ML042390469

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License Renewal Application - Response to Requests for Additional Information on Aging Management Programs (TAC Nos. Mc 1202 and Mc 1203)
ML042390469
Person / Time
Site: Cook  American Electric Power icon.png
Issue date: 08/19/2004
From: Jensen J
Indiana Michigan Power Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP:NRC:4034-13
Download: ML042390469 (26)
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Text

Indiana Michigan Power Company 500 Circle Drive Buchanan, Ml 49107 1395 INDIANA MICHIGAN POWER August 19, 2004 AEP:NRC:4034-13 10 CFR 54 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop O-PI-17 Washington, DC 20555-0001

SUBJECT:

Donald C. Cook Nuclear Plant, Units 1 and 2 Docket Nos. 50-315 and 50-316 License Renewal Application - Response to Requests for Additional Information on Aging Management Programs (TAC Nos. MC 1202 and MC 1203)

Dear Sir or Madam:

By letter dated October 31, 2003, Indiana Michigan Power Company (I&M) submitted an application to renew the operating licenses for Donald C. Cook Nuclear Plant, Units I and 2 (Reference 1).

During the conduct of its review, the Nuclear Regulatory Commission (NRC)

Staff identified areas where additional information was needed to complete its review of the license renewal application (LRA). This letter responds to Staff requests for additional information (RAIs) pertaining to the aging management program descriptions in the following LRA sections:

  • B. 1.5 - Bottom-Mounted Instrumentation Thimble Tube Inspection
  • B. 1.24 - Pressurizer Examinations
  • B. 1.26 - Reactor Vessel Integrity
  • B. 1.27 - Reactor Vessel Internals Plates, Forgings, Welds, and Bolting
  • B. 1.31 - Steam Generator Integrity This letter also responds to two RAIs pertaining to the following LRA section:

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U S. Nuclear Regulatory Commission Page 2 AEP:NRC:4034-13 The enclosure to this letter provides an affirmation pertaining to the statements made in this letter. Attachment 1 provides the additional information requested from the NRC Staff. Attachment 2 provides a regulatory commitment made in this letter in response to RAI B.1.27-2 for the new Reactor Vessel Internals Plates, Forgings, Welds, and Bolting Program. It is noted that this commitment supplement the commitments to implement the new Reactor Vessel Internals Plates, Forgings, Welds, and Bolting Program, as summarized on Page 6 of to the LRA cover letter (Reference 1).

Should you have any questions, please contact Mr. Richard J. Grumbir, Project Manager, License Renewal, at (269) 697-5141.

Sincerely, Site Vice President NH/rdw

Enclosure:

Attachments:

Affirmation

1. Response to Requests for Additional Information for the Donald C. Cook Nuclear Plant License Renewal Application
2. List of Regulatory Commitments

References:

1. Letter from M. K. Nazar, I&M, to NRC Document Control Desk, "Donald C.

Cook Nuclear Plant Units 1 and 2, Application for Renewed Operating Licenses,"

AEP:NRC:3034, dated October 31, 2003

[Accession No. ML033070177].

2. Letter from J. Rowley, NRC, to M. K. Nazar, I&M, "Request for Additional Information for the Review of Donald C. Cook Nuclear Plant, Units 1 and 2 License Renewal Application,"

dated June 30, 2004

[Accession No. ML041830088].

U. S. Nuclear Regulatory Commission AEP:NRC:4034-13 Page 3

3. Letter from J. Rowley, NRC, to M. K. Nazar, I&M, "Request for Additional Information for the Review of Donald C. Cook Nuclear Plant, Units I and 2 License Renewal Application,"

dated June 30, 2004

[Accession No. ML041840218].

4. Letter from J. Rowley, NRC, to M. K. Nazar, I&M, "Request for Additional Information for the Review of Donald C. Cook Nuclear Plant, Units I and 2 License Renewal Application,"

dated July 26, 2004

[Accession No. ML042210285].

C:

J. L. Caldwell, NRC Region III K. D. Curry, AEP Ft. Wayne, w/o attachment J. T. King, MPSC, w/o attachment J. G. Lamb, NRC Washington DC MDEQ - WHMD/HWRPS, w/o attachment NRC Resident Inspector J. G. Rowley, NRC Washington DC

Enclosure to AEP:NRC:4034-13 AFFIRMATION I, Joseph N. Jensen, being duly sworn, state that I am Site Vice President of Indiana Michigan Power Company (I&M), that I am authorized to sign and file this request with the Nuclear Regulatory Commission on behalf of I&M, and that the statements made and the matters set forth herein pertaining to I&M are true and correct to the best of my knowledge, information, and belief.

Indiana Michigan Power Company Site Vice President SWORN TO AND SUBSCRIBED BEFORE ME

5 7 -DAY

,2004 l /

Notary Public My Commission Expires 8 SS~

JULIE E. NEWMILLER Notary Public, Berrien County, MI My Cornmsslon Expires Aug 22,2004

Attachment I to AEP:NRC:4034-13 Page I Response to Requests for Additional Information for the Donald C. Cook Nuclear Plant License Renewal Application This attachment provides Indiana Michigan Power Company's (I&M's) responses to the Donald C. Cook Nuclear Plant (CNP) License Renewal Application (LRA) Requests for Additional Information (RAIs) pertaining to the Aging Management Program descriptions in the following LRA sections:

  • B.1.5 Bottom-Mounted Instrumentation Thimble Tube Inspection
  • B. 1.24 Pressurizer Examinations
  • B. 1.26 Reactor Vessel Integrity
  • B. 1.27 Reactor Vessel Internals Plates, Forgings, Welds, and Bolting

The RAls addressed in this attachment were received in two NRC letters dated June 30, 2004 (References 1 and 2) and a third letter dated July26, 2004 (Reference 3).

References

1. Letter from J. Rowley, NRC, to M. K. Nazar, I&M, "Request for Additional Information for the Review of Donald C. Cook Nuclear Plant, Units 1 and 2 License Renewal Application,"

dated June 30, 2004 [Accession No. ML041830088].

2. Letter from J. Rowley, NRC, to M. K. Nazar, I&M, "Request for Additional Information for the Review of Donald C. Cook Nuclear Plant, Units I and 2 License Renewal Application,"

dated June 30, 2004 [Accession No. ML041840218].

3. Letter from J. Rowley, NRC, to M. K. Nazar, I&M, "Request for Additional Information for the Review of Donald C. Cook Nuclear Plant, Units 1 and 2 License Renewal Application,"

dated July 26, 2004 [Accession No. ML042210285].

Attachment I to AEP:NRC:4034-13 Page 2 RAI 2.3.3.8-6:

The failure of the following components could affect the ability of their associated EDG to perform its intended function and are therefore in the scope of license renewal for meeting criteria 10 CFR 54.4(a)(2):

  • Exhaust silencer QT-104-AB and associated vent stack on LRA-1-5151B-0 at Location N7/8
  • Exhaust silencer QT-104-CD and associated vent stack on LRA-1-5151D-0 at Location N7/8
  • Exhaust silencer QT-104-AB and associated vent stack on LRA-2-5151B-0 at Location N6/7
  • Exhaust silencer QT-104-CD and associated vent stack on LRA-2-5JJ5D-0 at Location N6/7 The exhaust silencers and associated vent stacks are long-lived passive components and are therefore subject to an AMR.

The applicant is requested to confirm that the exhaust silencers and associated vent stacks are in scope and subject to an AMR and identify which "component type" on LRA Table 2.3.3-8 represents them or providejustificationfor their exclusion.

I&M Response to RAI 2.3.3.8-6:

The emergency diesel generator (EDG) exhaust silencers and associated vent stacks are nonsafety-related components whose only functions are to limit the noise created by the diesel engine and complete the transport of exhaust gas to the atmosphere. These components do not perform a function that meets the scoping criteria of 10 CFR 54.4(a)(1) or 10 CFR 54.4(a)(3).

Because they are located outside the EDG rooms and contain air and exhaust gases, they cannot impact safety-related components through spatial interaction as discussed in LRA Section 2.1.1.2.2, and do not meet the scoping criteria of 10 CFR 54.4(a)(2).

Therefore, the EDG exhaust silencers and vent stacks are not subject to aging management review.

RAI 2.3.3.8-7:

The failure of the following components could affect the ability of their associated EDG to perform its intendedfunction and are therefore in the scope of license renewal in accordance with 10 CFR 54.4(a)(2):

  • Centrifugal exhauster QT-140-AB and associatedpiping on LRA-1-515B-0 at Location L3
  • Centrifugal exhauster QT-140-CD and associatedpiping on LRA-1-515JD-0 at Location L3
  • Centrifugal exhauster QT-140-AB and associatedpiping on LRA-2-5151B-0 at Location M3
  • Centrifugal exhauster QT-140-CD and associatedpiping on LRA-2-5151D-0 at Location M3 The centrifugal exhausters and their associated flexible connectors and piping are long-lived passive components and are therefore subject to an AMR.

to AEP:NRC:4034-13 Page 3 The applicant is requested to confirm that the centrifugal exhausters and their associatedflexible connectors and piping are in scope and subject to an AMR and identify which "component type" on LRA Table 2.3. -8 represents them or provide justification for their exclusion.

I&M Response to RAI 2.3.3.8-7:

The EDG centrifugal exhausters and associated flexible connections and piping are nonsafety-related components in the crankcase breather subsystem.

This EDG subsystem maintains a slight vacuum in the crankcase to remove vapors and minimize oil leakage. This function is not required for diesel engine operation, and a failure of these components would not render the EDG inoperable.

Therefore, these components do not serve a license renewal intended function in accordance with the scoping criteria of 10 CFR 54.4(a)(1),

10 CFR 54.4(a)(2), or 10 CFR 54.4(a)(3). Furthermore, because these components contain only air and crankcase gases, their failure cannot adversely impact safety-related components through spatial interaction as discussed in LRA Section 2.1.2.2.2. Consequently, these components do not meet the spatial scoping criteria of 10 CFR 54.4(a)(2). Therefore, the centrifugal exhausters and associated flexible connections and piping are not subject to aging management review.

RAI B.1.4-1:

License Renewal Application (LRA) Section B.1.4, "Boric Acid Corrosion Prevention," states that the scope of Boric Acid Corrosion Prevention Program will be revised to include electrical components in addition toferritic steel. Identify all specific systems and components and their supports, inside and outside containment, that may be susceptible to boric acid corrosion/degradation. Provide information regarding provisions in this program for inspecting, detecting, or monitoring degradation of structures and components due to boric acid leakage and provisions for inspecting, detecting, or monitoring boric acid leakage in inaccessible locations and areas covered by external insulation surfaces.

I&M Response to RAI B.1.4-1:

The Boric Acid Corrosion Prevention Program is credited with managing loss of material and loss of mechanical closure integrity due to boric acid corrosion for component types as indicated in the LRA Section 3 aging management review results tables. This program applies to portions of systems and structures, both inside and outside containment, that are subject to aging management review and contain borated water or are subject to exposure to leaking borated water. This includes electrical connectors, as indicated in LRA Table 3.6.2-1.

Provisions in this program for inspecting, detecting, or monitoring degradation of structures and components due to boric acid leakage, and provisions for inspecting, detecting, or monitoring boric acid leakage in inaccessible locations and areas covered by external insulation surfaces are to AEP:NRC:4034-13 Page 4 consistent with the program described in NUREG-1801,Section XI.MI0, Boric Acid Corrosion.

When a Boric Acid Corrosion Prevention Program inspection detects leakage, the leakage path is followed to identify the source and all affected components along the path, including locations covered by insulation.

As discussed in the Statements of Consideration for the Final Part 54 Rule:

"Given the Commission's ongoing obligation to oversee the safety and security of operating reactors, issues that are relevant to current plant operation will be addressed by the existing regulatory process within the present license term rather than deferred until the time of license renewal. Consequently, the Commission formulated two principles of license renewal.

The first principle of license renewal was that, with the exception of age-related degradation unique to license renewal and possibly a few other issues related to safety only during the period of extended operation of nuclear power plants, the regulatory process is adequate to ensure that the licensing bases of all currently operating plants provides and maintains an acceptable level of safety so that operation will not be inimical to public health and safety or common defense and security.... The second and equally important principle of license renewal holds that the plant-specific licensing basis must be maintained during the renewal term in the same manner and to the same extent as during the original licensing term."

Consistent with the first and second principles of license renewal, on-going boric acid corrosion inspection and evaluation commitments made in support of current operations, including those made in response to NRC Generic Letter 88-05, "Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR Plants," will be carried forward through the period of extended operation.

RAI B.1.4-2:

LRA Section B.1.4 states that Boric Acid Corrosion Prevention Program continues to be improved based on operating experience, and program revisions have incorporated lessons learned from condition reports and industry guidance.

Provide information about these improvements as related to lessons learnedfrom the Davis-Besse vessel head degradation and the control rod drive mechanism penetration cracking discussed in Bulletins 2001-01, 2002-01, 2002-02, and Order EA-03-009. Also, provide a discussion on implementation of corrective actions in the program to prevent the recurrence of degradation caused by boric acid leakage, as required by GL [Generic Letter] 88-05.

to AEP:NRC:4034-13 Page 5 I&M Response to RAI B.1.4-2:

Boric acid corrosion control is an issue that has received, and continues to receive significant regulatory and industry attention. In accordance with the Statements of Consideration for the Final Part 54 Rule, the existing regulatory process, which includes consideration of industry operating experience, will ensure that the Boric Acid Corrosion Prevention Program can effectively monitor the condition of ferritic steel components on which borated water may leak.

Given that boric acid corrosion is relevant to current plant operations, I&M will carry forward, through the period of extended operation, regulatory obligations and commitments made in accordance with the existing regulatory process to address the issue of boric acid corrosion. If additional guidance or requirements are promulgated to address this issue prior to the end of the current licensing period, any regulatory obligations or on-going commitments made in response to those requirements will also apply to the period of extended operation.

Notwithstanding the above, the Control Rod Drive Mechanism and Other Vessel Head Penetration Inspection Program continues to be improved based upon operating experience, as evidenced by program improvements that incorporate lessons learned from the Davis-Besse vessel head degradation and the control rod drive mechanism penetration cracking discussed in Bulletins 2001-01, 2002-01, 2002-02, and NRC Order EA-03-009 and its successors. I&M's obligations to satisfy First Revised NRC Order EA-03-009 supersede the obligations to satisfy NRC Order EA-03-009 and commitments made in response to NRC Bulletins 2002-01 and 2002-02. I&M's current licensing basis for this issue is described in the referenced I&M response to RAI B. 1.9.2-1, dated August 11, 2004.

The Boric Acid Corrosion Prevention Program includes corrective actions to prevent recurrence of degradation caused by boric acid leakage, as required by GL 88-05.

An example was provided in LRA Section B. 1.4, which discussed a recent condition report documenting an accumulation of boric acid crystals on a heat exchanger flange. A brown stain was noted in the acid crystals, indicating corrosion of the carbon steel bolts. The bolts were replaced with stainless steel bolts to prevent recurrence.

Three I&M commitments made in response to NRC Generic Letter 88-05, Requirement No. 4, are documented in correspondence dated June 7, 1988 (Reference 2).

The first of these commitments was to develop a program for replacement of carbon steel packing follow studs on valves within the reactor coolant pressure boundary. The program was developed, and the studs were replaced on the valves that were identified as requiring replacement.

The second commitment was to ensure proper consideration is given to (a) reducing the probability of reactor coolant leaks at locations where they may cause corrosion damage and (b) the use of suitable corrosion resistant materials or the application of protective coatings/claddings.

This commitment was satisfied based on guidance that was incorporated into the boric acid corrosion control procedure. The third corrective action commitment was to review the training programs and procedures to ensure they contain adequate guidance on issues relevant to reactor coolant pressure boundary leakage and corrosion concerns. In July 2003, a biennial evaluation of this

Attachment I to AEP:NRC:4034-13

-Page 6 on-going commitment identified that the training requirements are being met, although the implementing documents (i.e., training modules) do not properly reference the commitment. An action has been entered into the Corrective Action Program to ensure this commitment is referenced properly in the appropriate training modules.

Generic Letter 88-05 actions were audited in July 1989 and found by the NRC to be adequately implemented, as documented in an audit report dated February 22, 1990 (Reference 3).

References for RAI B. 1.4-2

1. Letter from J. N. Jensen, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant, Units 1 and 2, License Renewal Application - Response to Requests for Additional Information on Aging Management Programs (TAC Nos. MC1202 and MC1203),"

AEP:NRC:4034-10, dated August 11, 2004.

2. Letter from M. P. Alexich, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant, Units 1 and 2, NRC Generic Letter 88-05: Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR Plants," AEP:NRC:1061, dated June 7, 1988.
3. Letter from J. G. Glitter, NRC, to M. P. Alexich, I&M, "Prevention of Boric Acid Corrosion at D. C. Cook Units 1 and 2 (Generic Letter 88-05) (TAC Nos. MC1202 and MC1203)," dated February 22, 1990.

RAI B.1.5-1:

LRA Section B.1.5, "Bottom-Mounted Instrumentation Thimble Tube Inspection, " was designed for the detection of wear, not cracking due to SCC. However, in LRA Table 3.1.2-1 "cracking" was listed as an aging effect requiring management for bottom-mounted instrumentation (BMI) thimble tubes and bullet plugs. If SCC is a credible degradation mechanism requiring aging management for BMI thimble tubes, explain how your proposed program is adequate to detect SCC or modify the thimble tube inspection program to include inspections for thimble tube cracking due to SCC. As part of your response, please also address whether the eddy current (ET) examination discussed in the LRA has been qualified to detect and size SCC. Alternatively, information demonstrating that the thimble tubes are not susceptible to SCC and LRA Table 3.1.2-1 should be revised accordingly.

Attachment I to AEP:NRC:4034-13 Page 7 I&M Response to RAI B.1.5-1:

LRA Table 3.1.2-1 lists BMI thimble tube and bullet plug aging effects requiring management (i.e., loss of material and cracking). Loss of material due to wear is managed by the Bottom-Mounted Instrumentation Thimble Tube Inspection Program, as discussed in LRA Section B. 1.5.

Loss of material due to pitting or crevice corrosion is managed by the Primary and Secondary Water Chemistry Control Program, which mitigates loss of material for stainless steel exposed to treated borated water, as discussed in the referenced response to RAI 3.1-3. Cracking due to stress corrosion cracking (SCC) is managed by the Primary and Secondary Water Chemistry Control Program and the Inservice Inspection Program (i.e., leak detection as part of Examination Category B-P). The Bottom-Mounted Instrumentation Thimble Tube Inspection Program eddy current testing (ECT) is not credited for detection and sizing of SCC. No changes to LRA Table 3.1.2-1 are needed.

Reference for RAI B.1.5-1 Letter from M. K. Nazar, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant, Units I and 2, License Renewal Application - Response to Requests for Additional Information on Aging Management Programs," AEP.NRC:3054-12, dated August 11, 2004.

RAI B.1.5-2:

LRA Section B. 1.5 provides the acceptance criteria of BMI thimble tubes as: (1) replacement or isolation of a thimble tube with 80 percent through-wall wear, (2) reposition of a thimble tube with more than 40 percent through-wall wear, provided that it is projected to remain under 80 percent until the next inspection, and (3) replacement, isolation, or reposition of a thimble tube with more than 40 percent through-wall wear if it is projected to exceed 80 percent by the next inspection. Using reposition as an option for Criterion 3 for a tube which is projected to exceed 80-percent wear by the next inspection is inadequate because the uncertainty of the tube wear rate at the selected location for the tube reposition in a certain time period might make the reposition ineffective. Provide a revision of the AMP by incorporating ET uncertainty in future wear measurements and by considering only replacement and isolation of tubes as options for Criterion 3 of the acceptance criteria.

I&M Response to RAI B.1.5-2:

The BMI inspection is based on recommendations provided in WCAP-12866, Bottom Mounted Instrument Flux Thimble Wear. The WCAP demonstrates that thimble tube percent wall loss varies at different core locations over several operating cycles. The current inspection procedure permits relocation of a BMI thimble tube from a location with wear predicted to equal or exceed 80% through-wall by the next inspection to a location that would not result in 80% wear by the to AEP:NRC:4034-13 Page 8 next inspection.

Therefore, a thimble tube can be repositioned to a core location that has historically demonstrated little or no thimble tube wall loss. The final relocation position of a thimble tube predicted to exceed 80% wear will be determined via the corrective action evaluation of the eddy current results. Alternatively, the affected thimble tube may be replaced or isolated. The use of WCAP-12866 for the BMI thimble tube inspection program basis is consistent with the McGuire and Catawba Nuclear Stations LRA in which the Corrective Action and Confirmation Process program element of the Bottom-Mounted Instrumentation Thimble Tube Inspection Program states: "Thimble tubes that are predicted to exceed the acceptance criteria may be capped or repositioned. Specific corrective actions and confirmatory actions are implemented in accordance with the corrective action program." This position was accepted by the staff in NUREG-1772, Safety Evaluation Report Related to the License Renewal of McGuire Nuclear Station, Units I and 2, and Catawba Nuclear Station, Units I and 2.

Reference for RAI B. 1.5-2 Letter from E. E. Fitzpatrick, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant, Units 1 and 2, Response to Confirmatory Action Letter No. RIII 97-011 NRC Architect Engineer (AE) Design Inspection August 1997," AEP:NRC:1260G3, dated December 2, 1997 RAI B.1.5-3:

LRA B. 1. 5 states that ET inspections are scheduled to be performed every third refueling outage.

Provide the basis for determination of this schedule using industry and plant-specifc ET inspection data and considering the anticipated operating conditions during the period of extended operation. It should be noted that the proposed thimble inspection every third outage is only acceptable if no wear has been discovered in the past three refueling outages for all thimble tubes. When wear appears, the inspection interval must be reevaluated based on the observed thimble tube-specific wear rates. Please provide a revised inspection schedule, anticipating wear and based on severity of wear. The UFSAR Supplement should be revised to include a description of this inspection schedule. In addition, discuss any mitigative measures, such as flushing of the tubes, taken during refueling outages. If SCC is determined to be a potential degradation mechanism for thimble tubes in your response to RAI 3.0.3.3-1, provide justification for the inspectionfrequencyfor detecting the SCCflaws.

I&M Response to RAI B.1.5-3:

In accordance with NRC Bulletin 88-09, "Thimble Tube Thinning in Westinghouse Reactors,"

all holders of operating licenses or construction permits for Westinghouse-designed nuclear power reactors that utilize BMI were required to establish and implement an inspection program to periodically confirm incore neutron monitoring system thimble tube integrity. In response to this bulletin, I&M established an incore thimble tube eddy current inspection program with ECT performed at the next refueling outages of both units. Inspections continued to be performed each refueling outage based on observed ECT wear results until changed as discussed below.

Attachment I to AEP:NRC:4034-13 Page 9 During the 1992 refueling outages, 15 thimble tubes were replaced in Unit 1 and 22 thimble tubes were replaced in Unit 2; replacement tubes were chrome-plated at axial locations corresponding to the lower core plate and fuel assembly lower nozzle area. In January 1998, I&M reported that after three cycles of operation, the chrome-plated thimble tubes showed no indications of wear and provided the NRC with an update of plans to replace the remaining design thimble tubes with chrome-plated tubes.

In 2000, I&M completed the replacement of thimble tubes with tubes that are chrome-plated from approximately twelve feet below the tip for a length of approximately fourteen feet. This encompasses the entire lower reactor internal region from immediately above the lower fuel nozzle into conduits below, which covers the most active wear location.

During the 2002 refueling outages, after one cycle of operation, ECT results showed no indication of wear on any of the thimble tubes, at any axial location (from the bullet tip of the thimble tube to the seal table).

ECT inspection results of CNP's chrome-plated thimble tubes have repeatedly demonstrated that the chrome plating effectively mitigates vibration-induced thimble tube wear. Inspection results have shown that three consecutive operating cycles will not degrade chrome-plated portions in the area of the lower nozzle/lower core plate region of the thimble tube. Therefore, based on ECT results for chrome-plated tubes, I&M revised the inspection frequency to every third refueling outage.

Projections of thimble tube wear are based on the applicable inspection interval and are verified through eddy current tests. On-going program requirements and commitments to NRC Bulletin 88-09 will be carried forward through the period of extended operation. This is consistent with the second principle of license renewal, as discussed in the Statements of Consideration for the Final Part 54 Rule, which states that the plant-specific licensing basis must be maintained during the renewal term in the same manner and to the same extent as during the original licensing term.

Should inspection results indicate that more frequent inspections are needed during the current term of operation or the period of extended operation, the ECT frequency will be revised in accordance with the Corrective Action Program. For clarification of the inspection schedule basis, the UFSAR Supplement for this program in LRA Section A.2.1.5 is revised to add the following sentence to the Bottom-Mounted Instrumentation Thimble Tube Inspection Program description:

"The inspection frequency is based on measured data and projected wear results."

Mitigative measures include flushing, drying, and lubricating the thimble tubes each outage. In addition, foreign material exclusion and cleanliness controls are required during maintenance activities.

Attachment I to AEP:NRC:4034-13 Page IO For consistency with other stainless steel components exposed to a treated (borated) water environment, LRA Table 3.1.2-1 lists cracking (including cracking due to SCC) as an aging effect for the component type "BMI thimble tubes and bullet plugs."

This aging effect is managed by the Primary and Secondary Water Chemistry Control Program and the Inservice Inspection Program. A review of operating experience indicated no instances of cracking by SCC of BMI thimble tubes at CNP.

RAI B.1.24-1:

LRA Section B. 1.24, "Pressurizer Examinations, " assesses the cladding and attachment welds to the cladding of the pressurizer. Identify all nickel-alloy welds which were used to attach various penetrations to the pressurizer; confirm that these welds are managed by this AMP and justify that your proposed examinations for them are adequate in terms of the proposed frequency, inspection method, and scope for managing the degradation associated with this type of weld.

I&M Response to RAI B.1.24-1:

There are no nozzles attached to the pressurizer with nickel-based alloy welds. All pressurizer cladding is stainless steel.

The surge, spray, relief, and safety nozzle-to-piping safe end connections are buttered with nickel-alloy weld material prior to attachment of the stainless steel safe ends with nickel-alloy weld material. As described in LRA Section B. 1. 1, aging effects for these welds will be managed by the Alloy 600 Aging Management Program, not by the Pressurizer Examinations Program. Justification, including codes and standards referenced, that the technique and frequency used in the Alloy 600 Aging Management Program will be adequate for managing the effects of aging effects on nickel-alloy welds is provided in I&M's response to RAI B. 1. 1.2-2 in the referenced letter.

Reference for RAI B.1.24-1 Letter from J. N. Jensen, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant, Units I and 2, License Renewal Application - Response to Requests for Additional Information on Aging Management Programs," AEP:NRC:4034-10, dated August 11, 2004.

RAI B.1.24-2:

The spray head and its associated components covered by LRA Section B. 1.24 may be subject to severe thermal cycling. Inadequate justification was provided to demonstrate that a VT-3 examination is sufficient to detect a potential flaw in the spray head which could lead to failure of the component. Provide justification for using VT-3 examination instead of VT-I examination for the one-time inspection of these components in either Unit I or Unit 2. In addition, provide information regarding acceptance criteria; the evaluation methodology for disposition of indications; and the needfor successive examinations for the one-time inspection of spray head,

Attachment I to AEP:NRC:4034-13 Page I11 spray head locking bar, and coupling. Also, please provide your commitment in the commitment list and in the UFSAR Supplement.

I&M Response to RAI B.1.24-2:

The pressurizer spray head and associated components are not pressure-retaining items. The primary aging effect of concern is cracking.

Reduction of fracture toughness of the cast austenitic stainless steel (CASS) spray head may contribute to accelerated crack growth. The one-time visual inspection (VT-3) of the spray head will detect cracking. If cracks are detected in the spray head, engineering analysis will determine corrective actions, which could include follow-up examinations or replacement of the spray head.

The acceptance standards for the visual examinations will be in accordance with American Society of Mechanical Engineers (ASME)Section XI VT-3 examinations. This approach is consistent with the Oconee Nuclear Station (ONS) Pressurizer Examinations Program for CASS spray heads, as accepted by the Staff in NUREG-1723, Safety Evaluation Report Related to the License Renewal of Oconee Nuclear Station, Units 1, 2, and 3, in Section 3.4.3.3 on page 3-115. As summarized in the ONS Safety Evaluation Report (SER), the Staff expects cracking of the spray head to be a slow acting aging effect and expects minimal cracking, if any, to be found. The use of a one-time visual inspection (VT-3) to detect cracking was found to be adequate for the ONS spray heads, which are similar in design and function to the CNP pressurizer spray heads.

The acceptance criteria and corrective actions to disposition identified flaws are currently stated in LRA Section B.1.24, and the related commitments are listed in Attachment 1 to the referenced LRA submittal letter. The LRA Updated Final Safety Analysis Report Supplement also includes the commitment to complete a one-time inspection of the spray head and associated components.

LRA Section A.2.1.27 states: "This program will also determine the condition of the internal spray head, spray head locking bar, and coupling by a one-time visual examination of these components in one CNP unit. This program requires enhancements that will be implemented prior to the period of extended operation." This description is consistent with the level of detail in other LRA Appendix A program descriptions. Because LRA Attachments A and B provide the requested information and commitments, no additional changes are required.

Reference for RAI B. 1.24-2 Letter from M. K. Nazar, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant, Units 1 and 2, Application for Renewed Operating Licenses," AEP:NRC:3034, dated October 31, 2003 [Accession No. ML033070177].

RAI B.1.24-3:

LRA Section B. 1.24 states that the volumetric inspections have been performed with inservice inspection techniques that have been proven effective within the industry at detecting cracking.

Provide plant-specific and industry operating experience regarding detection, sizing, and to AEP:NRC:4034-13 Page 12 disposition of cracking in the pressurizer cladding using volumetric examinations and cracking and loss of parts in spray head components using visual examinations consistent with the inspection discussed in LRA BJ.24.

I&M Response to RAI B.1.24-3:

The Pressurizer Examinations Program described in LRA Section B. 1.24 includes volumetric examinations of pressurizer items having the highest fatigue usage factors. The stainless steel clad item with the highest fatigue cumulative usage factor is the circumferential weld at the head to shell junction. In accordance with ASME Section XI, Examination Category B-B, volumetric examination of essentially 100 percent of the circumferential shell-to-head weld is performed each inspection interval. In addition, the weld metal between the surge nozzle and the vessel lower head is subjected to high stress cycles. Periodic monitoring of this area in accordance with ASME Section XI, Examination Category B-D, provides monitoring for cracking of the cladding that may extend into the underlying base metal. The volumetric inspections included as part of the Pressurizer Examination Program are performed in accordance with ASME Section XI using techniques that have been proven effective throughout the industry in detecting cracking. The use of ASME Section XI volumetric inspections to manage cracking of cladding that may extend into the base metal has been accepted by the NRC for the Arkansas Nuclear One, Unit 1, Pressurizer Examinations Program as documented in NUREG-1743, Safety Evaluation Report Related to the License Renewal of Arkansas Nuclear One, Unit 1, May 2001 in Section 3.3.2.2.2.2 on pages 3-61 through 3-63.

Also, as documented in WCAP-14574-A, License Renewal Evaluation: Aging Management Evaluation for Pressurizers, and its associated SER, non-destructive examination has been used to support the acceptable disposition of pressurizer cladding cracking at Haddam Neck Plant.

The plant-specific operating experience for the pressurizer volumetric inspections performed through the Inservice Inspection Program is detailed in LRA Section B.1.14.

The proposed VT-3 examination of the pressurizer spray head and associated components is a new inspection that will be in accordance with ASME Section XI, Paragraph IWA-2213. At present, ASME Section XI does not require visual inspection of the pressurizer spray head and associated components and CNP has no operating experience regarding visual examination of these components.

RAI B.1.26-1:

The staff reviewed documents supporting LRA Section B.1.26, "Reactor Vessel Integrity, " and foundfrom the most recent capsule withdrawal schedule for Unit I documented in WCAP-12483, Revision 1, "Analysis of Capsule U from the American Electric Power Company D. C. Cook Unit 1 Reactor Vessel Radiation Surveillance Program, " that Capsule Wwas formerly located at the 4° position and known as Capsule S and Capsule S was formerly known as Capsule W Please confirm that the LRA has reported the most recent information regarding capsule

Attachment I to AEP:NRC:4034-13 Page 13 identification. In addition, please provide the projectedfluence in n/cm2 and in EFPY relative to the fluence at the peak reactor pressure vessel (RPJ9 fluence location for Capsule Wfor Unit I and Capsule Sfor Unit 2 at the proposed time of their next withdrawal.

I&M Response to RAI B.1.26-1:

As specified in WCAP-12483, Revision 1, Table 7-1, footnote (d), current Capsule W was formerly located at the 40 position and known as Capsule S. Current Capsule S was formerly known as Capsule W.

Thus, for administrative purposes the original Capsule W was re-designated as Capsule S and the original Capsule S was re-designated as Capsule W. This represents the most recent information regarding capsule identification, as reported in the LRA.

This re-designation of capsules was done to be in agreement with Technical Specification Table 4.4-5 so that re-designated Capsule S will be removed at 32 EFPY.

In LRA Section B.1.26, the proposed enhancement for Program Element 5, Monitoring and Trending, identifies the Unit I capsule with the outdated "Capsule W' designation. This table entry should be changed to read as follows:

"I&M will pull and test one additional standby capsule for each unit between 32 EFPY and 48 EFPY to address the peak fluence expected at 60 years. A fluence update will be performed at approximately 32 EFPY when Capsule S in each unit is pulled and tested.

A subsequent fluence update will be performed when the standby capsules are pulled and tested between 32 EFPY and 48 EFPY."

The projected fluence and removal time for Unit 1 Capsule S (i.e., the original Capsule W that was relocated and re-designated as Capsule S) are estimated as 2.018 x 1019 n/cm2 and 32 effective full power years (EFPY), respectively. The projected fluence and removal time for Unit 2 Capsule S are estimated as 1.983 x 10'9 n/cm2 and 32 EFPY, respectively.

RAI B.1.27-1:

Because of the limited information provided in LRA Section B. 1.27, "Reactor Vessel Internals Plates, Forgings, Welds, and Bolting, " the staff could not verify that this program is consistent with GALL for most of the 10 elements.

For example, the LRA does not mention the identification of the most susceptible items, an Attribute I concern; the speciflc water chemistry guidelines used, an Attribute 2 concern; and whether enhanced visual VT-I examinations or ultrasonic testing will be employed in inspections for certain selected components and locations, an Attribute 4 concern. Provide information regarding whether all 10 elements of the program are in accordance with GALL Program XI.M16, "PWR Vessel Internals, " and whether your program contains any exceptions or enhancements.

Attachment I to AEP:NRC:4034-13 Page 14 I&M Response to RAI B.1.27-1:

As stated in LRA Section B. 1.27, the Reactor Vessel Internals Plates, Forging, Weld, and Bolting Program will be consistent with the program described in NUREG-1801,Section XI.M16, "PWR Vessel Internals."

In accordance with the standard license renewal application format, the information provided in LRA Section B. 1.27 is consistent with the level of detail provided for all programs that are consistent with NUREG-1801.

There are no exceptions to the NUREG-1801 program. As identified in LRA Section B.1.27, and discussed below in the Parameters Monitored or Inspected section, one enhancement to the NUREG-1801,Section XI.M 16, program is applicable.

Aging Management Program Elements of the Reactor Vessel Internals Plates, Forging, Weld, and Bolting Program are provided below.

Scope The Reactor Vessel Internals, Plates, Forgings, Welds, and Bolting Program will apply to internal reactor vessel stainless steel and nickel-based alloy components, as listed in LRA Table 3.1.2-2.

The scope of the program will be consistent with NUREG-1 801.

Preventive Actions As the Reactor Vessel Internals, Plates, Forgings, Welds, and Bolting Program will be a condition monitoring program, no actions to prevent or mitigate aging effects are applicable.

However, the Primary and Secondary Water Chemistry Control Program is an effective preventive program to deter SCC and localized corrosion of the stainless steel and nickel-based alloy reactor vessel internal components. The Primary and Secondary Water Chemistry Control Program includes periodic monitoring and control of contaminants in accordance with the guidelines in the Electric Power Research Institute (EPRI) document TR-105714 for primary water chemistry.

The preventive actions included in the program will be consistent with NUREG-1 801.

Parameters Monitored or Inspected The Reactor Vessel Internals, Plates, Forgings, Welds, and Bolting Program will monitor the following parameters through inspections for items comprised of plates, forgings, welds, and miscellaneous bolting:

to AEP:NRC:4034-13 Page 15

  • Detection and measurement of dimensional changes due to void swelling, and
  • Detection of the loss of bolted closure integrity due to stress relaxation.

The program will include activities for the management of distortion due to void swelling which is not included in NUREG-1801,Section XI.M16. (This is included as an enhancement in LRA Section B. 1.27.)

The parameters monitored and inspected by the program will be consistent with NUREG-1801.

Detection of Aging Effects The Reactor Vessel Internals, Plates, Forgings, Welds, and Bolting Program will detect cracking, reduction of fracture toughness, dimensional changes, and loss of preload prior to loss of the reactor vessel internals intended function(s) utilizing the following activities:

  • A visual inspection will be performed on plates, forgings, and welds to detect cracking caused by IASCC enhanced by reduction of fracture toughness by irradiation embrittlement and distortion due to void swelling.

Other demonstrated acceptable inspection methods will be utilized for bolted joints (core barrel bolts and thermal shield bolts), if deemed necessary.

  • For baffle bolts, a volumetric inspection of critical locations will be performed to assess cracking.

This program will supplement the normal inservice inspections conducted in accordance with the interval and acceptance requirements of ASME Section XI, Examination Category B-N-3.

The detection of aging effects included in the program will be consistent with NUREG-1801.

Monitoring and Trending Engineering evaluations will be performed for inspection results that do not meet established acceptance standards. The engineering evaluations will consider the extent of degradation to reasonably assure that timely corrective or mitigative actions are taken. The Unit 1 and Unit 2 reactor vessel internals are of similar designs and utilize similar materials of construction.

Further, the operating conditions (power level, fluence) of the two units are similar.

Reactor Power Baffle / Former Plate Baffle Bolt 4a EFPY Fluence Unit (MWt)

Material Material at Vessel Inside 1

3304 Type 304 stainless Type 347 stainless 2.83E19 steel steel 2

3468 Type 304 stainless Type 347 stainless 2.46E 19 steel I steel to AEP:NRC:4034-13 Page 16 Therefore, inspections on one unit will be representative of the other and inspections of both units will not be necessary in each inspection interval. Unit 1 will be inspected in the fifth inspection interval while Unit 2 will be inspected in the sixth interval, prior to the last year of the first license renewal period. Should industry data or evaluations of the Unit 1 inspection data indicate that the inspection interval for Unit 2 should be modified (or re-inspection is required for Unit 1), I&M will provide plant-specific justification for the modification.

The monitoring and trending included in the program will be consistent with NUREG-1 801.

Acceptance Criteria For the plates, forgings, welds, and bolting other than baffle bolts that will be visually inspected, critical crack size will be determined by analysis prior to inspection. Acceptance criteria for dimensional changes due to void swelling will be developed prior to the inspection.

For baffle bolts, any detectable crack indication is unacceptable for a particular baffle bolt. The critical number of baffle bolts needed to be intact and their locations will be determined by analysis as part of this program.

The acceptance criteria included in the program will be consistent with NUREG-1 80 1.

Corrective Actions Specific corrective actions will be implemented in accordance with the Corrective Action Program.

Required repairs and replacements will be completed in accordance with ASME Section XI.

The corrective actions included in the program will be consistent with NUREG-1801.

Confirmation Process This attribute is discussed in LRA Section B.O.3.

Administrative Controls This attribute is discussed in LRA Section B.0.3.

Operating Experience Compliance with the inspection requirements of ASME Section XI has been maintained at CNP since initial operation. In general, visual examinations have proven effective to detect cracking.

CNP also participates in the Westinghouse Owners Group (WOG) program for baffle/former bolting.

Most of the current industry activities addressing aging effects on reactor vessel

Attachment I to AEP:NRC:4034-13 Page 17 internals are conducted under the EPRI Materials Reliability Project (MRP). The MRP strategy is to evaluate potential aging mechanisms and their effects on specific reactor vessel internals parts by evaluating causal parameters such as fluence, material properties, and state of stress.

Critical locations can be identified and tailored inspections can be conducted on an integrated industry, nuclear steam supply system, or plant-specific basis. As these projects are completed, I&M will evaluate the results and factor them into the Reactor Vessel Internals, Plates, Forgings, Welds, and Bolting Program, as applicable.

In accordance with the commitment made in the referenced LRA submittal letter, "I&M will participate in industry-wide programs designed by the PWR Materials Reliability Project Issues Task Group for investigating the impacts of aging on PWR vessel internal components."

The referenced LRA submittal letter also includes I&M's commitment to establish the program based on the above program elements.

Reference for RAI B.1.27-1 Letter from M. K. Nazar, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant, Units 1 and 2, Application for Renewed Operating Licenses," AEP:NRC:3034, dated October 31, 2003 [Accession No. ML033070177].

RAI B.1.27-2:

The information provided in LRA Section B. 1.27 is insufficientfor the staff to determine whether the PWR Materials Reliability Project (MRP) Issues Group and Westinghouse Owners Group (WOG) programs discussed there address all key issues of this aging management program (AMP), i.e., crack initiation and growth due to stress corrosion cracking (SCC) or irradiation-assisted SCC, loss offtacture toughness due to neutron irradiation embrittlement, and distortion due to void swelling. Provide a description of all the tasks under the MRP program and their goals and an assessment of the relevance of these tasks to the three aging effects mentioned above. Provide the same for the WOG program for baffle and former bolting. Further, your participation in the MRP program should be included as a commitment in your LRA commitment list and in the UFSAR Supplement to be submitted to the NRC.

Also, please provide a commitment that the program to manage void swelling will be submitted for staff review and approval three years prior to the period of extended operation.

I&M Response to RAI B.1.27-2:

The status of EPRI MRP Reactor Internals (RI) Issues Task Group (ITG) (RI-ITG) initiatives was presented to the NRC on October 23, 2003. As described in the NRC Meeting Summary (Reference 1), the purpose of the meeting in October 2003 was to discuss work being managed by the RI-ITG, which also supports license renewal aging management programs referenced by various owners groups and utilities. The goal of the RI-ITG is to establish aging management to AEP:NRC:4034-13 Page 18 programs that will assure reactor vessel internals integrity through plant life, including 60+ years of operation. The group serves as an industry focal point for resolution of issues related to pressurized water reactor (PWR) internals materials degradation, performs research to identify aging mechanisms and their effects on reactor internals, and provides a focal point to support communication with the agency. The RI-ITG, which is coordinated by the EPRI MRP, includes the WOG Subcommittee, the Babcock and Wilcox Owners Group Subcommittee, the International IASCC Program, the Electricite de France RI Materials Reliability Program, the Boiling Water Reactors Vessel and Internals Project, and the EPRI Corrosion Research Program.

A summary of activities to address the specific aging effects and associated aging mechanisms listed in this RAI and to address management of baffle and former bolting aging effects is provided in Slide 13 of the October 23, 2003, meeting handout (Reference 2). As presented in Slide 15 of the handout, the associated timeline for completion of program tasks includes obtaining material aging data through 2005, developing aging management guidelines and analysis between 2005 and 2008, and developing inspection guidelines after 2008. Development of these inspection guidelines is intended to support extended operation for those plants pursuing license renewal.

Subsequent slides in the handout provide completed products status and research program results.

The Reactor Vessel Internals Plates, Forgings, Welds, and Bolting Program commitment will be revised to indicate that the program to manage void swelling will be submitted for staff review and approval three years prior to the period of extended operation.

Participation in the EPRI MRP and management of void swelling are currently addressed in LRA Section B. 1.27, and related commitments listed in the LRA submittal letter (Reference 3).

For clarification, the last paragraph of LRA Section A.2.1.30 is revised as follows:

"This program will provide visual inspections and non-destructive examinations of the reactor vessel internals. I&M will participate in industry-wide programs designed by the PWR Materials Reliability Project Reactor Internals Issues Task Group for investigating the impacts of aging on PWR vessel internal subcomponents. The Reactor Vessel Internals Plates, Forgings, Welds, and Bolting Program will be implemented prior to the period of extended operation."

[NOTE: The text added for clarification in response to RAI B. 1.27-2 is in italics.]

Attachment I to AEP:NRC:4034-13 Page 19 References for RAI B. 1.27-2

1. NRC Memorandum from P. C. Wen, Project Manager, Policy and Rulemaking Program (PRP), Division of Regulatory Improvement Programs (DRIP), NRR, to C. Haney, Program Director, PRP/DRIP/NRR, "Summary of October 23, 2003, Meeting with Nuclear Energy Institute (NEI) on Reactor Vessel Internals," dated November 6, 2003 [Accession No. ML033110130].
2. Meeting Handouts for the 10/23/2003 Meeting on NRM RI-ITG Program Results and Status, dated October 23, 2003 [Accession No. ML033080166].
3. Letter from M. K. Nazar, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant, Units I and 2, Application for Renewed Operating Licenses," AEP:NRC:3034, dated October 31, 2003 [Accession No. ML033070177].

RAI B.1.31-1:

This aging program is called "Steam Generator Integrity," but the program description addresses only tubes. It is therefore consistent with NUREG-1801, Xl.M19, "Steam Generator Tube Integrity, " which also addresses only tubes. However, the applicant credits this program, all or in part, for the following forms of aging other than tube degradation:

  • material loss of carbon steel tube wrappers in treated water;
  • cracking of carbon steel tube wrappers in treated water;
  • material loss of stainless steel tube support plates and anti-vibration bars in treated water;
  • cracking of stainless steel tube support plates and anti-vibration bars in treated water;
  • material loss of carbon steel tube support plate stayrods and spacers in treated water;
  • cracking of carbon steel tube support plate stayrod nuts in treated water;
  • loss of mechanical closure integrity of tube support plate stayrod nuts in treated water;
  • material loss of nickel alloy tubes support plate stayrod washers and A VB retaining rings in treated water;
  • cracking of nickel alloy tubes support plate stayrod washers and A VB retaining rings in treated water; and
  • material loss of carbon steel lattice grid ring arch bars in treated water
  • cracking of carbon steel lattice grid ring studs in treated water
  • loss of mechanical closure integrity of carbon steel lattice grid ring studs in treated water
  • material loss of stainless steel lattice grid bars, U-bendflat bars, and J-tabs in treated water
  • cracking of stainless steel lattice grid bars, U-bendflat bars, and J-tabs in treated water to AEP:NRC:4034-13 Page 20 The staff requests that the applicant discuss, according to the ten Aging Management Program Elements (NUREG-1800 Section A.1.2.3), how the Steam Generator Integrity Program manages aging of components other than tubes.

I&M Response to RAI B.1.31-1:

NUREG-1801,Section XI.M19, "Steam Generator Tube Integrity," is not specific only to steam generator tubes.

As described in NUREG-1800 XI.M19, Program Element 3, Parameters Monitored/Inspected, "The inspection activities in this program detect flaws in tubing or degradation of secondary side internals needed to maintain tubing integrity... Degradation of steam generator internals is evaluated for corrective actions." The components discussed in this RAI form the steam generator secondary side tube support structure.

As noted in LRA Table 3.1.2-5, these components perform the intended function of providing structural and/or functional support for in-scope components (i.e., the steam generator tubes), and are therefore subject to aging management review. The CNP Steam Generator Integrity Program includes secondary side visual inspections of the tubesheet region, the tube support structures, the U-bend region, and the feedwater distribution system to verify the overall structural integrity of the steam generator secondary side internals. These areas are visually inspected for evidence of degraded conditions, including component deformation, material loss (erosion-corrosion, pitting, wear), cracking, foreign object damage, loss of component integrity, and deposit buildup.

If foreign objects are found, the Steam Generator Integrity Program also prescribes corrective actions such as metallurgical testing of the part; categorization of probable causes, origin, and mitigation; and determination of the need to expand inspections. If degraded conditions or foreign objects are found, the condition is documented using the Corrective Action Program and the inspection scope in the area of interest is expanded until the condition is bounded.

RAI B.L31-2:

The UFSAR Supplement item A.2.1.34, Steam Generator Integrity Program, discusses the integrity only of tubes.

However, the Steam Generator Integrity Program is credited with managing aging of other components. Please change the UFSAR Supplement to reflect the full scope of the Steam Generator Integrity program and reference the NEI 97-06 Steam Generator Program Guidelines.

I&M Response to RAI B.1.31-2:

For completeness, LRA Section A.2. 1.34 is revised as follows:

"The Steam Generator Integrity Program, which is based on guidance provided in NEI 97-06, Steam Generator Program Guidelines, uses nondestructive examination techniques to identify tubes that are defective and need to be removed from service or repaired in accordance with the Technical Specifications.

In addition, the Steam

Attachment I to AEP:NRC:4034-13 Page 21 Generator Integrity Program uses visual inspections to manage the effects of aging on secondary side internals needed to maintain steam generator tube integrity."

[NOTE: The text added for clarification in response to RAI B. 1.31-2 is in italics.]

to AEP:NRC:4034-13 Page I LIST OF REGULATORY COMMITMENTS The following table summarizes the action committed to by Indiana Michigan Power Company (I&M) in this document. Any other actions discussed in this submittal represent intended or planned actions by I&M. They are described to the Nuclear Regulatory Commission (NRC) for information and are not regulatory commitments.

Commitment Date I&M Response to RAI B. 1.27-2:

The Reactor Vessel Internals Plates, Forgings, Welds, and Bolting Unit 1:

Program commitment will be revised to indicate that the program to October 25, 2011 manage void swelling will be submitted for staff review and approval three years prior to the period of extended operation.

Unit 2:

December 23, 2014

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