ML103140599

Site, Units 1, 2 & 3, Proposed License Amendment Request for the Reactor Vessel Internals Inspection Plan Number 2010-06

ML103140599
Person / Time
Site:	Oconee
Issue date:	11/08/2010
From:	Gillespie T Duke Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML103140599 (119)

Text

Duke T. PRESTON GILLESPIE, Jr.

Vice President Energy. Oconee Nuclear Station Duke Energy ON01 VP / 7800 Rochester Hwy.

Seneca, SC 29672 864-873-4478 November 8, 2010 864-873-4208 fax T.Gillespie@duke-energy.com U.S. Nuclear Regulatory Commission.

Attn: Document Control Desk Washington D. C. 20555-0001

Subject:

Duke Energy Carolinas, LLC Oconee Nuclear Site, Units 1, 2, and 3 Docket Numbers 50-269, 50-270, and 50-287 Proposed License Amendment Request for the Reactor Vessel Internals Inspection Plan License Amendment Request Number 2010-06 In accordance with 10 CFR 50.90, Duke Energy Carolinas, LLC (Duke Energy) proposes to amend Renewed Facility Operating Licenses (FOLs) DPR-38, DPR-47, and DPR-55 for Oconee Nuclear Station (ONS), Units 1, 2, and 3. Specifically, Duke Energy requests Nuclear Regulatory Commission (NRC) review and approval for the proposed adoption of the Reactor Vessels (RV) Internals inspection plan based on the use of Materials Reliability Program (MRP) 227, Pressurized Water Reactors Internals Inspection and Evaluation Guidelines.

A letter dated June 16, 2010 was submitted to the NRC stating Duke Energy's intent to adopt MRP-227, Pressurized Water Reactors Internals Inspection and Evaluation Guidelines.

The inspection plan contains a discussion of the background of the Babcock and Wilcox (B&W) designed plant RV Internals programs, first sponsored by the utilities through the B&W Owner's Group (BWOG) and later by the Pressurized Water Reactor Owner's Group (PWROG),

culminating in a submittal to the NRC through the Electric Power Research Institute (EPRI)

MRP. The ONS inspection plan also contains a discussion of operational experience, time-limited aging analyses (TLAAs), and relevant existing programs.

The RV Internals Aging Management Program (AMP) includes the inspection plan and demonstrates that the program adequately manages the effects of aging for RV Internals components. It also establishes the basis for providing reasonable assurance the RV Internals components will remain functional through the license renewal period of extended operation. provides the proposed RV Internals inspection plan. The regulatory commitments associated with the RV Internals inspection plan are provided in Attachment 2.

www.duke-energy.comr

U.S. Nuclear Regulatory Commission November 8, 2010 Page 2 Duke Energy requests approval of this LAR by June 30, 2012. The ONS RV Internals inspection plan will be revised as necessary following approval by the NRC. The ONS UFSAR will be updated per 10 CFR 50.71(e) as required.

In accordance with Duke administrative procedures and the Quality Assurance Program Topical Report, these proposed changes have been reviewed and approved by the Plant Operations Review Committee. Additionally, a copy of this LAR is being sent to the State of South Carolina in accordance with 10 CFR 50.91 requirements.

Inquiries on this proposed amendment request should be directed to Kent Alter of the Oconee Regulatory Compliance Group at (864) 873-3255.

I declare under penalty of perjury that the foregoing is true and correct. Executed on November 8, 2010.

Sincerely.

T. Preston Gillespie, Jr.

Vice President Oconee Nuclear Station

Enclosure:

1. EVALUATION OF PROPOSED CHANGE Attachments:

1. INSPECTION PLAN FOR THE OCONEE NUCLEAR STATION UNITS 1, 2, and 3 REACTOR VESSEL INTERNALS

2. REGULATORY COMMITMENTS

U.S. Nuclear Regulatory Commission November 8, 2010 Page 3 bc w/enclosures and attachments:

Mr. Luis Reyes, Regional Administrator U. S. Nuclear Regulatory Commission - Region II Marquis One Tower 245 Peachtree Center Ave., NE, Suite 1200 Atlanta, Georgia 30303-1257 Mr. John Stang, Project Manager Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop 0-8 G9A Washington, D. C. 20555 Mr. Andy Sabisch Senior Resident Inspector Oconee Nuclear Site Susan E. Jenkins, Manager, Radioactive & Infectious Waste Management SC Dept. of Health and Env. Control 2600 Bull St.

Columbia, SC 29201

ENCLOSUREI EVALUATION OF PROPOSED CHANGE

- Evaluation of Proposed Change License Amendment Request Number 2010-06 November 8, 2010

Subject:

Proposed License Amendment Request for the Reactor Vessel Internals Inspection Plan

1. Summary Description

2. Detailed Description

3. Technical Evaluation

4. Regulatory Safety Analysis 4.1 Significant Hazards Consideration 4.2 Applicable Regulatory Requirements/Criteria 4.3 Precedent

5. Environmental Consideration

6. References

- Evaluation of Proposed Change License Amendment Request 2010-06 November 8, 2010 1.0

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.90, Duke Energy Carolinas, LLC (Duke Energy) proposes to amend Renewed Facility Operating Licenses (FOLs) DPR-38, DPR-47, and DPR-55 for Oconee Nuclear Station (ONS), Units 1, 2, and 3.

The proposed License Amendment Request (LAR) provides the Reactor Vessel (RV) Internals Inspection Plan report. The LAR also provides a description of the inspection plan as it relates to the management of aging effects consistent with previous commitments. The inspection plan is based on MRP-227, Revision 0, "PWR Internals Inspection and Evaluation Guidelines" and describes using the ten Aging Management Program (AMP) elements in the current revision of NUREG-1801 "Generic Aging Lessons Learned" (GALL, Revision 1) report.

The inspection plan contains a discussion of the background of the Babcock and Wilcox (B&W) designed plant RV Internals programs, first sponsored by the utilities through the B&W Owner's Group (BWOG) and later by the Pressurized Water Reactor Owner's Group (PWROG),

culminating in a submittal to the NRC through the Electric Power Research Institute (EPRI)

Materials Reliability Program (MRP). The ONS inspection plan also contains a discussion of operational experience, time-limited aging analyses (TLAAs), and relevant existing programs.

The RV Internals AMP includes the inspection plan and demonstrates that the program adequately manages the effects of aging for RV Internals components and establishes the basis for providing reasonable assurance the RV Internals components will remain functional through the license renewal period of extended operation.

2.0. DETAILED BACKGROUND The RV Internals Inspection Plan, located in Attachment 1, provides the detailed background associated with this LAR.

3.0 TECHNICAL EVALUATION

The RV Internals Inspection Plan, located in Attachment 1, provides the technical evaluation for this LAR.

4.0 REGULATORY SAFETY ANALYSIS 4.1 Significant Hazards Consideration Pursuant to 10 CFR 50.91, Duke has made the determination that this amendment request does not involve a significant hazards consideration by applying the standards established by the NRC regulations in 10 CFR 50.92. This ensures that operation of the facility in accordance with the proposed amendment would not:

1) Involve a significant increase in the probability or consequences of an accident previously evaluated.

No. The proposed license amendment request provides the Reactor Vessel Internals Inspection Plan report. The report also provides a description of the inspection plan as it relates to the management of aging effects consistent with 1

- Evaluation of Proposed Change License Amendment Request 2010-06 November 8, 2010 previous commitments. The inspection plan is based on MRP-227, Revision 0, "Pressurized Water Reactors Internals Inspection and Evaluation Guidelines" and describes using the ten Aging Management Program (AMP) elements in the current revision of NUREG-1 801 "Generic Aging Lessons Learned" (GALL, Revision 1) report.

The inspection plan contains a discussion of the background of the Babcock and Wilcox designed plant Reactor Vessel Internals programs, first sponsored by the utilities through the Babcock and Wilcox Owner's Group and later by the Pressurized Water Reactor Owner's Group, culminating in a submittal to the Nuclear Regulatory Commission through the Electric Power Research Institute Materials Reliability Program. The inspection plan also contains a discussion of operational experience, time-limited aging analyses, and relevant existing programs.

The Reactor Vessel Internals Aging Management Program includes the inspection plan and demonstrates that the program adequately manages the effects of aging for Reactor Vessel Internals components and establishes the basis for providing reasonable assurance the Reactor Vessel Internals components will remain functional through the license renewal period of extended operation.

This license amendment request provides an inspection plan based on industry work and experiences as agreed to in Duke Energy's license renewal commitments for Reactor Vessel Internals Inspection. It is not an accident initiator; therefore, it will not increase the probability or consequences of an accident previously evaluated

2) Create the possibility of a new or different kind of accident from any accident previously evaluated.

No. The proposed Reactor Vessel Internals Inspection Plan does not change the methods governing normal plant operation, nor are the methods utilized to respond to plant transients altered. The revised inspection plan is not an accident / event initiator. No new initiating events or transients result from the use of the Reactor Vessel Internals Inspection plan.

3) Involve a significant reduction in a margin of safety.

No. The proposed safety limits have been preserved. The License Amendment Request requests review and approval for the Reactor Vessel Internals Inspection plan that Duke Energy committed to provide prior to commencing inspections.

4.2 Applicable Regulatory Requirements/Criteria

1. U. S. Nuclear Regulatory Commission, "Safety Evaluation Report Related to the License Renewal of Oconee Nuclear Station, Units 1, 2, and 3," NUREG-1723, March 31, 2000.

2. Letter (from D. Baxter) of Intent to adopt Materials Reliability Program 227, Pressurized Water Reactor Internals Inspection and Evaluation Guidelines, June 16, 2010.

2

- Evaluation of Proposed Change License Amendment Request 2010-06 November 8, 2010

3. UFSAR 18.3.20, Reactor Vessel Internals Inspection 4.3 Precedent

1. Letter from Constellation Energy to Nuclear Regulatory Commission - License Renewal Aging Management Reactor Vessel Internals Program, February 27, 2009.

2. Letter from Progress Energy to Nuclear Regulatory Commission - Reactor Vessel Internals Aging Management Program Inspection Plan, September 24, 2009.

5.0 ENVIRONMENTAL CONSIDERATION

Duke Energy Carolinas, LLC, has evaluated this license amendment request against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21. Duke has determined that this license amendment request meets the criteria for a categorical exclusion set forth in 10 CFR 51.22(c)(9). This determination is based on the fact that this change is being proposed as an amendment to a license issued pursuant to 10 CFR 50 that changes a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or that changes an inspection or a surveillance requirement, and the amendment meets the following specific criteria.

(i) The amendment involves no significant hazards consideration.

As demonstrated in Section 4.1, the proposed Reactor Vessel Internals Inspection plan does not involve significant hazards consideration.

(ii) There is no significant change in the types or significant increase in the amounts of any effluent that may be released offsite.

The proposed Reactor Vessel Internals Inspection plan will not impact effluents released offsite. Therefore, there will be no significant change in the types or significant increase in the amounts of any effluents released offsite.

(iii) There is no significant increase in individual or cumulative occupational radiation exposure.

The proposed Reactor Vessel Internals Inspection plan will not have an adverse impact on occupational radiation exposure. Therefore, there will be no significant increase in individual or cumulative occupational radiation exposure resulting from this action.

6.0 REFERENCES

The RV Internals Inspection Plan, located in Attachment 1, provides the references associated with this LAR.

3

ATTACHMENT 1 INSPECTION PLAN FOR THE OCONEE NUCLEAR STATION UNITS 1, 2, and 3 REACTOR VESSEL INTERNALS

.o 011 ANP-2951 Revision 001 October 2010 Inspection Plan for the Oconee Nuclear Station Units 1, 2, and 3 Reactor Vessel Internals by Sarah Davidsaver, AREVA NP, Inc.

and Rachel Doss, Duke Energy Carolinas, LLC David Whitaker, Duke Energy Carolinas, LLC Dorrington Williams, Duke Energy Carolinas, LLC Prepared for Duke Energy Carolinas, LLC Oconee Nuclear Station AREVA NP, Inc.

3315 Old Forest Road P.O. Box 10935 Lynchburg, Virginia 24506-0935 Page 1 of 108

ANP-2951, Rev. 001 Copyright © 2010 AREVA NP Inc.

All Rights Reserved Page 2 of 108

ANP-2951, Rev. 001 Nature of Changes Revision Date Changes 000 August 2010 Initial Release 001 October 2010 Complete Rewrite

-I- 4

+ +

Page 3 of 108

ANP-2951, Rev. 001 Table of Contents Page NATU R E O F CHANGES ............................................................................................................................ 3 LIST OF ACRONYMS AND ABBREVIATIONS ................................................................................... 7 1.0 INTRODUC TION ............................................................................................................................ 9 2 .0 BA CKGROUND .................................... ....................................................................................... 10 2.1 ONS License Renewal- Background ............................................................................ 10 2.2 ONS RV Internals Aging Management Review/Industry Program Background .......... 12 2.3 O NS RV Internals AM P Intent ...................................................................................... 13 2.4 ONS RV Internals Background .................................................................................... 13 2.5 RV Internals Inspection Commitment Change Letter ................................................... 15 3.0 PR O G RA M OW NER .................................................................................................................... 16 4.0 INDUSTRY AND ONS PROGRAMS AND ACTIVITIES .......................................................... 17 4.1 Industry Programs and Activities ..................................... 17 4.1.1 EPRI PWR MRP Activities .......................................................................... 17 4.1.2 PW R OG Activities ........................................................................... ................ 20 4.1.3 Time-Limited Aging Analyses ..................................... ................................... 20 4.1.4 Internals Bolting Surveillance Program ..................................................... 21 4.1.5 Joint Owners' Baffle Bolt Program ............................................................. 21 4.1.6 Fuel/Baffle Interaction Investigation .......................................................... 22 4.2 ONS Programs and Activities ....................................... 22 4.2.1 ASME B&PV Code Section XI In-Service Inspection Requirements ...... 22 4.2.2 Primary Water Chemistry Program ............................................................. 24 4.2.3 Vent Valve In-Service Test Program ......................................................... 24 4.2.4 Continuation of Use of Low-Leakage Cores ............................................... 24 4.2.5 Lower Thermal Shield Replacement Bolt ................................................... 24 4.2.6 Fabrication Records Search ..................................................................... 24 4.2.7 Volumetric (UT) Examinations of Upper Core Barrel Bolts ........................ 26 4.2.8 Core Clamping Measurements ................................................................... 26 4.2.9 Visual Examination of Baffle-to-Baffle and Baffle-to-Former Bolts ............ 27 4.2.10 ONS Unit-Specific Amendments to MRP-227, Rev. 0 Requirements ........ 27 4.3 Conclusions of Section 4.0 ........................................................................................... 27 5.0 ONS RV INTERNALS AMP ATTRIBUTE EVALUATION ....................................................... 29 5.1 AMP Element 1 - Scope of Program .......................................................................... 29 5.1.1 ONS Scope ........................................... 29 5.1.2 ONS RV Internals Components Subject to an AMR .................................. 29 5.1.3 C onclusion ................................................................................................... 29 5.2 AMP Element 2 - Preventative Actions ...................................................................... 29 5.2.1 Primary Water Chemistry Control Program .............................................. 30 5.2.2 Low-Leakage Cores ................................................................................. 30 5.2.3 C onclusion ................................................................................................... 30 Page 4 of 108

ANP-2951, Rev. 001 Table of Contents (Continued) 5.3 AMP Element 3 - Parameters Monitored or Inspected ................... .......................... 30 5.3.1 The ONS In-Service Inspection Program ....................................................... 30 5.3.2- MRP-227, Rev. 0 "Primary" and "Expansion" (Augmented) Inspections ....... 31 5.3.3 C onclusion .................................................................................................. 33 5.4 AMP Element 4 - Detection of Aging Effects ............................................................... 33 5.4.1 One-Time Physical Measurement .............................................................. 33

.5.4.2 ASME B&PV Code Section Xl Examination Category B-N-3 VT-3 Exam inations .............................................................................................. 33 5.4.3 Augmented VT-3 Examination to Detect Specific Aging Effects ................ 34 5.4.4 Augmented UT Examinations to Detect Cracking of Bolting ...................... 34 5.4.5 Justification by Evaluation for Inaccessible Bolts ....................................... 34 5.4.6 Conclusion .................................................................................................. 34 5.5 AMP Element 5 - Monitoring and Trending ................................................................. 34 5.5.1 Conclusion .................................................................................................. 35 5.6 AMP Element 6 -Acceptance Criteria ........................................................................ 35 5.6.1 Examination Acceptance Criteria ............................................................... 35 5.6.2 Functionality/Engineering Acceptance Criteria .......................................... 35 5.6.3 Conclusions ................................................................................................ 36 5.7 AMP Element 7 - Corrective Actions .......................................................................... 36 5.7.1 ONS PIP Program ............................................ 36 5.7.2 ONS Root Cause ........................................................................................ 36 5.7.3 Operability .................................................................................................. 37 5.7.4 Conclusion ................................................................................................. 37 5.8 AMP Element 8 - Confirmation Process ...................................................................... 37 5.8.1 Conclusion .................................................................................................. 37 5.9 AMP Element 9 - Administrative Controls ................................................................... 37 5.9.1 - Conclusion .................................................................................................. 37 5.10 AMP Element 10- Operating Experience ................................................................... 37 5.10.1 Incidents of Degradation in B&W RV Internals .......................................... 37 5.10.2 Conclusion ................................................................................................... 38 5.11 Program Conclusion .................................................................................................... 38 6.0

SUMMARY

AND CONCLUSIONS ........................................................................................... 39

7.0 REFERENCES

............................................................................................................................ 40 APPENDIX A: BAW-2248A AMR AND INDUSTRY PROGRAM COMPARISON .............. 42 APPENDIX B: TABLE 4-1 FROM MRP-227, REV. 0 ......................................................................... 58 APPENDIX C: TABLE 4-4 FROM MRP-227, REV. 0 ........................................................................ 63 APPENDIX D: TABLE 5-1 FROM MRP-227, REV. 0 ........................................................................ 66 APPENDIX E: NON-PROPRIETARY UCB AND LCB BOLT OCONEE UNIT-SPECIFIC TECHNICAL JU ST IFICAT ION .......... *................................................................................................................ 73 APPENDIX F: ONS UNIT-SPECIFIC AMENDMENTS TO MRP-227, REV. 0 ................................. 91 Page 5 of 108

ANP-2951 Rev. 001 Table of Contents (Continued)

Page 6 of 108

ANP-2951, Rev. 001 List of Acronyms and Abbreviations AMP -Aging Management Program AMR - Aging Management Review ASME - American Society of Mechanical Engineers B&PV - Boiler and Pressure Vessel B&W - Babcock & Wilcox B&WOG - B&W Owners Group CASS - Cast Austenitic Stainless Steel CFR - Code of Federal Regulations CLB - Current Licensing Basis CRGT - Control Rod Guide Tube CR Crystal River Unit 3 CSA - Core Support Assembly CSS - Core Support Shield Duke Energy - Duke Energy Carolinas, LLC EdF - Electricit6 de France EPRI - Electric Power Research Institute FD - Flow Distributor FMECA - Failure Modes, Effects, and Criticality Analysis FSER - Final Safety Evaluation Report GALL - Generic Aging Lessons Learned GLRP - Generic License Renewal Program HTH - High Temperature Heat-Treatment (Alloy X-750)

IBSP- Internals Bolting Surveillance Program I&E Guidelines - Inspection and Evaluation Guidelines (MRP-227, Rev: 0)

IASCC - Irradiation-Assisted Stress Corrosion Cracking IE - Irradiation Embrittlement IGSCC - Intergranular Stress Corrosion Cracking IMI - Incore Monitoring Instrumentation INOS - Independent Nuclear Oversight ISI - In-Service Inspection ITG - Issue Task Group (EPRI)

JOBB - Joint Owners' Baffle Bolt (Program)

LCB - Lower Core Barrel LOCA - Loss of Coolant Accident LRA - License Renewal Application Page 7 of 108

r ANP-2951, Rev. 001 List of Acronyms and Abbreviations

'(Continued)

LTS - Lower Thermal Shield MRP - Materials Reliability Program NDE - Non-Destructive Examination NRC - U.S. Nuclear Regulatory Commission OEP - Operating Experience Program ONS - Oconee Nuclear Station ONS Oconee Nuclear Station Unit 1 ONS Oconee Nuclear Station Unit 2 ONS Oconee Nuclear Station Unit 3 PIP - Problem Investigation Process PWR - Pressurized Water Reactor PWROG - Pressurized Water Reactor Owners Group QA - Quality Assurance RI-FG - Reactor Internals-Focus Group RFO - Refueling Outage RV - Reactor Vessel SCC - Stress Corrosion Cracking SER - Safety Evaluation Report S/N - Serial Number SSC - Structures, Systems, and Component SSHT - Surveillance Specimen Holder Tube TLAA - Time-Limited Aging Analysis TE - Thermal Embrittlement TJ - Technical Justification UCB - Upper Core Barrel UFSAR - Updated Final Safety Analysis Report U.S. - United States UT - Ultrasonic Testing (Nondestructive Examination Technique)

UTS - Upper Thermal Shield VT Visual Examination Page 8 of 108

ANP-2951, Rev. 001

1.0 INTRODUCTION

The purpose of this report is to document the Oconee Nuclear Station (ONS) Units 1, 2, and 3 (ONS-1, ONS-2, and ONS-3) Reactor Vessel (RV) Internals inspection plan for submittal to the United States (U.S.) Nuclear Regulatory Commission (NRC). This report provides a description of the ONS RV Internals inspection plan as it relates to the management of aging effects consistent with previous commitments. The ONS RV Internals inspection plan is based on "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227, Rev. 0)"'ul and described using the ten Aging Management Program 2 (AMP) elements in the current revision of NUREG-1801 "Generic Aging Lessons Learned" (GALL) report.i )

This ONS RV Internals inspection plan contains a discussion of the background of the Babcock and Wilcox (B&W)-designed plant RV Internals programs, first sponsored by the utilities through the B&W Owner's Group (B&WOG) and later through the PWR Owner's Group (PWROG), culminating in a submittal to the NRC through the Electric Power Research Institute (EPRI) Pressurized Water Reactor (PWR) Materials Reliability Program (MRP). The ONS RV Internals inspection-plan also contains a discussion of operational experience, time-limited aging analyses (TLAAs), and relevant existing ONS programs.

The ONS RV Internals AMP will include this ONS RV Internals inspection plan and will demonstrate that the program adequately manages the effects of aging for RV Internals components and establish the basis for providing reasonable assurance the RV Internals components will remain functional through the ONS license renewal period of extended operation.

Page 9 of 108

ANP-2951, Rev. 001

2.0 BACKGROUND

2.1 ONS License Renewal Background By letter dated July 6, 1998, Duke Energy Carolinas, LLC (Duke Energy hereafter) submitted the License Renewal Application (LRA) for ONS in accordance with Title 10, Part 54, of the Code of Federal Regulations (10 CFR 54).' Through the LRA, Duke Energy requested the NRC to renew the operating license for ONS-l (license number DPR-38), ONS-2 (license number DPR-47), and ONS-3 (license number DPR-55) for a period of 20 years beyond the original expiration of midnight February 6, 2013 (ONS-1), midnight October 6, 2013 (ONS-2),

and midnight July 19, 2014 (ONS-3). The renewed license was issued by the NRC on May 23, 2000.141 The safety evaluation report (SER) NUREG-17231 5' documented the technical review of the ONS-1, ONS-2, and ONS-3 LRA by the NRC Staff.

The Renewed Facility Operating License Numbers DPR-38, DPR-47, and DPR-55 for the oNS-1, ONS-2, and ONS-3 plants were granted, as documented in NRC letter of April 10, 2000[6] which identifies the technical basis for issuing the renewed license as being set forth in NUREG-1723.

Section 4.3.11 of the LRA(33 discusses the ONS RV Internals AMP for license renewal. Per the LRA, the proposed ONS RV Internals inspection plan includes the following activities:

a) Continue the characterization of the potential aging effects that have been identified in BAW-2248[71 ,

Demonstrationof the Management ofAging Effects for the Reactor Vessel Internals.The scope of the characterization includes, but is not limited to, the development of key program elements to address the following aging effects: cracking, reduction of fracture toughness, and loss of closure integrity.

b) After the characterization of the potential aging effects and prior to February 6, 2013, Duke Energy will develop an appropriate monitoring and inspection program, with attributes as defined in Section 4.2 [of the LRAE31]. This monitoring and inspection program will provide additional assurance that the RV Internals will remain functional through the period of extended operation.

Since the submittal of BAW-2248, the B&WOG (now incorporated into the PWROG) has periodically met with the NRC to discuss RV Internals aging management issues. The Joint Owners' Baffle Bolt (JOBB) program (discussed further in Section 4.1.5 of this report) was completed under the direction of the EPRI PWR MRP. In addition, the EPRI PWR MRP has taken on the industry initiative to provide inspection and evaluation (I&E) guidelines for PWR RV Internals. EPRI PWR MRP meets periodically with the NRC to provide updates. These meetings between the industry and the NRC comply with Duke Energy's commitment in the safety evaluation of BAW-2248A, as repeated in NUREG-1 723 (Section 3.4.3, Action Item 4 under Action Items from Previous Staff Evaluation of BAW-2248), to participate in industry RV Internals AMP and provide updates to the NRC on a periodic basis after completion of significant milestones commencing within one year of the issuance of the renewed license.

Section 4.2.5 of NUJREG-1723 identifies that the TLAAs from the LRA were reviewed. NUREG-1723 concludes the LRA identified and evaluated the TLAAs associated with RV Internals for ONS-I, ONS-2, and ONS-3 consistent with the requirements of 10 CFR 54.21. See Section 4.1.3 of this report for a further discussion of TLAAs.

Section 6 of NUREG-1723 concludes that, based on the evaluation of the application as discussed in NUREG-1723, the staff determined the requirements of 10 CFR 54.29 were met by the ONS-1, ONS-2, and ONS-3 application. The staff found reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the current licensing basis (CLB) for ONS-1, ONS-2, and ONS-3.

Table 2-1 summarizes the ONS RV Internals LRA commitments and their resolutions.

Page 10 of 108

ANP-2951, Rev. 001 Table 2-1. ONS RV Internals LRA Commitment Resolutions

- Commitment Commitment/Action Items Reference section describing the fulfillment reference section of the commitment NUREG-172315', Section Action Items 1-3, 5-10 from the NRC Staff NUREG-1723, Section 3.4.3 (pages 3-97 through 3-100) 3.4.3 (pages 3-97 evaluation of BAW-2248 contains Duke Energy's responses to Renewal Applicant through 3-100) Items and resolves the action items and states the action items have been resolved NUREG-1723I51, Section Action Item 4 from Staff Evaluation of NUREG-1723, Section 3.4.3 (page 3-98) contains Duke 3.4.3 (page 3-98) BAW-2248 states "The applicant must Energy's commitment commit to participation in the B&WOG RVIAMP, and any other industry programs Section 2.1 of this report describes how the commitment as appropriate, to continue the investigation was fulfilled of potential aging effects for RVI components, and to establish monitoring and inspection programs for RVI components.

The applicant shall provide the NRC with either annual reports or periodic updates (after completion of significant milestones) on the status of the RVIAMP, commencing within one year of the issuance of the renewal license."

NUREG-1723ts1, Section Combination of periodic in-service See Section 4.2.1 of this report for a discussion of the 3.4.3.3 (page 3-114) inspection required by ASME Section XI, ONS In-Service Inspection (ISI) program Subsection IWB and a flaw evaluation procedure specified in 1WB-3640 for Per an assessment of thermal aging and neutron "CASS Flaw Evaluation Procedure". embrittlement of cast austenitic stainless steel (CASS),

CASS items in the B&W designed PWR internals are redundant and/or potentially able to be analyzed for functionality in the anticipated degraded conditions.

Replacement of the degraded item o'r component is also a potential option. Thus, no fracture toughness properties would be required for fracture mechanics analyses.

NUREG-1723t1t, Section Commitment to manage the aging of RV. (1) See Section 4.2.1 of this report for a discussion of 3.4.3.3 (pages 3-120 Internals with the ONS ISI program through 3-122) (1) In-Service Inspection Plan and (2) This commitment is fulfilled by this report (2) Oconee RV Internals Inspection (3) See Section 4.1.5 of this report (3) The final report will contain the test -

results from the RVIAMP and the recommended inspection program for the RV Internals.

NUREG-1723r5 I, Section 1. Flow-induced vibration endurance limit The BAW-2248 evaluation of this TLAA was found to 4.2.5 (pages 4-23 assumptions be acceptable by NUREG-172351 t

, Section 4.2.5 (pages through 4-25) 4-23 through 4-24). No additional action by Duke Energy is required.

2. Transient cycle count assumptions for the The BAW-2248 evaluation of this TLAA was found to replacement bolting (Action Item II from be acceptable by NUREG-1723S1", Section 4.2.5 [pages NRC Staff evaluation of BAW-2248) 4-23 through 4-25). Duke Energy will continue to monitor and track occurrences of design transients.

3. Reduction in fracture toughness (Action An analysis has been performed for the ONS units for Item 12 from NRC Staff evaluation of this TLAA (see Section 4.1.3 of this report). The BAW-2248) analysis will be provided to the NRC to demonstrate the completion of this TLAA for the ONS units.

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ANP-2951, Rev. 001 Commitment Commitment/Action Items Reference section describing the fulfillment reference section of the commitment

4. Flaw growth acceptance Duke Energy's response for this TLAA was found to be acceptable by NUREG-1723'51 , Section 4.2.5 (pages 4-24 through 4-25). No additional action by Duke Energy is required.

ONS LRA1 31, Section Identification of activities which may be See discussion under (a) and (b) in Section 2.1 of this 4.3.11 (page 4.3-28) included in the ONS RV Internals AMP: report (a) Continued characterization of the potential aging effects identified in BAW-2248 and (b) Develop an appropriate monitoring and inspection program 2.2 ONS RV Internals Aging Management Review/Industry Program Background The ONS LRA was submitted in 1998 and the SER was granted in 2000; these license renewal documents predate NUREG-1 801. However, this ONS RV Internals inspection plan is defined using the ten AMP elements identified in the GALL report published in 2005.12]

The initial work performed, which supports the ONS RV Internals inspection plan, included an aging management review (AMR) documented in BAW-224871 1that was directed by the B&WOG Generic License Renewal Program (GLRP). The NRC final safety evaluation report (FSER) of BAW-2248 was attached to the NRC's letter to the B&WOG dated December 9, 1999.*1] The NRC's letter and FSER are included in the updated BAW-2248A report.1 91 The NRC identified 12 action items in the FSER to be addressed in the plant-specific LRA when incorporating BAW-2248A in a renewal application. Upon resolution of these action items, Duke Energy may rely on BAW-2248A to demonstrate there is reasonable assurance the ONS RV Internals components will perform their intended functions in accordance with the CLB.

As presented in BAW-2248A, Table 4-1, a combination of existing programs and additional work, to be identified by the "RV Internals Aging Management Program" was credited for aging management of the B&W operating plant RV Internals, including the ONS RV Internals.

An AMR was performed for the ONS units in 2001 in accordance with 10 CFR 54(a)(3).["°] The methodology Duke Energy used to ensure the ONS RV Internals components are bounded by BAW-2248A included three steps: 1) Comparison of RV Internals intended functions, 2) Comparison of RV Internals items subject to AMR, and 3) Review of ONS-specific operating history to ensure the aging effects identified in the generic report are applicable to the ONS RV Internals. This AMR is an ONS-specific application of the B&W generic AMR performed in BAW-2248A.

The additional industry work on the aging of the RV Internals, begun by the submittal of BAW-2248, culminated in the submittal of MRP-227, Rev. 0.J1 MRP-227, Rev. 0 was submitted in January of 2009 to the NRC for review and SER approval." "] Components requiring inspections are categorized as "Primary", "Expansion", or "Existing Programs". Components not requiring augmented inspections are categorized as "No Additional Measures". The industry program is intended to provide a consistent approach to the aging management of PWR RV Internals components across the PWR fleet. For additional information about MRP-227, Rev. 0 see Section 4.1.1.1 of this report.

Table A. 1 in Appendix A shows how the components identified in the BAW-2248A, Table 4-1 AMR were evaluated and characterized by the industry program. Justifications for ONS unit-specific amendments to MRP-227, Rev. 0 are found in Appendix F of this report.

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ANP-2951, Rev. 001 2.3 ONS RV Internals AMP Intent The ONS RV Internals AMP, which will include the ONS RV Internals inspection plan described in this report after it is approved by the NRC, utilizes a combination of prevention, mitigation, and condition monitoring.

Where applicable, credit is taken for existing programs (e.g., primary water chemistry and American Society of Mechanical Engineers [ASME] Boiler & Pressure Vessel [B&PV] Section XI inspections) and mitigation projects such as lower thermal shield (LTS) bolt replacement. The ONS RV Internals inspection plan then incorporates recommendations for augmented inspections provided by industry guidelines in MRP-227, Rev. 0, as modified by amendments to MRP-227, Rev. 0 (see Appendix F of this report). Augmented inspections are in addition to the requirements of ASME B&PV Code Section XI1"'2 ; the I&E guidelines do not reduce, alter, or otherwise affect current ASME B&PV Code Section XI inspections.

Aging degradation mechanisms that impact the RV Internals have been identified in MRP-227, Rev. 0. The overall outcome of the additional work performed by the Industry summarized in MRP-227, Rev. 0 is to ensure functionality of the RV Internals is maintained by detection of the effect of the degradation mechanism listed in Table 2-2. Therefore, this ONS RV Internals inspection plan is consistent with the industry work provided in MRP-227, Rev. 0 as modified by amendments to MRP-227, Rev. 0 (see Appendix F of this report).

Table 2-2. RV Internals Aging Degradation Mechanisms and Their Aging Effects Aging Degradation Mechanism Aging Effect Stress Corrosion Cracking (SCC) Cracking Irradiation=Assisted Stress Corrosion Cracking (IASCC) Cracking Wear Loss of Material Fatigue Cracking Thermal Aging Ernbrittlement Loss of Ductility and Unstable Crack Extension Irradiation Embrittlement (EE) Loss of Ductility and Unstable Crack Extension Void Swelling and Irradiation Growth Dimension Change, Distortion, and Cracking Thermal and Irradiation-Enhanced Stress Relaxation or Loss of Mechanical Closure Integrity leading to Cracking Irradiation-Enhanced Creep Ik Section 5.0 of this report uses the ten AMP elements ofNUREG-1801, Rev. 1 to describe the ONS RV Internals inspection plan and AMP as required by the NRC. The ONS RV Internals AMP, which will also include this ONS RV Internals inspection plan after it is approved by the NRC, incorporates programs and activities that are credited for managing the aging effects produced by the aging degradation mechanisms listed in Table 2-2. ONS RV Internals components within the scope of BAW-2248A, the LRA, and NUREG-1723 have been considered in this ONS RV Internals inspection plan.

Table A. 1 in Appendix A of this report shows how the components identified in the BAW-2248A, Table 4-1 AMR were evaluated and characterized by the industry program.

2.4 ONS RV Internals Background The intended functions of the ONS RV Internals are as follows[5 '91 :

• Support and orient the reactor core

" Support, orient, guide, and protect the control rod assemblies

• Provide a passageway to distribute the reactor coolant flow to the reactor core

• Provide a passageway to support, guide, and protect incore instrumentation

" Provide a secondary core support to limit downward displacement of core support structure

• Provide gamma and neutron shielding Page 13of 108

ANP-2951, Rev. 001 The ONS RV Internals consists of two structural subassemblies that are located within the RV: the plenum assembly and the core support assembly. Duke Energy has reviewed the design and operation of the ONS RV Internals using the process described in Section 2.4 of the ONS LRAp] and determined they are bounded by the description contained in BAW-2248A, with the exception of the thermal shield and thermal shield upper restraint.

Note that the thermal shield and thermal shield upper restraint were omitted from BAW-2248A; however these items support an ONS RV Internals intended function and were found to be subject to an AMR. The thermal shield surrounds the core barrel and is constructed of austenitic stainless steel. The thermal shield upper restraint is also constructed of austenitic stainless steel. These items were included&in the ONS AMRt 01" and the industry work which culminated in MRP-227, Rev. 0. The general arrangement of the ONS RV Internals is shown in Figure 2-1 .1 Figure 2-1. ONS RV Internals Plenum Cover Plenum i Assembly Assembly Vent Valve o0 E 7 Core Support Shield U')

t t J Plenum Cylinder

. .. . Assembly Upper Grid Assembly

< E L.-

W Tn 0

-Thermal Shield o Core Barrel U

Lower Grid Assembly 0

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ANP-2951, Rev. 001 2.5 RV Internals Inspection Commitment Change Letter By letter dated June 16, 2010 to the NRC")31, Duke Energy stated its intent to revise the existing license renewal commitment to inspect the RV Internals at each Duke Energy nuclear station, including the three ONS units. The existing inspection commitments are contained in Section 18.3.20 of the ONS Updated Final Safety Analysis Report (UFSAR). The UFSAR section contains an allowance that permits Duke Energy to modify or eliminate these inspections based on industry data or other evaluations if plant-specific justification is provided to demonstrate the basis for the modification or elimination.

Duke Energy is revising its commitments for RV Internals inspections from those that currently exist in the ONS UFSAR to the inspection guidelines provided in MRP-227, Rev. 0, as approved by the NRC. This ONS RV Internals inspection plan, as documented herein, will follow MRP-227, Rev. 0. After the anticipated release of MRP-227-A, Duke Energy will review and, if needed, revise this ONS RV Internals inspection plan.

The ONS RV Internals inspection plan will also be revised as necessary following review by the NRC. Once the ONS RV Internals inspection plan is approved, the ONS UFSAR will be updated as required.

Note: This ONS RV Internals inspection plan contains several amendments to MRP-227, Rev. 0 which are discussed in Section 4.2.10 and Appendix F of this report.

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ANP-2951, Rev. 001 3.0 PROGRAM OWNER The Oconee Reactor and Electrical Systems (Reactor Team) and the General Office Nuclear Technical Services are responsible for maintaining and implementing the ONS RV Internals inspection plan.

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ANP-2951, Rev. 001 4.0 INDUSTRY AND ONS PROGRAMS AND ACTIVITIES The ONS RV Internals inspection plan is based on the ONS LRA, NUREG-1723, and evaluations supporting MRP-227, Rev. 0. The ONS RV Internals AMP is demonstrated by implementation of MRP-227, Rev. 0 methodology and the continuation of the existing programs discussed in this section.

4.1 Industry Programs and Activities There are various industry programs and activities in which Duke Energy has been or is participating that support the aging management of the PWR RV-Internals; those discussed in this section include EPRI PWR MRP activities, PWROG activities, TLAAs, the internals bolting surveillance program (IBSP), the JOBB program, and the fuel/baffle interaction investigation for the B&W-designed units. Duke Energy will continue to participate in industry activities addressing PWR RV Internals.

4.1.1 EPRI PWR MRP Activities As part of the ONS License Renewal Program, Duke Energy made the commitment to participate in industry activities associated with the development of the standard industry guidance, which includes the EPRI PWR MRP activities which produced the guidelines and standards discussed below.

4.1.1.1 MRP-227, Rev. 0 The EPRI PWR MRP efforts have defined the required inspections and examination techniques for the RV Internals. The results of the industry recommended inspections, published in MRP-227, Rev. 0, serve as the basis for identifying any augmented inspections that are required to complete this ONS RV Internals inspection plan.

4.1.1.1.1 Development of MRP-227, Rev. 0 The MRP-227, Rev. 0 "Pressurized Water Reactor Internals Inspection and Evaluation Guidelines"'11 were developed by a team of industry representatives who reviewed available data and industry experience to identify and prioritize I&E requirements for RV Internals. MRP-227, Rev. 0 is the culmination of the industry work that began with BAW-2248A for B&W plants. The key sequential steps in the process included the following:

" The development of screening criteria, with susceptibility levels for the eight postulated aging degradation mechanisms relevant to reactor internals and their effects;

" An initial component screening and categorization, using the susceptibility levels and FMECA (failure modes, effects, and criticality assessment) to identify the relative ranking of the components;

" Functionality assessment of degradation for components and assemblies of components; and

• Aging management strategy development combining results of the functionality assessment with component accessibility, operating experience (OE), existing evaluations, and prior examination results to determine the appropriate aging management methodology, baseline examination timing, and the need for and the timing of subsequent inspections.

Through this process, the RV Internals for all three PWR designs in the U.S. were evaluated, and appropriate recommendations for aging management actions specific to each component were provided.

MRP-227, Rev. 0 utilized the screening and ranking process to aid in the identification of required inspections for "Primary" and "Expansion" components and credits existing component inspections when they were deemed adequate.

The basic description of each classification is as follows:

"Primary" Those PWR internals that are highly susceptible to the effects of at least one of the eight aging mechanisms were placed in the "Primary" group. The aging management requirements that are needed to ensure functionality of "Primary" components are described in these I&E guidelines. The "Primary" group also Page 17 of 108

ANP-2951, Rev. 001 includes components which have shown a degree of tolerance to a specific aging degradation effect, but for which no highly susceptible component exists or for which no highly susceptible component is accessible.

• "Expansion" Those PWR 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, were placed in the "Expansion" group. The schedule for implementation of aging management requirements for "Expansion" components will depend on the findings from the examinations of the "Primary" components at individual plants.

• "Existing Programs" Those PWR internals that are susceptible to the effects of at least one of the eight aging mechanisms and for which generic and plant-specific existing AMP elements are capable of managing those effects, were placed in the "Existing Programs" group.

Note there are no "Existing Programs" components in MRP-227, Rev. 0 for the B&W-designed PWRs.

" "No Additional Measures" Those PWR internals for which the effects of all eight aging mechanisms are below the screening criteria were placed in the "No Additional Measures" group. Additional components were placed in the "No Additional Measures" group as a result of FMECA and the functionality assessment. No further action is required by these guidelines for managing the aging of the "No Additional Measures" components.

The categorization and analysis processes used in the MRP-227, Rev. 0 approach are not intended to supersede any ASME B&PV Code Section XI requirements. Any components that are classified as removable core support structures, as defined in ASME B&PV Code Section XI IWB-2500, Examination Category B-N-3, have requirements that remain in effect and may only be altered as allowed by 10 CFR 50.55a.

The requirements of MRP-227, Rev. 0 are classified in accordance with the requirements of NEI 03-08 Guidelines. 1 41For the MRP-227, Rev. 0 guidelines there are one "Mandatory", three "Needed", and one "Good Practice" elements as follows:

"Mandatory" Each commercial U. S. PWR unit shall develop and document a PWR reactor internals aging management program (AMP) within thirty-six months following issuance of MRP-227-Rev. 0.

MRP-227, Rev. 0 was issued in December 2008 and submitted to the NRC in January 2009. Duke Energy will fulfill this "Mandatory" element by developing the ONS RV Internals AMP by the end of December 2011.

" "Needed" Each commercial U. S. PWR unit shall implement Tables 4-1 through 4-9 and Tables 5-1 through 5-3 for the applicable design within twenty-four months following issuance of MRP-227-A.

The applicable B&W tables contained in MIRP-227, Rev. 0, Table 4-1 ("Primary"), Table 4-4 ("Expansion"),

and Table 5-1 (Examination Acceptance and Expansion Criteria) are attached herein as Appendices B, C, and D. There are no "Existing Program" components in MRP-227, Rev. 0 for the B&W-designed PWRs. The ONS units have followed the MRP-227, Rev. 0 recommended inspections by the inspection activities already performed or planned as described in this ONS RV Internals inspection plan. The justification for ONS unit-specific amendments to MRP-227 Rev. 0 are discussed in Section 4.2.10 and Appendix F of this report. MRP-227, Rev. 0 has been submitted to the NRC with the ultimate goal of obtaining approval and issuance of an SER. After the release of MRP-227-A, Duke Energy will review and, if needed, revise this ONS RV Internals inspection plan. Therefore, implementation of this ONS RV Internals inspection plan will fulfill this "Needed" requirement for the three ONS units.

• "Needed" Examinations specified in these guidelines shall be conducted in accordance with the Inspection Standard MRP-228.

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ANP-2951, Rev. 001 Inspection standards developed under MRP-228""sl will be used by Duke Energy for the augmented inspections described in this ONS RV Internals inspection plan developed in accordance with MRP-227, Rev.

0 for the three ONS units. Implementation of this ONS RV Internals inspection plan will fulfill this "Needed" requirement for the three ONS units.

"Needed" Examination results that do not meet the examination acceptance criteria defined in Section 5 of the MRP-227 guidelines shall be recorded and entered in the plant corrective action program and dispositioned.

The ONS Corrective Action Program (PIP) will be applied as discussed in Section 5.7 of this report.

Implementation of this ONS RV Internals inspection plan will fulfill this "Needed" requirement for the three ONS units.

"Good Practice" Each commercial U. S. PWR unit should provide a summary report of all inspections and monitoring, items requiring evaluation, and new repairs to the MRP Program Manager within 120 days of the completion of an outage during which PWR internals are examined. The MRP template should be used for this report.

Duke Energy will make the best effort in providing summary reports to MRP of future ONS' inspection activities within 120 days of the completion of an outage during which PWR RV Internals are examined.

Implementation of this ONS RV Internals inspection plan will fulfill this "Good Practice" requirement for the three ONS units.

4.1.1.1.2 MRP-227, Rev. 0 Applicability to ONS The MRP-227, Rev. 0 guidelines are based on several general assumptions that were used for the analysis in the development of MRP-227, Rev. 0. These assumptions and their applicability to ONS units are listed below:

0 30 years or less 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.

The fuel management program for the three ONS units changed from a high to a low-leakage core loading pattern prior to 30 years of operation.

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

The three ONS'units operate as a base load unit.

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

MRP-227, Rev. 0 states that the recommendations are applicable to all operating U.S. PWR operating plants as of May 2007 for the three designs (i.e., B&W, and Westinghouse, CE) identified. No modifications have been made to the ONS RV Internals since May 2007. Fabrication records searches have been conducted for the ONS units as described in Section 4.2.6 of this report to verify the applicability of MRP-227, Rev. 0 recommendations to the ONS units. The justifications for ONS unit-specific amendments to MRP-227, Rev. 0 from the fabrication records search results are described in Section 4.2.10 of this report.

Based on the above review, MRP-227, Rev. 0 is applicable to all three ONS units, except for the amendments identified in Section 4.2.10 and Appendix F of this report.

4.1.1.2 MRP-228, Final Report "Materials Reliability Program: Inspection Standard for PWR Internals (MRP-228)" 51 was developed by the EPRI PWR MRP Inspection Issue Task Group (ITG) in cooperation with the reactor internals focus group (RI-FG). These inspection standards are intended to support MRP-227, Rev. 0 to detect the effects of aging Page 19 of 108'

ANP-2951, Rev. 001 degradation mechanisms. This report provides the PWR fleet with inspection procedure requirements for the RV Internals "Primary" and "Expansion" components included in MRP-227, Rev. 0 and offers a stable mechanism for documenting the capability of the evolving inspection technology.

MRP-228, Rev. 0 contains four "Needed" and two "Good Practice" requirements, which will be followed in accordance with the requirements of the NEI 03-08 Guidelines.1 1 41 Duke Energy will implement the MRP-228 requirements for the augmented inspections described in this ONS RV Internals inspection plan developed in accordance with MRP-227, Rev. 0 for the three ONS units.

4.1.2 PWROG Activities As part of the ONS License Renewal Program, Duke Energy made the commitment to participate in industry activities associated with the development of the standard industry guidance, which includes the activities performed by the PWROG. The PWROG activities provide continuous industry support and a strategic plan for the aging management of the PWR RV Internals through participation in technical meetings and industry forums.

Sections 4.2.7, 5.3, 5.4.5, and 5.6.2 of this report discuss past and ongoing PWROG activities.

4.1.3 Time-Limited Aging Analyses In the ONS LRA"', the identified three RV Internals applicable TLAAs, as listed below, were evaluated for the period of extended operation consistent with the requirements of 10 CFR 54.21.

I. Flow-induced vibration endurance limit assumptions The flow-induced vibration fatigue limit assumptions were increased from 1012 cycles for 40 years to 1013 cycles for 60 years. The stress values calculated were found to be less than the endurance limit, rendering the evaluation acceptable according to the requirements of 10 CFR 54.21. Therefore, this TLAA has been resolved.

2. Transient cycle count assumptions for the replacement bolting The ability to withstand cyclic loading without fatigue failure was evaluated using a cumulative usage factor methodology. In BAW-2248, for each utility, the number of transients accrued to date was conservatively extrapolated, and in all cases it was found that the number of design cycles would not be exceeded in the period of extended operation. The B&WOG reported that each of the participating utilities monitors occurrences of design transients and is thus managing the potential for cracking resulting from fatigue.

Therefore, Duke Energy will continue to monitor and track occurrences of design transients for all three ONS units during the extended license period.

3. Reduction in fracture toughness The TLAA described as "reduction in fracture toughness" is related to the acceptability of the RV Internals under loss of coolant accident (LOCA) and seismic loading. BAW-2248 states that BAW-10008, Part 1, Revision 11161 concludes "that at the end of 40 years, the internals will have adequate ductility to absorb local strain at the regions of maximum stress intensity, and that irradiation will not adversely affect deformation limits." BAW-2248 also states that this TLAA will be resolved on a plant-specific basis per 10 CFR 54.21 (c)(1)(iii) based on the results and conclusion of the planned B&WOG RV Intemals AMP. Duke Energy has stated that appropriate action will be taken in a timely manner to ensure continued validity of the design of the ONS RV Internals. Plant-specific analysis is required to demonstrate that, under LOCA and seismic loading and with irradiation accumulated at the expiration of the period of extended operation, the RV Internals have adequate ductility to absorb local strain at the regions of maximum stress intensity and will meet the deformation limits. The applicant must provide a plan to develop data to demonstrate that the RV Internals will meet the deformation limits through the period of extended operation. Duke Energy committed to perform the plant-specific analysis.

A bounding analysis applicable to the B&W-designed units including ONS-1, ONS-2, and ONS-3 was performed for the period of extended operation. The analysis concluded that at the end of a 60-year lifetime, the internals will have adequate ductility to absorb local strain at the regions of maximum stress intensity, and Page 20 of 108

ANP-2951, Rev. 001 the irradiation will not adversely affect deformation limits. The analysis will be provided to the NRC to demonstrate the completion of this TLAA for the ONS units.

There is a fourth TLAA discussed in NUREG-l 723 regarding flaw growth acceptance in accordance with the ASME B&PV Code Section XI In-Service Inspection (ISI) requirements. This TLAA is identified in BAW-2248 as requiring plant-specific evaluation. An open item (Open Item 4.2.5.3-2) was identified in the June 16, 1999 SERý 71and subsequently in a letter dated October 15, 1999[ýI8 Duke Energy responded that no flaws have been identified in the ONS RV Internals and hence no evaluation is required. The October 15, 1999 letter closes Open Item 4.2.5.3-2.

4.1.4 Internals Bolting Surveillance Program Starting in 1981, ultrasonic testing (UT) at several B&W units revealed the LTS bolt, upper core barrel (UCB) bolt, lower core barrel (LCB) bolt, flow distributor (FD) bolt, and surveillance specimen holder tube (SSHT) bolt locations had rejectable UT indications. Some of the bolts with rejectable UT indications were later determined to be cracked due to intergranular stress corrosion cracking (IGSCC) by laboratory examination. The failed bolts were fabricated from Alloy A-286, Condition A (ASTM A 453, Grade 660) material, except for the SSHT bolts, which were fabricated from Alloy A-286, Condition B (ASTM A 453, Grade 660) material.1 9" I As a result of the noted bolt failures, utilities began replacing bolts where needed. The B&WOG initiated the IBSP (which was completed by EPRI PWR MRP) to better assess the IGSCC susceptibility of the replacement bolts. The TBSP exposed scaled down replacement bolts to simulated PWR conditions in an autoclave and an actual PWR environment inside the RV specimen tube holder at an operating B&W unit. The scaled down bolts used in the testing were manufactured from Alloy A-286, Condition A and Alloy X-750, high temperature heat-treatment (HTH) condition materials. The scaled down bolts were tested in two surface conditions, peened and un-peened.

After the completion of the IBSP tests, several peened replacement Alloy A-286' scaled down replacement bolts developed IGSCC when loaded to a high stress, while the un-peened Alloy A-286 scaled down replacement bolts were free from IGSCC when loaded to the same high stresses for the test duration of 8 !/2 years. The Alloy X-750, HTH Condition scaled down replacements bolts of both peened and un-peened conditions were free from IGSCC when subjected to the same environmental and loading conditions as the Alloy A-286 bolts for the test duration of 8 /2 years.

Only the LTS bolts at the three ONS units have been replaced with Alloy X-750 HTH studs and nuts. The other locations such as UCB, LCB, upper thermal shield (UTS), and FD bolts are the original Alloy A-286 bolts at the ONS units. Most of the SSHT assemblies at the three ONS units, including the SSHT bolts, were removed from the RV Internals, and therefore no longer have an IGSCC concern.

A 2005 evaluation of the IBSP and industry experience resulted in PWROG Letter OG-06-1880, which makes recommendations for UT examinations of the high strength (Alloy X-750 or Alloy A-286) bolts in the B&W units. These recommendations were made in accordance with NEI 03-08. These UT inspection recommendations have been incorporated into MRP-227, Rev. 0 with the "Needed" recommendation being incorporated into the "Primary" category and the "Good Practice" recommendation being incorporated into the "Expansion" category.

The MRP-227, Rev. 0 inspections supersede these UT inspection recommendations.

4.1.5 Joint Owners' Baffle Bolt Program The JOBB Program stemmed from ultrasonic inspections of baffle-to-former bolts at several Electricitd de France (EdF) plants. Indications were noted under the bolt head in the head-to-shank fillet radius. The bolt failures were attributed to irradiation-assisted stress corrosion cracking (IASCC). Various tasks, including non-destructive examination (NDE) inspections, temperature, fluence, loading, and chemical composition comparisons of bolts, irradiation and mechanical testing, corrosion testing, helium effect evaluation, and microstructural evaluation were used to characterize the effect of irradiation on bolting materials under the JOBB.

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ANP-2951, Rev. 001 The JOBB is now being managed by EPRI with additional research on RV Internals material being performed under EPRI programs. The results of the JOBB program have been incorporated into EPRI PWR MRP documents through the screening criteria for IASCC, and specifically referenced in MRP-227, Rev. 0. In BAW-2248A, repeated in NUREG-1723, there is an action item to provide a final report that contains the test results from the RVIAMP and the recommended inspection program for the RV Internals. EPRI PWR MRP provides results to the NRC during meetings (see Reference 20 for an example), which fulfills this commitment.

4.1.6 Fuel/Baffle Interaction Investigation An investigation was conducted between 2004 and 2010 on the interaction between the baffle plates and fuel assembly grid straps in the B&W units by AREVA NP and the utilities with operating B&W-designed units.

Wear of fuel assembly spacer grid outer straps against RV Internals baffle plates has been observed since initial plant operation and has increased significantly with the switch from Alloy-718 to Zircaloy-4 grids in the 1980s.

The "Primary" requirement from MRP-227, Rev. 0 for a one-time physical measurement of the interference fit between the plenum cover weldment rib pads and the RV flange was performed at ONS between 2006 and 2008 in order to provide data to the investigation. Recommendations from the fuel/baffle interaction investigation were entered into the Duke Energy Problem Investigation Process (PIP).

4.2 ONS Programs and Activities ONS has a number of programs and activities that support the aging management of the RV Internals; these include the ASME B&PV Code Section XI In-Service Inspection program, primary water chemistry program, the vent valve in-service test program, implementation of low-leakage cores, LTS bolt replacement, a fabrication records search, UT examination of UCB bolts, core clamping measurements, visual examination of baffle-to-baffle and baffle-to-former bolts at each RFO, and ONS unit-specific amendments to MRP-227, Rev. 0.

4.2.1 ASME B&PV Code Section Xl In-Service Inspection Requirements The ONS ASME B&PV Code Section XI ISI requirements for examination of the RV interior, attachments, and internals are contained in ASME B&PV Code Section XI, Subsection IWB-2500-1 .l[J" Areas accessible during a refueling outage (RFO) of the RV interior (Examination Category B-N-I) are examined using visual examinations (VT-3) examination methods each period (approximately every3 years). RV interior attachments (Examination Category B-N-2) within the beltline region are examined using visual VT- 1 examination methods and interior attachments beyond the beltline region are examined using visual VT-3 examination methods each interval (approximately every 10 years). Removable core support structures are examined using visual VT-3 examination methods each interval (approximately every 10 years). Category B-N-l, B-N-2, and B-N-3 examinations will be performed during the next ONS ASME Section XI 10-year ISI examinations currently scheduled for the Fall 2012 RFO for ONS-1, the Fall 2013 RFO for ONS-2, and the Spring 2014 RFO for ONS-3.

The RV core guide lugs will receive a VT- I examination in accordance with Examination Category B-N-2 during the fourth and future 10-year ISI intervals at ONS. The remnants of the flow stabilizers and the incore monitoring instrumentation (IMI) nozzles will receive a VT-3 examination in accordance with Examination Category B-N-2 during the fourth and future 10-year ISI intervals at ONS. Removable core support structures (Examination Category B-N-3), which will receive a VT-3 examination during the fourth and future 10-year ISI intervals at ONS, are listed in Table 4-1 below. Relevant conditions for these examination categories are found in ASME B&PV Code Section XI, IWB-3520.

Table 4-1. ONS Removable Core Support Structure Components (Examination Category B-N-3)

Component Thermal Shield Thermal Shield Upper Restraint Assemblies Upper Thermal Shield Bolting Remnants of Surveillance Holder Tube Structures Page 22 of 108

ANP-2951 Rev. 001 Component Core Support Shield (CSS) Assembly CSS Top Flange, including Seating Surfaces CSS Outlet Nozzles CSS Outlet Nozzle Sealing Surfaces CSS Outlet Flow Deflectors Internals Vent Valves, Retaining Rings, Guide Blocks, Jack Screws and Locking Devices Core Support Assembly (CSA) Lifting Lugs CSA Keyways CSA Loss of Coolant Accident (LOCA) Bosses Upper Core Barrel Bolting CSA Baffle Plates CSA Former Plates Baffle Plate Bolting CSA Lower Grid CSA Lower Grid Pads Instrument Guide Tube Spiders Flow Distributor Bolting Interface between Upper Former Plate and Core Barrel and Adjacent Surfaces IMI Tubes and Guide Tubes Flow Distributor Head Guide Block Assemblies - Pairs Guide Block Bolting Shock Pad Assemblies Shock Pad Bolting Lower Core Barrel Bolting Lower Thermal Shield Bolting Liffing Lugs and Base Blocks Plenum Cover and Ribs Plenum Cover to Cylinder Bolted Connection Plenum Clamping Surfaces Plenum Cylinder to Upper Grid Bolted Connection Plenum Assembly Keyways Plenum Assembly Outside Surfaces Thermocouple Guide Tube Assemblies and Attachments Control Rod Guide Tube Assemblies (from top of plenum assembly)

Plenum LOCA Bosses and Welds Upper Grid Assembly (including bolting and grid pads)

Control Rod Guide Tube Assemblies (from bottom of plenum assembly)

Note: The ASME B&PV Code Section XI, Category B-N-3 ISI scope is defined by the owners (utilities) of the B&W units.

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ANP-2951, Rev. 001 4.2.2 Primary Water Chemistry Program The ONS Primary Water Chemistry Program limits the concentration of oxygen, halogens, and sulfate species in the primary water to prevent the coolant from becoming an environment favorable to stress corrosion cracking (SCC), and therefore greatly reduces the probability of SCC and IASCC. The limits imposed by the primary water chemistry 2

program meet the intent of the EPRI Pressurized Water Reactor Primary Water Chemistry Guidelines. [)

4.2.3 Vent Valve In-Service Test Program There is an existing ONS program that requires vent valve testing and visual inspection each RFO. The accessible surfaces of the vent valve are visually inspected, including the locking devices. Any observed surface irregularities on the valve body and disc seating surface are identified and evaluated. Additionally, vent valve operation is tested through manual actuation to verify that the lifting force required to fully open the vent valves does not exceed the specific limit.

4.2.4 Continuation of Use of Low-Leakage Cores As discussed in Section 3.4.3.3 of NUREG-1723, Duke Energy will continue to use low-leakage core loading patterns, which is considered a preventative action to lessen the effects of aging on the ONS RV Internals.

4.2.5 Lower Thermal Shield Replacement Bolt Failure of the original Alloy A-286 LTS bolts in the 1980s led to the replacement of all the original LTS bolts with replacement studs/nuts at the three ONS units. The replacement LTS nuts/studs on each unit are secured by tie plate and crimp locking cups in groups of two (two-hole design). The replacement studs are made from Alloy X-750 in the HTH condition. The compression nuts are also made of Alloy X-750 in the HTH condition. The tie plate and crimp cup are both made of Type 304 stainless steel. Note that the LTS bolts are categorized as an "Expansion" item in MRP-227, Rev. 0.'

4.2.6 Fabrication Records Search Two fabrication records searches were conducted by AREVA NP for the ONS RV Internals components listed as "Primary" and "Expansion" in MRP-227, Rev. 0. The goal of the first records search was to locate the chemical composition of the CASS items and, if possible, to screen them for susceptibility to thermal aging embrittlement, consistent with the screening criteria used by the EPRI PWR MRP. The goal of the second ONS-specific record search was to obtain a detailed description of the component, obtain fabrication records (heat numbers, CMTRs, etc.), obtain a description of the anticipated degradation mechanisms, and review of operational experience.

4.2.6.1 Cast Austenitic Stainless Steel Records Search In 2009 and 2010, a search of original fabrication records was made for several B&W CASS RV Internals items identified as susceptible to thermal aging embrittlement in MRP-227, Rev. 0. The thermal aging embrittlement susceptibility can be screened per the criteria in MRP-175t2 21, using Hull's equivalent factors in NUREG/CR-4513, Rev. 1[231, if chemical composition is known. The MRP-175 screening criteria for thermal aging embrittlement of CASS are identical to those in Section XI.M12 ofNUREG-1801. During the screening phase for MRP-227, Rev. 0, if the estimated ferrite content did not exceed the MRP-175 screening criteria for CASS thermal aging embrittlement, the CASS item was screened out. CASS items whose chemical composition was unknown were identified as susceptible to thermal aging embrittlement in MRP-227, Rev. 0.

The records search included the CSS vent valve discs, CSS outlet nozzles (ONS-3 only), and the CRGT assembly spacer castings. The following conclusions were made from this records search and corresponding calculation of ferrite content:

1. CSS vent valve discs, CF-8 (ONS-1, ONS-2, and ONS-3)

The CSS vent valve discs are categorized as "Primary" in MRP-227, Rev. 0. Each ONS unit has eight 14-inch CSS vent valves. The CSS vent valve discs were fabricated from Grade CF-8 castings. The chemical composition has been found for all 24 original CSS vent valves discs supplied to the three ONS units. The Page 24 of 108

ANP-2951, Rev. 001 ferrite content for all of the original CSS vent valve discs is under the 20% screening criteria for thermal aging embrittlement of CF-8 castings.

However, ONS was also supplied with spare CSS vent valves, in addition to the originally installed CSS vent valves. The spare CSS vent valve records were not found at AREVA NP during this records search. It is known that some spare CSS vent valves were installed (i.e., replaced the original CSS vent valves) in the late 1970's or early 1980's to fix the CSS vent valve jackscrew locking devices at ONS. To ascertain that each installed CSS vent valve disc is under the screening threshold, it was recommended that the CSS vent valve disc serial number (S/N) and CSS vent valve disc heat number for the currently installed CSS vent valves be identified. The S/N and heat numbers are stamped on the CSS vent valve disc surface.

The S/N of the eight CSS vent valves discs installed in ONS-1 were recorded during the Fall 2009 RFO. The S/N of the eight CSS vent valve discs installed in ONS-2 were recorded during the Spring 2010 RFO. Similar CSS vent valve disc identification is planned for ONS-3 in the Fall 2010 RFO. Based on the S/N and heat number identification, the CSS vent valve disc ferrite content of all currently installed CSS vent valves at ONS-1 and ONS-2 are confirmed to be below the screening threshold. Based on the records search and ONS-I and ONS-2 identification results, a similar finding is expected for ONS-3.

Therefore, a justification for an ONS unit-specific amendments to MRP-227, Rev. 0 requirements has been written (see Appendix F of this report) to recategorize all currently installed CSS vent valve discs at ONS-1 and ONS-2 from "Primary" to "No Additional Measures". No augmented inspection is required. This recategorization will apply to the ONS-3 installed vent valve discs after the ferrite is confirmed to be below the screening criteria. The existing inspection requirements for these items such as CSS Vent Valve In-Service Test Program, as discussed in Section 4.2.3 of this report, will continue to be performed.

2. CSS outlet nozzles, CF-8 (ONS-3 only)

The two CSS outlet nozzles in the ONS-3 RV Internals are categorized as "Primary" in MRP-227, Rev. 0 and are fabricated from Grade CF-8 castings. The records search confirmed the ferrite content calculated from the chemical composition of the CSS outlet nozzles at ONS-3 is below the 20% screening criteria for thermal aging embrittlement of a CF-8 casting. Therefore, a justification for an ONS unit-specific amendment to MRP-227, Rev. 0 requirements has been written (see Appendix F of this report) to recategorize the CSS outlet nozzles at ONS-3 from "Primary" to "No Additional Measures". No augmented inspection is required. The existing Section XI ISI for the CSS outlet nozzles will continue to be performed at ONS-3. The CSS outlet nozzles at ONS-I and ONS-2 are not fabricated from CASS and do not have a thermal aging embrittlement concern.

3. CRGT spacer castings, CF-3M (ONS-1, ONS-2, and ONS-3)

The CRGT spacer castings are categorized as "Expansion" in MRP-227, Rev. 0. Each ONS unit contains 690 CRGT spacer castings in its RV Internals fabricated from Grade CF-3M castings. Based on the chemical compositions found, the ferrite content for most of the CRGT spacer castings at the ONS units exceeds the 14% ferrite screening criteria for thermal aging embrittlement for CF-3M casting. Therefore, it is concluded that the CRGT spacer castings should remain applicable to the ONS units and cannot be recategorized to "No Additional Measures". Due to the recategorization of its primary linked items (CSS vent valve discs and CSS outlet nozzles [ONS-3 only]) to "No Additional Measures", the CRGT spacer castings are recategorized as "Primary" for the ONS units as described in Appendix F of this report.

4.2.6.2 ONS-1 Plenum Cover Weldment Rib Pad Items The records search for "Primary" and "Expansion" items in MRP-227, Rev. 0 also performed in 2009 and 2010 for the RV Internals for the three ONS units identified a feature unique to the ONS-I plenum cover weldment rib pads. Each of the 32 plenum cover weldment rib pads at ONS-1 are fastened to the plenum cover ribs with two Type 304 stainless steel screws and one Alloy X-750 dowel. The Alloy X-750 dowels, Type 304 screws, and their locking welds were unknown and were not screened for aging degradation mechanisms, nor evaluated for inclusion in MRP-227, Rev. 0.

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ANP-2951, Rev. 001 Using the MRP screening criteria and process, these items are categorized as Category "A" or "No Additional Measures". Therefore, no additional augmented inspection is required for this location. The evaluation for the ON S- I plenum cover weldment rib pad items is documented in Appendix F of this report.

4.2.7 Volumetric (UT) Examinations of Upper Core Barrel Bolts The most recent UCB bolt UT inspections were performed in April 2008 (ONS-I, 100%), October 2008 (ONS-2, 100%), and November 2007 (ONS-3, 100%) in accordance with a "Needed" NEI 03-08 recommendation in PWROG letter OG-07-43 (1/26/07) and also met the initial inspection requirements in MRP-227, Rev. 0. The results of the inspections at ONS-I and ONS-2 were zero UCB bolts rejected as in previous inspections. Two UCB bolts were rejected due to a lack of back wall reflection at ONS-3; these results are identical to the previous ONS-3 UT inspection results in 1984, 1985, and 1987. Note that at ONS-l, four of the 120 UCB bolts were removed for verification and better interpretation of bolt UT signals in the 1980s. Visual, ultrasonic, and fluorescent liquid penetrant examinations were performed in the laboratory on all four bolts. The examinations found no indications confirming the on-site UT results. These four UCB bolt locations at ONS-I are still empty.

The most recent LCB UT inspections were performed in June 1983 (ONS-1, partial), October 1983 (ONS-2, partial), and January 1987 (ONS-3, 100%) in response to the original B&W internal A-286 bolt failures. UT inspections for the LCB bolts are planned during the RV Internals inspections in 2012, 2013, and 2014, in compliance with the MRP-227, Rev. 0 recommended inspection requirement for the LCB bolts.

The AREVA NP UT examination procedure for the ONS UCB and LCB bolt examinations in 2007 and 2008 was validated by blind performance demonstration at EPRI in 2007 prior to the bolt inspections at ONS. The demonstration ensured the UT examination procedure's capability of determining the integrity of the UCB bolts.

This demonstration was documented by EPRI with essential elements identified.

After the ONS UCB bolt examinations were completed in 2008, an ONS-specific technical justification (TJ) in accordance ASME B&PV Code Section V Article 14 for both the UCB and LCB bolts examinations was created for both UCB and LCB bolts by compiling existing information in one document. In addition to providing a detailed explanation of the examination process and other influential parameters important to the overall performance of the examination system, the TJ contains a description of the component, manufacturing history, flaws of interest, and operating history. Appendix E of this report provides a nonproprietary version of the UCB and LCB bolt ONS-specific TJ. It is an NEI 03-08 "Needed" requirement that TJs be created for each .

examination procedure in accordance with Section 2.1 of MRP-228, except for visual examinations. ONS-specific TJs are being prepared for visual and UT examination methods to be used for inspecting "Primary" and "Expansion" RV Internals components.

The evaluation criteria for the most recent ONS UCB bolt examinations were based on the stress limits for threaded structural fasteners in Subsection NG of the ASME B&PV Code. Using an analytical tool developed .

under PWROG PA-MSC-0350, the ONS unit-specific analysis demonstrates large margins using the most recent UCB bolt UT inspection results.

4.2.8 Core Clamping Measurements Core clamping measurements were obtained by AREVA NP at ONS-I (2006), ONS-2 (2008), and ONS-3 (2007) during RFOs and satisfy the MRP-227, Rev. 0 requirements for a one-time physical differential height measurement of the plenum rib pad to RV seating surface. The reason for the timing of the measurements was a concern that loss of core clamping force could be a contributor to wear between baffle plates and fuel grids (see Section 4.1.6 of this report). The measurements at the three units found no evidence of wear occurring during the service period of operation and it was concluded there was no evidence that core clamping has been degraded.

Wear of the core clamping items in the plenum cover assembly and core support shield assembly will continue to be monitored via subsequent VT-3 examinations performed on the 10-year ISI interval per MRP-227, Rev. 0 requirements.

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ANP-2951, Rev. 001 4.2.9 Visual Examination of Baffle-to-Baffle and Baffle-to-Former Bolts In addition to the ASME B&PV Code Section XI ISI inspection requirement discussed in Section 4.2.1 of this report, Duke Energy has voluntarily performed visual inspection by underwater camera every RFO of the internal baffle-to-baffle bolts and baffle-to-former bolts at ONS units, per model work orders. The visual inspection is not required by the ASME B&PV Code and therefore is not conducted in accordance with the ASME B&PV Code.

No out of design configuration baffle-to-baffle bolts and baffle-to-former bolts have been observed during these visual inspections at ONS units.

The only abnormal condition for the baffle-to-baffle and baffle-to-former bolts in the B&W units was noted from the inspection of the baffle-to-former and internal baffle-to-baffle bolts at Crystal River Unit 3 (CR-3) in 2005.

Visual inspections indicated that three or four internal baffle-to-baffle bolts were not within the design configuration. The bolt heads extended beyond the baffle plate surface. This was an indication that the locking devices, and potentially the baffle-to-baffle bolts as well, had failed. A UT inspection of 100% of the baffle-to-former bolts at CR-3 was performed with no indications of broken baffle-to-former bolts. No UT inspection was performed on the internal baffle-to-baffle bolts. The abnormal baffle-to-baffle bolts at CR-3 have not been removed for laboratory examination to confirm the failures.

4.2.10 ONS Unit-Specific Amendments to MRP-227, Rev. 0 Requirements Due to new information found during an ONS fabrication records search and the development of TJs, six amendments to MRP-227, Rev. 0 were identified. The full description of needed ONS unit-specific amendments to MRP-227, Rev. 0 requirements and their bases are provided in Appendix F of this report. A synopsis of each amendment is given in the bullets below.

" The CSS vent valve discs at ONS-I and ONS-2 are recategorized to "No Additional Measures". No augmented inspection is required. This recategorization will apply to the ONS-3 installed vent valve discs after the ferrite is confirmed to be below the screening criteria.

• The CSS cast outlet nozzles at ONS-3 are recategorized to "No Additional Measures". No augmented inspection is required.

• The CRGT spacer castings at the three ONS units are recategorized to "Primary". The expansion link, inspection coverage, method, and examination frequency are listed in Appendix F of this report.

• The CSS vent valve disc shaft (hinge pin) is inaccessible for visual inspection. The examination method/frequency and examination coverage in MRP-227, Rev. 0 Tables 4-1 and 5-1 for the CSS vent valve disc shaft (hinge pin) at ONS-1, ONS-2, and ONS-3 are revised as described in Appendix F of this report.

" The aging effect (mechanism) in Table 4-1 of MRP-227, Rev. 0 for the UCB and LCB bolt and bolt locking devices is clarified as described in Appendix F of this report.

• The aging effect (mechanism) in Table 4-4 of MRP-227, Rev. 0 for the UTS, LTS, and FD bolt and bolt locking devices is clarified as described in Appendix F of this report.

4.3 Conclusions of Section 4.0 The ONS RV Internals inspection plan is based on the ONS LRA, NUREG-1723, and evaluations supporting MRP-227, Rev. 0. Inspections will consist of the ASME B&PV Code Section XI Examination Category B-N-3 inspections given in Table 4-1 of this report and the augmented inspections from MRP-227, Rev. 0, as modified by the amendments discussed in Appendix F of this report. Changes resulting from the NRC's review of this report and MRP-227, Rev. 0 will be incorporated as appropriate.

Table A.I in Appendix A of this report shows how the components identified in BAW-2248A, Table 4-1 AMR were evaluated and characterized by the industry program. Past and on-going activities by the EPRI PWR MRP, B&WOG, PWROG, and Duke Energy provide the needed clarification to the level of inspection quality necessary to determine the proper examination method and frequencies. The ONS RV Internals AMP includes existing ONS programs and ASME B&PV Code Section XI inspections combined with MRP-227, Rev. 0 augmented Page 27 of 108

ANP-2951, Rev. 001 inspections to provide reasonable assurance that the ONS RV Internals components will continue to perform their intended fuinctions through the period of extended operation.

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ANP-2951, Rev. 001 5.0 ONS RV INTERNALS AMP ATTRIBUTE EVALUATION The ONS RV Internals AMP, which will include this ONS RV Internals inspection plan, utilizes a combination of prevention, mitigation, and condition monitoring. Where applicable, credit is taken for existing programs (e.g.,

primary water chemistry and ASME B&PV Code Section XI inspections) and mitigation projects such as LTS bolt replacement. The ONS RV Internals inspection plan then incorporates recommendations for augmented inspections provided by industry guidelines in MRP-227, Rev. 0 as applicable to ONS.

This section uses the ten AMP elements from NUREG- 180 1, Rev. 1 to describe the ONS RV Internals inspection plan.

5.1 AMP Element I - Scope of Program The ONS RV Internals AMP is focused on managing the effects of the eight age-related degradation mechanisms given in Section 2.3 of this report and ensuring the RV Internals remain functional during the license renewal period. The ONS RV Internals AMP consists of not only the ONS RV Internals inspection plan ("Primary" and "Expansion" inspections from MRP-227, Rev. 0, as applicable to ONS, and ASME B&PV Code Section XI Examination Category B-N-3 inspections) but also credits programs such as the primary water chemistry program and the vent valve in-service test program and preventative actions such as the LTS bolt replacement. The ONS RV Internals inspection plan is focused on detecting possible degradation effects from the eight aging degradation mechanisms. The components chosen for inspection are the result of the LRA AMR refined by industry activity that culminated in MRP-227, Rev. 0.

5.1.1 ONS Scope The ONS RV Internals consist of two basic assemblies located inside the RV[31:

• Plenum Assembly The plenum assembly provides continuous guidance and protection of the control rods. In addition, the plenum assembly directs flow out of the core to the vessel outlet nozzles. The plenum assembly is removed every RFO to permit access to the fuel assemblies.

• Core Support Assembly (CSA)

The CSA remains in place in the RV and is only removed to perform scheduled inspections of the RV interior surfaces and attachments as well as the RV Internals.

A description of the ONS RV Internals intended functions is provided in Section 2.4 of this report. Additional RV Internals details are provided in the ONS UFSAR.

5.1.2 ONS RV Internals Components Subject to an AMR The components of the ONS RV Internals that were evaluated by the industry, culminating in MRP-227, Rev. 0, include those identified in the ONS LRA AMR and those contained in BAW-2248A. Table A. I in Appendix A of this report is included to show how the components identified in the AMR from BAW-2248A were evaluated and characterized by the industry program resulting in MRP-227, Rev. 0. The components evaluated also include the components identified in NUREG-1801, Rev. 1.

5.1.3 Conclusion The scope of the ONS RV Internals inspection plan includes the specific structures and components subject to an AMR and encompasses those components in the LRA AMR and those listed in Table IV.B4 of NUREG-1801, Rev. 1, thus satisfying the regulatory criteria in 10 CFR 54.

5.2 AMP Element 2 - Preventative Actions The ONS RV Internals AMP includes the following existing programs and activities that comply with the requirements of this AMP element. Maintaining high water purity reduces susceptibility to SCC and IASCC.

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ANP-2951, Rev. 001 Reactor coolant water chemistry is monitored and maintained in accordance with the EPRI Pressurized Water Reactor Chemistry Guidelines.121 Additionally, Duke Energy will continue to use low-leakage core loading patterns as a preventative action. A description of the Primary Water Chemistry Control Program, reference to the commitment to low-leakage cores, and their applicability to the ONS RV Internals inspection plan is provided in the following subsections.

5.2.1 Primary Water Chemistry Control Program The ONS Primary Water Chemistry Program, as implemented by the ONS primary water chemistry program limits the concentration of oxygen, halogens, and sulfate species in the primary water to prevent the coolant from becoming an environment favorable to SCC, and therefore effectively prevents SCC and greatly reduces the probability of IASCC. The limits imposed by the primary water chemistry program meet the intent of the Pressurized Water Reactor Chemistry Guidelines. 21' MRP-227, Rev. 0, the ONS LRA, and the GALL report credit the water chemistry AMP for mitigation of material loss due to crevice and pitting corrosion, as well as SCC.

5.2.2 Low-Leakage Cores As discussed in Section 3.4.3.3 of NUREG-1723, Duke Energy will continue to use low-leakage core loading patterns, which are considered a preventative action to lessen the effects of aging on the ONS RV Internals.

5.2.3 Conclusion The preventative actions for the ONS RV Internals AMP include the Primary Water Chemistry program as well as low-leakage core loading patterns, to mitigate the applicable aging effects, thus satisfying the regulatory criteria in 10 CFR 54.

5.3 AMP Element 3 - Parameters Monitored or Inspected The ONS RV Internals inspection plan monitors for the detectable effects of the eight aging degradation mechanisms outlined in Section 2.3 of this report. The ONS RV Internals inspection plan credits, and further augments, the ASME B&PV Code Section XI, Table IWB-2500-1 inspections with the inspections listed in MRP-227, Rev. 0, Tables 4-1 and 4-4 as applicable to ONS. For "Expansion" bolts that are inaccessible, the PWROG is performing a justification by evaluation under PA-MSC-0692. TLAAs identified in NUREG- 1723 for "flow-induced vibration endurance limit assumptions", "transient cycle count assumptions for replacement bolting", and "reduction in fracture toughness" have been resolved as discussed in Section 4.1.3 of this report.

The ONS RV Internals inspection plan uses UT, VT-3, and physical measurement to monitor for the detectable effects of the eight aging degradation mechanisms outlined.in Section 2.3 of this report. UT is used to detect cracking in bolts. VT-3 is used to identify the conditions detailed in ASME B&PV Code Section XI, IWB-3520.

5.3.1 The ONS In-Service Inspection Program The ONS ASME B&PV Code Section XI ISI Program (as discussed in Section 4.18 of the ONS LRA[3] and Section 4.2.1 of this report) is credited for inspection of numerous RV Internals components requiring aging management, as identified by the AMR and listed in Table A. 1 in Appendix A of this report. The Examination Category B-N-3 components to be examined during the upcoming 10-year RV ISI are listed in Table 4-1 of this report.

Though some "Primary" and "Expansion" components are also listed as Examination Category B-N-3 items in the ISI Program, no "Existing Programs" were identified for B&W plants in MRP-227, Rev. 0. This was done in order to provide further guidance with respect to examination coverage and relevant conditions, as well as highlight the possible need for further evaluations. If a component receives a VT-3 inspection to satisfy both the ASME B&PV Code Section XI ISI program and MRP-227, Rev. 0 "Primary" or "Expansion" examinations, the examinations may be completed at the same time.

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ANP-2951, Rev. 001 5.3.2 MRP-227, Rev. 0 "Primary" and "Expansion" (Augmented) Inspections MRP-227, Rev. 0 "Primary" and "Expansion" inspections will be listed as augmented (required non-ASME B&PV Code Section XI inspections) inspections in the ONS ISI program. MRP-227, Rev. 0, Table 4-1 "B&W Plants Primary Components" and Table 4-4 "B&W Plants Expansion Components" are included as Appendices B and C of this report, respectively. In addition to listing the components requiring inspections, the Tables provide the Applicability (Plant), Effect (Mechanism), Expansion/Primary (Link), Examination Method, Frequency (for "Primary" components), and Examination Coverage. Appendix D of this report contains MRP-227, Rev. 0, Table 5-1 "B&W Plant Examination NDE Acceptance and Expansion Criteria," which provides Applicability, Examination Acceptance Criteria, Expansion Link, Expansion Criteria, and Additional Examination Acceptance Criteria for the applicable ONS components.

5.3.2.1 MRP-227, Rev. 0 "Primary" (Augmented) Inspections MRP-227, Rev. 0 lists the following B&W "Primary" RV Internals bolts to be inspected with UT:

" UCB Bolts

• LCB Bolts

" Baffle-to-Former Bolts MRP-227, Rev. 0 lists the following B&W "Primary" RV Internals components to be examined with a VT-3:

• Plenum Cover Weldment Rib Pads

• Plenum Cover Support Flange

• CSS Top Flange

" CSS Cast Outlet Nozzles (ONS-3 only) - ONS-3 Amendment, recategorized as "No Additional Measures" (see Appendix F of this report)

" CSS Vent Valve Disc - ONS Amendment, recategorized as "No Additional Measures" (see Appendix F of this report)

" CSS Vent Valve Top Retaining Ring

• CSS Vent Valve Bottom Retaining Ring

• UCB Bolt Locking Devices

• LCB Bolt Locking Devices

• Baffle Plates

" Baffle-to-Former Bolt Locking Devices including Locking Welds

• Internal Baffle-to-Baffle Bolt Locking Devices including Locking Welds

• Alloy X-750 Dowel-to-Guide Block Welds

" IMI Guide Tube Spiders

" IMI Guide Tube Spider-to-Lower Grid Rib Section Welds During development of the TJ for the VT-3 visual inspection of the CSS vent valve disc shaft or hinge pin, it was discovered that the component was inaccessible and will need to be justified by evaluation with respect to the vent valve in-service test program or by replacement, as discussed in Appendix F of this report.

The following additional RV Internals component will be considered "Primary" via amendments and will be examined with a VT-3, as discussed in Appendix F of this report:

• CRGT Spacer Castings Page 31 of 108

ANP-2951, Rev. 001 MRP-227, Rev. 0 lists the following B&W "Primary" RV Internals components to be examined by physical measurement:

A one-time physical differential height measurement of the plenum rib pad to RV seating surface is required by MRP-227, Rev. 0. This measurement would indicate any change from the as-fabricated stacked height of the following components:

• Plenum Cover Weldment Rib Pads

" Plenum Cover Support Flange

" CSS Top Flange As discussed iii Section 4.2.7 of this report, the most recent inspections of the UCB bolts were performed in 2007 and 2008 in compliance with an NEI 03-08 "Needed" requirement from the PWROG. These UCB bolt inspections are being credited as the initial UT inspections required in MRP-227, Rev. 0, Table 4-1. The inspection results were the same as previous inspections performed in 1984 through 1987. No bolts were rejected at ONS-1 and ONS-2. ONS-3 had two bolts rejected due to a lack of back wall reflection.

The one-time differential height physical measurements for loss of core clamping of the plenum rib pads to RV -

seating surface were completed in 2006, 2007, and 2008 in support of the root cause for baffle-to-fuel wear. There was no evidence of loss of clamping due to wear.

Section 4.2.10 of this report provides a summary of the ONS unit-specific amendments to MRP-227, Rev. 0 and the justification for the ONS amendments to MRP-227, Rev. 0 is included in Appendix F of this report.

5.3.2.2 MRP-227, Rev. 0 "Expansion" (Augmented) Inspections MRP-227, Rev. 0 lists the following B&W "Expansion" RV Internals bolts to be inspected with UT:

• Upper Thermal Shield (UTS) Bolts

• Lower Thermal Shield (LTS) Bolts

(

• Flow Distributor (FD) Bolts MRP-227, Rev. 0 lists the following B&W "Expansion" RV Internals components to be examined with a VT-3:

• Alloy X-750 Dowel-to-Upper Fuel Assembly Support Pad Welds

• CRGT Spacer Castings - ONS Amendment, recategorized as "Primary" (see Appendix F of this report)

• Lower Fuel Assembly Support Pad

• Lower Fuel Assembly Support Pad-to Rib Section Welds

" Lower Fuel Assembly Support Pad Alloy X-750 Dowel

• Lower Fuel Assembly Support Pad Cap Screws and their Locking Welds

" Alloy X-750 Dowel-to-Lower Fuel Assembly Support Pad Welds VT-3 examinations of the locking devices are added to the "Expansion Items" for ONS units per the amendments in Section 4.2.10 of this report.

" UTS Bolt Locking Devices

• LTS Bolt Locking Devices

• FD Bolt Locking Devices MRP-227, Rev. 0 lists the following B&W RV Internals "Expansion" components as inaccessible:

• Core Barrel Cylinder (Including Vertical and Circumferential Seam Welds)

" Former Plates Page 32 of 108

ANP-2951, Rev. 001

• Baffle-to-Baffle Bolts (External)

• Core Barrel-to-Former Bolts

" External Baffle-to-Baffle Bolts Locking Devices, including Locking Welds

• Core Barrel-to-Former Bolts Locking Devices, including Locking Welds For baffle-to-baffle bolts and core barrel-to-former bolts that are inaccessible, the PWROG is performing a justification by evaluation.

As discussed in Section 4.2.10 of this report, the justification for the ONS amendments to MRP-227, Rev. 0 is included in Appendix F of this report.

5.3.3 Conclusion The parameters monitored or inspected in the ONS RV Internals inspection plan along with the TLAAs and evaluations of "Expansion" components are linked to the effects of aging on the intended functions of the particular structure and components. The inspections performed in accordance with ASME B&PV Code Section XI, the augmented inspections from MRP-227, Rev. 0 (as applicable to ONS), TLAAs, and evaluations provide reasonable assurance that the intended functions of the RV Internals discussed in Section 2.4 of this report will continue to be met through the period of extended operation.

The parameters to be monitored or inspected are the result of industry activity that satisfies the regulatory criteria listed in 10 CFR 54.21.

5.4 AMP Element 4 - Detection of Aging Effects The methods, coverage and schedule of the ONS RV Internals inspection plan along with the other programs credited in Section 5.1 of this report are designed to ensure that aging effects will be detected in a timely manner in order to maintain the intended functions of the RV Internals components. The methods used for detection are a one-time physical measurement, ASME B&PV Code Section XI Examination Category B-N-3 VT-3 examination, augmented VT-3 examination to detect specific aging effects, augmented UT examinations to detect cracking of bolting, and justification by evaluation for inaccessible components.

The MRP-227, Rev. 0 augmented inspections, as applicable to ONS, will be entered into the ISI program as augmented inspections and the results will be included in the 90 day outage report. There is also a NEI 03-08 "Good Practice" contained in MRP-227, Rev. 0 to provide a summary of inspection results to the MRP within 120 days (see Section 4.1.1.1.1 of this report).

The visual and UT examination data will be recorded and stored. Inspection reports will also be provided by the inspection vendor. These examinations have been scheduled to be performed at ONS during the 2012, 2013, and 2014 RFOs.

5.4.1 One-Time Physical Measurement A one-time physical differential height measurement of the plenum rib pad to RV seating surface at all three ONS units was conducted between 2006 and 2008 to ensure the clamping force holding the RV Internals was adequate, and thus rule out loss of clamping force as a possible cause of baffle plate to fuel grid wear as discussed in Section 4.1.6 of this report. There was no evidence of loss of clamping due to wear. The data collection for the one-time physical measurement was recorded by AREVA NP.

5.4.2 ASME B&PV Code Section Xl Examination Category B-N-3 VT-3 Examinations ASME B&PV Code Section XI Examination Category B-N-3 VT-3 examinations of accessible core support locations are conducted on a schedule in accordance with Table IWB 2500-1. The ASME B&PV Code Section XI VT-3 examination is intended to detect the type of general degradation condition identified in ASME B&PV Code Section XI IWB-3520.

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ANP-2951, Rev. 001 5.4.3 Augmented VT-3 Examination to Detect Specific Aging Effects The purpose of the augmented MRP-227, Rev. 0 "Primary" and "Expansion" VT-3 examinations is to detect aging effects of the eight aging degradation mechanisms outlined in Section 2.3 of this report. The method, coverage, schedule, and frequency of examinations are identified in MRP-227, Rev. 0, Table 4-1 and Table 4-4 and contained in Appendices B and C of this report, and for ONS unit-specific amendments to MRP-227, Rev. 0 are found in Appendix F of this report. Technical Justifications in accordance with ASME B&PV Code Section V, Article 14 Low Rigor are being prepared for each component receiving an augmented VT-3 examination.

Where fitting, additional analyses are being performed to complete the TJs and demonstrate why the examination method, schedule, frequency, and coverage are appropriate.

For VT-3 examinations performed within the core barrel region, the functionality analysis (finite element model) performed to support MRP-227, Rev. 0 indicates the inspections are being performed in a timely manner to prevent any loss of intended function. Additionally, ASME B&PV Code Section XI VT-3 examinations and B&W plant OE support the timeliness of the augmented VT-3 examinations.

5.4.4 Augmented UT Examinations to Detect Cracking of Bolting The purpose of the augmented MRP-227, Rev. 0 "Primary" and "Expansion" UT examinations is to detect cracking in bolts resulting from some of the eight identified aging degradation mechanisms outlined in Section 2.3 of this report. The method, coverage, schedule, and frequency of examinations are identified in MRP-227, Rev. 0, Table 4-1 and Table 4-4 and contained in Appendices B and C of this report. Technical Justifications in accordance with ASME B&PV Code Section V, Article 14 LowRigor are being prepared for each component receiving an augmented UT examination. Appendix E of this report is provided as a representation of what will be contained in the TJs for the ONS units. This document, along with the EPRI letter containing the demonstration results, will make up the TJ for the UCB and LCB bolts.

The timing of the B&W RV Internals structural bolt inspections (UCB, LCB, UTS, LTS, FD and SSHT bolts, note ONS units no longer have SSHT bolts) is based on past inspections performed due to failures of Alloy A-286 bolting that occurred in the 1980s. The MRP-227, Rev. 0 inspection requirements for B&W RV Internals structural bolts are similar to the B&WOG NEI 03-08 "Needed" requirements which are presently in effect. The functionality analysis (finite element model) for the baffle-to-former bolts performed to support MRP-227, Rev. 0 indicates the inspections are being performed in a timely manner to prevent any loss of intended function.

The latest UT examinations of the UCB bolts at ONS units were completed in 2007 and 2008 (see Section 4.2.7 of this report).

5.4.5 Justification by Evaluation for Inaccessible Bolts Augmented MRP-227, Rev. 0 "Expansion" components identified as inaccessibleare listed in MRP-227, Rev. 0 Table 4-4 and contained in Appendix C of this report. Additionally, the "Primary" vent valve disc shaft or hinge pin has been identified as being inaccessible (see Appendix F of this report). Additional analyses are being performed under PWROG PA-MSC-0692 "B&W Internals MRP-227 Phase II Functionality Criteria Core Barrel Assembly Modeling" to show functionality of the core barrel-to-former bolts and baffle-to-baffle bolts for the B&W plants through extended plant operation, regardless of potential failures in the linked "Primary" components.

5.4.6 Conclusion The ONS RV Internals inspection plan ensures detection of aging effects before there is loss of any structure and component intended function, including aspects such as method, frequency, coverage, and schedule.

5.5 AMP Element 5 - Monitoring and Trending RV Internals components at ONS are monitored and the results are used to develop potential trends in RV Internals aging management concerns.

The timing of one-time physical measurement or initial inspection along with inspection intervals are based on previous inspections, OE, and expert opinion. For components within the beltline region, results of the Page 34 of 108

ANP-2951, Rev. 001 functionality analysis performed on a typical B&W plant indicates the choice of components and timing of

• inspections confirms the prediction of the effects of aging and timely corrective or mitigative actions.

"Expansion" components have been defined if the scope of examination and re-examination needs to be expanded beyond the "Primary" group. Additionally, reporting requirements allow the industry to monitor and trend results, thus driving preemptive industry action through notifications and updating of the MRP-227 guidelines. EPRI PWR MRP will compile the summary of results into an overall history report which will track industry progress, aid in evaluation of significant issues, identify fleet trends, and determine any needed revisions to the MRP-227 guidelines. The industry (including Duke Energy) will be updated biennially by EPRI PWR MRP; the biennial report will serve to assist the monitoring and trending for AMPs established by the industry. Duke Energy will periodically review the industry inspection results and reports provided by EPRI PWR MRP.

See Section 4.1.1 of this report for a discussion of development of the industry program.

5.5.1 Conclusion The monitoring and trending provided in the ONS RV Internals inspection plan allows for prediction of the extent of the affects of aging and timely corrective or mitigative actions.

5.6 AMP Element 6 - Acceptance Criteria Two acceptance criteria will be discussed in this section. The first is examination acceptance criteria that define relevant conditions during component examinations. The second is functionality/engineering acceptance criteria, which dispositions relevant conditions, thus ensuring the particular structure and component intended functions are maintained under all CLB design conditions during the period of extended operation.

5.6.1 Examination Acceptance Criteria Examination acceptance criteria identify the visual examination relevant condition(s), signal-based level, or relevance of an indication that requires formal disposition for acceptability. Section 5 of MRP-227, Rev. 0 provides the examination acceptance criteria for the "Primary" and "Expansion" components; Table 5-1 of MRP-227, Rev. 0 is provided in Appendix D of this report.

In addition, the criteria for expanding the examinations from the "Primary" components to include the "Expansion" components are provided. The examination acceptance criteria include:

• Specific, descriptive relevant conditions for the visual (VT-3) examinations;

• Specific relevant indications for volumetric (UT) examination of bolting.

Additionally, TJs are being developed for the ONS RV Internals component inspections for.VT-3 and UT examinations. The TJs, where appropriate, will include further guidance with respect to examination coverage and relevant conditions.

Relevant conditions requiring corrective action for ASME B&PV Code Section XI Examination Category B-N-3 VT-3 examinations of RV Internals components are detailed in ASME B&PV Code Section XI, IWB-3520.

Any detected condition that does not satisfy these examination acceptance criteria must be dispositioned. See Section 5.6.2 of this report for a discussion of functionality/engineering acceptance criteria.

5.6.2 Functionality/Engineering Acceptance Criteria Based on the identified condition and supplemental examinations, if required, the disposition process results in an evaluation and determination of whether to accept the condition until the next examination or repair or replace the item. Relevant conditions identified during ASME B&PV Code Section XI Examination Category B-N-3 VT-3 examinations of RV Internals components are evaluated per ASME B&PV Code Section XI, IWB-3142. For augmented MRP-227, Rev. 0 inspections, the results of the PWROG projects discussed below will allow for the development of acceptance criteria used to determine whether to accept the condition until the next examination or repair or replace the components.

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ANP-2951, Rev. 001 The PWROG PA-MSC-0473 developed WCAP-17096, "Reactor Internals Acceptance Criteria Methodology and Data Requirements". For each of the "Primary" and "Expansion" components listed in MRP-227, Rev. 0, WCAP-17096 outlines the type of analyses required, required evaluation procedures, data required to support analysis, logic chart illustrating evaluation path and potential disposition options, and components that can be addressed on a generic basis.

PWROG PA-MSC-0350 developed the basis for functional acceptance criteria that consisted of researching and documenting the functional (design basis) requirements for the UCB and LCB bolts. Models that represent the UCB and LCB bolted connections of the RV Internals of the ONS units were developed to evaluate and disposition inspection results.

PWROG PA-MSC-0692 will develop the basis for functional acceptance criteria which will consist of researching and documenting the functional (design basis) requirements for selected items and loading conditions that may exist during operation. The locking devices for the bolts do not require a functionality analysis. The existing core barrel assembly (CBA) supermodel developed under PWROG PA-MSC-0350 will be modified for specific differences that are anticipated to cover the variations in the B&W-designed RV Internals. This will provide unit-specific tools that can be used to analyze the condition of the CBA. The items included in the model are (a) the "Primary" items baffle-to-former bolts and baffle plates; and (b) the "Expansion" items core barrel cylinder, former plates, baffle-to-baffle bolts, core barrel-to-former bolts. A safety assessment of the consequence of failed inaccessible bolts (core barrel-to-former bolts, baffle-to-baffle bolts) and associated locking devices will also be performed. Task 4 of this PA is to "Develop Plan and Phase III Scope Definition" for future work.

5.6.3 Conclusions Acceptance criteria for the ONS RV Internals inspection plan, against which the need for corrective action will be evaluated, will ensure the particular structure and component intended functions are maintained under all CLB design conditions during the period of extended operation.

5.7 AMP Element 7 - Corrective Actions In accordance with 10 CFR 50, Appendix B, Duke Energy has established a corrective action program for the ONS units.

Measures are established in the ONS corrective action program to ensure conditions adverse to quality are promptly identified and corrected. In the case of significant conditions adverse to quality, the measures ensure the cause of the condition is determined and corrective action taken to preclude repetition. The identification of the significant condition adverse to quality, the cause of the condition, and the corrective action taken shall be documented and reported to appropriate levels of management.

5.7.1 ONS PIP Program Duke Energy uses the PIP program and the Work Management System (WMS) to implement its Corrective Action Program per 10 CFR 50, Appendix B. The PIP program for the ONS units includes topics such as responsibilities, timeliness guidelines, action categories, reportability, and problem evaluation. The WMS and corrective maintenance function are established to identify and resolve the normal and expected degradation.

5.7.2 ONS Root Cause Cause analysis is an essential part of an effective corrective action program. Root cause analysis prevents repetitive or similar problems by the identification and correction of specific causes of failures. The ONS cause analysis program provides a systematic approach to identify the fundamental reason or cause for a problem that has occurred and includes topics such as responsibilities, qualifications of personnel, cause analysis process, and record retention requirements. This directive is used in conjunction with the ONS PIP program on degraded conditions identified as needing a root cause analysis.

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ANP-2951, Rev. 001 5.7.3 Operability The ONS Operability/Functionality program applies to degraded/non-conforming conditions and unanalyzed conditions associated with structures, systems, and components (SSCs) that perform specified functions as set forth in the CLB. This directive compliments the guidance in the ONS PIP program for the resolution of degraded and/or nonconforming conditions.

5.7.4 Conclusion Corrective actions, including root cause determination and prevention of recurrence, are timely.

5.8 AMP Element 8 - Confirmation Process Duke Energy Topical Report Duke-I-A, "Quality Assurance Program" [241 describes the Duke Energy Quality Assurance (QA) program for the operational phase of its nuclear power plants. The Duke Energy QA program conforms to applicable regulatory requirements such as 10 CFR 50, Appendix B and to approved industry standards such as ANSI N45.2-1977 and ANSI N 18.7-1976 and corresponding daughter standards or to equivalent alternatives. Duke Energy regularly reviews the status and adequacy of the QA program.

At ONS, Independent Nuclear Oversight (INOS) is assigned the QA functions of (1) assuring that an appropriate QA program is established and effectively executed; and (2) verifying, such as by checking, auditing, and inspecting, that activities affecting the safety related functions have been correctly performed.

5.8.1 Conclusion Duke Energy's QA program ensures preventative actions are adequate and appropriate corrective actions have been completed and are effective.

5.9 AMP Element 9 - Administrative Controls Administrative controls, including Duke Energy-specific documents used to implement this ONS RV Internals inspection plan, provide for a formal review and approval process.

5.9.1 Conclusion Duke Energy's administrative controls provide for a formal review and approval process.

5.10 AMP Element 10- Operating Experience MRP-227, Rev. 0 establishes new augmented inspection guidance for the B&W units. Accordingly, there is no direct programmatic history for the bulk of these inspections. The program is based upon industry OE that is contained in MRP-227, Rev. 0, MRP-23 1, research data, and vendor evaluations. Development of the program relied upon the consensus review and inputs of the MRP Reactor Internals Core and Focus Groups, which include representatives from utilities, research scientists, and vendors. The MRP-227, Rev. 0 guidelines will continue to evolve as additional operating experience and augmented inspection results are gained. PWR RV Intemals failures, both domestic and international, have been considered in the development of MRP-227, Rev. 0.

The ONS Operating Experience Program (OEP) defines and communicates Duke Energy's OE Program and management expectations for the use of OE information. This program also defines and communicates expectations for the receipt, evaluation, and distribution of OE information and the resolution of applicable OE items.

5.10.1 Incidents of Degradation in B&W RV Internals Relatively few incidents of PWR internals aging degradation have been reported in operating U.S. commercial PWR plants. In B&W units, the incidents observed have been limited to the baffle-to-baffle bolts (see Section 4.2.9 of this report), RV Internals structural bolting (see Sections 4.1.4 and 4.2.7 of this report), locking devices for the vent valve jackscrews (see Section 4.2.6.1 of this report), and interaction between the baffle plates and fuel assembly grid straps (see Section 4.1.6 of this report).

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ANP-2951, Rev. 001 5.10.2 Conclusion Operating experience has been used in the development of the ONS RV Internals inspection plan, thus ensuring the effects of aging will be adequately managed so the structure and component intended functions described in Section 2.4 of this report will be maintained during the period of extended operation.

5.11 Program Conclusion Section 5.0 of this report shows that the ONS RV Internals inspection plan complies with the ten AMP elements from NUREG-1801, Rev. 1; compliance with these ten AMP elements demonstrates adequacy of managing aging effects of the ONS RV Internals.

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ANP-2951, Rev. 001 6.0

SUMMARY

AND CONCLUSIONS This report documents and provides a description of the ONS-1, ONS-2, and ONS-3 RV Internals inspection plan and how it relates to the RV Internals AMP at ONS for management of aging effects consistent with previous commitments. This ONS RV Internals inspection plan is based on MRP-227, Rev. 0, as modified by the ONS unit-specific amendments identified in Section 4.2.10 of this report. Section 5.0 of this report has demonstrates that the ONS RV Internals AMP, which contains the ONS RV Internals inspection plan, will comply with the ten AMP elements described in NUREG-l1801, Rev. 1.

This ONS RV Internals inspection plan contains a discussion of the background of the B&W-designed plant RV Internals programs, including operational experience, TLAAs, and existing ONS programs.

The examinations required by ASME B&PV Code Section XI, MRP-227, Rev. 0, and ONS unit-specific amendments as described in this report have been scheduled to be performed at ONS during the 2012, 2013, and 2014 RFOs, and any relevant conditions will be documented and dispositioned in Duke Energy's corrective action program and reported to the industry.

The ONS RV Internals AMP will include this ONS RV Internals inspection plan and will demonstrate that the program adequately manages the effects of aging for RV Internals components and establishes the basis for providing reasonable assurance that the RV Internals components will remain functional through the ONS license renewal period of extended operation.

This ONS RV Internals inspection plan, along with the ONS RV Internals AMP, fulfills the approved license renewal methodology requirement to identify the most susceptible components and to inspect those components with an indication detection level commensurate with the expected degradation mechanism indication, thus meeting 10 CFR 54.

Once this ONS RV Internals inspection plan is approved, the ONS UFSAR will be updated as required.

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ANP-2951, Rev. 001

7.0 REFERENCES

1. Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-Rev.0), EPRI, Palo Alto, CA: 2008. 1016596.

2. NUREG-1801, Revision 1, "Generic Aging Lessons Learned (GALL) Report," Dated September 2005.

3. Letter from Duke Energy Corporation forwarding application for renewal of operating licenses for the Oconee Nuclear Station, Unit Nos. 1, 2, and 3. Requests extension of operating licenses to 20 years beyond current expiration dates. July 6, 1998. NRC Accession No.: 9807200136

4. Duke Energy Corporation, Oconee Nuclear Station Units 1, 2, and 3 Renewed Facility Operating License," May 23, 2000. NRC Accession No.: ML013510445

5. U. S. Nuclear Regulatory Commission, "Safety Evaluation Report Related to the License Renewal of Oconee Nuclear Station, Units 1, 2, and 3," NUREG- 1723, March 31, 2000. NRC Accession No.:

ML003697717

6. SECY-00-0081, "Oconee Nuclear Station, Units 1, 2, and 3 -Renewal of Full-Power Operating License," April 10, 2000. NRC Accession No.: ML003693303

7. BAW-2248, "Demonstration of the Management of Aging Effects for the Reactor Vessel Internals,"

July 1997. NRC Accession No.: ML003708443

8. U. S. Nuclear Regulatory Commission Letter, "Acceptance for Referencing of Generic License Renewal Program Topical Report Entitled, 'Demonstration of the Management of Aging Effects for the Reactor Vessel Internals'," BAW-2248, July 1997. NRC Accession No.: ML993490288

9. BAW-2248A, "Demonstration of the Management of Aging Effects for the Reactor Vessel Internals,"

April 2000. NRC Accession No.: ML003708443

10. Duke Power Company, Oconee Nuclear Station Units 1, 2, and 3 Reactor Coolant System Aging Management Review for License Renewal, OSS-0274.00-00-0004, Revision 1, August 2001.

11. Letter (from Christian Larsen) Report Transmittal: Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-Rev. 0), EPRI Palo Alto, CA, 2008, 1016596 (NRC Accession No. ML090160204), January 2009.

12. American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, Rules for In-Service Inspection of Nuclear Power Plant Components, 1998 Edition with 2000 Addenda, American Society of Mechanical Engineers, New York, NY.

13. Letter (from D. Baxter) Letter of Intent to adopt Materials Reliability Program 227, Pressurized Water Reactor Internals Inspection and Evaluation Guidelines, June 16, 2010.

14. Nuclear Energy Institute, "Guideline for the Management of Materials Issues," NEI 03-08, Revision 2, January 2010.

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ANP-2951, Rev. 001

15. Materials Reliability Program: Inspection Standard for Reactor Internals (MRP-228). EPRI, Palo Alto, CA: 2009. 1016609.

16. BAW-10008, Part 1, Revision 1, "Reactor Internals Stress and Deflection Due to Loss-of-Coolant Accident and Maximum Hypothetical Earthquake," Babcock & Wilcox, June 1970.

17. NRC letter (from D. Matthews) forwards safety evaluation report regarding licensee's July 6, 1998, application to NRC for renewal of Oconee Nuclear Stations 1, 2, and 3 operating license for additional 20 years. Open items must be resolved before NRC can make final determination on application. NRC Accession No. 9906210071.

18. Memorandum (signed by: J. Sebrosky) forwards Duke editorial comments on June 1996 SER related to license renewal application of Oconee Nuclear Station Units 1, 2, and 3. Duke intends to provide additional comments on SER of more substantial nature in separate letter. NRC Accession No.: 9910220085

19. Materials Reliability Program: Stress Corrosion Cracking of High Strength Reactor Vessel Internals Bolting in PWRs (MRP-88), EPRI, Palo Alto, CA: 2003. 1003206.

20. "Briefing Material from June 14, 2005 Mtg on Rx Vessel Internals from the EPRI MRP - HT Tang",

June 14, 2005. NRC Accession No. ML051710058.

21. Pressurized Water Reactor Primary Water Chemistry Guidelines: Volume 1, Revision 5, EPRI, Palo Alto, CA: 2003. 1002884.

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

23. NUREG/CR-4513, Rev. 01, "Estimation of Fracture Toughness of Cast Stainless Steels During Thermal Aging in LWR Systems," U.S. Nuclear Regulatory Commission, May 1994.

24. Duke Energy Carolinas Topical Report, Duke-l-A, "Quality Assurance Program," Revision 37, February 2010. NRC Accession No. ML100610381 Page 41 of 108

ANP-2951, Rev. 001 APPENDIX A: BAW-2248A AMR AND INDUSTRY PROGRAM COMPARISON Table A. l in Appendix A shows how the components identified in the BAW-2248A, Table 4-1 AMR were evaluated and characterized by the industry program.

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ANP-2951, Rev. 001 Table A.1. AMR and MRP-227, Rev. 0 Comparison Affected Parts Aging Effect Programs That Manage Applicable Aging Effects PLENUM ASSEMBLY Upper Grid Assembly Upper Grid Assembly Cracking of base metal and ASME Section XI ISI Programs welds Examination Category B-N-3, Removable Core Support Structures VT-3 visual examination of accessible surfaces - Every interval RVI Aging Management Program See Section 4.6.

MRP Category A, except for the No Additional Measures X-750 dowel locking weld MRP-189 Rev. L(Tables 4-1 and 4-2) and MRP-231, Rev, 1 (Tables 3-8 and 3-9)

Reduction of Fracture ASME Section XI ISI Programs Toughness of base metal and Examination Category B-N-3, Removable Core Support Structures welds VT-3 visual examination of accessible surfaces - Every interval RVI Aging Management Program See Section 4.6.

MRP Category A, MRP-189 Rev. 1 Tables 4-1 and 4-2 Fuel Assembly Support Pads Loss of Material ASME Section XI ISI Programs Examination Category B-N-3, Removable Core Support Structures VT-3 visual examination of accessible surfaces - Every interval.

MRP Category A, except for the Expansion X-750 dowel locking weld MRP-189 Rev. I (Tables 4-1 and 4-2) and MRP-231, Rev. 1 (Tables 3-8 and 3-9)

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ANP-2951, Rev. 001 Table A.1. AMR and MRP-227, Rev. 0 Comparison Affected Parts Aging Effect Programs That Manage Applicable Aging Effects PLENUM ASSEMBLY (continued)

Plenum Rib Pads Loss of Material ASME Section XI ISI Programs Examination Category B-N-3, Removable Core Support Structures VT-3 visual examination of accessible surfaces - Every interval.

MRP Primary, MRP-231, Rev. 1 (Tables 3-8 and 3-9)

Control Rod Guide Tube Assembly CRGT Flange to Upper Grid Loss of Closure Integrity ASME Section XI ISI Programs Screws Examination Category B-N-3, Removable Core Support Structures VT-3 visual examination of accessible surfaces - Every interval RVI Aging Management Program See Section 4.6.

MRP Category A, MRP-189 Rev. 1 (Table 4-1)

CRGT Assembly Spacer Reduction of Fracture RVI Aging Management Program Castings Toughness See Section 4.6.

MRP Expansion, MRP-23 1, Rev. 1 (Tables 3-8 and 3-10)

/

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ANP-2951, Rev. 001 Table A.1. AMR and MRP-227, Rev. 0 Comparison Affected Parts Aging Effect Programs That Manage Applicable Aging Effects CORE SUPPORT SHIELD ASSEMBLY Core Support Shield Cylinder Top Flange Loss of Material ASME Section XI ISI Programs Examination Category B-N-3, Removable Core Support Structures VT-3 visual examination of accessible surfaces - Every interval.

MRP Primary, MRP-231, Rev. 1 (Tables 3-8 and 3-9)

Core Support Shield to Core Barrel Bolts Cracking ASME Section XI ISI Programs Examination Category B-N-3, Removable Core Support Structures VT-3 visual examination of accessible surfaces - Every interval RVI Aging Management Program See Section 4.6.

MRP Primary, MRP-23 1, Rev. 1 (Tables 3-8 and 3-9)

Reduction of Fracture ASME Section XI ISI Programs Toughness Examination Category B-N-3, Removable Core Support Structures VT-3 visual examination of accessible surfaces - Every interval RVI Aging Management Program See Section 4.6.

MRP Category A, MRP-189 Rev. 1 (Table 4-1)

Loss of Closure Integrity ASME Section XI ISI Programs Examination Category B-N-3, Removable Core Support Structures VT-3 visual examination of accessible surfaces - Every interval RVI Aging Management Program See Section 4.6.

MRP Category A, MRP- 189 Rev. 1 (Table 4-1)

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ANP-2951, Rev. 001 Table A.1. AMR and MRP-227, Rev. 0 Comparison Affected Parts Aging Effect Programs That Manage Applicable Aging Effects CORE SUPPORT SHIELD ASSEMBLY (continued)

ONS-3 CSS Outlet Nozzles Reduction 'of Fracture ASME Section XI ISI Programs Toughness Examination Category B-N-3, Removable Core Support Structures VT-3 visual examination of accessible surfaces - Every interval Supplemented by extension of the evaluation procedures in IWB-3640 to CASS items as specified in Section 4.2 of BAW-2243A [30].

RVI Aging Management Program See Section 4.6.

MRP Primary, MRP-231, Rev. 1 (Tables 3-8 and 3-9)

Notes:

DB's outlet nozzles are also made of casting, but were not listed because DB was not a participant of BAW-2248A.

The cast outlet nozzles at ONS-3 and DB have been determined to be below the NUREG/CR-4513 Rev. 1 screening level for thermal aging after the issuance of MRP-227.

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ANP-2951, Rev. 001 Table A.1. AMR and MRP-227, Rev. 0 Comparison Affected Parts Aging Effect Programs That Manage Applicable Aging Effects CORE SUPPORT SHIELD ASSEMBLY (continued)

Vent Valve Bodies Reduction of Fracture ASME Section XI ISI Programs Toughness Examination Category B-N-3, Removable Core Support Structures VT-3 visual examination of accessible surfaces - Every interval.

Supplemented by:

Plant Technical Specifications (ANO-1. TMI-1)

Vent valve testing and inspection requirements each refueling

-outage Pump and Valve In-Service Test Program (ONS-1,-2,-3)

Vent valve testing and inspection requirements each refueling outage Further Supplemented by extension of the evaluation procedures in IWB-3640 to CASS items as specified in Section 4.2 of BAW-2243A [30].

RVI Aging Management Program See Section 4.6.

MRP Category A, MRP-189 Rev: 1 (Table 4-1)

Note:

The vent valve body castings have been confirmed to be below the NUREG/CR-4513 Rev. 1 screening level for thermal aging.

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ANP-2951, Rev. 001 Table A. 1. AMR and MRP-227, Rev. 0 Comparison

\Affected Parts Aging Effect Programs That Manage Applicable Aging Effects CORE SUPPORT SHIELD ASSEMBLY (continued)

Vent Valve Retaining Rings Reduction of Fracture ASME Section XI ISI Programs Toughness Examination Category B-N-3, Removable Core Support Structures VT-3 visual examination of accessible surfaces - Every interval.

Supplemented by:

Plant Technical Specifications (ANO- 1, TMI- 1)

Vent valve testing and inspection requirements each refueling outage Pump and Valve In-Service Test Program (ONS-1,-2,-3)

Vent valve testing and inspection requirements each refueling outage RVI Aging Management Program See Section 4.6.

MRP Primary, MRP-231, Rev. 1 (Tables 3-8 and 3-9)

[both top and bottom retaining rings]

Vent Valve Locking Devices (Modified) Cracking ASME Section XI ISI Programs Examination Category B-N-3, Removable Core Support Structures VT-3 visual examination of accessible surfaces - Every interval.

MRP "No Additional Measures" (see Section 4.2.10), MRP-23 1, Rev. 1 (Table 3-9 Note 1)

Supplemented by:

Plant Technical Specifications (ANO- 1, TMI- 1)

Vent valve testing and inspection requirements each refueling outage Pump and Valve In-Service Test Program (ONS-1,-2,-3)

Vent valve testing and inspection requirements each refueling outage Page 48 of 108

ANP-2951, Rev. 001 Table A.1. AMR and MRP-227, Rev. 0 Comparison Affected Parts Aging Effect Programs That Manage Applicable Aging Effects CORE SUPPORT SHIELD ASSEMBLY (continued)

Vent Valve Locking Devices (Original) Loss of Material ASME Section XI ISI Programs Examination Category B-N-3, Removable Core Support Structures VT-3 visual examination of accessible surfaces - Every interval.

MRP "No Additional Measures", MRP-23 1, Rev. 1 (Table 3-9 Note 1)

Supplemented by:

Plant Technical Specifications (ANO- 1, TMI- 1)

Vent valve testing and inspection requirements each refueling outage Pump and Valve In-Service Test Program (ONS- 1,-2,-3)

Vent valve testing and inspection requirements each refueling outage CORE BARREL ASSEMBLY Core Barrel Assembly Cracking of base metal and ASME Section XI ISI Programs welds Examination Category B-N-3, Removable Core Support Structures VT-3 visual examination of accessible surfaces - Every interval RVI Aging Management Program See Section 4.6.

MRP Primary and Expansion, MRP-231, Rev. I (Tables 3-8, 3-9, and 3-10)

Reduction of Fracture ASME Section XI ISI Programs Toughness of base metal and Examination Category B-N-3, Removable Core Support Structures welds VT-3 visual examination of accessible surfaces - Every interval RVI Aging Management Program See Section 4.6.

MRP Primary and Expansion, MRP-231, Rev. 1 (Tables 3-8, 3-9, and 3-10)

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ANP-2951, Rev. 001 Table A.1. AMR and MRP-227, Rev. 0 Comparison Affected Parts Aging Effect Programs That Manage Applicable Aging Effects CORE BARREL ASSEMBLY (continued)

Dimensional Change RVI Aging Management Program MRP See Section 4.6.

MRP No Additional Measure, MRP-231, Rev. 1 (Tables 3-8)

Baffle-to-Baffle Bolts Cracking ASME Section XI ISI Programs Examination Category B-N-3, Removable Core Support Structures VT-3 visual examination of accessible surfaces - Every interval RVI Aging Management Program See Section 4.6.

MRP No Additional Measure and Expansion, MRP-231, Rev. 1 (Tables 3-8 and 3-10)

Reduction of Fracture ASME Section XI ISI Programs Toughness Examination Category B-N-3, Removable Core Support Structures VT-3 visual examination of accessible surfaces - Every interval RVI Aging Management Program See Section 4.6.

MRP Expansion, MRP-231, Rev. 1 (Tables 3-8 and 3-10)

Loss of Closure Integrity ASME Section XI ISI Programs Examination Category B-N-3, Removable Core Support Structures VT-3 visual examination of accessible surfaces - Every interval RVI Aging Management Program See Section 4.6.

Expansion, MRP-231, Rev. 1 (Tables 3-8 and 3-10)

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ANP-2951, Rev. 001 Table A.1. AMR and MRP-227, Rev. 0 Comparison Affected Parts Aging Effect Programs That Manage Applicable Aging Effects CORE BARREL ASSEMBLY (continued)

Dimensional Change RVI Aging Management Program See Section 4.6.

MRP No Additional Measure, MRP-23 1, Rev. 1 (Tables 3-8)

Baffle-to-Former Bolts Cracking ASME Section XI ISI Programs Examination Category B-N-3, Removable Core Support Structures VT-3 visual examination of accessible surfaces - Every interval RVI Aging Management Program See Section 4.6.

MRP Primary, MRP-231, Rev.. 1 (Tables 3-8 and 3-9)

Reduction of Fracture ASME Section XI ISI Programs Toughness Examination Category B-N-3, Removable Core Support Structures VT-3 visual examination of accessible surfaces - Every interval RVI Aging Management Program See Section 4.6.

MRP Primary, MRP-231, Rev. 1 (Tables 3-8 and 3-9)

Page 51 of 108

ANP-2951, Rev. 001 Table A.1. AMR and MRP-227, Rev. 0 Comparison Affected Parts Aging Effect Programs That Manage Applicable Aging Effects CORE BARREL ASSEMBLY (continued)

Loss of Closure Integrity ASME Section XI ISI Programs Examination Category B-N-3, Removable Core Support Structures VT-3 visual examination of accessible surfaces - Every interval RVI Aging Management Program See Section 4.6.

MRP Primary, MRP-23 1, Rev. 1 (Tables 3-8 and 3-9)

Dimensional Change RVI Aging Management Program See Section 4.6.

MRP No Additional Measure, MRP-23 1, Rev. 1 (Tables 3-8)

Lower Internals Assembly to Core Barrel Cracking ASME Section XI ISI Programs Bolts Examination Category B-N-3, Removable Core Support Structures VT-3 visual examination of accessible surfaces - Every interval RVI Aging Management Program See Section 4.6.

MRP Primary, MRP-231, Rev. 1 (Tables 3-8 and 3-9)

Page 52 of 108

ANP-2951, Rev. 001 Table A.1. AMR and MRP-227, Rev. 0 Comparison Affected Parts Aging Effect Programs That Manage Applicable Aging Effects CORE BARREL ASSEMBLY (continued)

Reduction of Fracture ASME Section XI ISI Programs Toughness Examination Category B-N-3, Removable Core Support Structures VT-3 visual examination of accessible surfaces - Every interval RVI Aging Management Program See Section 4.6.

MRP Category A, MRP- 189 Rev. I (Table 4-I)

Loss of Closure Integrity ASME Section XI ISI Programs Examination Category B-N-3, Removable Core Support Structures VT-3 visual examination of accessible surfaces - Every interval RVI Aging Management Program See Section 4.6.

MRP Category A, MRP- 189 Rev. 1 (Table 4-1)

Core Barrel to Thermal Shield Bolts Cracking ASME Section XI ISI Programs Examination Category B-N-3, Removable Core Support Structures VT-3 visual examination of accessible surfaces - Every interval RVI Aging Management Program See Section 4.6.

MRP Expansion, MRP-231, Rev. 1 (Tables 3-8 and 3-10)

Page 53 of 108

ANP-2951, Rev. 001 Table A.I. AMR and MRP-227, Rev. 0 Comparison Affected Parts Aging Effect Programs That Manage Applicable Aging Effects CORE BARREL ASSEMBLY (continued)

Loss of Closure Integrity ASME Section XI ISI Programs Examination Category B-N-3, Removable Core Support Structures VT-3 visual examination of accessible surfaces - Every interval RVI Aging Managzement Program See Section 4.6.

MRP Category A,MRP-189 Rev. I (Table 4-1)

LOWER INTERNALS ASSEMBLY Lower Grid Assembly Cracking of base metal and ASME Section XI ISI Programs welds Examination Category B-N-3, Removable Core Support Structures VT-3 visual examination of accessible surfaces - Every interval RVI Aging Management Program See Section 4.6.

MRP Note: The dowel locking weld for the guide blocks are "Primary" for SCC.

MRP-231, Rev. I (Tables 3-8 and 3-9)

Category A, except for the "Expansion" for the dowel locking welds for the lower grid fuel assembly support pads MRP-189 Rev. 1 (Tables 4-1 and 4-2) and MRP-231, Rev. 1 (Tables 3-8 and 3-10)

Page 54 of 108

ANP-2951, Rev. 001 Table A.1. AMR and MRP-227, Rev. 0 Comparison Affected Parts Aging Effect Programs That Manage Applicable Aging Effects LOWER INTERNALS ASSEMBLY (continued)

Reduction of Fracture ASME Section XI ISI Programs Toughness of base metal and Examination Category B-N-3, Removable Core Support Structures welds VT-3 visual examination of accessible surfaces - Every interval RVI Aging Management Program See Section 4.6.

MRP Category A, except for the for the "Expansion" items and welds associated with the lower grid fuel assembly support pads MRP-189 Rev. 1 (Tables 4-1 and 4-2) and MRP-231, Rev. 1 (Tables 3-8 and 3-10)

Lower Internals Assembly to Thermal Cracking ASME Section XI ISI Programs Shield Bolts Examination Category B-N-3, Removable Core Support Structures VT-3 visual examination of accessible surfaces - Every interval RVI Aging Management Program See Section 4.6.

MRP Expansion, MRP-231, Rev. 1 (Tables 3-8 and 3-10)

Loss of Closure Integrity ASME Section XI ISI Programs Examination Category B-N-3, Removable Core Support Structures VT-3 visual examination of accessible surfaces - Every interval RVI Aging Management Program See Section 4.6.

MRP Category A,MRP-189 Rev. I (Table 4-1)

Page 55 of 108

ANP-2951, Rev. 001 I Table A.i. AMR and MRP-227, Rev. 0 Comparison Affected Parts Aging Effect Programs That Manage Applicable Aging Effects LOWER INTERNALS ASSEMBLY (continued)

Fuel Assembly Support Pads Loss of Material ASME Section XI ISI Programs Examination Category B-N-3, Removable Core Support Structures VT-3 visual examination of accessible surfaces - Every interval.

MRP Category A, MRP-189 Rev. 1 (Table 4-1)

Note: The items and welds associated with the lower grid fuel assembly support pads are "Expansion" for SCC and/or irradiation embrittlement.

MRP-231, Rev. 1 (Tables 3-8 and 3-10)

Guide Blocks Loss of Material ASME Section XI ISI Programs Examination Category B-N-3, Removable Core Support Structures VT-3 visual examination of accessible surfaces - Every interval.

MRP Category A, MRP-189 Rev. 1 (Table 4-1)

Note: The dowel locking weld for the guide blocks are "Primary" for SCC.

MRP-231, Rev. 1 (Tables 3-8 and 3-9)

Shell Forging to Flow Distributor Bolts Cracking ASME Section XI ISI Programs Examination Category B-N-3, Removable Core Support Structures VT-3 visual examination of acceisible surfaces - Every interval RVI Agin2 Management Program See Section 4.6.

MRP Expansion, MRP-231, Rev. 1 (Tables 3-8 and 3-10)

Loss of Closure Integrity ASME Section XI ISI Programs Examination Category B-N-3, Removable Core Support Structures VT-3 visual examination of accessible surfaces - Every interval RVI Aging Management Program See Section 4.6.

MRP Category A, MRP-189 Rev. 1 (Table 4-1)

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ANP-2951, Rev. 001 Table A.1. AMR and MRP-227, Rev. 0 Comparison Affected Parts Aging Effect Programs That Manage Applicable Aging Effects LOWER INTERNALS ASSEMBLY (continued)

Lower Grid Rib to Shell Forging Screws Loss of Closure Integrity ASME Section XI ISI Programs Examination Category B-N-3, Removable Core Support Structures VT-3 visual examination of accessible surfaces - Every interval RVI Aging Management Program See Section 4.6.

MRP No Additional Measure, MRP-23 I, Rev. I (Tables 3-8)

Incore Guide Tube Spider Castings Reduction of Fracture RVI Aging Management Program Toughness See Section 4.6.

MRP Primary, MRP-231, Rev. 1 (Tables 3-8 and 3-9)

Page 57 of 108

ANP-2951, Rev. 001 APPENDIX B: TABLE 4-1 FROM MRP-227, REV. 0 Appendix B contains Table 4-1, "B&W Plants Primary Components" from MRP-227, Rev. 0. The "Primary" component inspections listed in the ONS RV Internals inspection plan are based on this table, as modified by the ONS unit-specific amendments identified in Section 4.2.10 and Appendix F of this report, Page 58 of 108

ANP-2951, Rev. 001 Item Applicability Effect (Mechanism) Expansion Link Examination Frequency Examination Coverage (Note 2) Examination Frequency Examination Coverage Plenum Cover Assembly & All plants Loss of material and None On-time physical Determination of differential Core Support Shield associated loss of core measurement no later than height of top of plenum rib Assembly clamping pre-load (Wear) two refueling outages from pads to reactor vessel seating Plenum cover weldment rib the beginning of the license surface, with plenum in pads renewal period, reactor vessel.

Plenum cover support flange CSS top flange Perform subsequent visual (VT-3) examination on the 10-year ISI interval.

Core Support Shield ONS-3, DB Cracking (TE), including CRGT spacer Visual (VT-3) examination 100% of accessible surfaces.

Assembly the detection of surface - castings during the next 10-year ISI.

CSS cast outlet nozzles irregularities, such as Core Support Shield All plants damaged or fractured Subsequent examination on 100% of accessible surfaces.

Assembly material the 10-year ISI interval.

CSS vent valve discs (Note 1)

Core Support Shield All plants Cracking (TE), including None Visual (VT-3) examination 100% of accessible surfaces.

Assembly the detection of surface during the next 10-year ISI.

CSS vent valve top retaining irregularities, such as ring damaged, fractured, or Subsequent examinations on CSS vent valve bottom missing items the 10-year ISI interval.

retaining ring CSS vent valve disc shaft or hinge pin (Note 1)

Page 59 of 108

ANP-2951, Rev. 001 Item Applicability Effect (Mechanism) Expansion Link Examination Frequency Examination Coverage

_______ ________ (Note 2)

Core Support Shield All plants Cracking (SCC) LCB bolts Volumetric examination 100% of accessible bolts.

Assembly (Note 3) (UT) of the bolts within two Upper core barrel (UCB) refueling outages from bolts and their locking UTS, LTS, and FD 1/1/2006 or next 10-year ISI devices bolts interval, whichever is first.

SSHT bolts (CR-3 Subsequent examination to be determined after and DB only) evaluation the baseline results.

Lower shock pad bolts (TMI-1 only) Visual (VT-3) examination of bolt locking devices on the 10-year ISI interval.

Core Barrel Assembly All plants Cracking (SCC) UTS, LTS, and FD Volumetric examination 100% of accessible bolts.

Lower core barrel (LCB) bolts (UT) of the bolts during the bolts and their locking next 10-year ISI interval devices SSHT bolts (CR-3 from 1/1/2006.

and DB only)

Subsequent examination to Lower shock pad be determined after bolts (TMI-a only) evaluation the baseline results.

Visual (VT-3) examination of bolt locking devices on the 10-year ISI interval.

Core Barrel Assembly All plants Cracking (IASCC, IE, Baffle-to-baffle Baseline volumetric 100% of accessible bolts.

Baffle-to-former bolts IC/ISR/FatiguelWear, bolts examination (UT) no later Overload) Core barrel-to- than two refueling outages former bolts from the beginning of the license renewal period with subsequent examination after 10 to 15 additional years.

Page 60 of 108

ANP-2951, Rev. 001 Item Applicability Effect (Mechanism) Expansion Link Examination Frequency Examination Coverage (Note 2) Examination Frequency Examination Coverage Core Barrel Assembly All plants Cracking (IE), including Core barrel Visual (VT-3) examination 100% of the accessible Baffle plates the detection of readily cylinder (including during the next 10-year ISI. surface within 1 inch around detectable cracking in the vertical and each flow and bolt hole.

baffle plates circumferential Subsequent examinations on seam welds) the 10-year ISI interval.

Former plates Core Barrel Assembly All plants Cracking (IASCC, IE, Locking devices Visual (VT-3) examination 100% of accessible baffle-to-Locking devices, including Overload), including the for the external during the next 10-year ISI. former and internal baffle-to-locking welds, of baffle-to- detection of missing, non- baffle-to-baffle baffle bolt locking devices.

former bolts and internal functional, or removed bolts and baffle-to- Subsequent examinations on baffle-to-baffle bolts locking devices or welds former bolts the 10-year ISI interval.

Lower Grid Assembly All plants Cracking (SCC), Alloy X-750 dowel Initial visual (VT-3) 100% of accessible locking Alloy X-750 dowel-to-guide including the detection of locking welds to examination no later than welds of the 24 dowel-to-block welds separated or missing the upper and two refueling outages from guide block welds.

locking welds, or missing lower fuel the beginning of the license dowels assembly support renewal period.

pads Subsequent examination on ten-year interval.

I ncore Monitoring All plants Cracking (TE/IE), CRGT spider Initial visual (VT-3) 100% of accessible top Instrumentation (IMI) including the detection of castings examination no later than surfaces of 52 spider castings Guide Tube Assembly fractured or missing two refueling outages from and welds to the adjacent IMI guide tube spiders spider arms or separation Lower fuel the beginning of the license lower grid rib section.

IMI guide tube spider-to- of spider arms from the assembly support renewal period.

lower grid rib section welds lower grid rib section at pad items: pad, the weld pad-to-rib section Subsequent examinations on welds, Alloy X- ten-year. interval.

750 dowel, cap screw, and their locking welds (Note: the pads, dowels, and cap screws are included because of TE/IE of the welds)

Page 61 of 108

ANP-2951, Rev. 001 Notes:

1. A verification of the operation of each vent valve shall also be performed through manual actuation of the valve. Verify that the valves are not stuck in the open position and that no abnormal degradation has occurred. Examine the valves for evidence of scratches, pitting, embedded particles, variation in coloration of the seating surfaces, cracking of lock welds and locking cups, jack screws for proper position, and wear. The frequency is defined in each unit's technical specifications or in their pump and valve in-service test programs (see Appendix A).

2. Examination acceptance criteria and expansion criteria for the B&W components are in Table 5-1 (of MRP-227, Rev. 0).

3. Expansion to LCB applies if the required Primary examination of LCB has not been performed as scheduled in this table.

Page 62 of 108

ANP-2951, Rev. 001 APPENDIX C: TABLE 4-4 FROM MRP-227, REV. 0 Appendix C contains Table 4-4, "B&W Plants Expansion Components" from MRP-227, Rev. 0. The "Expansion" component inspections listed in the ONS RV Internals inspection plan are based on this table, as modified by the ONS unit-specific amendments identified in Section 4.2.10 and Appendix F of this report.

Page 63 of 108

ANP-2951, Rev. 001 Item'. Applicability Effect (Mechanism) Primary Link (Note 1) Examination MethodCoverage

_ __ _ _ __ _ _ _ _ _ _ _ _ _ _ __ __ _(Note 1) ExaminationCoverage Upper Grid Assembly All plants Cracking (SCC) Alloy X-750 dowel-to- Visual (VT-3) 100% of accessible dowel Alloy X-750 dowel-i6-upper (except DB) including the detection of guide block welds examination, locking welds.

fuel assembly support pad welds separated or missing locking welds, or missing dowels Control Rod Guide Tube All plants Cracking (TE), including CSS cast outlet nozzle, Visual (VT-3) 100% of accessible surfaces Assembly the detection of fractured CSS vent valve disks, or examination, at the 4 screw locations CRGT spacer castings spacers or missing screws IMI guide tube spiders (every 90')

(limited accessibility)

Core Barrel Assembly All plants Cracking (SCC) UCB and LCB bolts Volumetric examination 100% of accessible bolts.

Upper thermal shield bolts (UT)

Core Barrel Assembly CR-3, DB Surveillance specimen holder tube (SSHT) studs/nuts (CR-3) or bolts (DB)

Core Barrel Assembly All plants Cracking (IE) including Baffle plates Justify by evaluation or by Inaccessible Core barrel cylinder (including readily detectable replacement vertical and circumferential cracking seam welds)

Former plates Core Barrel Assembly All plants Cracking (IASCC, IE, Baffle-to-former bolts Internal baffle-to-baffle N/A Baffle-to-baffle bolts IC/ISR/Fatigue/Wear, bolts:

Core barrel-to-former bolts Overload) No examination requirements, Justify by evaluation or replacement External baffle-to-baffle Inaccessible bolts, Barrel-to-former bolts:

No examination requirements, Justify by evaluation or replacement Page 64 of 108

ANP-2951, Rev. 001 Item Item____ Applicability Applicability Effect Effect (Mechanism) Primary Link (Note 1) Examination Method (Mechanism) PrimaryLink(Note1)(Note 1) Examination Coverage Core Barrel Assembly All plants Cracking (IASCC, IE) Locking devices, Justify by evaluation or by Inaccessible Locking devices, including including locking welds, replacement locking welds, for the external of baffle-to-former bolts baffle-to-baffle bolts and core or internal baffle-to-barrel-to-former bolts baffle bolts Lower Grid Assembly All plants Cracking (IE), including IMI guide tube spiders Visual (VT-3) 100% of accessible pads, Lower fuel assembly support the detection of separated and spider-to-lower grid examination dowels, and cap screws, and pad items: pad, pad-to-rib or missing welds, missing rib section welds associated welds.

sections welds, Alloy X-750 support pads, dowels, cap dowel, cap screw, and their screws and locking locking welds welds, or misalignment or the support pads (Note: the pads, dowels, and cap screws are included because of TE/IE of the welds)

Lower Grid Assembly All plants Cracking (SCC), Alloy X-750 dowel-to- Visual (VT-3) 100% of accessible dowel Alloy X-750 dowel-to-lower including the detection of guide block welds examination welds.

fuel assembly support pad welds separated or missing locking welds, or missing dowels Lower Grid Assembly TMI-1 Cracking (SCC) UCB and LCB bolts Volumetric examination 100% of accessible bolts Lower grid shock pad bolts (UT)

Lower Grid Assembly All plants Cracking (SCC) UCB and LCB bolts Volumetric examination 100% of accessible bolts Lower thermal shield (LTS) (UT) bolts Flow Distributor Assembly Flow distributor (FD) bolts Notes:

I. Examination acceptance criteria and expansion criteria for the B&W components are in Table 5-1 (of MRP-227, Rev. 0).

Page 65 of 108

ANP-2951, Rev. 001 APPENDIX D: TABLE 5-1 FROM MRP-227, REV. 0 Appendix D contains Table 5-1, "B&W Plants Examination Acceptance and Expansion Criteria" from MRP-227, Rev. 0. The inspections listed in the ONS RV Internals inspection plan will use the.acceptance and expansion criteria in this table, as modified by the ONS unit-specific amendments identified in Section 4.2.10 and Appendix F of this report.

Page 66 of 108

ANP-2951, Rev. 001 Examination Acceptance Expansion Additional Item Applicability Criteria (Note 1) Link(s) Expansion Criteria Examination Acceptance Criteria Plenum Cover All plants One-time physical None N/A N/A Assembly & Core measurement. In addition, a Support Shield visual (VT-3) examination is Assembly conducted for these items.

Plenum cover weldment rib pads The measured differential Plenum cover support height from the top of the flange plenum rib pads to the vessel CSS top flange seating surface shall average less than 0.004 inches compared to the as-built condition.

The specific relevant condition for these items is wear that may lead to a loss of function.

Core Support Shield ONS-3, DB Visual (VT-3) examination CRGT spacer Confirmed evidence of relevant The specific relevant Assembly castings conditions for a single CSS cast condition is evidence of CSS cast outlet nozzles The specific relevant condition outlet nozzle shall required that fractured spacers or is evidence of surface VT-3 examination be expanded to missing screws.

irregularities, such as damaged include 100% of the accessible or fractured nozzle material, surfaces at the 4 screw locations (at every 900) of the CRGT spacer castings by the completion of the next refueling outage.

Core Support Shield All plants Visual (VT-3) examination. CRGT spacer Confirmed evidence of relevant The specific relevant Assembly castings conditions in two or more CSS condition is evidence of CSS vent valve discs The specific relevant condition vent valve discs shall require that fractured spacers or is evidence of surface the VT-3 examination be missing screws.

irregularities, such as damaged expanded to include 100% of the or fractured disc mat*erial. accessible surfaces locations (at at theof4the every 900) screw CRGT spacer castings by the completion of the next refueling outage.

Page 67 of 108

ANP-2951, Rev. 001 Examination Acceptance Expansion Additional Item Applicability Criteria (Note 1) Link(s) Expansion Criteria Examination Acceptance Criteria Core Support Shield All plants Visual (VT-3) examination. None N/A N/A Assembly CSS vent valve top The specific relevant condition retaining ring is evidence of damaged or CSS vent valve bottom fractured material, and missing retaining ring items.

CSS vent valve disc shaft or hinge pin Page 68 of 108

ANP-2951, Rev. 001 Item Applicability Examination Acceptance Criteria (Note 1) Expansion Link(s) Expansion Criteria Additional Examination Acceptance Criteria Core Support Shield All plants 1) Volumetric (UT) LCB bolts (Note 2) 1) Confirmed unacceptable 1) The acceptance Assembly examination of the UCB indications exceeding 10% of criteria for the UT of Upper core barrel (UCB) bolts. UTS, LTS, and FD the UCB bolts shall require the expansion bolts and their locking The examination bolts that the UT examination be bolting shall be devices acceptance criteria for the expanded by the completion established as part of UT of the UCB bolts shall of the next refueling outage to the examination be established as a part of SSHT bolts (CR-3 include: technical the examination technical and DB only) justification.

For all plants justification. 100% of the accessible UTS, 2) The specific relevant

2) Visual (VT-3) examination Lower grid shock LTS, and FD bolts, condition for the of the UCB bolt locking pad bolts (TMI-1 expansion of the VT-Additionally for TMI-1 devices. only) 3 locking devices is The specific relevant UT examination to include evidence of broken 100% of the accessible lower condition for the VT-3 of or missing bolt the UCB bolt locking grid sock pad bolts.

locking devices.

devices is evidence of Additionally for.CR-3 and DB broken or missing bolt U't examination to include locking devices. 100% of the accessible SSHT bolts.

2) Confirmed evidence of relevant conditions exceeding 10% of the UCB bolt locking devices shall require that the VT-3 examination be expanded by the completion of the next refueling outage to include:

For all plants 100% of the accessible UTS, LTS, and FD bolt locking devices Additionally for TMI- 1 100% of the accessible lower grid shock pad bolt locking devices, Additionally for CR-3 and DB 100% of the accessible SSHT bolt locking devices A .1.

Page 69 of 108

ANP-2951, Rev. 001 Examination Acceptance Expansion Additional Item Applicability Criteria (Note 1) Link(s) Expansion Criteria Examination

__ _Acceptance Criteria Core Barrel Assembly All plants 1) Volumetric (UT) UTS, LTS, and FD 1) Confirmed unacceptable 1) The acceptance Lower core barrel (LCB) examination of the LCB bolts indications exceeding 10% of criteria for the UT bolts and their locking bolts. the LCB bolts shall require of the expansion devices The examination that the UT examination be bolting shall be acceptance criteria for the. expanded by the completion established as part UT of the LCB bolts shall of the next refueling outage to of the examination be established as part of the include: technical examination technical For all plants: justification.

justification. 100% of the accessible UTS, 2) The specific

2) Visual (VT-3) examination LTS, and FD bolts relevant condition of the LCB bolt locking Additionally for TMI-1 for the expansion of devices. 100% of the accessible lower the VT-3 locking The specific relevant grid shock pad bolts devices is evidence condition for the VT-3 of of broken or Additionally for CR-3 and DB the LCB bolt locking missing bolt devices is evidence of 100% of the accessible SSHT locking devices.

broken or missing bolt bolts locking devices. 2) Confirmed evidence of relevant conditions exceeding 10% of the LCB bolt locking de,vices shall require that the VT-3 examination be expanded by the completion of~the next refueling outage to include:

For all plants 100% of the accessible UTS, LTS, and FD bolt locking devices Additionally for TMI-1 100% of the accessible lower giid shock pad bolt locking devices, Additionally for CR-3 and DB 100% of the accessible SSHT bolt locking devices L L ___________ __________________ L Page 70 of 108

-ANP-2951, Rev. 001 Examination Acceptance Expansion Additional Item Applicability Criteria (Note 1) Eixnk(s) Expansion Criteria Examination Acceptance Criteria Core Barrel Assembly All plants Baseline volumetric (UT) Baffle-to-baffle Confirmed unacceptable N/A Baffle-to-former bolts examination of the baffle-to- bolts, Core barrel-to- indications in greater than or former bolts, former bolts equal to 5% (or 43) of the baffle-to-former bolts, provided that none The examination acceptance of the unacceptable bolts are on criteria for the UT of the baffle- former elevations 3, 4, to-former bolts shall be and 5, or greater than 25% of the established as part of the bolts on a single former plate, shall examination technical require an evaluation of the justification. internal baffle-to-baffle bolts for the purpose of determining whether to examine or replace the internal baffle-to-baffle bolts. The evaluation may include external baffle-to-baffle bolts and core barrel-to-former bolts for the purpose of determining whether to replace them.

Core Barrel Assembly All plants Visual (VT-3) examination, a) Former plates a and b. Confirmed cracking in a and b. N/A Baffle plates. b) Core barrel multiple (2 or more) locations in The specific relevant cylinder the baffle plates shall require condition is readily (including expansion, with continued operation of former plates and the detectable cracking in the vertical and core barrel cylinder justified by baffle plates. circumferential evaluafion or by replacement by seam welds) the completion of the next refueling outage.

Core Barrel Assembly All plants Visual (VT-3) examination. Locking devices for Confirmed relevant conditions in N/A Locking devices, the external baffle- greater than or equal to 1% (or 11) including locking welds, The specific relevant condition to-baffle bolts and of the baffle-to-former or internal of baffle-to-former bolts is missing, non-functional, or barrel-to-former baffle-to-baffle bolt locking and internal baffle-to- removed locking devices, bolts devices shall require and baffle bolts evaluation of the external baffle-to-baffle and core barrel-to-former bolt locking devices for the purpose of determining continued operation or replacement.

Page 71 of 108

ANP-2951, Rev. 001 Examination Acceptance Expansion Additional Item Applicability Criteria (Note 1) Link(s) Expansion Criteria Examination Acceptance Criteria Lower Grid Assembly All plants Initial visual (VT-3) Alloy X-750 dowel Confirmed evidence of relevant The specific relevant Alloy X-750 dowel-to- examination locking welds to the conditions at two or more condition is separated or guide block welds upper and lower fuel locations shall require that the VT- missing locking weld, or assembly support 3 examination be expanded to missing dowel.

pads include the Alloy X-750 dowel locking welds to the upper and lower fuel assembly support pads by the completion of the next refueling outage.

Incore Monitoring All plants Initial visual (VT-3) a) CRGT spider a. Confirmed evidence of relevant a. For the CRGT spacer Instrumentation (IMI) examination castings conditions for two or more IMI castings, the specific Guide Tube Assembly guide tube spider locations shall relevant conditions are IMI guide tube spiders The specific relevant conditions b) Lower fuel require that the VT-3 examination fractured spacers or IMI guide tube spider-to- for the IMI guide tube spiders assembly be expanded to include 100% of missing screws.

lower grid rid section are fractured or missing spider support pad the accessible surfaces at the 4 welds arms. items: pad, pad- screw locations (at every 90') of b. For the lower fuel to-rib section the CRGT spacer castings by the assembly support pad The specific relevant conditions welds, Alloy X- completion of the next refueling items (pads, pad-to-rib for the IMI spider-to-lower grid 750 dowel, cap outage. section welds, Alloy X-rib section welds are separated screw, and their 750tiowels, ap s -

or missing welds. locking welds b. Confirmed evidence of relevant and their locking welds),

conditions at two or more IMI th guide tube spider locations or IMI guide tube spider-to-lower grid rib conditions are separated section welds shall require that the or missing welds, VT-3 examination be expanded to missing support pads, include lower fuel assembly dowels, cap screws and ditems by the locking welds, or support pad itm ytemisalignment of the completion of the next refueling support pads.

outage Notes:

1. The examination acceptance criterion for visual examination is the absence of the specified relevant condition(s).

2. Expansion to LCB applies if the required Primary examination of LCB has not been performed as scheduled in Table 4-1 (of MRP-227, Rev. 0).

Page 72 of 108

ANP-2951, Rev. 001 APPENDIX E: NON-PROPRIETARY UCB AND LCB BOLT OCONEE UNIT-SPECIFIC TECHNICAL JUSTIFICATION Appendix E contains the non-proprietary UCB and LCB Bolt Technical Justification report for the ONS units.

The references and numbering for the section, figures, and tables in Appendix E are from the original TJ report.

Page 73 of 108

ANP-2951, Rev. 001 20004,015 (09/3012008)

AREVA AREVA NP Inc.,-

Siemens company an AREVA andi Engineering Information .Record Technical Justification for Upper and Lower Core Barrel Bolting Volumetric (Ultrasonic) Examinations- at Oconee Nuclear Station Page 74 of 108

ANP-2951, Rev. 001 Table of Contents Page SIGNATURE BLOCK ................................... ................... ........ 2 RECO RD O F REVISIO N .3.. .... ....... ...................................................................... ........... 3 LIST OF FIGURES ...........................................-................. 5 1 '0 P U RPO S E .................... ................. I......................... :..................- ;..................... . ........ ................  ;...

.6 2;0 DES(RIPTION OF COMPONENTiFLAWS TO BE EXAMINED ............................

2.1 Upper-and Lower Core Barrel Bolting Design ...... 66....................................................

2.2 Bolting Configuration, Fabrication, and Size..... .............. ............

I 1_--

A4 2.3 Previous Inspections and Flaws of Interest ..... .................... ................................. :16 2.4 Material Heats Common to Other B&W Operating Units ................... .. 18 2;5 Critical Flaw Size and Crack Growth Rate for Alloy A-286 Bolting ............... 18 2.6 Minimum Number of Bolts Required for Operation ......... ............ .... .....1.........., 9 3.0 OVERVIEW OF EXAMINATION SYSTEM ..................................... 9 4;0 DESCRIPTION OF INFLUENTIAL PARAMETERS ......... ........... .....

5.0 DESCRIPTION

OF EXAMINATION TECHNIQUES ....................... ................................... 20 6;0 DESCRIPTION OF EXAMINATION MODELING ........................... .. 20

7.0 DESCRIPTION

OF PROCEDURE EXPERIENCE .................................................................. 20 8.0 S UMMAR Y ................................................................................................................... ......... 2......

20 9.0 REFER ENC ES ................ . ................................................

.............. ..... 20 Page 75 of 108

ANP-2951, Rev. 001 1.0 PURPOSE The'purpose of ibis document-is to provideoa detailed explanation of the examination process, including the theory ofithe:examinationitechnique (as-applied to reactor-intemals inspcctions),, the essential variables of the procedure, other influential parameiers important to the overall performance of the examination-system and fild experience and/or mqockup demonstrations suppoling:*the capabilities of the NDE systen for volumetric (ultrasonic testing, LUT) examinations ofihe upper and lower core barrel.bolting in the Oconce Nuclear Station (ONS) Units1, I and 3'reactor vcSselinitmals..

Thisidocument is prepared in accordance with the requiremefit..detailed in Section V, Article 14, Examination

,System Qualificationsof the APmerican Society of Mechanical-Engineers Boiler and:Pressure Vessel Code.[))

ThiS*document is the deliverablefor Task12 .1.4i "Update ofthe Beta Tcsting: for UCB and.LCB Bolts," ini ARE.VA NIP's proposal:GC2008246 to DukeEnergy ( AREVA NP contract number A0002579). This document is thenonprpr yyietay vqerion of ARE-VA NPdocument 51-9112756ý-00.

2.0. DESCRIPTION OFCOMPONENTIFLAWS TO BE EXAMINED 2.1 Upper and Lower Core Barrel Bolting Design The fiveostructural joint locations iniithe seven operating B&W, I77-FA (Fuel Assembl ) reaetors (Ocofiee Unit'l

[ONS-1I], Oconee Unit 2 IONS-2J, Oconee Unit 3 [ONS-31, Crystal River Unit 3 [CR-3], Arkansas Nuclear One U0nit i [ANO-I 1.,Dais-Besse [D-BJ,.and Three Mile Island.Unit i [P[-:])!are shown in Figure 2-L In addition toMth felctns e shown in:Figure 2-I a.sixt structural bolting locatýion xist at two of the units(D-Band CR-3). This-is a redesigned surveillance specimen holder tube (SSI-IT), which is attached to the side of the thermaltshield. Remnant. o0f the0original SSHTs (Figure 2-2.and Figtirc 2-3)an dsomeof the: bolting still exist at

each of the ONS units and.are also believed to exist at the remaining twro.units:(ANO-l and'TMI-).

Upper core banil-to-core.support shield boiting (a.ka. upper core barrel bolts, UCB .blts) fastens the bottoim fflange of the core support shield to the top of the core barrel cylinder. Ihere are atotal o120 f UCBbolt locations. The UC;Bjoint carries the entireweight of the core and majority ofthe weightofthe reactorvessel interfials. The typical UCB~bolst configurations for.ONS-1, ONS-2, and ONS-3'areshown-in Figure 2-4. Figure 2-.5 shows avideo captre image of re presentativyeUCB bolts atONS. Note that the ordering drawing allows that the bolt head:design may eitherlbe a six-pointed or twelve-pointed hed configuiatio'n.

Thelower core barreI-to-lower grid bolting (a.k.a. lower core barrel bolts&.LCD bolts) fastens the bottom of the core barel cylinde to the lower grid assembly flange. There are atotal.of 108 LCBD bolt locations. The L*CB joint carries the weight of the core, but not the weightof the. core barrel. The typical LCB configuration for ONS-1, ONS-21 and ONS-3 is given in Figure 2-6 (also shown are the replacement lower thermal shield stds/nUts).

Figure 2-7,shows a video capture image of representativc.LCB bolts at ONS. Note that,the ordering drawing allo6ws' that the bolt head design may'either betassi-x-pointed or twelve-pointed head configuration.

Of the six structural joints in the B&Wm-design reactor vessel internals, only the UCB andmLCB joints have a core support function and, therefore, represent any potentia.safety concerns (Reference [2]).

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ANP-2951, Rev. 001

-;ppý oeBr~~t L-CbreBwrrei. Joint Figuire 271 Strucýtral Bolting~rLocations in B&4W-Design RatrVse~nenl Page 77 of 108

ANP-2951, Rev. 001 Figurie2-2: Location 6oOriginal Surveillance Specimen Holder.Tube in B&W-Design ReactoreVe!el Internals Page 78 of 108

ANP-2951, Rev. 001

.CORE SUPPORT

- SHIELD.

- SHIELD TUBE SHIELDED REGION SHIELD TUBE WELDED TO

'SUPPORT BLOCKS. AT THREE LOCATIONS THERMAL SHIELD PARALLEL FLOW REGION LOCK PIN IN UPPER PINTLE PERFORATED SECTION LOWER PINTLE Fig6e 2-3: Original Sur"eillniriteSpeciien Holder Tub:e Designin B&W-Design ReActor Vessel nternals Page 79 of 108

ANP-2951, Rev. 001 Figuire 2-4: Upper Core.Barrel BoltAConfiguration atONS un'its (Note: Iocking'devices not modieled;`bo*th hexx!'boheadaand `12-point bolt6head were allowed during ihe originallfartcation)

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ANP-2951, Rev. 001 Figure 2-5: Represaitatve Upper Core Barde:Bolt ýC figurdtion t ONS K

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ANP-2951, Rev. 001 Figure'24: Typical LowerCore Barrel BoltConfiguration at ONS Units (Note: locking devices not modeled;.bothbhex bolt head and12-point bolt head were allowedduring the original -fabrication).

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ANP-2951, Rev. 001 Figiure 2-7: Represmaiiitive'Loww Coe ad n"io Otafg atON Page 83 of 108

ANP-2951, Rev. 001 The:original Alloy A-286 UCD3 and LCB bolts were identical (MK256) at-six of the operating B&W 177-FA units. TMII 4is ,the exception, having used Alloy X-750 for all of the structural b6lting locations. ,All the original Alloy A7286 UCB and LOB bolts were fabricated by the same process. A single heatof Alloy A-.286 bar stock material was used for both UCB andLCB bolting locations at each unit; however, sevoralibeats ofmaterial wore used to fabricate thercquired-originalAlloy .A-286.MK-256 bolts used at the six B&W-- units. The.distributionof bolting material heatsý for ONS is provided in Reference [2].

At ONSl only, theýLCB joint also contains 12 extra Type 304 stainless steel bolts (ASTM A 193-65 Grade BS),

which are located bchirid the 12;guide lug shock pads, and therefore they are iiacccssible and not visible during inspections.

2.2 Bolting Configuration, Fabrication, and Size All UCB and LCB bolts*itnoperation at the three ONS units arethel originally installed¢ blting; none have been replaced to date. ONS-i began commrcia operation onApril 19,1973; ONSw2 began commercia operation on November 11, 1973; and ONS-3 began,commercial' operation on' September 5, 1974.

The original MKm256 UCB-and LOB bolts are fabricated from ASTMA 453 Grade 660 (ak-a. Alloy A-286),

Ceondition A imaterial. Condition A material was solution-annealed as 1i650"F for-two hours' and age-hardened at, 13259F for 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />. All material is believed to have been 15-20% cold-worked (as inferredby practices described by the bolt fabricator) by the material supplier before botling fabrication. Material was: hot-headed,,solution-annealed, andage-ha nedwith n0oassvtionperformed, followed by thrcad-rolling by the bolttfabricator.'The

,bar stock was ultrasonically examined and 100% of the finished bolt's surface was examined using fluorescent liquid penetrant. Details of the fabri6ation processing steps arc provided in Reference [2]. No rcord of fabrication flaws was found.

A:summary ofidcntified changes to~the original bolting joint configurations at the ONS units is as folows:

• Four UCB bolts, at, ONS-l1 were. removed for verification-and better interp~rettin of bolt ultrasonic signals in the 1980's (Reference [3]). At the end-of-cycle (EOC)-6 (August 1981,)two bolts were removed and shipped to the B&W hot cell laboratory. At EOC-7 (June 1983), two0additional bolts wore removed and shipped to the BS&W hot cell laboratory. Visual, ultrasonic, and fluorescent liquid penetrant examinations were performed in thelaboratory on all four bolts. The examinations confirmed ihe on-site

.ultrasonicdexaminafion result in-these four bolts (References [4j and [51).

• AtONS- l, there isamissingguideblock. It was idetntified as missing: during the Nideo inspecfion conducted at the EOC-7 outage in 1981 (Reference [61). Further investigations concluded-that the guide block had been missing from the CSA since the 1976 SSHT removal.effort-during the EOC-2 outage.

Figure 2-8 shows the typical guide block configiuation at ONS.

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ANP-2951, Rev. 001 FIgure,2-: Typical Shock Pad Configuration at ONS-4, ONS-2, and ONS-3

" At ONS-2,during the EOC 5 (December 1991)outage, in additionto failed lower thermal shield bolts, the Type 304 stainless steel attachment bolts for a shock pad were found broken (Reference [71). This shock pad .sumreoved and has never been reinstalled, Figure 2-9 shows a video capture image of the missing shock pad at ONS-2.

" At ONS-3,.two UCB bolts have shown unusual ultrasonic signals:in'prior examination.campaigns and have conservatively been assumed to have failed (Reference f2]); howsever, theyrremain captured bythe locking device and have ot been removed. The most recent ultrasonic examination (11/2007) has also identified indeterminate flaw indications at these two locations,(Reference [S1).

• At ONS-3. three LCB bolts have shown unusual ultrasonic signals during the 1987 emminations and have conservatively been assumed to have failed (Appendix C of Reference [91), however, they remain captured by the locking device and have not been removed.

The UCB, and LCB bolts tvere preloaded to various stress levels prior to service. Reference [2] contains the detailed prelbad information.for each ofthe ONS units. The nominal peak stress for these tw.o bolt locations at.

each oflthe ONS units includes the stress concentration factor (SCF),in the head-to-shank area. The yield~strength values for both thebUCB and LCB bolt heats of material are also provided in Reference 121.

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ANP-2951, Rev. 001 FlgurO*2ý9: Mmissng.Shock.Pad Conlfiguratlonwat ONS-2 2.3 Previous .inspecbonsl and Flaws of interest Refer*eces [9., 101 pr"vid ea summary of pastinspection informationmfor the UCB and LCB bolting locationslat each ofthec B&W*opcrating units.

Various rejected bolts were scnto tho'B&W hot cell forfailurc analyses. Laboratory examinations of those rejected bolts sent to the hot cell showed that all fracturesweri primarily camuedby intergranular stress corrosion' cracking (IGSCC) in or near the fillet between the head and theshank. Numerous branched~inteirganular cracking also intersected the fracture surfaces. Almost all fractures'initiated by intergranularcracing, and completed by ti-insgra-ularufafigue (some:bolts showing-morctr .nsgranular fatiguc than others). Thisled'to some belief that corrusion fatigue also might have played a~role in theIfailures.

The-crack*iitiat.ion locationscoincidedwiththe peak stress location-of the head-to-shank fillet region. Bolt material and fabrication factors. also contributed:to theI GSCC that included:chiomium content on the low end of theallowabl specification :range, heav.y cold-working, and hot-headihg. The observed fractures show cracking that curves upward slightly from the:fillet area'into.t'hehe*ad region, as shown schematically in Figure 2-10 and in a ro-ss-sectional _ii *viwin Figuite2- ! 1.

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ANP-2951, Rev. 001 Sý(ETQ1 OF UIPER CORE .BARREL BOLT SHWKNG LOCATION OF FRACTURE lOT HEADING METALLURGICAL TRANlSITION LOCATION OF, PInCIAL FRACTURE AM)

SECONDARY CRAC1CD.G Fligure-2-10: Schematic ofTypical Fracture Orientatin, Observed on Faled. StructuralBoiag at B&W Units Page 87 of 108

ANP-2951, Rev. 001 Figure2-11: Typical Fracture Orientation Observed durin gLaboratory Examinations~on Failed Structural Bolting at B&W,Units (Note:.OriginalNiagnification was 3,1X)ý Reference 71 2.4 Material Heats Common to Other'B&W.Operating Units

.skssummarized'above.in Section 2. 1, several heatsl f material wer used during the manufacture ofthe;ONS UCB and LCB bolts. These heats;of material are also-common with several ýother operating B&Wa-design units.

2.6 Critical Flaw Size and Crack Growth Rate for AlloyA-286 Bolting Reference [11] summarizes a review of crack growth rates for,.Alloy A-286 miitrial in a primary water enVirohment. The informati6nigatherd iridicatesthdat.relatiýv.ely:fast crackgrýwthirate ,wouldbeanticipated.

F or such a crack growth rate, failure of an,UCB or LOB boltrwithin an. 18-month cycle of operation wouldeasily occur. Therefore. an estimation of crack initiation time is betieved'to be ofmore importance and determination of

.a critical crack size is not relevant for these bolts. Assessment of operability needs to be determined by the Page 88 of 108

ANP-2951, Rev. 001 number'ofWIbts bfoken,,their location, and the iate6at which additi6nal bolts Will brak befoie the following inspection period.

2.6 Minimum Number of Bolts Required for Operation The UCB and LCDBjoints have large;structural margins. This conservatism greatlyretdcs the, likelilhood hat either-the UCB or LCB joint might fail. Acceptance criteria are based 6in stress limits forthre.ded.structural fasten6rs given in Subsection NG of tie ASME B&PVCode,'Reference [12]:.

Six, conditiohs of remoed/rejected boltshaVe been evaluated forieach analyzed core sugport structures corresponding to all B&Wýdesignmloprating units.

The firstscenarioanalyzed foreiach unit reptesentscurrently existing conditions. Tlie7remainringseenarios are hypothetical.

The'analyzed scenatios:foi theithrec ONS Units, along withthe 'niumberiand lcaitions far the.upper a ildow.ercore barrel bolts arc, summarized in Reference [13].

Maximum bolt stresses are calculatedlandcompared w~ith theiri.allbwable valuc per thetrcquirehents 6f ASME Code Subsection;NG. Thesresults of thehypothetical cases are given inReference [13].

3.0 OVERVIEW OF EXAMINATION SYSTEM The examination system consists ofAREVA.NP procedure 54-ISI-165.l UltUhrasiciEkcainationo0fP.WR Internal Bolting [14], an ultrasonic scope, coaxialca!xle, a round transducer applicable-tothe 0bolt size mounted in a spring-oadedlixture,,a suitble liquid cupiant suchas bo.rad de .nealize odisfi.&waler, and0sofiware to acquire anwimage of the screen. The UCBibolt exýam istypicall, pelbIned 1owesiagthe'tiasducerfixture xby on a series of poles until the: transducer is seated on the bolt head. The LCB exam is typically performed by delivering the transducer to the bolt heads from below using either'a crwler orapo stick typ manm-handled pole that reacts o0T of the deep-end-floor. However, this does notlimit thepbtenilfor theleioeiopmentof alternative transducer delivery systemnioptions in the future. The ultrasonic data collecition antdanalysmi.s the same for the UCB and LCB bolts.

A transducer monitors the back-wall and boltusignals during the acquisition process. Aklosswofback-walltor multiple indicationsignalsequAly¢-spacd M time. are;considered recordableindications. Thiscijfiique basbically rejectsia bolt ifa flaw is recordedwithoutfregard to flaw size.

4.0 - DESCRIPTION OF INFLUENTIAL PARAMETERS The anticipated damage mechanism for UCB or LCBxbolting is IGSCC,.basedon.lpast~failuresfand laboratory data.

The essential.variables for the :bolting exams can be found~in AREVA NP procedure 54-1SI-165-1 1Ultras6nic Examination.of PWR Intemal: lolting,[l4] Theesslential variables include tra,.nsdcrsszrnucf " frequenlcy, cable length, number.of-intermediate connectors,. LT scope, and UTscopesettings. Shoufldany. f these.essential variables'change, AREVA NP would be required to perform ainew demonstraft.nusiihg the'e.new essential, variables.

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ANP-2951, Rev. 001 Thce xaminer shall be qualified at a minimum to Level II (see [1 51.:for definition) irinaccordance with the AREVA and shall acceptUthe.

NP writtenmpractice. The Level II or III.shall be responsible-for interpretation ofthedata resultsisftltelexaminati6n. The x*amine sha veadditional documented traihing:in ttt.chniques described within AREVA NP procedure 54-ISI-165-I1 Ultrasonic Eza~mination of PWR Internal Bolting.

5.0 DESCRIPTION

.OF EXAMINATION TECHNIQUES The UCB andLCB bolts'are interrogated for a direct reflection ofsound from theflaw or alack o0fback-wall response. The technique is not intended to determine flaw size,.otientaiion or percent ofremainingligamentin the*bolt. Te gainilevel is adjusted up and down toobNeve the rcsponse from the bolt.

Thistechniquc was demonstrated at :EPRI on bolts represcntativc of the UCB:anid LCB bolting in'the ONS units.

EDM notches were machined into some of the bolts. The technique used onthese Alloy A-286'bolts detected the aniticipaied.degradationf[161 When evaluating flaws, there are several points to consider. This, technique-basically rcjects:abolt if a flaw is reo6rded without regard to the sizec 6.0 DESCRiPTIONWOF EXAMINATION MODELING Since these examinations have* een perfortnedduring thepast 28 years and have recentlyben validated by blind perforanqc demostration at EPRI,.no: modeling is required.

7.0 DESCRIPTION

WOF PROCEDURE.EXPERIENCE UT examination;of Upp and Lower Core Barrel Bolting has been pe-rformed at onetime or another-at all of the operating B&W units since 1981. These examinations wrce pcerfrmed using carlicrvrisions of,544S,1.'l65 with thc'sam0e.s9esnial variables as the current revision.

8.'0

SUMMARY

An.explanation ofite UT examination technique for use in examining reactor vesselintemals upper~and lower core barrel structural.bo1ting at ONSis given in this document. In addifion. detailsofithe bolting designs cuently inuseand pri'or!feldex-perince With the UT cxaminations are also provided. In conclusion, 54-ISS-1651i Lis capable, as demonstrated, of determining the integrity of the Alloy A-286 upper and lowercore barrel bolts in the B&W-design reactor vessel internals at ONS.

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ANP-2951, Rev. 001 APPENDIX F: ONS UNIT-SPECIFIC AMENDMENTS TO MRP-227, REV. 0 Appendix F contains the justification for ONS unit-specific amendments to MRP-227, Rev. 0. The references and numbering for the section, figures, and tables in Appendix F are from the original report.

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ANP-2951, Rev. 001 AREVA NP Inc.,

an AREVA ahd Siemens "company Engineering Information Record Justification for ONS Unit-Specific Amendments to MRP-227-Rev. 0 Requirements Page 92 of 108

ANP-2951, Rev. 001 Table of Contents Page SIGNATURE BLOCK ....... 2...........

ý2 RECORD OF REVISION .......... .. ................ 3

1.0 BACKGROUND

AND PURPOSE ......................................... 5 2.0 ONS CASS AMENDMENTS . ................... ........... . ........ 66.............................................

2.1 MRP-227-Rev. 0ORequirement ........... ................... . ...... ......... . . 6

.2 2;2 Ba..frC sis forCASS.A S....t............

endhi ts ................ . .............. 6 2.3 Amendments.,- ....... ..........

.. ...... ................................  ;.7 2:3.1 CSS Vent Valve Disc Amendment .. I......................................... 7 2.3.2 CSSOutlet Nozzde,(QNS-3:only) Amendment ...... s ............. I ................ 7...............

7 SC 2;.3'3 CRGTýSpacer C tingwA A..........sAn...................................8.......................8 endment ...........

3.0 PLENUM COVER WELDMENT RIBWPAD ITEMS AT ONS*- ONLY .................................... 10 3.1 Screeh ing .............................................................. ........................................................ 10 3.1. AlloyX-750 Dowel. ........... ..................................

................ 10 3.1.2 Alloy X-750%Dowel Locking Weld .................................................................................... 10 3.A.13 Type 3O4Screws. . ..... .......................... ........................ .... .... 11 3.1.4 Type 364Screw Locking Weldc ............. ........................... 12 3'.2 Assessmentr ,ofAlloy ,X,-750.Dowel Locking Weld............... ............. ........ 12 4.0 CSS VENT VALVE.DISC SHAFT (HINGE PIN) ... ..................... ........ 14 4.1 MRP-227-Rev. 0 Requirement........................................ 14 4.2 42.Accssibilty.............. ..  ; . ........... ..........

A cdessibilib/ ................................................................................................  ;........... 14

...... ............... :14 4.3 Amendment. . ............. .. I 14 5.0 LOCKING DEVICES FOR HIGHSTRENGTH BOLTING ...................................................... 16

5.1 Clarification

to MRP-227-Rev. 0 Table 4-1 ...... ....... .................. 16

.5.2 :Clarification to MRP-227-Rev. 0 Table 4-4 ............. .................................................... 16 6.0 S UMMARY .............. ....... ............ .............................................................................................. 18

7.0 REFERENCES

..................... ....................................... ....... 20 Page 93 of 108

ANP-2951, Rev. 001

1.0 BACKGROUND

ANDPURPOSE In December 2008* Electric Power Research Institute (EPRI) issued Rev. 0 of the Material Reliability Program (MRP) MRP-22701 Inspection and Evaluation (l&E)Guidelines for managing long-term aging of pressurized water reactor (PWR) reactor vessel (RV) internals inmthe U.S. MRP-227-Rev. 0 provides~generic augmented inspection requirements for thelcdurreat operating fleet of U.S: PNRs, MRP-227-Rv.,0 Sectioin 7.3 categorizes the implementation of the augmiented inispections as 'Needed, in acordance with Nuclear Energy Instlitue-(NEI) 03-08 Guidelines4."

The generic augmerited inspection requirements for the Babcock and Wilcox (B&W) PVVRs are listed in MRP-227-Rev: 0, Tables 4-1., 4-4 and5-1i. MRP-227-Rev..0 requires implementation of the internals program requirements listed in these tables within 24 months following issuance" fMRP-227-A and completion of the examinations within the applicable-defined inspection timelines. MRP-227-Rev. 0 recognizes that.eadier implementation may be required by plant-spec-ific regulatory commitments. Plants implementing MRP-227 prior to issuance of the U-S. NRC approved MRP-'227 (MRP-'227-A)`version will thus-implement the requirements in accordance with the current published.version of the guidelines,(MRP-227-Rev. 0).

Since the publication of MRP-227-Rev. 0, AREVA NP has performed additional records sSearch:for the "Primary' and Exparniont items for Duke Energy's Oconee B&W PWR units (ONS-1, 6NS&2, and ONS-3). The records search is for prepanng the ONS unit-specific RV interails inspectiohnplan and aging management program (AM P) that will comply With the MRP-227-Rev, 0 requirements. I Fabrication records for several cast austenitic stainless steel (CASS) items were notravailable during the screening and evaluation performed for MRP-227-Rev. 0. Therefore, these CASS items were conservatively categorized as "Primary" for thermal aging embrttlement in MRP-227-Rev. 0. The fabrication records found recently have-shown that these CASS items are not susceptible to thermal aging embrittlement and therefore can be re-categorized to the 'No Additional Measures' category (see Section 2).. In addition, the records search also has identified a new Alloy. X-750 dowel location at ONS-1 thlatis not generic to the B&W RV intemrals-and not previously known. Thus, it was not listed in MRP-227-Rev. 0 (see Section,3).

Recently; it hasalso beendiscovered that the vent valve disc shaft is completely inaccessible to visual inspection.

The ,V!-3 examination requirement was therefore incorrectly specified in MRP-227-Rev. b Table 4-1 and Table: 5-1 for the vent valve disc shaft. In addition, ithas been found thatthe aging degradation and VT-3 requirement in MRP-227-Rev. 0 Tables 4-1 and 4-4 for the locking devices of high-strength bolts are imprecise or inconsistent In light of the findings, amenddments to the MRP-227-Rev. 0 requirements forthe affected items are needed for the ONS units in order to meet the intent of MRP-227-Rev, 0. This document provides written-technical justification for the needed amendments tothe MRP-227-Rev. 0 requirements-for several affected RV internals items at ONS units.

It is intended that;the technical justification in this document be used to satisfy appropriate;procedural processes in order to incorporate the needed amendments into the.ONS unit-specific:RV internals inspection plan and the AMP. At this time, it has not been decided what procedural process will be used Regardless of the procedural process to be used, this document provides thaewritten technical justification for the needed. amendments to MRP-227-Rev. 0 for theaONS RV intemals inspection plan and the AMP.

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ANP-2951, Rev. 001 ZO ONSCASS AMENDMENTS-Three CASS items in the B&W units are identified in MRP-227-!Rev.. 0 as "Primary, and "Expansion' due to the

,potential for thermal agingýembrittlement This section provides the ONS uhit-specific CASS amendments and justification, 2:1 MRP-227-Rev. 0 Requirement The following three CASS. items in the ONS RV intemals aredefined in MRP-227-Rev,.0 as "Primary' or

_Exparision" duetto thermal aging embrittlement:

Prirrir/

" CSS assembly vent valve disc atONS-1, ONS-2, and Of'JS-3 ONS-3

" CSS assembly castoutlet,,nozzlesat Expansion

- Control rod guide tube.(CRGT')assembly spacer castings at ONS-1,.C)NS-2, and ONS-3 The MRP-227-Rev. 0 augmented inspection reqUiramntsW for the abbve'CASSitems are listed in Table 4-1. Table 4-4,;and Table5-1 of MRP-227-Rev. 0 During theprparatoywork leading to MRP-227-Rev 0, the generic andkhown unit-specific B&W RV intemrals componernswere~screened for all applicable aging deradation mechanisms. The screening process and results are docurented inrMRP-189-R&v 1. ForCASS iterswith unkiow ,cherrical compositions. the ferrite contents were conservatively assumed to exceed.the screening criterion. Due:to a lack of fabrication records, the above three CASStenis ere n tvely screend as susceptible to thermral aging emnbittlement and eventually.categonzed as:MRP-227-Rv. O =Prinary" otExpnsion'*

There is-a fourth CASSitem in the2B&W'R.V internals~that also requires augmiet-ed irsp*ection b*MRP-227-Rev.

0,.the spider castings for the iricore instrurmehtatibnm(IMI),guide' tubeassembly. The spider castings are categorized as "Primary',in MRP-227-Rev: 0 dutetoboththearmal aging ýembrittlement (TE) 'a irr*diation agirg, I nbtfttklment (IE): Therefore the spiderbcastirngs cannot bere-catbgoizled byVuing therscreenhin§gcdteria for thermal aging ernbrittlement- However, 'TE' in MRP-'227-Rev. 076 ble 4T1 is ho longer appicable to the spider castings.

2.: Basis for CASSAmendmenits Since the publication of MRP-227-Rev. , AREVA NP has performeda records search for the aboyeCASS items.

The records search results are documented in References 4.and: 5. The records were, not available during development of MRP-227 and these, CASS componentswere therefore considered to be0susieptible it thermal aging embrittlement when MRP-227-Rev.- Owas being developed:

The records search found the appilcable, CASS hat'numrbers and*ertified material test reports,(CMTRs), which allows screening in accordance with the MRp5-1 751 screening criteria. The MRP-175 screnigcrifteria.are identical tq those recommended by SecGion Xl.IM12 of NUREG1 801-Rev. 1 for thermal aging ebrnbttlenent of CASS.

The*CASS ferrite content is estimated with Hullisequivalent factirs in NUREG-CR,-4513; Rev. 1.i, Because the nitrogen content is not listed in the CMTRs, nitrogen is assumed to be 0.04% as recommended by NUREG/CR-4513, Rev'I. The technical background for the CASS screening criteria is detailed in MRP-1 75, NUREG-180i.-

Rev. 1, and NUREG/CR-4513, Rev. 1.

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ANP-2951, Rev. 001 2.3 Amendments 2:3.11 CSS Vent ValveDisc.Amendment Each ONS. unit has eight vent valves installed in the StS cylinder. Each vent valve contins a hinged*

ve~ntvave disc and a vent valvebody. E~achvent va*le*can be remotel, hanynoe as a unit for removal or iallation. The hinged disc inc*udesa*device. or remot testing veriffunction. rand pressure in theqnterior of thPS asmly during a large break LOCA, preventing backpressurefmmreversing coolant flow through the=core. Duringg all:other operating conditionrs theyvent yalves.are closedandprpeYent bypass reactor co*lant.flow from the reactor vessel annulus to the:core outlet region The vent valve dic are fabricated from Grade CF8 of. ASTM 7A351I-65N

,The heat~numbers and CMTRs of the vent valve discs have been retrieved for all 24 original vent valves supplied to:the ONS units. The heat numbers and ferrite contents based on CMTR chemical composition with Hulh.ls equivalent factors are documented in RIef. 4. The ferrite contents for all'gil v ve ow screening criterion'for Grade CF8 material.

The ONS units were also supplied with spare vent valves ahn;some6sres*were nstallef(ie., replaced the original:vent valves), Some vent valvbswere exchanged between the ONS units. Therefore, Duke Energy recently ide tified the S/N and heat number of c.urie"ly instaledsatONS-i'(Fall20) and ONS&2:(SpJrng 2010) using a remote underwater Video CaTmera..Their idenirfication and ferrite content are documented in Ref. 5. All currently installed vent valveadiscs at ONSi 'and ONS-2are below the.20% ferriti screening criterion for G de C 8. Asimilar identification f the O talle vent valve discs is planned for Fall 2010. Based on the records search and ON$1l and ONS-2identification results, a .similar findingis expected for ONS-3._

if these records had been available during initial screening,, thevent valve discs would have been, categorized as Category "N (below screenin gvalues) and, no aua'etedinspectionwquld have been required by MRP-227-Rev. o0Therefore, the yent-valve dirsscatONS are re-categorized from the

'PirTary"'in MRP-227-Rev. 0 to"No Additional Measures, and theLaugmeted inspect.ohnin:MRP-227-Rev.. Table 4-1 isnot required.

It should be noted that this ONS vent valve disc amendment.to the MRP-227-Rev:0 requirement does not affect thefollowing requirernents:

• Note1 to MRP-227-Rei 0, Table.4-1, vent-valtve exercising* and visualin'spectio.

-1 MRP-227-Rev. 0, Table.4-1 inspection requirement for the.vent valve top and bottom retaining rings, and disc shaft (hinge pin)-

, ASME Section XI, Category B-N-3 VT-3 examination of the vent valves (see notelow).

Note: TheASME Section X1, Category B-N-3 inser'vce- inspeton (ISI).scope is defined by the Owners (utilities) of the B&W units-2i312 CSS Outlet Nozzle (ONS73 only) Amendment Each of the three-ONS units has two outlet nozzles in the CSS cylinder. The outlet nozzles are used to direct the RCS flow out of the core to the two reactor vessel outlet nozzles. Only the CSS outlet'nozzJes at ONS-3 are fabricated from GradeýCFofASTM A,351665." The CSS outlet:nozzles at.ONS-1.and ONS-2 are fabricated from Wrought stainless steel and do not havela thernal aging embrittlementeconcem.

The heat numbers and CMTRs have beenfound for the two 0NS-3-odSSotlet nbzzles. Thelheat numbers and ferrite contents based on CMTR chemical composition with Hullsequivalent factors are Page 96 of 108

ANP-2951, Rev. 001 documented in Ref: 4. The ferrite contents of both outlet nozzles are below the'20% ferrite screening criterion for Grade CF8 material.

If these records had been during initial screening, theO.NS-3 SS oultlet nozzles would have

.been categorized as"'A . available and no0augrnented inspection wou(d have been required by MRPID27-Re .

Therefore, the ONS*3 .CSS outlet nozzles are re-categorized from the 'Primary' in MRRP-227-Rev. 0 to

'No Additional Measures', and the apgmentedrinspction in MRp-227-Rev. 0' Table 4 is not relquired.

It should be noted thatlthis.CSS outlet nozzle (ONS-3 only) amendment to MRP-227-Rev*0 r0equirement does not, a ffect the Ifollowi ng requirernrt:

• ASME Section Xl,.Category B1f-3 VT-3 examinatibns-ofithe outlet nozzles:(seernote below)

NoteL Th.e ASME Section XI, Categýory.B-N-3 inserviceinspection (ISI) sope is-fineddt'6e owners (utilities) of.the B&W units-2.'3.3' CRGT Spacer Casting Amendment Each of the three ONS units has 69 CRGT.assemblies welded to the plenum cover pIatea.nd b.bltedto.

the uppr grid. Inside each CRGT assembly isa brazeme.nt subassembly co Insisting of ten parallel' horizonal spacer castiOngs. Each spacer casting is brazed.to 12 lerflrated vertical rod guide tubes and 4 pa rs of,vertI rodguide sectors, also called 'I"--tubes." A. tIhe ten spacer elevations, the CRGT housing rica pipe s drilled at 'four elqually spaced circumferentiai locations to acc'ommodate four cap screws that hold each spacer casting in place. The 10Ospacer castings provide structural, support to the12 perforated vertica ro guide tubes and 4 pairs of vertical rod guide sectors within'each CRGT~asembly andckeep them aligned with the guide tubes in the fuel assembly below.

Thespacer castings-are fabricated from'Grade CF3M of ASTM A 351-,65.he heat numbers And CMTRs have been found for most spacer castings at ONS units, A majortyof thes~espaqer castings exceed'the "4% screeningrcritenon for Grade CF3M. Therefore'the CRGT spacercasti*igsh as a group, cannot t re-categorized to t~he"N-o`Addition-al Measures' cat'egory from their current MkP-227-Rev.- 0 &Epansion category.

The CRGT spacer castings are linked in MRP-227-Rev, 0 to the following 'Prinary- items:

0 CS0Svent valVeadiscs,

• outlet nozzles (ONS-3 only), and CSS S!Mguidg e tube spier casings.

The amendments in this document re-categorize the CSS vent valvediscs and CSS outlet nozzles to,'No Additionai Measures", and therefore removed their link to the. CRGT spacer casting. Inspecting tth'e.CSS vent valveý disc. and CSS outlet nozzles will' not meet the "Expansion" link inrtent becase t two items are not susceptible for thermal aging embrttlement based on the!screening results.

The amendments leave:the IMI guide tube spider castingsýas theý only Pnrmary'"linikfr the CRGT spacer castings at ONS.units. However, the spider castings alone.do not meet the intent.'of MRP-227-Rev. 0dto providledetection'of the aging effect before it affects the CRGT spacer casingsdue to the.following' considerations:

The"CRGT, T spacer. castings arefabricated from Grade .CF3M ( 6ntaining molybdennu) while the spider ,castings.are fabricated from Grade CF8 material.

• The spider ferrite content per Hull's equivalent factors (Ref. 5)*is below the 20%:criterion for G.ade CF8 forthermal aging embrittlement; the potential f6r irradiation embrittlement'iSwhat prevents the spider castings from being re-categorizedtod'No AdditionallMeasures'.

Page 97 of 108

ANP-2951, Rev. 001 Thelspacer castings are exposed to the corneexit reactor coolanttemperature while the spider castingsare..exposed to cold leg reactor coolant temperature.

Therefore; the MRGT spacer castings at ONS unitslare re-categorized.from'Expansion' in MRP-227IRev.

0 toPmri'ay With th.e following: equiremrnents,

• Effect (Medhanism): Cracking(TE), ihcluding detection of fractured spacers or missing screws.

Note: The effect ,is unchanged from MRP-227-Rev. 0, Table .4-4.

• Expansion LinkR:None:

Note: The CRGT.spacer casting expansion link to the spider castings in MRP-227-Rev. 0 Tables 4-1 and 5-1 issremoved by this amendment.

Examination Method/Frequency.: Visual (VT-3) examination during the next 10-year ISI; subEsqueht ekamiinations6n the 10-year ISI interval.

Note. VT-3 nfhethodis unchanged'frorm MRP-227'Rev. 0, Table 4-4 Frequency is, identical to MRP-227-Rev. 0, Table,4-1 for the CSS vent valve discs and:CSS-outlet nozzle&s

. Examination Coverage: 100% of accessible surfaces at the.4 screw locations.(at everyn900)

(limited accessibility).

Note: The coverage: is unchanged from MRP-227-Rev; 0, Table4-4:

Page 98 of 108

ANP-2951, Rev. 001 3.0 PLENUM ICOVERWELDMENT RIB PAD ITEMS,.AT ONS-1 ONLY Duning the records search, a&feature unique toONS-1 was identified foethe plenumrncover weldment rib pads.

Each of the 32 plenumio6ver rib pads at ONS-1 is fastened t the6 plenumcove' nbs with to Type 304 sres and oný Alloy.X-750 dowel. At the other ONS units, the:plenumr icover rib pads are welded t the plenurm cover weldmhent rbs -

Since this unique feature at ONS-1 was unknown during the preparatory work leading to MRP-227-Rev. 0, the Alloy:X-750 dowel and Type 3041 screws at this location-werenot screened for all applicableaging degradation, mechanisms, nor evaluated for inclusion in MRP-227-Rev. 0. This section provideesan ,evaluation using the MRP criteria and process to determine:if-additional;augmented inspection is-required for-thIisloocationat ONS-1.

3.1 Screening Ifthiv feature at ONS-1 had been"known, it-would: have been'screened*for all applicable aging degradation mechainisms. Therefiore, th~is ldcatiqn is screene beo Wit th scenn i sarid-crter'ia. that has sam been-usled for other locationsmand documented in MRP-189-Rev. 1 PIThe fabricationi records related to this locatido'aresurr narized in Section,9'of. Ref, 5.

3.1.1 Alloy X-750 Dowel The environmental parameters (temperatureAnd neutron exposure) are the- samret as for the'plenumn cover weldment rib pads(P~.11) in MRP-1 89 Rev. 1 Table 3-2. The Alloy X74 :dowel body an crrythe possible shear load introduced by, the relative thermal moverment of theclsum head and plenum cover.

However, the dowels are under compressive stress at the surfaces exposed to primary cont Therefore, 5CC-of the Alloy.X-750 dowel is not a concern. The dowel lockingwdi not a multplepass weld. Applying .th*e"mnateria.l anrd parameters against the MRP-1 75 criLeia prod** ete following:

screeiihg, results:

• StresskCdrrosion'Crackjng;(SCC) below screeung., Cate "A"

" Irradiation-Assisted SCC (IASOC), beloW screening, Categ/ry. 'A?

IIrradiation Stress. R eaxation and CIree (ISRilC),. below screening, Caeg.y t A"-

" Wear, below-screening, CateEgory "A

" FatigLu,: below;screening, Category *A

" TherrmalAgirigEmbrittiement6. below screenirng, Categ*ory IA

• Irradiation Aging Emnbrittlement eo sening CategrA

• Void'Swelling, below screening, Category 'A*

Therefore, the overall screening category for ,the-AIIoy X-750 dowel is'Category "A" and hnbaugmented inspectiomnis required, 3.1.22 Alloy !X-750 Dowel Locking Weld The environmental parameters (temperature and neutron exposure).are-the sameias for theAlloy X-750 dowels As documented in Ref. 5, ther.AllybYX-,750,dwels for thi QNS! plemnrucover weldmentrib pads have the same diameter as for the Alloy X-75Gdowels used for-the upper grid and lower grid fuel assembly support pads at ONS-1. The lockng weld for the Alloy X-750:dowels for the .ONS- plenum Page 99 of 108

ANP-2951, Rev. 001 cover rib pads is'identical to the locking weld;for the.Alloy X-750 dowels forthe upper and lower fuel assembly support pads at ONS-1 in the following ways:

• Same Allby69 (INCO 69) weld metal

• Same weld process

• Same heat of weld wire SSarne wel]d size Appying the material and environmental parameterst against the MRP-175 critena produces the. following sc*reening results:

• Stre-ssCorrosion'iCracking (SCC), above screfening, Ca*tegory "Not.,A'

• Irradiatigo-Assisted.SCC (IASCC), below screening, Catregory."A

• Irradation Stress Relaxation and Creep (ISRJIC), below screening, Category 'A-

• Wear, below screening, Category "A'

• Fatigue;: below.screening; Category:'A"

• Thermal;Aging Embrittlement, below screening, Category"A" IrradiationAging Embrittlement, below screening, Category"'A

. VoidSwelling,., below screening, Category"A" Because this locking weld used nickel-base material, it is potentially susceptible to primary wateriSCC (PWCC).similar to the nickel-base locking welds used in'the upper and lower grid fuel assembly support pads,-and lower grid g-uideblocks Therefore, the overallhscreening category fo the AllyX-7506dowel l

locking weld isCategory."Not A" dueito PWSCC. This locking weld is fuither assesseddinSection3.2:of this.document.

3.1.3 TypeI304.Screws.

The environerfnetalparameters.(temperature and neutron Oeposiue) are the sameas' fr the plenui cover weldmentrib pads (P.1.1) in MRPX189 Rev. 1 Table,3-2. The screws are fabnicated from Type.304 austenitic stainless. Therscrew locking weld is not a multiple pass weld Applying the material and paramreers against the MRP-175iteritiproduces the following screeningres-ults:

• Stress4Corrosion Cracking (SCC), below screening, Category WA'

• Irradiation-A.ssisted SCC (IASCC), below screening, Ca*tegqry "A" IIrradiation Stress Relaxation-and.Creep (ISR/IC), below scriening, Category "A"

Wear, below screening, Category "A*

• Fatigue, below.screening, Category "A"

• Thermal Aging Embrittlement, below screening, Category"A"

• Irradiation Aging Ern.ittlement, below screening, Category."A"

• Void Swelling, below, screening, Category "A" Therefore, the overall screening category for theType 304 screwsistCategory "A'kand no augmented inspection isirequired.

Page 100 of 108

ANP-2951, Rev. 001 3-i.4 Type 304 Screw Locking Weld The environmihntal parameters (temperatur-eaand neutron exposure) are the same as for the plenum cover weldment rb padsý (P.1.1) in MRP-1 89,Rev. 1 Table 3-2. The lockingw.eld used Type 308 austenitic without usingý .a stainiess weld metal. Th .lockigdgweld directly welds the:screw head to the rib pads locking.bar or a locking~cup. Applying~the mraterialand.pararmeters against the MRP-175c ritena p5roduces.

the f611-winigrweenihg retults:.

• Stress CorrosionýCracking (SCC), below screening, Category "A"

• Irradiation-Assisted SCC (IASCC), below screening, Category"A"

• Irradiation Stress Relaxation and;Creep (ISR/C), below screening, Category WA

M ari below screening, Category "A"

• Fatigue;,-bel 0Wscreening, Category"A"

• ThermalAging Embrittlement, below screening, Category "'A Irradiaon Aging Embrittlernent. below screening, Category "A'

• Void Swelling, below screeriing; Category WA Therefore, the~overall screeningcategory:for'the.Type 304 screw locking weld isZCategory WAarad: no augmented inspection-is required 3.2, Assessment of AlloyX-750 Dowel Locking We~ld This nickel-based Alloy 69 (INCO 69) loc..ng weld is susceptible to: PWSCCý. Each of the 32 plenum cover weldrmnt rib pads at ON81 'is fastened to the plenum cover weldment fibs with two Type.304 screws and one AIJoyX-750dowel. The~dowel is located inrthe center.of each rib pad while the two screws are at-the-two ends0of each rib pad.

The plenum cover weldment-db pads are part of the RV internals holddown.stack,. which provides clampingforce to stabilize and.significantly restrict the rigid body pendulum motion of the core support and plenum assemblies.

In other words, the clampingaction prevents rigid-body rotation at~the interface area. The clampingactiomndoes not have a direct core supportsafety function. Loss of clamping would undoubtedly lead to corebarrel motions:

that-wouldeventually lead;to a.reactor'shut dowrn Dueto thewear concern, the plenum cover weldment rib pads are categorized as. MRP-227-Rev. 0 "Primary-.

Thedowel lockting weld serves as4a loose:part prevention device. The dowel locking weld will retain the.dowel if.any portionof the weldis in~place: The only loading onithe dowel is shear, whichis in~the'dowel body and not near the locking Weld. In addition, the,Alloy X-750 dowels on the rib pads are completely coveredby the reatofr veel head flange during plant operation and therefore cannot back out even ifthe-dowelilockirngweld is completed

.cracked The Alloy -x-750,dowels and Type 304-screws are .Category*'* components, .arndarenot affected:by any aging degradation mechanismss. Therefore, c*acking of the Alioy'X-750 doweldockihgweld will not affect the functionality of the plenum cover weldment ribs-1ased ohnthe above functionality assessment, the locking weld for the ONS;1 plenumcover weldment, rib pad:

Alloy.'X-750 dowel'is categorized as."No Additional Measures". Therefore, no-additional augmented inspection is required for this location.

Note: the following requirements for the plenum cover rib pads~are not affected by theascreening in Section 3.1 and assesment in rSection 3.2-of this document.

e MRP-227-Rev. 0, Table 4-1, plenum cover weldment rib pads.(Primary).

Page 101 of 108

ANP-2951, Rev. 001 AS.ME SectionId,'Category. ON-3 VT-3 examination of the.plenum cove weidmenrrib pads.(see note below).

Note: The ASME Section XI, Category B-N-3 inservice inspection (ISI) scope is defined by the owners (utilities) of the B&W units.

Page 102 of 108

ANP-2951, Rev. 001 4.0 CSS VENT VALVE.DISC SHAFT (HINGE PiN)

Each of the'three.ONS-units has eight vent valves installed inlthe CSS cylinder. Each vent-valve consists of a hinged disc, aivalve body with sealigshrtces, a split-retainiing- ng, ahd jacI s:that hold the retaining rings in place6to support theperiheter ofthe valve:assertbly:

4.1 MRP-227ýRev. 0 Requirement The, SS vent Yalve disc shaft (also'called hinge pin) is listed in MRFP-227-Rey. 0 as 'Primary' due to-thermal agingembrittletninet The disc sht is made of A.STM A 276, T pe 431 martensitic stainless steel. The vent valve disc shaft insp*ction requirements-in MRP-227T(Rev. 0) Table:4-1land Tabl 5&-1 are'the followiri:g Visual (VT-3) .examinatih.Lduring the, next 10-year ISI,.Subsequent examinations on the 10-year ISI interval. Exaniinatibi cover e is 100% of. accessible surfaces.

4.2 Accessibility The hinre assembly consistsofavaYent, valve disc shaft, four ilargd shaft journals (bushings) and two journal receptales The .'nt~vahie dic'cittairn an integral exercise ILig.for rem6te exercising, Alter the vent valve disc Was hinged toýth* vent*a*lve bodte oftheJounal, receptacles were' cvered with welded co*ver plates.1 10' As illustrat *edin Figure 41, the ;Ventvalve ;disc shaft is~corrnletely, enlos by theintegral exercise lug, four flanged shaft jourmTals,;two valvediscjoumal receptaclesand theircover ,lates.Therefore, the vent valve disc shaft is inaccessible for visual ,ispection wi6out disassebling the nt valve.

4.3 A*fmeiidfehnt The verntvalve disc shaft-is completely enclosed andinaccessible to visual-insPection. The visual (VT-3),examination requirementrwas incorrectly~specified in MRP-227-Rev. OTable 4-1 and.Table 5-1 for this item. Therefore, the MRP-

227Rev. 0 Table4-l requirement for the vent valvediscshaft- is revised to the following for the ONS units:

• Phrimary: The yeýnt alve ,disc shaft remain~s"P. ~rma0r) unless~justified by evaluation.

" Effect (Mechanism): Cracking (TE),irncluding the detection of surface ir*eLarities; such as damaged, fractuid material, or missin'gitems.

" Expansion: Link: None..

• Examina.iio0nMethodrequen~cy' No examination requirements, justify by evaluationmor by replacement Examinationi Covdrage: Inmaiccesible N6te: The abovechanges dolnot apply to: the CSS vent valve top and bottom retaining ring.

In addition, theiSuaV3eamination' under Examinatlon Acceptahce Criteria' in MRP-227-Rev.O Table 5-1 for thesventvalve disc shaft'i.srevisedto 'lnaccesible., iUstifvl evaluation or by replacement" Page 103 of 108

ANP-2951, Rev. 001 FigureA4-1I: Typical CSS vent valve -outside-view The vent Vale disc'shaft (also called hinge ,pi)A-iscomn*pletely hiddemn aid inaccessible without disassembling the vent valve.

Page 104 of 108

ANP-2951, Rev. 001 5.0 LOCKING DEViCES FOR)HIGH-STRENGTH BOLTING Thissection clarifies the description aging degradation effect:and.examinationmethod in MRP-227-Rev. 0 Tables 4L1 and:4-4ff the high-strercthObolts andttheirlocking device. The high-strength bolts.inside the RV initemals of ONS units are the following bolts:

• MRP-227-Rev. 0 Table -1, Primary o Upper Core Barrel (UB)*Bolts including cking devices'and looking welds o: Lower xe Barrel (1LOBi)Bolts, nciuding rliong decesaýnd locking weldsý

" MRP-227-Rev. .0 Table, 4-4,,Expansion o Flow Distributor(FD) Bolts, inicluding locking dev\'nesand lclkirn welds o Upper Thermal Sheld (UTS) Bolts, includinglocing devicesand locking welds o Lower Theeral Shield (LTS) Bolts, including lockihg.de and locking welds 5.1 Clarification tozMRP-227-Rev. 0 Table4-i MRP-227-Rev. 0 Table 4-1 (Primary Items) requires UT insp*dttionf t&e.UCB~anId .LCBbolts, and VT.3 inspection of their lockingdevices. Table 74-1 lists SCC,(strescorrosion cracking) under Effec' (Mechanism)" for the UCS and LCB bo6ls and their lockingdevices.

The UCB and LCB bolts are susceptible to 50;i the bolt locking dgvices arernotsusceptible to SCOI However, dam aged locking dee a0videe of pan y failed UCB anLOBbsdue toSOC., Due to this considation,

.MRP-227-Rev. 0 also requires .Výr3%inspe.Ctn of the locking deviceýs.

Therefore, the folloWig cldrification is provided to the UCB and,LCOBbots inrMRP-227-Rev OTable 4-1:

• The aging effect.(mechanism) for bolt is crackIng:(SO%

• The aging effect I(mectianism)'for ,bolt locking d.e~vices is loss oif mnaterial, damaged, distorted,,or missing locking.devices (wear or-fatigue amage failed bolt).

5.2 Clarification to MRP-227-Rev. 0 Table 4.4 In MRP-227-Rev. 0 Table 4-4, th-e listing for the "Expansionrfboltslined to.UCEand.LOB bolts:does not explicitly list the bolt locking devices or the VTw3 visual inspection of the locking devices.; This error originated from MRP-231.Rev.

011111Table*3-10, which was the underlying,rce for MRP-22,7TR&v;O'Table 4-4. The:omission of the locking~devices and VT-3 requirement in MRP-2274Rev. 0.Tablel,4-4was inadvlerlteas MRP-227'Rev. 0 Table,5-1 cdearly lists the VT-3 requirement for the locking.devices of-the "Expansion" bolts linked to the "Primary" UCB and.LCB bolts.

Therefore, the following clarification is provided to0the"Expansion" UTS, LTS, and FD bolts in MRP-227-Rev. 0 Table 4-4 that are linked to the "Primary" UCB and LCB bolts:

• Each item is revised to include:, and, ý*ir locking deylces."

• The aging effect (mechanism) for bolt. iscracking (SCO) a The aging effect (niecanism) for'bolt looking devices isloss of material, damaged; distorted, or missing locking devices (wear or fatigue damage by failed bolt):

" Each.pnmary link is revised toinclude "and their locking devices."

• The examinatibon method for bolts is Volumetric exarnination (UT).

Page 105 of 108

ANP-2951, Rev. 001

• The examination method or boit locking devi es isvisual(IT-3) exarninatbon.

• The examination coverage is 100% of accessible bolts and locking devices.

Page 106 of 108

ANP-2951, Rev. 001 6.0

SUMMARY

Amendments to MRP-227-Rev. 0 requirements for-the ONS units:

1. MRP-227-Rev. 0 Primary: CSS assemblyventvalve discs 6 The vent valyvediscs atQNS-1 and ONSo2 arere-categorized~to"NoAdditional Measures*. No:

augmented inspection is required.

Note: All currently installed vent!valve~discs~at*ONS-.1and.dONS-2 have been verified to0bebelow the.-20. ferrite.screening criterion for.Grade CF8. A similarverification for theoNS'-3 installed vent valve discs is plannied for Fall 2010 This amendment Will-apply to the oNS-3: installed vent valve discs after theferrite is confirmed to be below thescreeningciteria.

,2. MRP-227-Rev. 0 Primary: CSS.assembly cast; outletnozzles :(ONS-3 only),

o The outlet nozzles at ONS-3 are re-categorizedto "No Additional Measures". No augmented inspection is required.

3. MRP-227-Rev. 0 Expansion: CRGT spacer castings o The.CRGT spacer castings at:ONS-1, O)NS2 and'ONS-3 are re-categorized to Primary,

" There is no "Expansion' link-

- Examination Method/Frequerncy: ,Viral (VT-3).examinationdunrngthe next 10-year I$S; Subsequent examinations onthe 10-yeaf I..Iinterval,

" Effect (Mechanism) and examination coverage.are unchanged fromVMRPA227ýRev 0 Table 44.

Notes:

4 The CRGT.spacer casting expansiaon link to the IMI, guide tube spider castings.in MRP-227-Rev. 0, Tables 4-1 andrl-is removed by, this amendm*nt

- 'TE! in MRP-227TRev. 0 Table 4-1:is no longe.r applicable to,the IMI guide tube&spider castings. However, the sp.der castihgs remain.as!.Prima ry' due to'lE, and are linked to the "Expanrsion' of lower grid fuel assembly supportpad items.

4. MRP-227-Rev. 0 Primary: CSS Vent Valve Disc Shaft (Hinge Pin) o The examination method/frequency'and examinationmcoverage in:MRP-227-Rev. 0 Table 4-1 and Table 5-1 for CSS~vent valve disc, shaft (hinge pin) at ONS1, 'ONS-2, and ONS-3 are revised to thefollowing'

" Examination Method/Frequency inTable 4-1 :-No examination requirementsjustify by evaluation or by. replacement

" Examination Coveragein Table 4-1 Inaccessible. I

" Examination Acceptance Criteria inTable 5-1:4naccessible, justify by evaluation or by replacement

5. MRP-227-Rev, 0 Table,4-1, UCB and LOB Bolts o The aging effect (mechanism) for bolt is cracking (SCO).

o The aging effect (mechanism)for boit loc*kng device isloss of material, damagd, distorted, or missing lockingdevices (wear or fatigue damage by failed bolt).

6. MRP-227-Rev. 0 Table 4-4, 'Expansion" UTS, LTS; and FD Bolts Linked to UCB and LCB-bolts.

Page 107 of 108

ANP-2951, Rev. 001

o Each itern is revised to include: 'and their !ockrng devices o Thetaging effect (Mechanism) for bolti*ecrackding (SCC).

o The agingieffect (mechanism) for boltlocking devices is]iss of material, damageid, distorted, or missing kbcking.devices (wear or fatigue damage bVfbiled,bolt).

o. Each primar link is revised to include: "and.their lockng deyioes, o -Theexamirntbon-methd for bolts is1volumetric exannriatn (UT).

o The examinUtion method for bolt locking devices is viual ,(VT-3) exarninatin.

o Theexamination.coveragetis 100% of accessible bolts and locking devices.

ONS-i P.le'mri C.ei'rWeldrrnt Rib Pad Items Duridg the records search, a feature'unique to.ONS-1 was~identifiildfor the plenum coyer weldrent rib pads, The AlloyXJ-750 dowels, Type 304lscrews, and their Iocking weids were unknownandwere not screened for a~ingrdegradation mechanisms, nor evaluatedfor inclusion in MRP-227ýRev. 0. Using.the MRP;screenknhgcriteria and p these:itersare categorizedlas.Category 'A6ior "NoAdditioral

,oc~s, Measures.. Therefore; no additional augiented inspection is required fir this~locati6hn The MRP-227-Rev, 0 report is current[ybeing reviewed by the U.S. NRC for a safety evaluation re'ort-(SER)Y.If necessary, the ONS armendments in this document Will be updated for'complianice with the U.S. RC approved MRP227 (MRWP-227-A),after the SER is granted.

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ATTACHMENT 2 REGULATORY COMMITMENTS

- Regulatory Commitments Page 1 The following table identifies the regulatory commitments in this document. Any other statements in this submittal represent intended or planned actions. They are provided for information purposes and are not considered to be regulatory commitments.

Commitment Due Date After approval of MRP-227, Duke Energy will review and, if 90 days after issuance of needed, revise the ONS RV Internals inspection plan. MRP-227-A The ONS ISI program will be updated to include the items from 90 day outage report the NRC-approved ONS RV Internals Inspection Plan as augmented inspections and the inspection results will be submitted.