ML11355A156

Reactor Vessel Internals Inspection Plan Review Request

ML11355A156
Person / Time
Site:	Kewaunee
Issue date:	12/12/2011
From:	Price J Dominion Energy Kewaunee
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
11-603
Download: ML11355A156 (44)

Text

Dominion Energy Kewaunee, Inc. .....

5000 Dominion Boulevard, Glen Allen, VA 23060 DomlnEONj DEC 12Z2 I ATTN: Document Control Desk Serial No.11-603 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 REACTOR VESSEL INTERNALS INSPECTION PLAN REVIEW REQUEST Pursuant to the provisions of Renewed Operating License DPR-43, Dominion Energy Kewaunee, Inc. (DEK) hereby requests NRC approval of the attached inspection plan for reactor vessel internal (RVI) components at Kewaunee Power Station (KPS).

Renewed Operating License DPR-43, Section 2.C(15)(b), requires that certain activities be completed in accordance with Appendix A of NUREG-1958, "Safety Evaluation Report Related to the Kewaunee Power Station," dated January 2011. These activities are described in the KPS Updated Safety Analysis Report (USAR), Chapter 15, Table 15.7-1, "License Renewal Commitments." Items 1 and 2 of the required activities (commitments) are as follows:

1. The ASME Code 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 staff for review and approval to augment the current inspections.

2. The ASME Code Section Xl Inservice Inspection, Subsections IWB, IWC, and IWD Program will be enhanced to include identification of the limiting susceptible cast austenitic stainless steel (CASS) reactor vessel internal components from the standpoint of thermal aging susceptibility, neutron fluence, and cracking. For each identified component, a plan will be developed that accomplishes aging management through either a supplemental examination or a component-specific evaluation. The plan will be submitted for staff review and approval, not less than 24 months before entering the period of extended operation.

The attachment to this letter transmits the proposed inspection plan as required by the above commitments. DEK requests review and approval of the proposed inspection

Serial No.11-603 Page 2 of 3 plan by October 2012. DEK plans to perform the proposed inspections over the course of the next three refueling outages, commencing with the spring 2012 refueling outage.

Commitment 1 above is based on EPRI Materials Reliability Program (MRP) Report 1016596 (MRP-227), Revision 0, "Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines." Subsequent to the creation of Commitment 1, the NRC staff issued their final Safety Evaluation (SE) for MRP-227. In conjunction with this SE, the NRC also issued Regulatory Issue Summary (RIS) 2011-07, "License Renewal Submittal Information for Pressurized Water Reactor Internals Aging Management," to provide information on acceptable changes to existing license renewal commitments in order to allow licensees to submit their RVI inspection plan based on the guidance for the forthcoming version of MRP-227 approved by the staffs SE (MRP-227-A). This RIS stated that licensees such as DEK may modify their commitments to submit their RVI inspection plan no later than October 1, 2012.

However, KPS License Condition 2.C(15)(b) requires that DEK submit the RVI inspection plan in accordance with Appendix A of NUREG-1958, which would necessitate a submittal date no later than December 21, 2011.

Therefore, DEK is submitting the attached inspection plan in accordance with the KPS license condition. In order to meet the intent of RIS 2011-07, the attached inspection plan incorporates information based on the NRC staffs SE of MRP-227, using the most recent information available prior to the date of this letter.

If you have questions or require additional information, please contact Mr. Jack Gadzala at 920-388-8604.

Very truly yours, J. an edce Vi e esident- Nuclear Engineering

Attachment:

1. Kewaunee Power Station Inspection Plan for the Augmented Inservice Inspection Program for Examination of Reactor Vessel Internals Commitments made by this letter: No new commitments are made. This letter fulfills Items 1 and 2 (USAR Chapter 15, Table 15.7-1, License Renewal Commitments) to submit an inspection plan for reactor internals (including CASS components) to the staff for review and approval to augment the current inspections.

Serial No.11-603 Page 3 of 3 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.11-603 ATTACHMENT I KEWAUNEE POWER STATION REACTOR VESSEL INTERNALS INSPECTION PLAN REVIEW REQUEST INSPECTION PLAN FOR THE AUGMENTED INSERVICE INSPECTION PROGRAM FOR EXAMINATION OF REACTOR VESSEL INTERNALS KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

Serial No.11-603 Attachment 1 Page 1 of 10 INSPECTION PLAN FOR THE AUGMENTED INSERVICE INSPECTION PROGRAM FOR EXAMINATION OF REACTOR VESSEL INTERNALS INTRODUCTION The American Society of Mechanical Engineers (ASME) Code Section XI Inservice Inspection (ISI) (Reference 1), Subsections IWB, IWC, and IWD program is described in the KPS Updated Safety Analysis Report (USAR) Section 15.3.2, "ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD." As stated in the USAR, this program corresponds to NUREG-1801,Section XI.M1, "ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD." Program details are contained in Technical Report KLR-1309, "License Renewal Project, Aging Management Program, ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD."

Enhancements to the ASME Section Xl ISI, Subsections IWB, IWC, and IWD program for managing aging effects on reactor internals and on limiting susceptible cast austenitic stainless steel (CASS) reactor vessel intemals components are detailed in Technical Report KLR-1309A, "License Renewal Project, Aging Management Program, ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD, Reactor Vessel Internals Inspections." These enhancements are being made in accordance with two license renewal commitments described in the KPS Updated Safety Analysis Report (USAR), Chapter 15, Table 15.7-1, "License Renewal Commitments" (Commitments 1 and 2). These two commitments are as follows:

1. The ASME Code 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 staff for review and approval to augment the current inspections.

2. The ASME Code Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Program will be enhanced to include identification of the limiting susceptible cast austenitic stainless steel (CASS) reactor vessel internal components from the standpoint of thermal aging susceptibility, neutron fluence, and cracking. For each identified component, a plan will be developed that accomplishes aging management through either a supplemental examination or a component-specific evaluation. The plan will be submitted for staff review and approval, not less than 24 months before entering the period of extended operation.

Augmented examinations are those examinations that are performed outside the scope of the requirements of ASME Boiler and Pressure Vessel Code Section XI (and are

Serial No.11-603 Attachment 1 Page 2 of 10 instead governed by the USAR and Technical Specifications) or that are required to be performed by ASME/ANSI OM Standard Part 4 (as referenced in ASME Boiler and Pressure Vessel Code Section XI).

The Augmented ISI Program inspection plan for examination of reactor vessel internals is organized into four (4) groups of tables for examinations as primary components, expansion components, existing programs, and cast austenitic stainless steel (CASS) components.

The four program groups are defined as follows (The first three groups are associated with Commitment 1. The fourth group is associated with Commitment 2).

Group 1 Primary Components - provides a listing of items to be inspected consistent with Table 4-3, "Westinghouse Plants Primary Components (MRP-227-Rev-0)." This group has been supplemented to include TRC-2 from the NRC SER dated June 22, 2011.

Group 2 Expansion Components - provides a listing of items to be inspected if degradation is verified from group 1 inspections consistent with Table 4-6, "Westinghouse Plants Expansion Components (MRP-227-Rev 0)," and Table 5-3, "Westinghouse Plants Examination Acceptance and Expansion Criteria."

This group has been supplemented to include TRC-1 from the NRC SER dated June 22, 2011.

Group 3 Existing Programs - provides a listing, for information only, of items historically inspected consistent with Table 4-9, 'Westinghouse Plants Existing Programs Components."

Group 4 Reactor vessel internal components fabricated from cast austenitic stainless steel (CASS). Each of these CASS components has been reviewed by Westinghouse Electric Company to determine whether they should be classified as primary, expansion, existing, or no additional measures.

Commitment 1 requires augmented inspections associated with the first three groups, which are based on the guidance in EPRI Materials Reliability Program (MRP) Report 1016596 (MRP-227), Revision 0, "Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines." Applicable acceptance criteria for the MRP-227 inspections are provided in WCAP-1 7096 (Reference 3), WCAP-1 5425 (Reference 4),

and WCAP-1 7020-P (Reference 5). In developing the inspection plan for the first three groups, DEK used the information provided in the NRC staffs final Safety Evaluation (SE) of MRP-227, dated June 22, 2011. A summary of the applicability of each topical report condition and licensee action item discussed in the SE to the KPS reactor vessel internal components is provided in the program description below.

Serial No.11-603 Attachment 1 Page 3 of 10 Commitment 2 requires augmented inspections associated with the fourth group, CASS components. Guidance for the associated inspections of the fourth group is taken from NUREG-1 801, XI.M13, Inspection Plan Cast Austenitic Stainless Steel (CASS) Reactor Vessel Internal Components (Reference 2).

This attachment submits the plan, contained in Tables 1 through 4 below, for conducting augmented ISI Program inspections of reactor vessel internal components, organized into the four groupings discussed above. These inspections are planned to start during the fourth inspection interval (June 2004 - June 2014).

INSPECTION PROGRAM DESCRIPTION - GROUPS 1, 2, AND 3 The Augmented ISI Program for examination of reactor vessel internals associated with primary components, expansion components, and existing programs (Groups 1, 2, and

3) is based on EPRI Materials Reliability Program (MRP) Report 1016596 (MRP-227),

Revision 0, "Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines" (Reference 9). Subsequent to the creation of Commitment 1, the NRC staff issued their final Safety Evaluation (SE) of MRP-227 (Reference 10). In conjunction with this SE, the NRC also issued Regulatory Issue Summary (RIS) 2011-07, "License Renewal Submittal Information for Pressurized Water Reactor Internals Aging Management," to provide information on acceptable changes to existing license renewal commitments in order to allow licensees to submit their RVI inspection plan based on the guidance of the forthcoming version of MRP-227 approved by the staffs SE (MRP-227-A). This RIS stated that licensees such as DEK may modify their commitments to submit their RVI inspection plan no later than October 1, 2012.

However, KPS License Condition 2.C(15)(b) requires that DEK submit the RVI inspection plan in accordance with Appendix A of NUREG-1958, which would necessitate a submittal date no later than December 21, 2011.

Therefore, DEK is submitting this inspection plan in accordance with the license condition. In order to meet the intent of RIS 2011-07, the inspection plan incorporates information based on MRP-227, Revision 0, as augmented by the NRC staffs SE for MRP-227, using the most recent information available prior to the date of this letter.

Conformance of MRP-227 Inspection Plan to NRC SE The NRC staff issued their final SE for MRP-227 on June 22, 2011 (Reference 10). The SE contains seven (7) topical report conditions (TRC) and eight (8) licensee action items (LAI). As discussed in RIS 2011-07, these TRCs and LAIs are to be incorporated into the approved version of MRP-227, designated MRP-227-A. However, MRP-227-A may not be published prior to the date that DEK is required to submit the RVI inspection plan specified in KPS License Condition 2.C(15)(b). Therefore, to meet the intent of

Serial No.11-603 Attachment 1 Page 4 of 10 RIS 2011-07, the TRCs and LAIs in the SE were reviewed and incorporated, as applicable, into the proposed inspection plan.

This section provides a summary of the applicability of each TRC and LAI discussed in the SE, as it relates to the KPS reactor vessel internal components (applicable only to Groups 1 through 3).

TRC-1 When a surface breaking flaw is confirmed by EVT-1 on the upper core barrel flange welds or control rod guide tube flange welds, then expansion of the EVT-1 examination is required to the lower support forging and upper core plate. The expansion examinations are to be completed by the end of the next refueling outage (following flaw confirmation). TRC-1 is applicable to KPS.

TRC-2 Add EVT-1 examination of the core barrel girth welds as a Primary item. There are a total of four (4) circumferential welds in the KPS core barrel: the upper core barrel flange weld (Primary), Core Barrel Mid Plane Weld (NRC SER), Core Barrel Lower Mid Plane Weld (NRC SER), and Core Barrel Lower Bottom Weld. Therefore, EVT-1 examination of the Core Barrel Mid Plane Weld and Core Barrel Lower Mid Plane Weld is required no later than two refueling outages from the beginning of the license renewal period and subsequent examination on a ten-year interval. The examination volume is 100% of one side of the accessible surfaces of the selected weld and adjacent base metal. TRC-2 is applicable to KPS.

TRC-3 TRC-3 pertains to the Support Column Welds of Combustion Engineering Plants and is not applicable to KPS.

TRC-4 This item relates to the minimum coverage that applies to examination of expansion items. Per NRC SER the minimum examination coverage applicable to examination of expansion items at KPS will be 75%. TRC-4 is applicable to KPS.

TRC-5 This item deals with the required re-inspection frequency for ultrasonic examination of the baffle bolts. MRP-227, Revision 0 proposes that re-inspection would occur on a 10 to 15 EFPY frequency. However, the NRC SER requires that ultrasonic inspection be performed on a 10-year frequency. TRC-5 is applicable to KPS.

Serial No.11-603 Attachment 1 Page 5 of 10 TRC-6 This item deals with how often re-inspections will occur to expansion items once they are initially performed because the acceptance criteria of the Primary component requires expansion to another component. Per the NRC SER the re-inspection frequency is 10 years for the expansion items. TRC-6 is applicable to KPS.

TRC-7 This item deals with the 10 programmatic elements identified in the GALL document.

The existing KPS Aging Management Program (AMP) for the reactor vessel internals, KLR-1309A, identifies the 10 programmatic elements listed in GALL Revision 1. These program elements remain the same as the program elements included in Appendix A of MRP-227, Revision 0 and Appendix A of NUREG-1800, Revision 1, "Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants." Therefore, no change is required by TRC-7 since the existing KPS AMP contains the same 10 programmatic elements. At KPS, the program scope will be consistent with these program elements as implemented by Procedure ER-AA-AMP-1003, "Aging Management Programs for License Renewal Structures, Systems, and Components."

LAI-1 This item deals with the functionality analyses and supporting aging management strategies in MRP-232. Section 2.4 of MRP-227 requires that the following assumptions be validated for each Westinghouse reactor:

30 years of operation with high leakage core loading patterns (fresh fuel assemblies loaded in peripheral locations) followed by implementation of a low-leakage fuel management strategy for the remaining 30 years of operation;

• Base load operation, i.e., typically operates at fixed power levels and does not usually vary power on a calendar or load demand schedule; and

• No design changes beyond those identified in general industry guidance or recommended by the original vendors.

KPS has validated that these assumptions are applicable, as referenced in KLR-1309A.

LAI-2 This item deals with ensuring that the reactor internals components have been considered in the scope of license renewal. KPS contracted Westinghouse Electric Company to review the KPS reactor vessel internals. As part of the review process Westinghouse Electric Company reviewed and classified each of the intemals as Primary, Expansion, Existing, or No Additional Measures. Therefore, each of the reactor vessel internals components have been classified into the appropriate aging management group based upon industry recommendations outlined in MRP-227. This item is complete for KPS.

Serial No.11-603 Attachment 1 Page 6 of 10 LAI-3 This item deals with management of the Westinghouse guide tube support pins (split pins). Originally, the split pins at KPS were fabricated from type X-750 inconel. Some of the original type X-750 inconel split pins at other Westinghouse sites failed as a result of higher than desired stresses in the head to shank region and heat treatment that was not fully optimized. The split pins at KPS were replaced in 2004 with type 316 stainless steel. Replacement with type 316 stainless steel. split pins is expected to resolve the cracking issue observed in the type X-750 inconel split pins seen at other sites. The KPS type 316 split pins will only have been in service approximately 29 years through the end of the license renewal period (2033). Therefore, this item is complete for KPS.

LAI-4 This item deals with B&W Core Support Structure Upper Flange Stress Relief. This item is not applicable to KPS.

LAI-5 This item deals with physical measurement of the Westinghouse hold-down springs fabricated from type 304 stainless steel. The hold-down springs at KPS are fabricated from type 403 stainless steel materials. Type 403 stainless steel is not subject to the aging mechanism of concern. This item is not applicable to KPS.

LAI-6 This item deals with evaluation of inaccessible B&W components. The KPS reactor internals are designed by Westinghouse Electric Company. Therefore, LAI-6 does not apply to KPS.

LAI-7 This item deals with the need for a plant-specific evaluation of CASS materials for the Westinghouse lower support column bodies. At KPS, the lower support column bodies are not fabricated from CASS. Therefore, LAI-7 does not apply to KPS.

LAI-8 This item deals with the need for submittal of information for NRC review and approval.

Commitment 1 in the Updated Safety Analysis Report (USAR), Chapter 15, Table 15.7-1, "License Renewal Comments," indicates that KPS will submit an inspection plan for the reactor internals to the NRC staff for review and approval to augment the current inspections not less than 24 months before entering the period of extended operation.

LAI-8 applies to KPS.

INSPECTION PROGRAM DESCRIPTION - CASS COMPONENTS (GROUP 4)

Serial No.11-603 Attachment 1 Page 7 of 10 Reactor vessel internals are visually inspected in accordance with ASME Code Section XI, Subsection IWB, Category B-N-3. This inspection is augmented to detect the effects of loss of fracture toughness due to thermal aging and neutron irradiation embrittlement of CASS reactor vessel internals. This CASS reactor vessel internals inspection program includes the following two aspects.

1. Identification of susceptible components determined to be limiting from the standpoint of thermal aging susceptibility (i.e., ferrite and molybdenum contents, casting process, and operating temperature) and/or neutron irradiation embrittlement (neutron fluence); and,

2. For each "potentially susceptible" component, aging management is accomplished through either a supplemental examination of the affected component based on the neutron fluence to which the component has been exposed as part of the 10-year ISI program during the license renewal term, or a component-specific evaluation to determine its susceptibility to loss of fracture toughness.

CASS components in the KPS reactor vessel internals include:

" Upper Internals Mixing Devices [CF8]

• Upper Instrumentation Conduit Supports [CF8]

" Upper Instrumentation Clamps [CF8]

• Upper Support Column Bases [CF8]

• Upper Support Thermocouple Stops [CF8]

• BMI Column Cruciforms [CF8]

The program piovides screening criteria to determine the susceptibility of CASS components to thermal aging on the basis of casting method, molybdenum content, and percent ferrite. The screening criteria are applicable to primary pressure boundary and reactor vessel internal components constructed from SA- 351 Grades CF3, CF3A, CF8, CF8A, CF3M, CF3MA, CF8M, with service conditions above 250 0 C (482 0 F). The screening criteria for susceptibility to thermal aging embrittlement are not applicable to niobium-containing steels; such steels require evaluation on a case-by-case basis. For "potentially susceptible" components, the program provides for the consideration of the synergistic loss of fracture toughness due to neutron embrittlement and thermal aging embrittlement. For each such component, DEK can implement either (a) a supplemental examination of the affected component as part of a 10-year ISI program during the license renewal term; or, (b) a component specific evaluation to determine the component's susceptibility to loss of fracture toughness.

Based on the criteria set forth in the May 19, 2000 letter from Christopher Grimes (NRC) to Douglas Walters (NEI) (Reference 6), the susceptibility to thermal aging embrittlement of CASS components is determined in terms of casting method,

Serial No.11-603 Attachment 1 Page 8 of 10 molybdenum content, and ferrite content. For low-molybdenum content steel (0.5 wt.%

max.), only static-cast steel with > 20% ferrite is potentially susceptible to thermal embrittlement. Static-cast low-molybdenum steel with < 20% ferrite and centrifugal-cast low-molybdenum steel is not susceptible. High-molybdenum content (2.0 to 3.0 wt.%)

steel, static-cast steel with > 14% ferrite, and centrifugal-cast steel with > 20% ferrite are potentially susceptible to thermal embrittlement. Static-cast high-molybdenum steel with < 14% ferrite and centrifugal-cast high-molybdenum steel with

• 20% ferrite are not susceptible. In the susceptibility screening method, ferrite content is calculated by using the Hull's equivalent factors (described in NUREG/CR-4513, Revision 1) (Reference 7).

The program specifics depend on the neutron fluence and thermal embrittlement susceptibility of the component. Each of the CASS items has been classified by Westinghouse Electric Company as "No Additional Measures" per the guidance in MRP-227 and MRP-232. EPRI MRP-175, "Materials Reliability Program: PWR Internals Material Aging Degradation Mechanism Screening and Threshold Values - EPRI Report 1012081," 2005, indicates the neutron fluence threshold for CASS as greater than 1020 n/cm 2 (E>1 MeV) (Reference 8). For CASS items classified as "No Additional Measures," the inspection program monitors the effects of loss of fracture toughness on the intended function of the component by identifying the CASS materials that have a neutron fluence of greater than 1020 n/cm 2 (E>1 MeV) and are determined to be susceptible to thermal aging embrittlement. For such materials, the program consists of either supplemental examination of the affected component based on the neutron fluence to which the component has been exposed, or component-specific evaluation to determine the component's susceptibility to loss of fracture toughness.

For reactor vessel internal CASS components classified as "No Additional Measures" that have a neutron fluence of greater than 1020 n/cm 2 (E>1 MeV) and are determined to be susceptible to thermal embrittlement, the 10-year ISI program during the renewal period includes a supplemental inspection covering 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). One example of a supplemental examination is enhancement of the visual VT-1 examination of Section Xl IWA-2210. A description of such an enhanced visual VT-1 examination could include the ability to achieve a 0.0005-inch resolution, with the conditions (e.g., lighting and surface cleanliness) of the inservice examination bounded by those used to demonstrate the resolution of the inspection technique. Another example of a supplemental examination is an EVT-1 visual examination. An enhanced visual (EVT-1) examination has the same requirements as the ASME Code Section Xl visual (VT-1) examination, with additional requirements given in MRP-228, "Materials Reliability Program: Inspection Standard for PWR Internals." The inspection schedule for CASS items at KPS is provided in Table 1.

Alternatively, in lieu of performing a supplemental enhanced visual examination, DEK may perform a component-specific evaluation, including a mechanical loading

Serial No.11-603 Attachment 1 Page 9 of 10 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.

Failure to meet this criterion requires continued use of the supplemental inspection program. For each CASS component that has been subjected to a neutron fluence greater than 1020 n/cm 2 (E>1 MeV) and is potentially susceptible to thermal aging, the supplemental inspection program applies; otherwise, the existing ASME Section Xl inspection requirements are adequate if the components are not susceptible to thermal aging embrittlement.

An enhanced visual inspection will not be required for KPS reactor internal CASS items that are shown to have either a neutron fluence less that 1020 n/cm 2 (E>1 MeV); delta ferrite less than 20%; or, loading in compression or low enough (< 5 ksi) to preclude fracture. Accessible surfaces of CASS items that screen out as not susceptible will continue to be inspected to the extent possible using a VT-3 method if required by ASME Section XI, Subsection IWB, Categories B-N-i, B-N-2, and B-N-3.

Extent of Examination A supplemental enhanced visual examination is performed on accessible surfaces of CASS items that have a neutron fluence greater than 1020 n/cm 2 (E>1 MeV) and delta ferrite greater than 20% and tensile loading (> 5 ksi).

Relevant Conditions While an enhanced inspection technique will be used for detection of relevant conditions, the inspection results will be assessed using ASME Section Xl. Any of the following relevant conditions shall be unacceptable for continued service unless the requirements of ASME Section Xl, IWB - 3142 are met.

• Structural distortion or displacement of parts to the extent that component function may be impaired.

• Loose, missing, cracked, or fractured parts, bolting, or fasteners.

• Foreign materials or accumulation of corrosion products that could interfere with control rod motion or could result in blockage of coolant flow through fuel.

• Corrosion or erosion that reduces the nominal section thickness by more than 5%.

" Wear of mating surfaces that may lead to loss of function.

" Structural degradation of interior attachments such that the original cross-sectional area is reduced by more than 5%.

If relevant conditions are found during the enhanced visual inspections, ASME Section Xl, IWB-3142 states that the affected components cannot be returned to service until deemed acceptable per code requirements.

Serial No.11-603 Attachment 1 Page 10 of 10 REFERENCES

1. ASME Section Xl, Rules for Inservice Inspection of Nuclear Power Plant Components, ASME Boiler and Pressure Vessel Code, 1998 edition including the 2000 Addenda.

2. NUREG 1801, Volume 2, Revision 1, Chapter XI.M13.

3. WCAP-1 7096, "Reactor Internals Acceptance Criteria Methodology and Data Requirements," Revision 0, July 2009.

4. WCAP-15425, "Determination of Acceptable Baffle-Barrel-Bolting for Kewaunee and Prairie Island Plants," Revision 0, May 2001.

5. WCAP-17020-P, "Point Beach Unit 1 Upper Internal Guide Tube - Guide Card Wear Evaluation," Revision 0, September 2009.

6. Letter from Christopher I. Grimes (NRC) to Douglas J. Walters (NEI), "License Renewal Issue No. 98-0030, Thermal Aging Embrittlement of Cast Stainless Steel Components," May 19, 2000. (ADAMS Accession No. ML003717179)

7. NUREG/CR-45113, "Estimation of Fracture Toughness of Cast Stainless Steels during Thermal Aging in LWR Systems," Revision 1, August 1994.

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

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

10. Final Safety Evaluation (SE) of EPRI Report, Materials Reliability Program Report 1016596 (MRP-227), Revision 0, "Pressurized Water Reactor (PWR)

Internals Inspection and Evaluation Guidelines," dated June 22, 2011.

Serial No.11-603 Table 1 Reactor Vessel Internals Inspection Plan MRP-227 Westinghouse Plants Primary Components (14 pages)

KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

Serial No.11-603

• ~TABLEI KEWAUNEE-POWER STATION

-, FURTH-ANgD '7FýT.H iNifERVALISLCHEDULE Examination Category: MRP-227

Description:

TABLE 4-3 WESTINGHQ(USE PILANTS PIR. AR, C.MENTCONTROLROD.*DRIVE TUBES,(CRGT) GUIDE PLATE CARDS

.........k......N.., Ira. * .WinN ,. C* ,, ...

Item No. Parts Examined 11- DrawigNo:* MqQ'pmdnt. iNT -Rel

'~ 3~ ShOI ~oV 2 Su~Vis r~eiefComments Reactor Vessel Internals CRGT Guide Plate Cards Attachment Position 2F N X Visual (VT-3) examination no Figure 2 later than 2 refueling outages from the beginning of the license renewal period, and no earlier than two refueling outages prior to the start of the license renewal period. 20% examination of the number of CRGT assemblies, with all guide cards within each selected CRGT assembly examined. A total of 29 locations. It is suggested that the population selected for initial inspection coincide with the inlet nozzle locations.

Attachment CRGT Guide Plate Cards Figure 2 Position 2H N X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 3E N X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 3G N X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 31 Y X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 4D N X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 4J Y X Same as above Table 1, Page 1 of 14

Serial No.11-603

.. .... . . - *:,,.:*.*.*..**- i i:,...

.:: .:**:*':*T B L E :I..** " =.-*-*...... ........

KEWANEE OWERSTATION.,

~FOURTH AND, FIFT;H NTERVAL I5 SI;SCHEDULE-Examination Category: MRP-227, DeScri*ptildn' TABLE .'3 WESTINGHOUSE-: PLANTS PRIMARY.COMPONENI;TS.SGONTROL RODDDRIVE TUBES (CRGT) GUIDE PLATE CARDS Eiaminatio'n 'Exemption, ExaminedforPeio *;,Ea~iDei~t*No.;

'l, Code Case,....

Item No. Parts Examined INT..'-*,t-o P 1i'Io

,-1,raw.n;1No.Equipment.No *-r"Relief:Comments AttachmentNo. - -

CRGT Guide Plate Cards Figure 2 Position 5C N X Same as above Attachment I CRGT Guide Plate Cards Figure 2 Position 5E N IX Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 50 N X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 5K X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 6B N X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 6F N X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 6F N X Same as above CRGT Guide Plate Cards Attachment-Figure 2 Position 6L N I - X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 7C N X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 7E N' X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 7G N X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 71 N X Same as above Table 1, Page 2 of 14

Serial No.11-603 TARLFA Examinationia Category

-I(WNANE R WE TRPS2 ROD DRIVE-TUBES (CRGT) GUIDE PLATE CARDS

- ~ ~Examination Exerrft on, Examination Perio~d- - Me~tho-d~s' Item No. Parts Excamined IISI DrawingNN;g.. Equipment',No. INT -T ___ pd ae Comments

.Sch I. . 3.. E':. "Vol Sur.- Vis. . .--- est Attachment CRGT Guide Plate Cards Figure 2 Position 7K N X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 8B N X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 8F N X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 8H N X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 8L N X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 9C Y X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 9E N X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 9G N X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 91 N X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 9K N X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 10D Y X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 10J N X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 11iE Y X Same as above Table 1, Page 3ofl14

Serial No.11-603 Examination Category: MRP-227 GUIDE PLATE CARDS Item N0. Parts Examined Itmo.-:-i.Comments

• 1--

Attachment CRGT Guide Plate Cards Figure 2 Position 11 G N X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 111 N X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 12F N X Same as above Attachment CRGT Guide Plate Cards Figure 2 Position 12H N X Same as above Category Notes:

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

Table 1, Page 4 of 14

Serial No.11-603

. . . EWAUN.EE POWER STATiON ExaminationCategory: *MRP227HFIONUR1,AL8 I CSCHEDIULE I.. ..

Examination Category: MRP-227

Description:

-TABLE 4-3 WESTINGHOUSE PLkNITiýPRIMMRYkCOMBONENTS d NTROL ROD DRIVE TUBES (CRGT) LOWER FLANGE WELDS Parts Examid ItemNo. o. E N Comments NS. Exaine DangN- Equippient 6o: INT. u Vi oreid SSc h- I 2 E Vol S Vt Re1uef Reactor Vessel Internals CRGT Lower Flange Welds Attachment Position 2F Y X Enhanced visual (EVT-1)

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

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

Attachment CRGT Lower Flange Welds Figure 4 Position 2H Y X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 3E Y X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 3G X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 31 Y X Same as above Table 1, Page 5 of 14

Serial No.11-603 TAB3LE 1 KEWAUNE E POWER STATION FOURTH AND FIFTH INTERVAL IS1 SCHEDULE -

Examination Category:. MRP227 Descrpription: TABLE 4-3 WESTINGHOUSE PLANTS PRIMARY COMPONENTS CONTROL ROD DRIVE TUBES (CRGT) LOWER FLANGE WELDS Exainarition' Exemption, Examination Period Methods Code CaseCmet Item No. Parts Examined ." 1 Drawing No. Equipment No. INT-..,- . - - oroRelief

-Sch. 1 2 .3.E0I Vol Sur.. Vis Request Attachment CRGT Lower Flange Welds Figure 4 Position 4D Y X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 4J Y X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 5C Y X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 5E N X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 5G N X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 51 N X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 5K Y X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 6B Y X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 6F N X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 6H N X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 6L Y X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 7C N X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 7E N X Same as above Table 1, Page 6 of 14

Serial No.11-603 TABLE 1 KEWAUNEE POWER STATION; FOURTH AND: FIFTH INTERVAL 1iS SCHEDULE .

Examination Category: MRP-227

Description:

TABLE 4-3 WESTINGHOUSE PLANTS PRIMARY COMPONENTS CONNTRoLROD DRIVE TUBES (CROT) LOWER FLANGE WELDS

. aExamnationit Exemption, Ite No SI PrtsExmind o.Methods Code Case, C Item No. Parts Examined :lsi*Drawing-No. Equipment No. INT. "--.- or. Relief Comments Sch 1 2 13~ E01 Vol; Sur Vis Request..

Attachment ________ __________

CRGT Lower Flange Welds Figure 4 Position 7G N jX Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 71 N X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 7K N X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 8B X Same as above Attachment I CRGT Lower Flange Welds Figure 4 Position 8F N X Same as above Attachment CRGT Lower Flange Welds Figure 4 'Position 81-1 N X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 8L Y X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 9C 8 X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 9G N X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 9G N X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 9K N X Same as above Attachment CRGT Lower Flange Welds Figure 4 Attachment Position 91 X Same as above IL CRGT Lower Flange Welds Figure 4 Position 10D YXSame as above Table 1, Page 7 of 14

Serial No.11-603

• ... : .  ::!:. ' .'.-T"13LE....  ;!

E'XWAUNEE POWER-S~TA1lON FOURH ANFIF~THINTERVAL ISI SCHBOUInE ExainaionCategory: MIRP-227

Description:

TABLE 4-3 WESTINGHOUSE. PLANTS PRIM ARY-.COM PONENTS, CONTRO RO D1RIVE TUBES (CRGT) LOWER FLANGE WELDS 6 ~ ka ation E item.No... .i Examla.la....xem ... n Case, EaemMRne 1 Decrai-Euipment No. INT RODDRIE.TBE Case, Comments

~'Sh 1 O o ur Request Attachment CRGT Lower Fiange Welds Figure 4 Position 10OJ VXSame as above Attachment CRGT Lower Flange Welds Figure 4 Position 11iE Y X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 11 G N X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 11l Y X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 12F Y X Same as above Attachment CRGT Lower Flange Welds Figure 4 Position 12H Y X ISame as above Category Notes:

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

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

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

Table 1, Page 8 of 14

Serial No.11-603 Table 1, Page 9 of 14

Serial No.11-603

Y - UPPER CORE BARREL FLANGE

..mption, '"

1_1-Relief

• Comments aquest Periodic enhanced visual (EVT-1) examination, no later than 2 refueling outages from the beginning of the license renewal period and subsequent examination on a ten-year interval. 100% of one side of the accessible surfaces of the selected weld and adjacent base metal. No expansion required.

Periodic enhanced visual (EVT-1) examination, no later than 2 refueling outages from the beginning of the license renewal period and subsequent examination on a ten-year interval. 100% of one side of the accessible surfaces of the selected weld and adjacent base metal. No expansion required.

Category Notes:

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

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

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

4. NRC SER TRC-2 does not apply to the Core Barrel Lower Bottom Weld since it is not a flange weld.

5. A minimum of 75% of the total weld length (examined + unexamined), including coverage consistent with the Expansion criteria in Table 5-3, must be examined from either the inner or outer diameter for inspection credit, Table 1, Page 10 of 14

Serial No.11-603 E'POWUA 0 ER'STATIQN ..

FORHABFIFTH INEVL W EPULE Examination Category: MRP-227

Description:

TABLE *.3

"'"'ý St1KEIsTJHSE PLiJANTS: PRIMARY;Q A.., ... *""iii

'** *.*. ' -, ':" .....* FORMERWASSEMBLY.-

MP-ONENTS.BAFFLE .. ":' - - I* 1,"

ýý -....ý BAFFLE:EDGE

" .. . . .... BOLTS

' . *..-.-.* Examination

• Exetmption,

_a-i-lodV Examlatn 'leuýase,*:. Mehd Co~d~ase Item No. Parts Examined IS'DrawingNo.,Euipment*, . ... ' -"- Comments

.le U es , ... . . . : . . *

, .i * . _*j Reactor Vessel Internals j

Core Barrel Baffle-Edge Bolts WCAP- 13266, R1 688 Edge Y X Visual (VT-3) examination, with Baffle- Figs 6.1, 6.2, 6.3 Bolts baseline examination between 20 Former and 40 EFPY and subsequent Assembly examinations on a ten-year interval. Bolts and locking devices on high fluence seams.

100% of components accessible from the cere side. Reference Figure 4-23 from MRP-227. No

_i expansion required.

Category Notes:

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

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

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

Table 1, Page 11 of 14

Serial No.11-603 Table 1, Page 12 of 14

Serial No.11-603 TABLE .

KWALJNEE POWIER STATIPN INTER-VAL.ISIS:

D,FI: .FTH HEULE Examination Category: MRP-227 Descriptton ThA 3L WESTINGlOUJSE'ANTS PRIMARYCOMPONENTS BAFLE . FORMER ASSEMIBLY - ASSEMBLY

'P Exempti1onto Examination Pdio to ihY xmtin Item No. Parts Examined :II Drawig NoN Equip ment",.INT . C Comments EOI dir~elief, Sch, 2: E0 Vol.. Sur VI Request Reactor Vessel Internals Core Barrel Assembly WCAP-13266, R1 Baffle Former Y X Visual (VT-3) examination to Baffle-Former Figs. 6.1, 6.2, 6.3 check for evidence of distortion, Assembly with baseline examination Attachment between 20 and 40 EFPY and Figure 4 subsequent examinations on a ten-year interval. Inspections are performed on the core side surface. Reference Figures 4-24, 4-25, 4-26, and 4-27 of MRP-227.

I__- No expansion required.

Category Notes:

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

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

Table 1, Page 13 of 14

Serial No.11-603

...TABLE 1.

KEWAUNEE POWER STATION S FOURTH ANID FIFTH IN TPRVAL IS1 SCHDL Examination Category: MRP-227 De.scription: TABLE 4-3 WESTINGHOUSE PLANTS PRIMARY COMPONENTSa THERMAL SHIELD ASSEMBLY - THERMAL SHIELD FLEXURES Examination Period Examination Exemption,

- Methods CoeCae Item No. Parts Examined ISI Drawing No. Equipment No.; INT. - - - - o*r Relif Comments

" - ": : 1" 2 3..EOI Vol Sur Vis qReuest Reactor Vessel Internals Core Barrel Thermal Shield Flexures Attachment Thermal Shield Y X Visual (VT-3) examination no Thermal Figure 4 Flexure 00 later than 2 refueling outages Shield from the beginning of the license Assembly renewal period. Subsequent examinations on a ten-year interval. 100% of thermal shield flexures. Reference Figures 4-29 and 4-36 of MRP-227. No expansion required.

Core Barrel Thermal Shield Flexures Attachment Therma! Shieid Y X Same as above Thermal Figure 4 Flexure 900 Shield Assembly Core Barrel Thermal Shield Flexures Attachment Thermal Shield I X Same as above Thermal Figure 4 Flexure 1800 Shield Assembly Core Barrel Thermal Shield Flexures Attachment Thermal Shield Y X Same as above Thermal Figure 4 Flexure 2700 Shield Assembly Categorv Notes:

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

Table 1, Page 14 of 14

Serial No.11-603 Table 2 Reactor Vessel Internals Inspection Plan MRP-227 Westinghouse Plants Expansion Components (7 pages)

KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

Serial No.11-603 T.-A BLE 2 KEWAUNEE POWERS TATION.

F*OURT*HAND FIFTHI.I VE!AL I SCHVED ULE .

Examination Category: MRP-227 -Dbsscription!*TABLE`4-66WESTINGHUSEO.LANTýS.EX*PANS"IN cOMPONEN*TS"CORE.BARREL FORMER ASSEMBLY - BARREL-FORMER BOLTS

_ A; 'xammation . Exemption,.

.ivriod ..... ... M * - aseC Item No. Parts Examined .ISl DrNoaWingNNootments INT"..- I .. "<;- -

-~~6 - .~oRelief

_.  :.*::I. ... .* *  :.I-*.

,. ,:.,-'..Sc.h. *k4.

P... i*..'  :** : .* : E:*

O:!*.. o I Su.r*.'.'

Vis:-.:

Reactor Vessel Internals Core Barrel- Barrel-Former Bolts WCAP-13266, R1 344 Barrel- N X Primary Link- Baffle-Former Former Figs 6.1, 6.2, 6.3 Former Bolts Bolts. Volumetric (UT)

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

of accessible bolts.

Accessibility may be limited by presence of thermal shields or neutron pads. Reference Figure 4-23 of MRP-227. Expansion Link - Lower support column bolts and Barrel-former bolts.

Category Notes:

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

and 52.1 EFPY at End of Life Extension.

2. Reference WCAP-13266, Revision I for details. The barrel-former bolts are fabricated from Type 347 stainless steel.

3. Examinations are scheduled per the Corrective Action Process if the number of indications on the Baffle-Former Bolts and Lower Support Column Bolts exceed the threshold.

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

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

6. If expansion is needed/invoked then re-inspection is required on a 10-year frequency per NRC SER TRC-6.

7. Minimum coverage for expansion items is 75% per NRC SER TRC-4.

Table 2, Page 1 of 7

Serial No.11-603

~..~~~~~~~.

. . . .. TA t 2,, .. , ,-....-*....... . .. .. ...

-- KE~WAUNIEEý POWR TATO Examination Category:: MRPI;227 Des.r.. _ PO NTS WER UF`6RORT;ASSEMBL- LOWERSUPP6ORTCO.LUMN

~Exam inati on.Penod0 MChg I Item No. Parts Examined'.. .-ISI.DrawiklgNb.' .*Equi""entNo..- .. . .. .o iie , . Comments

____.. ..,_ *-'.. __ __ EO &V ' is Keu.s Reactor Vessel Internals Lower Lower Support W Drawing N X Primary Link- Baffle-Former Support Column Bolts 882D685 Bolts. Volumetric (UT)

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

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

Category Notes:

1. End of Original License Is December 21, 2013. The KPS RV is projected to reach 33 EFPY at End of Original License, 34.5 EFPY at KR-34 (Spring 2015),

and 52.1 EFPY at End of Life Extension.

2. Examinations are scheduled per the Corrective Action Process if the number of indications on the Baffle-former bolts and Lower Support Column bolts exceed the threshold.

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

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

5. If expansion is needed/invoked then re-inspection is required on a 10-year frequency per NRC SER TRC-6.

6. Minimum coverage for expansion items is 75% per NRC SER TRC-4.

Table 2, Page 2 of 7

Serial No.11-603

" .... TABLE-2

~EWNAUE ~E STATION FO RTH R'~ITH INTERVAL 1SiS 6NEI-U Examination Category: MRP-2327* Desýsrlptio' I4ABLEAý6aWESTINGHOUSEP4ATS EXRANSION COMPONEIN S.QRE .BARREL ASSEMBLY. CORE BARREL- FLANGE. CORE

. .TB.

I iEi TILET N ESG'CRE BAREi9*SAE*'YINJ EfýTI °NNGz-zIESAND LOWE -ORE BARREL FLANGE WELD n." ..'nationxuesmtl . . .

Item No. Parts Examine IS Draw ng o ,_';-'EquExantinatiINT Perio Exepton Comments Reactor Vessel Internals Core Barrel Core Barrel Flange (1), N X Primary Link- Upper Core Barrel Assembly Core Barrel Outlet Flange Weld. Enhanced visual Nozzles(2), (EVT-1) examination, with initial sI examination dependent on the Safety Injection examination results for upper Nozzles (2) core barrel flange.

Re-inspection on a 10-year frequency. 100% of one side of the accessible surfaces of the selected weld and adjacent base metal. Reference Figure 4-34 of MRP-227.

Category Notes:

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

and 52.1 EFPY at End of Life Extension.

2. Examinations are scheduled per the Corrective Action Process if a surface breaking indication with a length greater than two inches is observed in the upper core barrel flange weld.

3. The confirmed detection and sizing of a surface-breaking indication with a length greater than two inches in the upper core barrel flange weld shall require that the EVT-1 examination, and any supplementary UT examination, be expanded to include the core barrel-to support plate weld by the completion of the next refueling outage. If extensive confirmed indications in the core barrel-to-support plate weld are detected, further expansion of the EVT-1 examination shall include the remaining core barrel assembly welds.

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

5. If expansion is needed/invoked then re-inspection Is required on a 10-year frequency per NRC SER TRC-6.

6. Minimum coverage for expansion items is 75% per NRC SER TRC-4.

Table 2, Page 3 of 7

Serial No.11-603

TA-BLE2ýý

-KEWAUNEE POWER STATION  :

MRP227FOURTH,ýND- FIFTH .:NTERVAL ISI SCHEDULE Examination Category: MRP__227..:Description: TABLE:44-.WESTINGHOUSEPLANTSEXPANSION COMPONENTS**LOWER SUPPORT ASSEMBLY - LOWER SUPPORT COLUMN

____ ýEbmnatcnPe~rsod - tn 'Eepin Item No. Parts Examined ISI Dawin*gi ý..Equdipment No.. INT. . ,- -. leC Comments Reactor Vessel Internals Lower Lower Support W685J896 N X Primary Link- Upper Core Barrel Support Column Bodies Flange Weld. Enhanced visual Assembly (Non Cast) (EVT-1) examination, with initial examination dependent on the examination results for upper core barrel flange. Re-inspection on a 10-year frequency. 100% of accessible surfaces. Reference Figure 4-34 cf MRP-227.

Category Notes:

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

2. Examinations are scheduled per the Corrective Action Process if a surface breaking indication with a length greater than two inches is observed in the upper core barrel flange weld.

3. The confirmed detection and sizing of a surface-breaking indication with a length greater than two inches in the upper core barrel flange weld shall require that the EVT-1 examination, and any supplementary UT examination, be expanded to include the core barrel-to support plate weld by the completion of the next refueling outage. If extensive confirmed indications in the core barrel-to-support plate weld are detected, further expansion of the EVT-1 examination shall include the remaining core barrel assembly welds.

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

5. If expansion is needed/invoked then re-inspection is required on a 10-year frequency per NRC SER TRC-6.

6. Minimum coverage for expansion items is 75% per NRC SER TRC-4.

Table 2, Page 4 of 7

Serial No.11-603 TABLE 2 KEWAUNEE POWER STATION FOURTH AND FIFTH-INTERVAL 151 SCHEDULE Examination Category: MRP-227. Descdription: TABLE 4-6 WES1NGHO USE PLANTS -EXPANSION COMPONENTS LOWER SUPPORT ASSEMBLY - LOWER SUPPORT COLUMN

'BODIES(C-AST)

Item No.,Parts..Examined,... ,-.

Examination Period

,,:Methods:

IEmiaon xemption, Co e Case, ommes PartsExai i Drawing No *.'SchEquipment No. INT ... Co-.orRl

.1; 22. .3.. EOi1 Vol Sur -MVis Request Reactor Vessel Internals Lower Lower Support W685J896 N X Lower Support Column Bodies Support Column Bodies are not cast at Kewaunee Assembly (Cast) Power Station. This expansion item [from the CRGT lower flange welds] is Not Applicable to KPS. Primary Link- Control Rod Guide Tube Lower Flanges. Visual (EVT-1) examination. 100% of accessible support columns.

Reference Figure 4-34 of MRP-227.

Categorv Notes:

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

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

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

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

5. If expansion is needed/invoked then re-inspection is required on a 10-year frequency per NRC SER TRC-6.

6. Minimum coverage for expansion items is 75% per NRC SER TRC-4.

Table 2, Page 5 of 7

Serial No.11-603

• EWAL Examination Category: fTION SYSTEM - BMI Item No. Parts E Comments Reactor Vessel 4-Internals 4 4- 4- 4-a-~-4.-4-4-4 4-4 Reactor Vessel Bottom-Mounted W685J896 N X Primary Link- Control rod guide Bottom Instrumentation tube lower flanges. Visual Mounted (BMI) Column (VT-3) examination of BMI Instrumentation Bodies (36) column bodies as indicated by System difficulty of insertion/withdrawal of flux thimbles. Flux thimble insertion/withdrawal to be monitored at each inspection interval. 100% of BMI column bodies for which difficulty is detected during flux thimble insertion/withdrawal.

Re-inspection on a 10-year frequency. Reference Figure 4-35 of MRP-227.

Category Notes:

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

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

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

4. If expansion is needed/invoked then re-inspection is required on a 10-year frequency per NRC SER TRC-6.

5. Minimum coverage for expansion items is 75% per NRC SER TRC-4.

Table 2, Page 6 of 7

Serial No.11-603

-  : ~~BL2 -2

-KEWAUNEERPGWERISTATION I-URkiHAND'FIFTHINTERA -51SCHEDULE Examination Category: SupplemehttiTablee4:66:NRC4Refere.nceSE'RTR'G'.* '.,,.

==

Description:==

wEsTINGHEUSEIP IOiESiUC~IRONENTSRRg

,ýLOWNSI)*G*/M. m*+*ER*SuRP.ORTIORGING,& UPPER CORE PLAETE

_U Examlnation Exemnption, Zl~~ Period Methoýs___ Code' Case, Item No. Parts Examined . Eqnlmeliefo. IoT ellDrawin*6' 2 3- -ElVlsuir~ Vi - Fe`4es't Reactor Vessel Internals Lower Lower Support M-1199 N X Primary Link- Control Rode Support Forging I Guide Tube Flange Welds.

Forging *iVisual (EVT-1) examination.

Re-inspection on a 10-year frequency. 75% per NRC SER TRC-4.

Upper Core Upper Core Plate M-1 199 N X Primary Link- Control Rode Plate Guide Tube Flange Welds.

Visual (EVT-1) examination.

Re-inspection on a 10-year frequency. 75% per NRC SER TRC-4.

Category Notes:

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

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

3. Upper Core Barrel Flange Welds and Control Rod Guide Tube Flange Welds. Confirmation of surface breaking indications shall require EVT-1 examination of the lower suppo~t forging & upper core support within three fuel cycles following the initial observation.

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

5. If expansion is needed/invoked then re-inspection is required on a 10-year frequency per NRC SER TRC-6.

6. Minimum coverage for expansion items is 75% per NRC SER TRC-4.

Table 2, Page 7 of 7

Serial No.11-603 Table 3 Reactor Vessel Internals Inspection Plan MRP-227 Westinghouse Plants Existing Programs Components (3 pages)

KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

Serial No.11-603 TABLE 3

,  :.KEWAIUNE PO ERSTATION -

FORT ADFITH ItNTERVAL 151k SCHEULE 4,-

Examination Category: .MRP-227 D~escrpon PRGAM -0ML~ S - BrN3.X.OREý SUPPORT STRUCTURES amnai Ena ati~nj Exenption, Item No. Parts Examined. ISI Dr ?n'N. Eqimýn Nao-:nero INT.' ase Comments

~Sh1 2' 2 3 ~E ,v lVOSUP. ~ Reus Reactor Vessel Internals ________ ________ ___

B13.70 Core Barrel Assembly Attachment Reactor Vessel Y X x2VT-3 examination. Loss of Core Barrel Flange Figure 4 Core Barrel material (wear).

XK-67866 B13.70 Upper Internals Attachment Reactor Vessel Y X X X X VT-3 examination. Cracking Assembly Upper Figure 4 and Upper Internals (IASCC, Fatigue)

-Support Ring or Skirt Figure 5 Assembly XK-67866 Upper Support Ring or Skirt B13.70 Lower Internals Attachment Reactor Vessel Y X X VTr-3 examination of the lower Assembly Lower Core Figure 4 and Lower Internals core plate to detect evidence of Plate Figure 5 Assembly distortion and/or loss of bolt XK-67866 Lower Core integrity. Cracking (.ASCC, Plate Fatigue)

B13.70 Lower Internals Attachment Reactor Vessel Y X X VT-3 examination. Loss of Assembly Lower Core Figure 4 and Lower Internals material (wear).

Plate Figure 5 Assembly XK-67866 Lower Core XK676 Plate B13.70 Alignment and Attachment Reactor Vessel Y X X X X VTr-3 examination. Loss of Interfacing Figure 2 Upper Internals material (wear).

Components Upper XK-67866 Upper Core Core Plate Alignment Plate Alignment Pins ________ Pins ___j___ ___________________

Category Notes:

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

2. Examinations are performed when the core barrel is removed typically once per interval.

Table 3, Page 1 of 3

Serial No.11-603

. -KEWAUNEE P~OV RSTATION E' - W....ABI.NG H6 6.*

AB !ý, ......

I- P.. =..... . .. .. . . : ..

FOURTH ANDTFIFTH-fINTERVL.IISHDL Examination Category:

MRP-227 tUEEXIq.NG&PROGRAMC4MPONENTS--B WING -N-1. INTERIOR.OF REACTOR VESSEL

'Ex. -a i iom ______

Examination Period Examlnaton.. Exemtilon, Item No. Parts Examined ISI Drawing No. Equipment NO. INT a Comments

-~ 2'or Relief Reactor Vessel Internals B13.10 Alignment and Attachment Reactor Vessel Y X VT-3 examination once per interval Interfacing Figure 4 Interior Surface when the lower internals is Components Clevis removed.

Insert Bolts I I Category Notes:

1. End of Original License is December 21, 2033.

2. The clevis insert bolts are located on the reactor vessel below the lower internals.

3. Per B-N-I, areas to be examined shall include the spaces above and below the reactor core that are made accessible for examination by removal of the components during normal refueling outages. The lower internals is typically removed once per interval.

Table 3, Page 2 of 3

Serial No.11-603

..TABLE .3 KEWAUNEE POWER 'STATION FOURTH AND FIFTH INTERVAL ISI SCHEDULE Examination Category: MRP-227

Description:

TABLE 4-9 WESTINGHOUSE EXISTINGPROGRAMS COMPONENTS - IEB-88-09 -REACTOR VESSEL. BOTTOM MOUNTED INSTRUMENTATION SYSTEW~FLUX-THIMBLE TUBES Examination Ecote-s No N IT ammiation Period MeExehtion Item No. Parts Examined: 15 .. D*awing No. Equipment NOorf. . Code Case, Comments Sc~h 1 2 I3: EO[ Vol Sur Vis.. Request Reactor Vessel Internals IEB 88-09 Reactor Vessel Flux Thimble Y X Eddy Current Examination of the Bottom Mounted Tubes (36) Flux Thimble Tubes Once Every Instrumentation Five Years.

System Flux Thimble Tubes Category Notes:

1. End of Original License is December 21, 2033.

Table 3, Page 3 of 3

Serial No.11-603 Table 4 NUREG-1801, XI.M13, Inspection Plan Cast Austenitic Stainless Steel (CASS) Reactor Vessel Internal Components (2 pages)

KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

Serial No.11-603

..: :*,TABLE 4 (Pa__e,.:1..ý-.

I -EWAUNEE POWER-STA-TION qYURTH,,QOFIFTH INTERVL 151 SCHEDULE 6

Examination Category: NUREG-i 801.ýXLMi 3, _-ýDescription -'AST AUSENITIC STAINLESS ýSTEEL.ITEMS

- ~ <. ' Examiriatidrin. - xiiptdn Item No. Parts Examin~ed 5 Draiývng. K. ~ J od Cse Comments_

IT,-upet -~ ~ -oeief W Cmmnt

~

NO N:1 sch' i 31EOI, ~YI' Sur Ri Reactor Vessel Internals ___________________

BMI Columns BMI Column M-1 199 Y12X x3EVT-1 of Accessible Surfaces.

Assemblies Cruciforms X10193The BMI Column Cruciforms are (160-196 classified as NAM.

Upper Internals Mixing Devices M-1 199 Y2X x3EVT-1 of Accessible Surfaces. The Mixing Devices (34) upper internals mixing devices are classified as NAM.

Upper Internals Supports M-1 199 Y2X x3EVT-1 of Accessible Surfaces.

Instrumentation '19) XK-100-1961 The upper internals instrumentation supports are classified as NAM.

Upper Clamps M-1 199 Y2X x3EVT-1 of Accessible Surfaces.

Internals (28) XK-100-1961 The upper internals Instrumentation Instrumentation clamps are classified as NAM.

Upper Support Bases M-1 199 X x3EVT-1 of Accessible Surfaces.

Column (16) The upper support column Assemblies assemblies' bases are classified as i NAM.

IL-.

Serial No.11-603

- ~ .TABLE4 (Page~o2 KEWAUNEE POWER STATION_

FOURTH INTERVAL ISI SCHEDULE Examination Category: NUREG-1801 XI.M13

Description:

CAST AUSTENITIC STAINLESS STEEL ITEMS

~., ExExminatio Exemption, Item No. Parts Examined

.. :nISI Drawing Equipment No. INT. .- odeI Comment No.~ Sch~ 1 2 3 EO1 Vol Sur Vis orelo

__________Request: ____________

Upper Support Thermocouple M-1199 y2 X x3 EVT-1 of Accessible Surfaces.

Column Stops The upper support column Assemblies (39) assemblies' thermocouple I__stops are classified as NAM.

Category Notes:

1. Examination of BMI column bodies including the CASS-BMI column cruciform's are invoked on an as-needed basis through the corrective action process under MRP-227, as an Expansion Component, when difficulty is detected during flux thimble insertion/withdrawal. Flux thimble insertion/withdrawal to be monitored at each inspection interval.

2. The Upper Core Plate Mixing Devices, Upper Instrumentation Conduit Supports, Upper Instrumentation Clamps, Upper Support Column Bases, Upper Support Thermocouple Stops (at mixing flow Devices), and BMI Column Cruciform's have been classified by Westinghouse Electric Company as NAM. Accessible surfaces of the Mixing Devices, Instrument Columns (Conduit Support and Clamp), and Support Columns Bases, Thermocouple Stops, and BMI Column Cruciform's are currently inspected under ASME Section XI, Category B-N-3, each interval.

3. EVT-1 inspection is not required if screening or evaluation described in GALL NUREG-1801, Rev 1, Chapter XI.M13 is satisfied for fluence < 10E17 n/cm 2 [or <10E20 n/cm 2 if agreement is reached with the NRC], ferrite content (Hull's equivalent factor from NUREG/CR-4513, Rev 1), loading as compressive or less than 5 ksi, or if a component specific evaluation to determine the component's susceptibility to loss of fracture toughness is successful.

4. Target dates for inspection include either KR 32 (Spring 2012), KR 33 (Fall 2013), or KR 34 (Spring 2015).