TMI-13-161, Submittal of Inspection Plan for Reactor Internals. Areva Report ANP-2952Q1NP Enclosed

Submittal of Inspection Plan for Reactor Internals. Areva Report ANP-2952Q1NP Enclosed

ML13317A931
Person / Time
Site:	Three Mile Island
Issue date:	11/06/2013
From:	Jim Barstow Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML13317B534	List: TMI-13-161, Submittal of Inspection Plan for Reactor Internals. Areva Report ANP-2952Q1NP Enclosed
References
TMI-13-161 ANP-2952Q1NP, Rev 0
Download: ML13317A931 (53)

Text

PROPRIETARY INFORMATION

-WITHHOLD UNDER 10 CFR 2.390 10 CFR Part 54 TMI-13-161 November 6, 2013 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Three Mile Island Nuclear Station, Unit 1 Renewed Facility Operating License No. DPR-50 NRC Docket No. 50-289

Subject:

Submittal of Inspection Plan for Reactor Internals

References:

1) NUREG-1928, "Safety Evaluation Report Related to the License Renewal of Three Mile Island Nuclear Station, Unit 1," dated October 2009 2) Letter from M. D. Jesse (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "Submittal of Inspection Plan for Reactor Internals," dated April 16, 2012 3) Letter from D. P. Helker (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "Submittal of Inspection Plan for Reactor Internals," dated April 17, 2013 4) Internal Memorandum from R. Ennis (Senior Project Manager, U.S. Nuclear Regulatory Commission) to V. Rodriguez (Acting Chief, U.S. Nuclear Regulatory Commission), "Three Mile Island Nuclear Station, Unit 1, Draft Request for Additional Information (TAC NO. MF1459)," ML13269A179, dated September 25, 2013 In the Reference 2 letter, Exelon Generation Company, LLC (Exelon) submitted the inspection plan for the Three Mile Island (TMI), Unit 1 Reactor Vessel Internals (RVI) ("Inspection Plan for the Three Mile Island Unit 1 Reactor Vessel Internals," ANP-2952, Revision 1, March 2012). The Reference 3 letter provided updates to commitments contained in the Reference 2 letter. As a result of further review and a change in scheduled inspection outages, updates to these commitments are being provided below. Attachment 3 transmitted herewith contains Proprietary Information.

When separated from attachments, this document is decontrolled.

Submittal of Inspection Plan for Reactor Internals November 6, 2013 Page 2 Additionally.

in Reference 4, the U.S. Nuclear Regulatory Commission requested additional information concerning the inspection plan. Attachment 2 is our response to Question 2 and Question 4, Parts 3, 4, and 5 of that request. Attachment 3 contains proprietary responses to the remaining questions as provided by AREVA NP Inc. Attachment 4 contains non-proprietary responses to the remaining questions as provided by AREVA NP Inc. In the Reference 2 letter the included inspection plan identified that all initial MRP-227-A inspections were planned for the fall 2015 refueling outage. Exelon has subsequently determined that a change to the examination schedule is appropriate and allowed by MRP-227-A.

Examination of RVI components that are accessible only while the Core Support Assembly is removed will be performed during the 2015 refueling outage. Other RVI components will be examined during the 2017 refueling outage. Refer to the Attachment 2, Question 4, Part 4 response for further discussion.

This schedule completes all initial MRP-227-A examinations within two refueling outages following entry into the period of extended operation.

Commitment:

2. Applicant/Licensee Action Item 6 (Table 0-2 of 77-2952-001), based upon Section 4.2.6 of the SER, requires a licensee to justify the acceptability of certain inaccessible components that Table 4-4 of MRP-227-A identifies as expansion components.

The SER action requires that a licensee submit an analysis of the acceptability of these components for continued service or a schedule for replacement of the subject components with the NRC submittal documenting the intent to implement the requirements of MRP-227-A.

TMI is currently working with the Pressurized Water Reactor Owners Group (PWROG) to address the action item. Exelon will submit an update of the PWROG progress in evaluating these components and a schedule for showing when Exelon will submit an evaluation for continued service or a schedule for replacement by April 19, 2013. Response:

The analyses of the inaccessible components identified in Table 4-4 of MRP-227-A are being pursued as TMI, Unit 1 plant-specific analyses.

As the inaccessible components are defined as expansion components under MRP-227-A, their inspection (analysis) is only required if the primary component inspection does not meet MRP-227 -A acceptance criteria contained in Table 5-1. TMI, Unit 1 will either submit a detailed analysis, a replacement schedule, or a justification for some other alternative process within one year of the initial inspection (Fall 2017) of the linked MRP-227-A primary component items, if the inspection results in indications beyond the threshold for expansion criteria presented in Table 5-1. This schedule is consistent with current NRC and Industry proposed schedules concerning topical report WCAP-17096-NP, Revision 2. Commitment:

3. Applicant/Licensee Action Item 7 (Table 0-2 of 77-2952-001), based upon Section 4.2.7 of the SER, requires that the licensee develop plant-specific analyses to demonstrate that there is not a loss of functionality of the Incore Monitoring Instrumentation (IMI) guide tube assembly spiders and Control Rod Guide Tube (CRGT) spacer castings due Submittal of Inspection Plan for Reactor Internals November 6, 2013 Page 3 to loss of fracture toughness.

TMI is currently working with the PWROG to address the action item. Exelon will submit an update of the PWROG progress in evaluating these components and a schedule for showing when Exelon will submit an evaluation for continued service or a schedule for replacement by April 19, 2013. Response:

The analysis and evaluation of the Control Rod Guide Tube (CRGT) Spacer Castings is in progress with the PWROG, with an estimated completion date of December 31,2013. The Incore Monitoring Instrumentation (IMI) Spider Castings evaluation will be pursued as a TMI, Unit 1 plant-specific analysis and evaluation, separate from the CRGT Spacer Casting analyses being performed by the PWROG. TMI will submit the CRGT Spacer Castings and IMI Spider Castings evaluations to the NRC by October 30,2016, which is approximately one year prior to the inspection outage (2017). Additionally, the vent valve retaining ring analysis will also be submitted by October 30, 2016. Attachment 3 contains information proprietary to AREVA NP Inc. AREVA NP Inc. requests that the document be withheld from public disclosure in accordance with 10 CFR 2.390(b)(4).

Attachment 4 contains a non-proprietary version of the AREVA NP Inc. report. An affidavit supporting this request is contained in Attachment

5. If you have any questions concerning this letter, please contact Tom Loomis at (610) 765-5510.

Respectfully, Barstow Director -Licensing

& Regulatory Affairs Exelon Generation Company, LLC Attachments:

1) Summary of Commitments

2) Exelon Response to Requests for Additional Information (Question 2 and Question 4, Parts 3, 4, and 5) 3) "Response to NRC Requests for Additional Information on the Three Mile Island Unit 1 Reactor Vessel Internals Inspection Plan," ANP-2952Q1 P, Revision 0, Proprietary Version 4) "Response to NRC Requests for Additional Information on the Three Mile Island Unit 1 Reactor Vessel Internals Inspection Plan," ANP-2952Q1 NP, Revision 0, Non-Proprietary Version 5) Affidavit cc: Regional Administrator, Region I, USNRC USNRC Senior Resident Inspector, TMI USNRC Project Manager, [TMIJ USNRC Attachment 1 Summary of Commitments Summary of Commitments The following table identifies commitments made in this document. (Any other actions discussed in the submittal represent intended or planned actions. They are described to the NRC for the NRC's information and are not regulatory commitments.)

COMMITMENT COMMITTED COMMITMENT TYPE DATE OR ONE-TIME Programmatic "OUTAGE" ACTION (Yes/No) (Yes/No) The analyses of the December 15, No Yes inaccessible components 2018 identified in Table 4-4 of MRP-227-A are being pursued as TMI, Unit 1 plant specific analyses.

As the inaccessible components are defined as expansion components under MRP-227-A, their inspection (analysis) is only required if the primary component inspection does not meet MRP-227-A acceptance criteria contained in Table 5-1. TMI, Unit 1 will either submit a detailed analysis, a replacement schedule, or a justification for some other alternative process within one year of the initial inspection (Fall 2017) of the linked MRP-227-A primary component items, if the inspection results in indications beyond the threshold for expansion criteria presented in Table 5-1. This schedule is consistent with current NRC and Industry proposed schedules concerning topical report WCAP-17096-NP, Revision 2.

Summary of Commitments November 6, 2013 Page 2 COMMITMENT The analysis and evaluation of the Control Rod Guide Tube (CRGT) Spacer Castings is in progress with the PWROG, with an estimated completion date of December 31, 2013. The Incore Monitoring Instrumentation (lMI) Spider Castings evaluation will be pursued as a TMI, Unit 1 plant specific analysis and evaluation, separate from the CRGT Spacer Casting analyses being performed by the PWROG. TMI will submit the CRGT Spacer Castings and IMI Spider Castings evaluations to the NRC by October 30, 2016, which is approximately one year prior to the inspection outage (2017). Additionally, the vent valve retaining ring analysis will also be submitted by October 30,2016. COMMITTED COMMITMENT TYPE DATE OR ONE-TIME Programmatic "OUTAGE" ACTION (Yes/No) (Yes/No) October 30, 2016 Yes No Attachment 2 Exelon Response to Requests for Additional Information (Question 2 and Question 4, Parts 3, 4, and 5)

Question:

Attachment 2 Response to Request for Additional Information Page 1 "In Section 4.1.7 of the RVI inspection plan, the licensee described RVI inspections that have been performed in the past at TMI-1. These include vent valve inspections, UCB bolt ultrasonic examinations (UT), and core clamping measurements.

One hundred percent of the UCB bolts were examined in 2009 via UT with no recordable indications found. One hundred percent UT examination of the UCB bolts was also performed in 1991 with no recordable indications.

MRP-227-A specifies UT examination of the UCB bolts within two refueling outages of January 1,2006, or the next ten-year inservice inspection (lSI) interval, whichever comes first. Please provide the following information:

1) Confirm that the UCB bolt examinations performed in 2009 constitute the initial MRP-227-A inspection requirement for the UCB bolts. 2) Will the UCB bolts be reinspected during the Fall 2015 refueling outage, when MRP-227-A inspections are scheduled in conjunction with the ASME Section XI inspections of the RVI?" Response:

1) The Upper Core Barrel (UCB) bolt examinations conducted during the 2009 refueling outage constitute the initial inspections as required by MRP-227-A within two refueling outages of January 1, 2006. The 2009 refueling outage was the second refueling outage after January 1,2006. MRP-227-A and MRP-228 were not completed in time to be fully implemented at TMI, Unit 1 prior to this inspection.

The additional requirements for training and equipment as designated in MRP-228 were not fully applied. The examination technique used in 2009 was demonstrated at EPRI in Charlotte, North Carolina.

A Technical Justification was completed for the examination process that was in accordance with MRP-228, Revision 0 and ASME Section V, Article 14,2004 Edition. The examination technique was determined to be capable of detecting flaws greater than 28% cross sectional area with no false calls. 2) TMI, Unit 1 plans to re-inspect the UCB bolts during the 2015 outage when the Core Support Assembly (CSA) is removed for the 10-year inservice inspections.

This examination will be compliant with MRP-227-A and MRP-228. Question: "NUREG-1801, Revision 2, "Generic Aging Lessons Learned Report," (GALL Report, Revision 2), Chapter XI.M16A, "PWR [pressurized water reactor] Vessel Internals Program," states, under "Detection of Aging," that the VT-3 visual methods may be applied for the detection of cracking only when the flaw tolerance of the component or affected assembly, as Attachment 2 Response to Request for Additional Information Page 2 evaluated for reduced fracture toughness properties, is known and has been shown to be tolerant of easily detected large flaws, even under reduced fracture toughness conditions.

In the licensee's evaluation of consistency with the "Detection of Aging Effects" attribute of GALL, Revision 2, Chapter XI.M16A, the licensee stated that VT-3 examinations will be used to detect cracking only after evaluation of the flaw tolerance of the component or affected assembly, under reduced fracture toughness conditions, has been shown to be tolerant of easily detected flaws. The NRC staff notes that there are six TMI-1 Primary components and three Expansion components for which visual VT-3 examination is specified for detection of cracking (consistent with MRP-227-A).

License Renewal Interim Staff Guidance LR-ISG-2011-04, "Updated Aging Management Criteria for PWR Reactor Vessel Internal Components," (Reference

6) modified the statement regarding VT-3 examination such that a flaw tolerance evaluation would only be required for non-redundant components.

Reference 1 also stated that VT-3 visual methods are acceptable for the detection of cracking in redundant RVI components (e.g., redundant bolts or pins used to secure a fastened RVI assembly).

Please provide the following information:

1) Which TMI-1 RVI components require flaw evaluations to justify VT-3 examination?

2) Are any of the components for which VT-3 examination is specified redundant components that do not require a flaw evaluation?

If so, justify considering the components redundant.

3) How will the flaw tolerance evaluations of the TMI-1 Primary and Expansion components be documented and communicated to the NRC to support the NRC staff's review? 4) Confirm that these flaw tolerance evaluations will be completed prior to the initial inspections of the TMI-1 Primary components, scheduled for the fall 2015 refueling outage. 5) Will the flaw tolerance evaluations of the control rod guide tube spacer castings and incore monitoring instrumentation guide tube spiders be part of the plant-specific evaluation of cast austenitic stainless steel components that the licensee committed to submit to the NRC by fall 2015?" Response 1) and 2) The responses to these items are provided in Attachments 3 and 4. 3) As required by MRP-227-A and MRP-228, TMI, Unit 1 has committed to develop the applicable flaw tolerance evaluations discussed in items 1 and 2. MRP-227-A and the Safety Evaluation Report do not require that these flaw tolerance evaluations be submitted to the NRC. The evaluations will be available on site prior to the refueling outage in which they are required.

4) As stated in item 3, the flaw tolerance evaluations for the applicable components, as identified in items 1 and 2, will be completed and available on site prior to the performance of the MRP-227-A examinations on those components.

TMI, Unit 1 Attachment 2 Response to Request for Additional Information Page 3 currently plans to perform the required MRP-227-A inspections across two outages (2015 and 2017), and the applicable flaw tolerance evaluations will be available prior to the outage of examination performance in which they are required.

Table 4.1 below provides the currently planned schedule for inspections.

This is a change in schedule from that identified in the April 16, 2012 submittal but maintains compliance with the scheduling required by MRP-227-A and TMllicense Renewal Commitments.

2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 Attachment 2 Response to Request for Additional Information Page 4 Table 4.1: Breakdown of MRP-227-A Inspections at TMI, Unit 1 UTS (Upper Thermal Shield), Upper Core Barrel (UCB) Volumetric 2015 L TS (Lower Thermal Shield), Bolts (UT) and Lower Grid Shock Pad Bolts UTS, L TS, and Lower Grid UCB Bolt Locking Devices Visual (VT -3) 2015 Shock Pad Bolt Locking Devices Lower Core Barrel (LCB) Volumetric 2015 UTS, L TS, and Lower Grid Bolts UT Shock Pad Bolts UTS, L TS, and Lower Grid LCB Bolt Locking Devices Visual (VT -3) 2015 Shock Pad Bolt Locking Devices Flow Distributor Bolts Volumetric 2015 UTS, L TS, and Lower Grid UT Shock Pad Bolts Flow Distributor Bolt Locking UTS, L TS, and Lower Grid Visual (VT-3) 2015 Shock Pad Bolt Locking Devices Devices Alloy X-750 Dowel-to-Guide Alloy X-750 Dowel locking Visual (VT-3) 2015 welds to Lower/Upper Grid Block Welds Su art Pads Visual (VT-3) 2017 N/A Visual (VT-3) 2017 N/A Visual (VT -3) 2015 N/A Visual (VT-3) 2017 N/A Visual (VT -3) 2017 N/A Visual (VT -3) 2017 N/A Volumetric Inaccessible:

Baffle-to-Baffle Baffle-to-Former Bolts (UT) 2017 and Core Barrel-to-Former Bolts Baffle Plates Visual (VT -3) 2017 Inaccessible:

Core Barrel C linder and Former Plates Baffle-to-Former and Internal Inaccessible:

External Baffle-to-Baffle Bolt Locking Visual (VT-3) 2017 Baffle-to-Baffle and Core Barrel-to-Former Bolt Devices and Welds Lockin Devices Incore Monitoring Lower Grid Fuel Assembly Instrumentation (IMI) Spider Visual (VT -3) 2017 Support Pad Items Castin s IMI Spider-to-Lower Grid Rib Visual (VT -3) 2017 Lower Grid Fuel Assembly Section Welds Su art Pad Items Attachment 2 Response to Request for Additional Information Page 5 5) The flaw tolerance evaluations of the Control Rod Guide Tube (CRGT) Spacer Castings and Incore Monitoring Instrumentation (IMI) Spider Castings will be performed as discussed in response to Question 4, Parts 1 and 2 (Attachments 3 and 4). These evaluations will not be developed as part of the Cast Austenitic Stainless Steel (CASS) fracture toughness evaluations that TMI, Unit 1 committed to providing to the NRC prior to inspection of the components, but will be developed separately as needed and available on site at TMI, Unit 1 prior to the outage of performance.

Attachment 4 "Response to NRC Requests for Additional Information on the Three Mile Island Unit 1 Reactor Vessel Internals Inspection Plan," ANP-2952Q1NP, Revision 0, Non-Proprietary Version A AREVA Response to NRC Requests for Additional Information on the Three Mile Island Unit 1 Reactor Vessel Internals Inspection Plan October, 2013 AREVA NP Inc. (c) 2013 AREVA NP Inc. ANP-2952Q1 NP Rev i sion 0 Copyright

© 2013 AREVA NP Inc. All Rights Reserved AREVA NP Inc. Response to NRC Requests for Additional Information on the Three Mile Island Unit 1 Item Rev. 0 Section(s) or Page(s) All Nature of Changes Description and Justification Original Issue ANP-2952Q1 NP Revision 0 AREVA NP Inc. ANP-2952Q1 NP Revision a Response to NRC Requests for Additional Information on the Three Mile Island Unit 1 Contents LIST OF TABLES ..........................................................................................................

IV LIST OF FIGURES ..........................................................................................................

V NOMENCLATURE

.........................................................................................................

VI ABSTRACT ...................................................................................................................

VII

1.0 INTRODUCTION

AND SUMMARy ...................................................................

1-1 2.0 REQUESTS FOR ADDITIONAL INFORMATION (RAIS) AND RESPONSES

....................................................................................................

2-1 2.1 RAJ-1 .......................................................................................................

2-1 2.1.1 Statement of RAJ-1 .......................................................................

2-1 2.1.2 Response to RAJ-1 .......................................................................

2-2 2.1.2.1 Response to Part 1: .......................................................

2-2 2.1.2.2 Response to Part 2: .......................................................

2-2 2.2 RAI-2 .......................................................................................................

2-2 2.2.1 Statement of RAJ-2 .......................................................................

2-2 2.2.2 Response to RAI-2 .......................................................................

2-3 2.3 RAI-3 .......................................................................................................

2-3 2.3.1 Statement of RAI-3 .......................................................................

2-3 2.3.2 Response to RAI-3 .......................................................................

2-4 2.4 RAJ-4 .....................................................................................................

2-14 2.4.1 Statement of RAI-4 .....................................................................

2-14 2.4.2 Response to RAI-4 .....................................................................

2-16 2.4.2.1 Response to Part 1: .....................................................

2-16 AREVA NP Inc. ANP-2952Q1 NP Revision 0 Response to NRC Requests for Additional Information on the Three Mile Island Unit 1 2.4.2.2 Response to Part 2: .....................................................

2-17 2.S RAI-S ....................................................................................................

2-22 2.S.1 Statement of RAI-S .....................................................................

2-22 2.S.2 Response to RAI-S .....................................................................

2-22 2.S.2.1 Response to Part 1: .....................................................

2-22 2.S.2.2 Response to Part 2: .....................................................

2-22 2.S.2.3 Response to Part 3: .....................................................

2-23 2.6 RAI-6 .....................................................................................................

2-24 2.6.1 Statement of RAI-6 .....................................................................

2-24 2.6.2 Response to RAI-6 .....................................................................

2-24

3.0 REFERENCES

..................................................................................................

3-1 AREVA NP Inc. ANP-2952Q1 NP Revision 0 Response to NRC Requests for Additional Information on the Three Mile Island Unit 1 List of Tables Table 2-1: Vent Valve Locking Device Material Identification

......................................

2-7 Table Vent Valve Locking Device Screening Parameters

....................................

2-8 Table 2-3: Vent Valve Locking Device Initial Screening Results .................................

2-9 Table 2-4: Vent Valve Locking Device FMECA Results ............................................

2-10 Table 2-5: Initial Categorization of Vent Valve Component Items .............................

2-12 Table 2-6: Flaw Tolerance Evaluation Requirements for VT-3 Examinations to Detect Cracking ................................................................................................

2-17 AREVA NP Inc. Response to NRC Requests for Additional Information on the Three Mile Island Unit 1 List of Figures ANP-2952Q1 NP Revision 0 Figure 2-1: Original Vent Valve Locking Device ..........................................................

2-5 Figure 2-2: Original Vent Valve Locking Device Cross-Section View ..............

.........

2-6 AREVA NP Inc. Acronym ASME B&W CRGT CSS FMECA IE IMI lSI ISR LCB LOCA NRC NSSS PWR PWROG RAI RV RVl a SCC SE(R) SS TE TMI-1 UCB UT VT (-1, -2, -3) Nomenclature Definition American Society of Mechanical Engineers Babcock & Wilcox Control Rod Guide Tube Core Support Structure Failures Modes and Effects Criticality Analysis Irradiation Embrittlement Incore Monitoring Instrumentation In-Service Inspection Irradiation-enhanced Stress Relaxation Lower Core Barrel Loss of Coolant Accident United States Nuclear Regulatory Commission Nuclear Steam Supply System Pressurized Water Reactor Pressurized Water Reactor Owner's Group Request for Additional Information Reactor Vessel Reactor Vessel Internals Stress Corrosion Cracking Safety Evaluation (Report) Stainless Steel Thermal Embrittlement Three Mile Island, Unit 1 Upper Core Barrel Ultrasonic Testing Visual Testing (Level 1, Level 2, Level 3) ANP-2952Q1 NP Revision 0 a RVI is used in the direct quotes from the NRC to indicate Reactor Vessel Internals.

In the AREVA responses, other forms such as RV internals or reactor vessel internals are used to avoid confusion with Reactor Vessel Integrity (RVI), as is commonly used in other AREVA reports.

AREVA NP Inc. h'<!nnln"", to NRC Requests for Additional Information on the Three Mile Island Unit 1 ABSTRACT ANP-2952Q1 NP Revision 0 AREVA document ANP-2952, Revision 1, "Inspection Plan for the Three Mile Island Unit 1 Reactor Vessel Internals," was prepared by AREVA for Exelon Nuclear and subsequently submitted to the Nuclear Regulatory Commission (NRC) by Exelon Nuclear. The NRC has issued Requests for Additional Information (RAls) on this submittal, and this report provides the answers for RAls 1, 3, 4 (Parts 1 and 2), 5 and 6.

AREVA NP Inc. Response to NRC Requests for Add!tionallnformation on the Three Mile Island Unit 1

1.0 INTRODUCTION

AND

SUMMARY

ANP-2952Q1 NP Revision 0 AREVA document ANP-2952, Revision (1), "Inspection Plan for the Three Mile Island Unit 1 Reactor Vessel Internals," was prepared by AREVA for Exelon Nuclear and subsequently submitted to the Nuclear Regulatory Commission (NRC) by Exelon Nuclear (2). The NRC has issued Requests for Additional Information (RAls) (3) on this submittal and this report provides the answers to those RAls assigned to AREV A. Upon receipt of the draft RAls, Exelon Nuclear and AREVA reviewed the RAls and determined who would answer each RAI. For completeness, this document lists all 6 RAls; however, the responses for those RAls that were assigned to Exelon Nuclear only say "Assigned to Exelon Nuclear by the Division of Responsibility."

AREVA NP Inc. Response to NRC Requests for Additional Information on the Three Mile Island Unit 1 ANP-2952Q1 NP Revision a 2.0 REQUESTS FOR ADDITIONAL INFORMATION (RAls) AND RESPONSES The NRC RAls are reproduced from Reference (3) in Sections 2.1.1 through 2.6.1. The reference numbers cited in these sections are for the references listed in Reference (3). The AREV AlExelon Nuclear responses are in Sections 2.1.2 through 2.6.2. 2.1 RAI-1 2.1.1 Statement of RAI-1 In Section 4.1.6 of the RVI 1 inspection plan (Reference 2), the licensee indicated that it has performed analyses to determine the tolerance of the RVI to broken or degraded upper core barrel (UCB) and lower core barrel (LCB) bolts. The analyses used the stress limits of the ASME Code,Section III, Subsection NG for threaded fasteners.

Five different hypothetical combinations of degraded bolts were analyzed, which showed that large numbers of degraded bolts could be tolerated provided the degraded bolts are not adjacent.

If the degraded bolts are adjacent, the number of degraded bolts that can be tolerated is less. Section 5.6.2 of the RVI inspection plan states that the component degradation that exceeds the examination acceptance criteria will be evaluated per MRP-227-A, but will also be evaluated considering the supplemental guidance in WCAP-17096, including any additional guidance resulting from the ongoing NRC review of that document.

Please provide the following information:

1) Are the UCB and LCB bolt analyses consistent with the recommended methodology and acceptance criteria of WCAP-17096-NP, Revision 2 1 RVI is used in Reference (3) to indicate Reactor Vessel Internals.

In the AREVA responses, other forms such as RV internals or reactor vessel internals are used to avoid confusion with Reactor Vessel Integrity (RVI), as is commonly used in other AREVA reports.

AREVA NP Inc. Response to NRC Requests for Additional Information on the Three Mile Island Unit 1 ANP-2952Q1 NP Revision a (Reference 3), as modified by the NRC staff's draft safety evaluation of that report? 2) If not, will the analyses be revised to be consistent with the approved version of WCAP-17096-NP if degradation is found in the bolts? 2.1.2 Response to RAI*1 2.1.2.1 Response to Part 1: The methodology and acceptance criteria used in the analysis referenced in Section 4.1.6 of ANP-29S2, Revision 1 is consistent with the methodology and acceptance criteria of WCAP-17096-NP, Revision 2 as modified by the NRC's draft safety evaluation of that report. 2.1.2.2 Response to Part 2: This Part is not applicable, based on the response to Part 1. 2.2 RAI*2 2.2.1 Statement of RAI-2 In Section 4.1.7 of the RVI inspection plan, the licensee described RVI inspections that have been performed in the past at TMI-1. These include vent valve inspections, UCB bolt ultrasonic examinations (UT), and core clamping measurements.

One hundred percent of the UCB bolts were examined in 2009 via UT with no recordable indications found. One hundred percent UT examination of the UCB bolts was also performed in 1991 with no recordable indications.

MRP-227-A specifies UT examination of the UCB bolts within two refueling outages of January 1, 2006, or the next ten-year inservice inspection (lSI) interval, whichever comes first. Please provide the following information:

AREVA NP Inc. Response to NRC Requests for Additional Information on the Three Mile Island Unit 1 ANP-2952Q1 NP Revision 0 1) Confirm that the UCB bolt examinations performed in 2009 constitute the initial MRP-227 -A inspection requirement for the UCB bolts. 2) Will the UCB bolts be reinspected during the Fall 2015 refueling outage, when MRP-227-A inspections are scheduled in conjunction with the ASME Section XI inspections of the RVI? 2.2.2 Response to RAI*2 Assigned to Exelon Nuclear by the Division of Responsibility.

2.3 RAI-3 2.3.1 Statement of RAI-3 In Attachment 2 to Exelon's submittal dated April 16, 2012, the licensee committed to perform an evaluation of the RVI vent valve locking devices, which are a TMI-1-specific component not covered by the generic MRP-227-A aging management recommendations.

The licensee provided the results of the review of the vent valve locking device in its letter dated April 17, 2013 (Reference 4). In Reference 4, the licensee stated that the Pressurized Water Reactor Owner's Group (PWROG) proposes to accommodate the vent valve locking devices as an existing program within Table 4-7, "B&W [Babcock & Wilcox] plants Existing Programs components," of MRP-227-A.

The licensee stated that the vent valve locking devices shall be addressed by ASME Code,Section XI Examination Category B-N-3, per BAW-2248-A (Reference 5), and that these inspections shall require a VT -3 examination of 100% of accessible surfaces of the vent valve locking devices during each 10-year lSI interval.

The licensee further stated that TMI-1 1 will examine the vent valve locking devices under the ASME Section XI lSI program. The licensee finally stated that this commitment is complete.

Please provide the following information:

Provide a summary of the evaluation of the RVI vent valve locking devices, including the material type, screening parameters (temperature, neutron fluence, stress) results of the screening for degradation mechanisms, Failure Modes, Effects and Consequence AREVA NP Inc. Response to NRC Requests for Additional Information on the Three Mile Island Unit 1 ANP-2952Q1 NP Revision 0 Analysis results, initial categorization, and how the final inspection category of the locking devices was determined (Existing Programs).

2.3.2 Response to RAI-3 There are two types of vent valve locking devices installed in B&W units, "original" and "modified".

By review of inspection video, TMI-1 is confirmed to have only "original" locking devices, shown in Figure 2-1 and Figure 2-2. The evaluation of these "original" locking devices is summarized as follows: material type (Table 2-1), screening parameters (temperature, neutron f1uence, stress) (Table 2-2), results of the screening for degradation mechanisms (Table 2-3), Failure Modes and Effects Criticality Analysis (FMECA) results (Table 2-4) , initial categorization (Table 2-5), final inspection category (text following Table 2-5). The applicable screening criteria from Table 3-1 of MRP-189, Revision 1 (4) were used to screen for degradation mechanisms.

Operating stresses for the locking devices were not found; therefore, they are assumed to exceed the applicable screening criteria.

AREVA NP Inc. Response to NRC Requests for Additional Information on the Three Mile Island Unit 1 Figure 2-1: Original Vent Valve Locking Device ANP-2952Q1 NP Revision a AREVA NP Inc. Response to NRC Requests for Additional Information on the Three Mile Island Unit 1 ANP-2952Q1 NP Revision 0 Figure 2*2: Original Vent Valve Locking Device -Cross-Section View AREVA NP Inc. ANP-2952Q1 NP Revision 0 Response to NRC Requests for Additional Information on the Three Mile Island Unit 1 Table 2*1: Vent Valve Locking Device Material Identification Drawing Part Number Part Description Material AREVA NP Inc. Response to NRC Requests for Additional Information on the Three Mile Island Unit 1 Reactor Vessel Internals Inspection Plan Table 2-2: Vent Valve Locking Device Screening Parameters Material Type, Neutron Cold Grade, Exposure Operating Work Part Part Material Material or Temp (n/cm 2 , Stress Number Description Category Spec Class (OF) E>1MeV (ksi) 20% Multi-Pass Weld ANP-2952Q1NP Revision 0 Page 2-8 Fatigue Usage Factor Boltin Note a -Operating stresses for the vent valve locking device parts were not located during the preparation of this task. Note b -Alloy A-286 spring wire is assumed to have surface cold work following spring manufacturing.

Note c Ferrite content is estimated to be between 5 and 15% (5)

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'" N ,," N Eeo. Cons. Safety Risk Band '" ----" ., ----Eeo. Risk Band ;:-;:-> > ;:-OJ OJ OJ OJ Category 0-" ::J Z 0"'0 AREVA NP Inc. Response to NRC Requests for Additional Information on the Three Mile Island Unit 1 ANP-2952Q1 NP Revision 0 The categorization of component items uses the initial screening results (Table 2-3) and the FMECA results (Table 2-4) to separate the component items into Category A, B, and C. Definitions of Categories A, B, and C and their respective criteria are listed in Section 5 of MRP-189, Rev. 1. The results are shown in Table 2-5.

AREVA NP Inc. Response to NRC Requests for Additional Information on the Three Mile Island Unit 1 ANP-2952Q1 NP Revision 0 Table 2*5: Initial Categorization of Vent Valve Component Items I/) Q. E 0) --0) I/) CD c: c: c: e! CD CD '0, CD c: CD c: t-.:t:: -(J (J CD E 0 E U) -t-:::s -( := <ii c: 0 (J (J ctJ CD CD "0 Part Description (J CD 0) E ctJ ::: c: c: CD 0) U) 3: := ctJ :0 U) 0 -( c: CD U) S ctJ E ';: ';: := 0 -ctJ ctJ U. t-.Q t-.Q "0 ,!! >< Q. (J CD E t-E '0 "0 ctJ E .r:. w W > ctJ -a; 0 l-t-t-o:: (J [ ] A A B A A A A A B [ ] A A B A A A A A B [ ] A B A A A A A B [ ] A A B A A A A A B [ ] A A A A A A A A A [ ] A A A A A A A A A [ ] A A A A A A A A A [ [ ] A A A A A A A A ] A A A A A A A A The component items screened as non-Category A component items in Table 2-5 are as follows: . [ ] for wear A A VVearbetweenthe

[ ] was the driving factor for the initial replacement of the original vent valves near the outlet nozzle locations, During video inspections of the valves at B&VV units in the late 1970s, damage to the original vent valves was seen [ ] TMI-1 did not observe this condition and subsequent examinations have not identified such degradation of the vent valves, Since the cause of the damage was determined to be from the vent valves within close proximity to the outlet nozzles due to the high flow in those locations it is unlikely that AREVA NP Inc. Response to NRC Requests for Additional Information on the Three Mile Island Unit 1 [ ] ANP-2952Q1 NP Revision 0 ] have limited As seen in Figure 2-1, the [ accessibility for inspection.

The [ jackscrew is [ ] is inaccessible.

The function of the ] The function of the vent valve locking device is [ ] The retaining rings are considered primary component items within the MRP-227 program at TMI-1 and are to be inspected via visual VT-3 inspection on the 10-year lSI interval.

If the function of [ ] were degraded during service, the visual inspection of the retaining rings, in the vicinity of these component items, would indicate that the jackscrew

[ ] The observation that the [ ] is not damaged due to loss of material and the confirmation that the jackscrew

[ ] is verification that the locking devices are acceptable.

Additionally, TMI-1 performs a visual inspection each refueling outage and will perform a VT-3 inspection of the locking devices during the 10-year ASME B&PV Code service inspection (lSI) intervals of the vent valves. The accessible surfaces of the vent valves are visually inspected, including the locking devices. Previously, this wear damage to the original vent valves was also detected via visual inspection.

If similar wear does occur, it will likely be detected via the visual inspection performed each refueling outage and the VT-3 inspection during the 10-year lSI intervals.

The SE to BAW-2248 states that the staff considers visual inspection (VT-3) adequate for AREVA NP Inc. Response to NRC Requests for Additional Information on the Three Mile Island Unit 1 ANP-29S2Q1 NP Revision 0 detecting wear of components in the scope of BAW-2248 (which includes the original locking devices).

[ ] For these reasons, the original vent valve locking device [ ] are further categorized as "Existing Programs" for TMI-1. It is noted that Table 4-7 of MRP-227 -A does not currently identify the existing program in place for vent valve locking device aging management at TMI-1. AREVA has established interim guidance to account for this discrepancy and incorporate the appropriate information in the next revision of MRP-227. 2.4 RAI-4 2.4.1 Statement of RAI-4 NUREG-1801, Revision 2, "Generic Aging Lessons Learned Report," (GALL Report, Revision 2), Chapter XI.M16A, "PWR (pressurized water reactor) Vessel Internals Program," states, under "Detection of Aging," that the VT-3 visual methods may be applied for the detection of cracking only when the flaw tolerance of the component or affected assembly, as evaluated for reduced fracture toughness properties, is known and has been shown to be tolerant of easily detected large flaws, even under reduced fracture toughness conditions.

In the licensee's evaluation of consistency with the "Detection of Aging Effects" attribute of GALL, Revision 2, Chapter XI.M16A, the licensee stated that VT -3 examinations will be used to detect cracking only after evaluation of the flaw tolerance of the component or affected assembly, under reduced fracture toughness conditions, has been shown to be tolerant of easily detected flaws. The NRC staff notes that there are six TMI-1 Primary components and three Expansion AREVA NP Inc. Response to NRC Requests for Additional Information on the Three Mile Island Unit 1 ANP-2952Q1 NP Revision 0 components for which visual VT-3 examination is specified for detection of cracking (consistent with MRP-227-A).

License Renewal Interim Staff Guidance LR-ISG-2011-04, "Updated Aging Management Criteria for PWR Reactor Vessel Internal Components," (Reference

6) modified the statement regarding VT -3 examination such that a flaw tolerance evaluation would only be required for non-redundant components.

Reference 1 also stated that VT -3 visual methods are acceptable for the detection of cracking in redundant RVI components (e.g., redundant bolts or pins used to secure a fastened RVI assembly).

Please provide the following information:

1) Which TMI-1 RVI components require flaw evaluations to justify VT-3 examination?

2) Are any of the components for which VT-3 examination is specified redundant components that do not require a flaw evaluation?

If so, justify considering the components redundant.

3) How will the flaw tolerance evaluations of the TMI-1 Primary and Expansion components be documented and communicated to the NRC to support the NRC staff s review? 4) Confirm that these flaw tolerance evaluations will be completed prior to the initial inspections of the TMI-1 Primary components, scheduled for the fall 2015 refueling outage. 5) Will the flaw tolerance evaluations of the control rod guide tube spacer castings and incore monitoring instrumentation guide tube spiders be part of the specific evaluation of cast austenitic stainless steel components that the licensee committed to submit to the NRC by fall 2015?

AREVA NP Inc. Response to NRC Requests for Additional Information on the Three Mile Island Unit 1 2.4.2 Response to RAI-4 ANP-2952Q1 NP Revision 0 Parts 3, 4, and 5 are assigned to Exelon Nuclear by the Division of Responsibility.

2.4.2.1 Response to Part 1: The six TMI-1 Primary components and three Expansion components for which visual VT-3 examination is specified for detection of cracking and their required flaw evaluations are listed in Table 2-6. Redundancy arguments are justified in the Response to Part 2 of this RAI.

AREVA NP Inc. ANP-2952Q1 NP Revision 0 Response to NRC Requests for Additional Information on the Three Mile Island Unit 1 Category Primary Expansion Table 2-6: Flaw Tolerance Evaluation Requirements for VT-3 Examinations to Detect Cracking Flaw Tolerance Item Evaluation Required?

Control Rod Guide Tube Assembly:

CRGT No; Item is redundant (See spacer castings Response to Part 2 of this RAI) Core Support Shield Assembly:

Vent Valve No; Item is redundant (See Top and Bottom Retaining Rings Response to Part 2 of this RAI) Core Barrel Assembly:

Baffle Plates Yes (See Exelon Nuclear responses to Parts 3 through 5 of this RAI) Core Barrel Assembly:

Locking devices, No; Item is redundant (See including locking welds, of baffle-to-former Response to Part 2 of this bolts and internal baffle-to-baffle bolts RAI) Lower Grid Assembly:

Alloy X-750 dowel-No; Item is redundant (See to-guide block welds Response to Part 2 of this RAI) Incore Monitoring Instrumentation (IMI) No; Item is redundant (See Guide Tube Assembly:

IMI guide tube Response to Part 2 of this spiders and IMI guide tube spider-to-Iower RAI) grid rib section welds Upper Grid Assembly:

Alloy X-750 dowel-No; Item is redundant (See to-upper grid fuel assembly support pad Response to Part 2 of this welds RAI) Lower Grid Assembly:

Lower grid fuel No; Item is redundant (See assembly support pad items: pad, pad-to-Response to Part 2 of this rib section welds, Alloy X-750 dowel, cap RAI) screw, and their locking welds Lower Grid Assembly:

Alloy X-750 dowel-No; Item is redundant (See to-lower grid fuel assembly support pad Response to Part 2 of this welds RAI) 2.4.2.2 Response to Part 2: Yes, many of the items identified as susceptible to cracking due to aging are considered redundant.

Redundant has two interpretations.

One interpretation is that several items support each other in performing a given function and the second is that an item has such a minor functional requirement that its degradation will not lead to loss of function.

AREVA NP Inc. Response to NRC Requests for Additional Information on the Three Mile Island Unit 1 ANP-2952Q1 NP Revision 0 While TMI-1 specific component functional evaluations have not yet been performed, given the similarity of TMI-1 's RV internals to RV Internals at other units, TMI-1 specific component functional assessments are expected to confirm the redundancy of these items. The items are listed below with a brief discussion on the basis of the redundant classification.

a) CRGT Spacer Casting-The spacer castings are a portion of the brazement assembly that provides the defined guide path for the control rod assemblies.

There are 69 brazement assemblies that contain ten spacer castings each. The actual guide path is provided by the continuous rod guide tubes and rod guide sectors that are brazed to the spacer castings.

The spacer castings are attached by four screws at the quadrant locations to the pipe weldment to form the 69 control rod guide tube assemblies.

The spacer castings provide no support function for the fuel. The spacer castings do provide the alignment features to allow the control rods to pass freely through the brazement.

The spacer castings are considered redundant items. The reactivity analysis assumes that one control rod may not trip on command and the brazement guide path has been evaluated for suitability in the event a spacer casting fractures.

Any given spacer casting within a brazement assembly can fracture without loss of function of the guide path and multiple spacer castings within a brazement can fracture if they are not axially adjacent to each other without loss of function.

Fractures would be the result of original existing flaws in the presence of high stresses induced by the screw preloading and reduced fracture toughness due to thermal embrittlement.

The spacer castings all have essentially the same low probability of failure regardless of location.

The VT-3 examination is adequate to verify that no spacer castings have a through thickness fracture.

b) CSS Vent Valve Retaining Rings-There are eight vent valves on the TMI-1 unit that are closed during normal operation to control bypass flow and are designed to open when a reverse pressure is imposed on the valve disc to enable proper core cooling. The vent valves are basically two parts. One part is the valve body and disc assembly and the second part is the retaining rings/jackscrew AREVA NP Inc. Response to NRC Requests for Additional Information on the Three Mile Island Unit 1 ANP-2952Q1 NP Revision 0 assemblies to hold the valve body in place. The eight vent valves are independent and not assumed to be redundant items. The retaining rings are low stressed and have a high tolerance for fracture even in a reduced toughness state with the exception of the threaded bosses. The threaded bosses transfer the axial loading of the jackscrews that hold the retaining ring in place. The retaining ring threaded bosses are redundant due to the fact that the loading of one jackscrew will keep the retaining rings and consequently the valve body in place. No vent valve function is lost as long as the valve body is in place. The VT -3 examination is sufficient to verify that the jackscrew and consequently the retaining ring threaded boss are in the design configuration.

c) Locking Welds of Baffle-fa-Former and Internal Baffle to Baffle Bolts-The locking weld functions are to retain the bolt head in the event of a bolt failure and to prevent backing out of a bolt due to vibration.

The locking welds are redundant due to the minor functional requirement and the fact that the bolts themselves are evaluated for random failures.

The driving stress to initiate IASCC or other causes of cracking is the residual stress in the weld at assembly.

Cracking of the weld alone does not cause loss of function.

Function is only lost if there is complete severance of the weld from the baffle plate. VT-3 examination is adequate to detect complete severance of the weld or absence of the bolt itself. d) Alloy X-7S0 Dowel to Guide Block Weld-There are 12 sets of two guide blocks. Each guide block has one screw and one Alloy X-7S0 dowel that secure the guide block to the lower grid assembly.

The Alloy X-7S0 dowel is in the horizontal position and the weld is to keep the dowel from backing out under operating conditions.

The dowel weld is considered redundant for at least two reasons. The weld does not have to be crack free. Complete severance of the weld from the guide block counterbore opening is required to allow the dowel to lose engagement.

In addition all 24 guide blocks are not required to perform the function of stabilizing the core on the core guide lugs in the event of a severe loss of the normal core support load path. The guide blocks have no function AREVA NP Inc. Response to NRC Requests for Additional Information on the Three Mile Island Unit 1 ANP-2952Q1 NP Revision 0 during normal operation.

The VT-3 examination is sufficient to verify the presence of the locking weld and consequently the Alloy X-750 dowel. e) IMI Spider Casting-There are 52 IMI spider castings.

The configuration is composed of a spider casting with four extensions that are welded to the lower grid pad pockets by two fillet welds on each extension.

The IMI spider castings are not core support items. The IMI spider castings align the upper portion of the incore instrument guide tube to the fuel assembly opening for the incore instrument.

Each incore instrument location is independent of the others and is treated individually.

The redundancy of the IMI spider casting is supported by the evaluation that one extension of the IMI spider (including or in addition to the complete failure of one set of fillet welds on that extension) can be failed and the IMI spider casting can still perform its function.

The evaluation has also concluded that failure of one leg does not increase the likelihood of failure of another extension. The conditions leading to possible failure are an original defect present in a reduced fracture toughness region subjected to the forces caused by residual stresses on the fillet welds. The VT-3 examination is sufficient to identify a failed extension on an IMI spider casting. f) Alloy X-750 Upper Grid Dowel Welds The upper grid pad dowel welds are an expansion item from the Alloy X-750 guide block dowel welds. There are 384 upper grid pads with two dowels and corresponding locking welds at each pad. Each pad functions independently but one dowel in a pad is sufficient to perform the shear loading resistance function.

The locking weld prevents the dowel from backing out of the grid pad opening and losing engagement with the upper grid rib. Cracking of the weld does not result in loss of function.

Complete severance of the weld from the grid pad is required to allow the movement of the dowel. Therefore redundancy is achieved by the presence of two dowels and the requirement of a specific failure to lose function.

The VT-3 examination is sufficient to verify the presence of the locking weld and consequently the Alloy X-750 dowel.

AREVA NP Inc. Response to NRC Requests for Additional Information on the Three Mile Island Unit 1 ANP-2952Q1 NP Revision 0 g) Alloy X-750 Lower Grid Dowel Welds-These welds are an expansion item from the Alloy X-750 guide block welds and have identical features of the upper grid dowel welds with the exception of the orientation.

The lower grid dowels cannot fall out of the lower grid pads but must be pulled out during a re-fueling operation if complete severance of the locking weld is present. Therefore the same arguments apply as for the upper grid pad dowels with the added argument that the missing dowel would probably be detected during the refueling operation.

In addition the lower grid pads have fillet welds to the lower grid ribs along each end of the pad. These welds would also serve as a redundant load carrying path if one or both dowels were missing. The VT-3 examination is sufficient to verify the presence of the locking weld and consequently the Alloy X-750 dowel. h) Lower Grid Support Items-This group is an expansion from the IMI spider casting for the aging mechanism irradiation embrittlement.

The Alloy X-750 dowel weld is repeated due to a different aging mechanism but the discussion would be the same as above for stress corrosion cracking.

The remaining items of the grid pad-to-rib section welds and cap screw locking welds are considered to be redundant items per the following discussion.

The grid pad-to-rib section welds were added as a precaution after the first B&W unit sustained substantial damage during hot functional testing due to flow induced vibration issues. In effect, the grid pad-to-rib section welds are a backup for the Alloy X-750 dowels and the stainless steel cap screws. One or the other is needed but not both. The stainless steel cap screws have no function in normal operation but to hold the lower grid pads in place during the refueling operation if the fuel assembly does not freely lift from the lower grid pads. The cap screw locking welds prevent the cap screw from turning out of the threaded hole during normal operation.

The VT-3 examination is sufficient to verify the presence of the locking welds for the cap screws and consequently the presence of the cap screw. The VT-3 examination of the grid pad-to-rib section weld is sufficient to verify cracking.

Any observed cracking is treated as a complete failure of the weld.

AREVA NP Inc. Response to NRC Requests for Additional Information on the Three Mile Island Unit 1 2.5 RAI-5 2.5.1 Statement of RAI*5 ANP-2952Q1 NP Revision 0 Appendix A of MRP-227-A discusses operating experience under various degradation mechanisms for the RVI components of all nuclear steam supply system (NSSS) designs, including B&W plants. Please provide the following information:

1) Identify the MRP-227-A, Appendix A experience that occurred at TMI-1. 2) Describe any TMI-1 experience with RVI component degradation that is not discussed in Appendix A of MRP-227 -A, and is not discussed in Section 4.1.7 of the RVI inspection plan. 3) Describe any changes to the RVI inspection plan made as a result of the operating experience described in the response to Part 2 of this RAI question.

2.5.2 Response to RAI-5 2.5.2.1 Response to Part 1: None of the operating experience of aging degradation described in Appendix A of MRP-227-A (6) occurred at TMI-1. 2.5.2.2 Response to Part 2: No additional operating experience of aging degradation has occurred in the TMI-1 reactor vessel internals subsequent to the release of MRP-227-A or Section 4.1.7 of the reactor vessel internals inspection plan. Upset metal has been observed on core support shield vent valve jackscrew locking devices and gouges and scratches have been observed on the core support shield round bars (or, "Ioss-of-cooling-accident (LOCA) lugs"). It is known that the plenum assembly can impact the core support shield if the plenum assembly is out of level AREVA NP Inc. Response to NRC Requests for Additional Information on the Three Mile Island Unit 1 ANP-2952Q1 NP Revision 0 when removed or installed.

Therefore, these observations are not considered aging degradation.

No other operating experience of non-aging related degradation is known for components in the scope of the TMI RV Internals program. The following operating experience is available for components outside of the scope of the RV internals aging management plan. 1) A 5° to 10° bend was observed in several thermocouple guide tube assembles.

This condition is known to have occurred prior to plant startup. As discussed in BAW-2248-A (7), the thermocouple guide tube assemblies are part of the plenum assembly and mate with penetrations in the reactor vessel closure head. The thermocouple guide tube assemblies serve no function.

These assemblies are not identified in the TMI-1 License Renewal Application Aging Management Review and are therefore out of the scope of the RV internals aging management plan. 2) During the first refueling outage (1976), damage was observed in the surveillance specimen holder capsules and holder tubes and was concluded to have been caused by flow induced vibration (8). The damaged components were removed during the outage and have not been replaced.

Several remnants of the holder tube assembly exist but they serve no function.

These remnants are not identified in the TMI-1 License Renewal Application Aging Management Review and are therefore out of the scope of the RV internals aging management plan (7). 2.5.2.3 Response to Part 3: There are no changes to the RV internals inspection plan based on the response to Part 2.

AREVA NP Inc. Response to NRC Requests for Additional Information on the Three Mile Island Unit 1 2.6 RAI-6 2.6.1 Statement of RAI-6 ANP-2952Q1 NP Revision 0 Appendix 0 to the RVI inspection plan summarizes the TMI-1 response to the Applicant/Licensee Action Items from MRP-227-A.

With respect to Action Item 2, the licensee stated that the only TMI-1 component not addressed by MRP-227-A was the vent valve locking devices. However, the discussion of Applicant/Licensee Action Item 2 does not address whether any components were fabricated from materials not consistent with "Materials Reliability Program: Screening, Categorization, and Ranking of 8&W-Designed PWR Internals Component Items (MRP-189-Revision 1)" (Reference 7). Confirm that the materials of the TMI-1 RVI components are consistent with those listed in MRP-189. If any materials are different, discuss whether additional aging mechanisms screened in for the components.

Provide plant-specific aging management recommendations for these components as necessary to address the additional aging mechanisms.

2.6.2 Response to RAI-6 In the RV Internals inspection plan (1), Appendix 0 references Appendix E for a detailed comparison of components and welds 2 in the scope of License Renewal relative to components in MRP-189, Revision 1 (4). Therefore the MRP-189, Revision 1 material specification and material Type, Grade, or Class were identified for all components in Appendix E, except the vent valve locking devices, control rod assembly and fuel assembll.

The MRP-189, Revision 1 material information was then compared against the material information identified in the TMI-1 fabrication records. 2 Welds are identified in Appendix E by reference to Table 4-2 of MRP-189, Revision 1 in the "Supporting Document and Location" column. 3 The vent valve locking devices are addressed in the Response to RAI 3; the control rod assembly and fuel assembly are "short lived" components and have no associated aging management programs.

AREVA NP Inc. Response to NRC Requests for Additional Information on the Three Mile Island Unit 1 ANP-295201 NP Revision 0 Minor material fabrication discrepancies were identified for several component items: A 240 Type 304 (plate product form) vs A 276 Type 304 (bar product form) for the Lower Grid Assembly Shock Pads and A 240 Type 304 vs SA-240 Type 304L for the Control Rod Guide Tube Assembly Rod Guide Tubes and Rod Guide Sectors. In numerous cases, Type 30B (or 30BL or both 30B and 30BL) weld filler metal was used at TMI-1 where MRP-1B9, Revision 1 reports Type 30BL (or 30B or both 30B and 30BL) weld filler metal. Similarly, Alloy 69 was occasionally used at TMI-1 where MRP-1B9, Revision 1 reports Alloy B2. In all these cases, the screening criteria in MRP-1B9, Revision 1 do not differentiate among these material differences.

Therefore, no additional aging degradation mechanisms are screened in as a result of these differences and no additional aging management recommendations are necessary.

No material fabrication differences were found for any other components reviewed.

Therefore no additional aging management recommendations are necessary.

AREVA NP Inc. Response to NRC Requests for Additional Information on the Three Mile Island Unit 1

3.0 REFERENCES

ANP-2952Q1 NP Revision 0 1 ANP-2952, Revision 1, (AREVA Document ID 77-2952-001), "Inspection Plan for the Three Mile Island Unit 1 Reactor Vessel Internals," March, 2012. 2 Exelon letter, TMI-12-069 to the NRC dated April 16, 2012 (ADAMS Accession No. Ml12108A029), Attachment 1, "Inspection Plan for the Three Mile Island Unit 1 Reactor Vessel Internals," AREVA Document 77-2952-001, ANP-2952, Revision 001, dated March 2012. 3 Memorandum from R. B. Ennis to V. M. Rodriguez, "Three Mile Island Nuclear Station, Unit 1, Draft Request for Additional Information (TAC No. MF1459), dated September 25, 2013 (NRC ADAMS Accession Number Ml 13269A 179). 4 Materials Reliability Program: Screening, Categorization, and Ranking of Designed PWR Internals Component Items (MRP-189-Rev.

1). EPRI, Palo Alto, CA: 2009. 1018292. 5 Materials Reliability Program: PWR Internals Material Aging Degradation Mechanism Screening and Threshold Values (MRP-175).

EPRI, Palo Alto, CA: 2005. 1012081. 6 Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A).

EPRI, Palo Alto, CA: 2011. 1022863. 7 BAW-2248-A, (AREVA Document ID 43-2248A-00), "Demonstration of the Management of Aging Effects for the Reactor Vessel Internals," March 2000. 8 Metropolitan Edison Company letter GQl 0378, "Event Report 76-12/1T", March 18, 1976, NRC ADAMS Accession Number 7910260668.

Attachment 5 Affidavit AFFIDAVIT STATE OF WASHINGTON ) ) 55. COUNTY OF BENTON ) 1. My name is Alan B. Meginnis.

I am Manager, Product Licensing, for AREVA NP Inc. and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by AREVA NP to determine whether certain AREVA NP information is proprietary.

I am familiar with the policies established by AREVA NP to ensure the proper application of these criteria.

3. I am familiar with the AREVA NP information contained in the report ANP-2952Q1 PRevision 0, "Response to NRC Requests for Additional Information on the Three Mile Island Unit 1 Reactor Vessel Internals Inspection Plan," dated October 2013 and referred to herein as "Document." Information contained in this Document has been classified by AREVA NP as proprietary in accordance with the policies established by AREVA NP for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA NP and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure.

The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4)

'Trade secrets and commercial or financial information." 6. The following criteria are customarily applied by AREVA NP to determine whether information should be classified as proprietary: (a) The information reveals details of AREVA NP's research and development plans and programs or their results. (b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service. (c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA NP. (d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA NP in product optimization or marketability. (e) The information is vital to a competitive advantage held by AREVA NP, would be helpful to competitors to AREVA NP, and would likely cause substantial harm to the competitive position of AREVA NP. The information in the Document is considered proprietary for the reasons set forth in paragraphs 6(b), 6(d) and 6(e) above. 7. In accordance with AREVA NP's policies governing the protection and control of information, proprietary information contained in this Document have been made available, on a limited basis, to others outside AREVA NP only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. AREVA NP policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief. SUBSCRIBED before me this --.;;;:::..-'is.:..-f'-_

day of 0 t:) &/ ,2013. \ Susan K. McCoy NOTARY PUBLIC, STATE OF WASRt GTON MY COMMISSION EXPIRES: 1/14/2016