation of interior Information that links the paremeters to be attachments such that the original cross.cta*onai monitored or inspected tl the aging effects being areala reduced more than 5%.

managed. Visual (VT-1) Exemination -VT-1 examination is definad In the ASME Code Section )i as an examIoadon "conducted to detect discontlanuftus end huperfeolons an the surface ofcomponents.

Including such conditions te cracks, wor. corronlon.

oremwson!

Enhanced Visual (EV-I1) Exarmnation - EVT-1 examination has the some requliremnts VT-1

=xamn n, with additonul requirements intended to lrrrove the detection andchauraetsdzatlo of_

KLR-1309A, Rev. 3 Page 55 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION XI ISI - RVI INSPECTIONS Attachment 5 Detailed NUREG-1800 Program Evaluation Att riute Attrib~ute D escrption Attribute R espo nse Document Document Purpose Document Change Number Number setion discontiulif*s taling Into account the remote visual aspect of reacter Intomals examlnatlons.

Suface Eanlinallon - Surface ET (eddy current) exarnlnaton Isspeclfid as an altemnatve or as a supplement to visual exuarlnations and provides for detection and length ondngof surfce-broakdng or near-surface mrackh.

Volumelo .Earln*lia*n - VT eaxmtlnatona are spedoled for bolts to detect planar detects.

The RVI tnspections program Inspectins ean performed Inaccordance ,ith the ASME Sectim XI Inservice Inspection, Subsections IWB, IWC, and IWO program. which Inctudes the procedures for performing the Inspecllons. the acceptance clteda for each examination technilque, and the review and disposltion of inspection results.

MRP-227 categorizes te RVI Componentsbint the following four categories;

- Primary- those RVI that are tdhtyausceptibto to the effects of at least one of the eight aging mechanisn.

SExpanslon - those RVI that am fighly or moderately suscepl*le to the effects of at least one of the eight aging mecduismo but for whiri1 functionality essessmnent has showo a degree of tolerance is those elfeots.

based ran Nudearpower plants are licensed redundancy. diversity, and defene-tn.-depth - Existhng Progrars (Exdstng) - those RVI that are prlncipleas.A degraded or fulled component reduces suscepUble to the effects of at leasal one of the the reliability of the system, chaltenges safety eighteagig mech airdm= and for dch ed sthng systems. and contributes to plant ith. Thuso the genera and plant-spedfic aging management None effestsofagIng oneastructrureiimpoeentishould programs are capabte of managing those effects. The purpose of this procedure Is to provide a be managed to ensure Its availability to peefouimIts - No Additonal Measures (NAM)- hese RVI for processltf the applicatron of Vr3 visual Intended functsone) as designed wlen called upon. whih the effects of atl eight aging mechardsms ER. AS esaminatmonlechniques for thu earminatan of 2

In thMsway. all system levai Intended func*lon(a), am below the ocreearng criteria. NDE-VT-BO A ExamInation Category B-N-I, Interior of Reactor Inudegreunanydlerlt, nddeene-n.

diversity, and dense. The Te examination i techniques and schedule foro noVessel In accordance ASME Section XI. with the requlrementa of hIncluing redundancy, depth conalsstenlrlth the plants CLB, would be Inspeotionof the RVI corepornt Is the Primary maintained for license renewal.A program based and Expansion categories Yal be conisistent wih the solely on detecting structure and component foluare reqtirements of MRP-227 and are provided In should not be considered as en effecive aging Tables 2 and 4.

management program for license reneeal. EPRI Report 1015009 (MRP-228) provides additional speclifi guidance to the Inspectien methodologles and requirements for the NDE examnlnathrr of the components in the Primary and Expanslon calegores. The hoplemaering procedures must be consistent Mith the MRP-228 guidance. The personnel perforning the Inspectlons will be trained Inaccordanca with the guidance In The personnel perfoenaing Inspections of reactor MRP-220. vessel lnteinals for MRP-227 will be trained In The RVI components In the Existing end NAM accordance with the guidance In hMRP-228.

categories will be In aocordance with the Page 58 of 62 KLR-l3ogA, Rev.

KLR-1309A, Rev. 33 Page 56 of 62

LICENSE RZNEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION XI ISI - RVI INSPECTIONS Attachment 5 Detailed NUREG-1800 Program Evaluation Attribute OO uruant Duourmnt Number Attribute Desocrption Attrtbute Rasprian Number Section Purpea Doettinent Chargo FAI CAPICR requiremenrs OftheAStE Code Inspection& per SectionXI.ExaminallonCategory B-N-3.for removable internalstructures, as outlined Intie ASIdESectlon XIInservlcetrtspecOon, Subsections IWB.IWO.an IWDprogram. For Exambiation category B.N-3, ER-AA-NDE.VT-.o0 spedies a VT.

3 examinaeton of ailaccessIbte surfaces of reactor cOresupport SurUCturas that can be removed from lte reactor velel. The zantIrvnak are performed once during VieInspeocton intervalwhen the core barrel Is removed, The visual VT-3exomTnalton deWmilnes the general mechani.l aendetricturni conditon of the componenls byInepedsg for stctural d1stortion or displacement of parts. loosailmlening, cracked or fractured parts, loooe,"rsing boiir*gor fasterr8s, debris, orrosloe or erosion, and wear of mretnr surfaces. The Inrpedjons are performed by quatlited personnel fdllowig procedures consistetrwitf the ASMECode and 10CPRPart 50 Appendtx 8.

tnaecordancenith tfieASMESection XI Irrwnvtc Inspecton, Subsectons IWO.,IWC.ard FWD program enhancemenLs,. apleet-spoeifl Inspection pTanfor the RVIcomponents willbe developed end wlt address the requirements established by MR-227 for Wesdrghoase designed PWRs, as approved or modafed by the NRC.The plan wil also define anyproposed altematives determined to be uecessary.

Theplant.spedrz Ilnspection plan willbe developed to oddreee the requirements astaivlbhed by MR-221 forWestinghouse designed PWRus. as approved or modiled by theNRC,Indcudig any proposed altemativee, and wil be submitted for NRCreview andapproval afleast24months frir to entlryIntol the period of extended operation.

Note:MRP-227 has been submitted to the NRC for review.FollowingNRCreviewand approval, MRP-22will be rearitl$ to incorpoerteanyomnoessory changes to the guldelines and reissued as MRP-227-A.The Reactor VeeltIniemels Inspections programwit be revised, as nosessary, to New NnNoeNDE Iow proadrirsae willbe developed to perform (he examlna~ors of the cmonarit in th Primary LA001042 PNoiedu re end Expanrton categodes ea Identitied In Tables 2 LA001043 and 4.

3 Thir program nd 'how" element program deacriias data "when, wre collected -wter, (L~e.,all , Tre when,"'wtrere nfraThe whnI" discuse,"erd'how' In*

"hawn program 4,AtribtegrNnaNnmNn Nn aspects of a orivities to ollect data as prtoe IntonwilondusundtnEinnemt4,Attrbutes I None Nose None Nose Nore Nose program). . I I II

'the method or technique and frequency may be The examination mefredo arid requemcis em None NoneNone None None linked to plont-specific or Industry.wide operating consistent with Industry guidance. _!!raN F I Page 57 of 62 KLR-1309A, Rev.

KLR-1309A, Rev. 3 3 Page 57 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION X1 ISI - RVI INSPECTIONS ion Nrumer Attribute 0escrIptIon Attribute Response Purpose Doeument Cange experience. Provide justificatlon, including codes ASdiscussed InElement 4, Attribute1, the and standards referenced, that the technlque and examinaion techntques to be used in the frequency ore adequate to detect the aging effects examtn trtin al the RVI components are the before a los of SC intended fuoction. A program aocepted techniques used In the ASME Section XI based stiely on detecting SC failures Is not Insorvice Inspection, Subsections IWE, tWC, and considered an effective agitg management NO program.

program. As discussed in Element 4. Attribute 2, the inspection schedule Is consistent with either MRP-227 orthe ASME Section XAInservtce Inspection, Subseottons VWB.IWC, eandIWD program, as applicabte.

When sampling Is used to Inspect a group of SCs.

provide the basis for the Inspection population and sample size. The Inspection population should be bused cm such aspects of the SCs as a simtuartty of rmatertata of construcuitn, flobrlcaslon prownremenr, design, Instaltatton, operating eonvlrniment.or aging effects. The smrple size should be based on such aspects of the SCs as the specotlo aglng effect, The Reactor Vessel nternats Inspections program location. euisting technlcal information, system and does not use sampling to characterize a group of None None structure design, materitas of construction, Oervtce components.

environment, or previous (oiftte history. The samples should be biased toward locations most susceptIbleto the specfli aging effect of concern In the period of extended operation. Provasons should also be included on expanding the sample stoe when degradation Is detected In the Initialsample, rorar lesp:.Mlntrp eordis'> i*Y ' .;. -

Monltoring and trending ectiviltes should be described, and they should provide predIctability of As discussed In Element 4. Atlibute 2. the the extent of degradation and thus effect timely Inspection schedute Is consistent with either MRP-correotive or mitigative actions. Plant-sepecitl end/or 227 or the ASME Section XI thservice Inspect*on, None None Industy-wltde operating experience maybe Subsections tWO. IWC,end IWO prograr, a considered In evaluating the appropreteness of the applicabre.

technique and frequenoy.

The acceptance creritla and Inspection expansion criteria are provided In Tables 2 und 4 for the RVI components In the MRP-227 Primary end Expansion This program element describes 'how' the data categories.

collected are evaluated and may also Include trending for a forward ole.oThis Includes an In addition to the Primary components, Expansion components have been defined should the scope of evaluateon of the results against the acceptance criteria and a prediction regarding the rate of examination and re-examlnatien need to be degradation In ordnr to contirm tlht iinog ot the expended beyond the Primary group due to detetion of significant aging effects.

nod scheduled Inspection will occur before a loss of SC Intended function. Although aging Indicators may Tables 2 and 4 Identify the reactor vessel internals 2 be quantitativa or qualitative, aging Indicators should components In the program, the aging effects None None be quantlied, to the extent possible, to allow epplicabia to each component. the examrnati.on trending. The parameter or Indlcator trended should techniques used to detect the aging effec(e), aned be described. The melhodaoogy for usnalong the the examin aeon schedule for the Primary and Inspection or test results against the acceptance Expansion categor components.

crlteria should be described. Trending Is a The RVI components In the Existing and NAM comparison of the cerent orrolaoring results with categories wil be evaluated In ascordence with the previous monitoring results In order to make requirements or the ASMfECode for Section XI, predictions Forthe future. Examination Category B-N-". for removable Intemal structures, as outlined In the ASME Setion XI Iesrvice Inspecloarr Subsections tWt. (WC. and IWD program.

Page 58 0162 KLR-1309A, Rev.

KLR-1309A, Rev. 3 3 Page 58 of 62

LUCENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTON XAISI - RVI INSPECTIONS Attachment 5 Detailed NUREG-1800 Program Evaluation Attribute Number Attribute Description Attribute Response D o Number .ment Dct urent Purpose Document Change Section Prgm~n qAd tac=e Ciei. . , j ~

The cceptahnca riteria of the program end Ite baste should be described. The acceplance criteria, As discusned hr Element 4, Attribute 1, the RVI against which the need W onrrective ealons wdilbe thspections progran is ImplementedIn accoardanca evaluated, should ensure that the etrndtu end with the ASME Section Xt Insoenca Inspection, component Intended functon(s) am maintained Subsections WB, IWC,and IWD program. The under ali CLB desegn conditions during tie period of program Irudes tle Inspection peocedure. tire extended operation. The program should Include a acceptance critnea foeeach exoan(roUon tecrhnlgue, mrlahoddrogy for analyzing t results agatint andthe process Woevauating unreefatlcaory applicable acceptance criteria. Inspection roeudle. For ASME Code Section XA For example, carbon steel pe walthinning may Inspectlone. Indicatioeraend relvent conditions oor under cartain conditions due to eroslon. detected during enalrnatin n are required to be cornosion.An eging management program far evaluated In aCcorderva with ASME Section XI, I erodson~ooreslonmay consist of perodicaly Atidle IWB-3500. Nore None No-e measuring the pipe wag thisknese aend compaeing The acceptance cdeda aed Inspectlan expanston that to a spadlic minimum wall aceptance clterirA criteria a*e provided In Tables 2 rid 4 tor the RVI e

Correcthe action taken, such as piping components In the MRP-227 Pehlay, end Es*paeron replacement, before reaching this accaptaree categories.

criterion. ThIS ptphrgmay be designed rcr thenlral The RVI componenis hr the Existing end NAM presseur, deadweight. seismic. end other loads, and categories wll be evaluated In acrdance with the tiNsdecptlane toneon mont be appreoprtae to reqduramntof the ASME Code for Section XJ, ensure that the thInned piping would be able to carry Enumlnatior Category 8-N-3. for removable Mtemel these CLB design loade. This acceptance criterion eJructroaw,as outlinedIn the ASME Secion XA should provide far timely oaorctive adion before Inservice trnrpeeullor Subwclons tWB. Plo,, and loss of intended function under these CLB dolge, WV program.

inada Acceptance criteriacorid be spedfic numerical values, or could cor=rstof a dsosclson oe the pmecene foe calculatlng spepdlc nurmericl values co 2 ounditional a=teptance crteia to ensure iUh the Refer to tlernt 6. Atribute 1. None Noe None None structur* and component Intended functeon(e) mil be maintained under el CL. design condiltons.

irformoati, from available referencee moy be died.

it Is not necessary to justify any acceptance citeria talen directly from the design bsis Iinformation thit Is Induded Ih the FSAR because that Is o port of the CLB. Also. It Is not necessary to disouas CLB design loado tI the acceptance tlrtera do not pernt Reler to Element 8, At 1. None Nane None None degradation because a structure and cmnponent R0 without degradaution s*ould continueto function as orighally deslgnod. Acceptance cdilda, which do permit degradotIot. aoe based on maintaining the intended functior under eti CLB design loads.

Qualtativ nhspecons should be performred to sarne pedetermined crldera 4 quantitatveainepacranahy RetertoElement4,Atttlbs 1. None Non None None pereonnel In accordance with ASME Code and through approved she sped&in prograrnr.

Topical Report DOM-OA.1. Dominfon Nuclear Feclity QualltyAssurance Program Derocption" and Actions to be tWoen w*res tie acceptance criteria ere sub-tjer procedures fulfilland Implement the This topisI report provides the esty asunrance be described. Corrective actions, of requiremonlt of 10 CFR Part 50. Appendir S. The M-QA-1 All program description (APD) for Dombrdon's nudear includingshould nIloMt root cause delenninatlon und prevention corrective actien process proeides reasonable power stations and Irdeperdent spent fuel storage None recurrence. should be timely. oassur*noe that events or cordtions poaerrielty [n.WatllaoS.

adverse to quaitly ore promptly Identified, evaluated, end corr*ected.

Page 69 of 62 KLR-1309A, Rev. Rev, 3 3 Page 59 of 62

LICENSE RENEWAL PROJECT KEWAUNEE PoWER STATION AGING MANAGEMENT PROGRAM ASME SECTION X1 IS[ - RVi INSPECTIONS Attachment 5 Detailed NUREG-1800 Program Evaluation Attibute Attribute Oescription Attribute Response NDoumert bou t Purpose Document Chnge FAI CAPlCR TNhs procedure establiutes the e.xpetatUons for the Identficateon end ieporting of condtions tharet adverse to quality. potentaly adverse to qualty,N PI-AA-200 Al taffect fe, effect persornenelu nucdear sfety, affnt NOne Noe None plat relibility. adverse trends,or other conditions that do not meet expectatlorm If onvolve actions permtt entlysis vtttout repei or 2 replacement.

etruture and tIheenelysis Intended compoarrt should ensure that Use funtonlen() wll be ....e ......... NNone N

, None None None None None maintainedonoslatent vwththe CL8.

The Processshouldb mationic"nfr o Action.program, ..a preoand Thleonottotilnon IeOCeneshouldhe sh

~dsscrbOd.

ensure that preventiveactionsameeual U The Coredoti n

atr t p anresenWfad elot aetr This topicalreport provides the quaWlty suoarasmo onhonditions adverne to o r "e use DOM-OA-1 At pognramdnecron (APDW ) forOom*drfs nuclear None None None endthatopprorate correctIve ectionrs hom been determ*notion D -A1power slotion andInodspandand spent fuelstorage and that corrective actiosn we t completed aid are effective. to preclude reoccurrence.i innionled .

This procedure establishes thmexpectations for the Identilictionm ndrepoting or *rlittios thatam adverse to quality.poiafoily adverse to qulity, None N None affect personnel safety, affec nuclear safety, efftect plantreliablitty,advoie trends,orther ourdi:oos that 0orot meet expectations.

The effectiveness of proveriton and m~gation program shoculdbe verilaed periodically. For 1 examp~le,tn mearwling inlerrJ com'OSlo of plplrigýa 2 miligagton to mninmizaprogram

=suscptibility ocrroilon.tray (watertochemist*y) be used4t However. Refer to Elnmern Corwitive 8, AthributeI for Adeon Program. discussion of tUe None ona None None Nora None may also be necessaryto have a condition mcating programt (uler rd,*c Inpection) to verify/

that corroson Is Indeed Insignificant, iI VInarticorrectiveactions are snecssary.

tismeehsbckt

ctivitiesto confirmthat the correclt" be dlow-ejp A t discussion ot the None Nam None Non None None 3 were completed, the metcase5 ctisons Refer to OElement 8. Attribtelo¶ determnteatdn was performed, end reciurnoe Is Corrective Action Program.

prevented.

Topical Report DOM-OA-1, "Oolatont Nuclear Facility QJality Asumnii Program DescelpOW and The administrative controls of the program should. vub-tier procedlures fulfil and implerment the This topical reortx proviies the quali ty assurance be described. They should provide a formal review requirernents of 10 CFR Part 50. Appendix S- DM.A1 A prograum desr, iptan (CMPD)for Domitrlan's nuclear None None None and Approval p10oeOL. Aximthlestrtive anzd tirpieernrmlng procedures 8are 0-A. Ag powverstations wW Independent spent Wuel ,orage navlo ,dapproved, anldmaintaned aa controlled Initallatoem.

docmerds In accordance v~tl the procedura control process._._.. ... . ...

Th*sprocedure provides thi remqulrent for the GNP- A pmptral*om review, approval, and doliburtlon and Of No None None 03.01o.01 A drectirves Impt-menlr-ng documents, procedures.

This directive providonInstrnion for neust*lblshng mepousilblitles end requresments forthe mornagement oi Nuclear Administrative 0(rectmvon, r1 None None Dominion RoeetProedures, NMCReet Procedures, and silo proceduresor other implemenllng KLR-1 309A, Rev. 3 Page 60 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION XI ISI -RVl INSPECTIONS Attachment 5 Detailed NUREG-1800 Program Evaluation .

Document Document Purpose Document Change FAI CAPICR Number section do*uments.

Thisprocedure esteblishes fte herarchical plrocedural retationshps; esteablLshes the program requlrements to BcfrninlntrmttvdYcontrol Dominion nuclear common administrative and techndcal Procedures:provides a descripdton of the process necessary to develop and malntatl the quality of DNAP-001 Al DomInlon nuclear common administrUllve end None None None technical pmocduoev; provides guidance and requirements for Implementation mid adherence for Dominion nude= cormmon Adminisraetive and technicat procedures; and describes the methods by which Domrnion nuclear common administratlve Procedureforms re cntrotoed, This procedure defthes (he major elements of the adminislrative contrls program; eslabln*es the standard monagementconhole as required to A A define Mre support poircies. programs management controlsand proceases;

ýrla the Non None None Inmplementationtode, Inclufdni the document

.ercby. nd Insdltutonalzst standardization throughout the business processes.

The purpose of tis Program Description Is to establish and Implement the Reactor Intomols Inspection Pr"ogram (RIIP) at all Domion nudeer RA.Rl- Al sites. The RUP Is based upon reqrdrementm from the None Noss None 10 Materials ReliabTIly Program: Pressurized Water Reactor intemoasInspection and Evaluation Guoideline (RIRP-227-Rav. 0) end site specifi license renewal commitments.

The purpose of this procedure Is to estabtsh and Irnqreerft the Reactor Internme lnoyeaton Pogra*n (RIIP) at al Dominion nuclear sites. Th"]eRtIP Is ER M WI- based upon requorements fram the Materials None Norm 101 All r RerliabiliyProgmramPressurized Water Reeacor Intemual Inspecton end Evaluaton Gideoteo -

(MRP-227* lRv. 01 and sie spednc license renewal comndtmeent.

t"ense renewal should have regulatory and admbinstrative controls. That Is ft basis for10 CFR 54.21(d) to requirethat the FSAR supplement 2 indudeea summerydesolpteon of*he progracrnand Referto Elemert 9. Afflbta 1 fordlcusmden ffthe None None None actlvities far manoging the effeclt of aging for adrimniltrathm controls None NOn' None license renewal. Thue,any Informal p*ogram* railed on to manage aging for license rarlewa! must be adminustrtively controled and IncludedIhthe FSAR supplemrnent.

nit~I O~ietirg~EdoerCnee.~ ~n~Jr~,CV r:,.v ~ACIi. d~>V..........*. .'~K. 7-. ~ ...

'C n..-r;;.C.i~I~..vCJlC;~>.-.;;....Cn.C.r<r. '7O~~'

Page 61 of 62 KLR-1309A, Rev. 3 3 Page 61 of 62

LICENSE RENEWAL PROJECT KEWAUNEE POWER STATION AGING MANAGEMENT PROGRAM ASME SECTION XI IS[ - RVI INSPECTIONS Attachment 5 Detailed NUREG.1 800 Proaram Evaluation Attribute Document Document Purpose Document Change Number Attribut Description Attribue Response Nub Section Operating exxtrlence with existing pograrm should be discussed. The operoting experience of aging management prograus, Including pestcorcective acilon&resultingin program enhancements or rtprograrms should be onsldered. A past additional faium would not necessarily Irnyldate, an aging management program because the feedback from opemring e*perience should he resulted In This document prehrdes expecations for use of approgra ram enhancements or new Operating Isdiscussed I Sectior 4.10. Otdsnce Experience (0E) In daly work actItes.

operating prgpnm This Informaton can show where anc 1007 A2 direction on evaluating OE information ond guidance None existing program has succeeded and where it has tn sharing a Nuclear Network Operating Experiance failed ff at 8li) In intercepting aging degradtoion In a (NN 0E) entry with the Industry.

timely marrer.This information should provide obJectiveevidence to support the conclusion that tie effects of aging will be managed adequately so that the slmcturemid componer Intendled function(s) wilt be maintained during the period Of extended operatins.

An appplcant may have to cotrrrndtto providing 2 operating experience in the fubtr for new pegramns Operating Experience Isdiscussed In Section 4.10. None None None None to courn their effectiveness.

Page 62 of 62 KLR-1309Aj Rev.

KLR-1309A, Rev. 33 Page 62 of 62

v t e Letter Sequence Response to RAI
TAC:ME7727, (Approved, Closed)
Initiation Request Acceptance... Administration Questions 15 May 2012 2 August 2012 21 August 2012 2 March 2012 13 April 2012 17 May 2012 17 July 2012 Answers 2 April 2012 2 April 2012 Results Approval Other: ML12011A123, ML12206A078, ML12214A583, ML12268A026, NRC-2012-0197, Exemption
MONTHYEAR2012-05-17 [Table View]

ML12101A231

Text

Enclosure:

References:

SUMMARY

SUMMARY

7.0 CONCLUSION

8.0 REFERENCES

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