August 19, 2004 AEP:NRC:4034-13 Washington, DC 20555-0001 · 2012. 11. 19. · (TAC Nos. MC 1202...

Indiana MichiganPower Company500 Circle DriveBuchanan, Ml 49107 1395

INDIANAMICHIGANPOWER

August 19, 2004 AEP:NRC:4034-1310 CFR 54

U. S. Nuclear Regulatory CommissionATTN: Document Control DeskMail Stop O-PI-17Washington, DC 20555-0001

SUBJECT: Donald C. Cook Nuclear Plant, Units 1 and 2Docket Nos. 50-315 and 50-316License Renewal Application - Response to Requests forAdditional 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. CookNuclear 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 itsreview of the license renewal application (LRA). This letter responds to Staffrequests for additional information (RAIs) pertaining to the aging managementprogram descriptions in the following LRA sections:

* B. 1.4 - Boric Acid Corrosion Prevention

* 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:

* 2.3.3.8 - Scoping and Screening Results - Emergency Diesel Generator

The RAIs addressed in this letter were received in two NRC letters datedJune 30, 2004 (References 2 and 3) and a third letter dated July 26, 2004(Reference 4).

I Dq

U S. Nuclear Regulatory CommissionPage 2

AEP:NRC:4034-13

The enclosure to this letter provides an affirmation pertaining to the statementsmade in this letter. Attachment 1 provides the additional information requestedfrom the NRC Staff. Attachment 2 provides a regulatory commitment made inthis letter in response to RAI B.1.27-2 for the new Reactor Vessel InternalsPlates, Forgings, Welds, and Bolting Program. It is noted that this commitmentsupplement the commitments to implement the new Reactor Vessel InternalsPlates, Forgings, Welds, and Bolting Program, as summarized on Page 6 ofAttachment 1 to the LRA cover letter (Reference 1).

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

Sincerely,

Site Vice President

NH/rdw

Enclosure:

Attachments:

Affirmation

1. Response to Requests for Additional Information for theDonald 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 OperatingLicenses," AEP:NRC:3034, dated October 31, 2003 [AccessionNo. ML033070177].

2. Letter from J. Rowley, NRC, to M. K. Nazar, I&M, "Request for AdditionalInformation for the Review of Donald C. Cook Nuclear Plant, Units 1 and 2License Renewal Application," dated June 30, 2004 [AccessionNo. ML041830088].

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

3. Letter from J. Rowley, NRC, to M. K. Nazar, I&M, "Request for AdditionalInformation for the Review of Donald C. Cook Nuclear Plant, Units I and 2License Renewal Application," dated June 30, 2004 [AccessionNo. ML041840218].

4. Letter from J. Rowley, NRC, to M. K. Nazar, I&M, "Request for AdditionalInformation for the Review of Donald C. Cook Nuclear Plant, Units I and 2License Renewal Application," dated July 26, 2004 [AccessionNo. ML042210285].

C: J. L. Caldwell, NRC Region IIIK. D. Curry, AEP Ft. Wayne, w/o attachmentJ. T. King, MPSC, w/o attachmentJ. G. Lamb, NRC Washington DCMDEQ - WHMD/HWRPS, w/o attachmentNRC Resident InspectorJ. 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. NEWMILLERNotary 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 theDonald C. Cook Nuclear Plant License Renewal Application

This attachment provides Indiana Michigan Power Company's (I&M's) responses to theDonald C. Cook Nuclear Plant (CNP) License Renewal Application (LRA) Requests forAdditional Information (RAIs) pertaining to the Aging Management Program descriptions in thefollowing LRA sections:

* B.1.4 Boric Acid Corrosion Prevention

* 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 attachment also provides the responses to two RAIs pertaining to LRA Section 2.3.3.8,Scoping and Screening Results - Emergency Diesel Generator. The RAls addressed in thisattachment were received in two NRC letters dated June 30, 2004 (References 1 and 2) and athird letter dated July26, 2004 (Reference 3).

References

1. Letter from J. Rowley, NRC, to M. K. Nazar, I&M, "Request for Additional Information forthe 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 forthe 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 forthe 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 toperform its intended function and are therefore in the scope of license renewal for meetingcriteria 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 aretherefore subject to an AMR.

The applicant is requested to confirm that the exhaust silencers and associated vent stacks are inscope and subject to an AMR and identify which "component type" on LRA Table 2.3.3-8represents 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 arenonsafety-related components whose only functions are to limit the noise created by the dieselengine and complete the transport of exhaust gas to the atmosphere. These components do notperform 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 cannotimpact safety-related components through spatial interaction as discussed in LRA Section2.1.1.2.2, and do not meet the scoping criteria of 10 CFR 54.4(a)(2). Therefore, the EDGexhaust 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 toperform its intendedfunction and are therefore in the scope of license renewal in accordancewith 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-livedpassive components and are therefore subject to an AMR.

Attachment 1 to AEP:NRC:4034-13 Page 3

The applicant is requested to confirm that the centrifugal exhausters and their associatedflexibleconnectors 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 arenonsafety-related components in the crankcase breather subsystem. This EDG subsystemmaintains a slight vacuum in the crankcase to remove vapors and minimize oil leakage. Thisfunction is not required for diesel engine operation, and a failure of these components would notrender the EDG inoperable. Therefore, these components do not serve a license renewalintended 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 onlyair and crankcase gases, their failure cannot adversely impact safety-related components throughspatial interaction as discussed in LRA Section 2.1.2.2.2. Consequently, these components donot meet the spatial scoping criteria of 10 CFR 54.4(a)(2). Therefore, the centrifugal exhaustersand 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," statesthat the scope of Boric Acid Corrosion Prevention Program will be revised to include electricalcomponents in addition toferritic steel. Identify all specific systems and components and theirsupports, inside and outside containment, that may be susceptible to boric acidcorrosion/degradation. Provide information regarding provisions in this program forinspecting, detecting, or monitoring degradation of structures and components due to boric acidleakage and provisions for inspecting, detecting, or monitoring boric acid leakage ininaccessible 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 andloss of mechanical closure integrity due to boric acid corrosion for component types as indicatedin the LRA Section 3 aging management review results tables. This program applies to portionsof systems and structures, both inside and outside containment, that are subject to agingmanagement review and contain borated water or are subject to exposure to leaking boratedwater. 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 andcomponents due to boric acid leakage, and provisions for inspecting, detecting, or monitoringboric acid leakage in inaccessible locations and areas covered by external insulation surfaces are


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 isfollowed to identify the source and all affected components along the path, including locationscovered 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 ofoperating reactors, issues that are relevant to current plant operation will be addressed bythe existing regulatory process within the present license term rather than deferred untilthe time of license renewal. Consequently, the Commission formulated two principlesof license renewal.

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

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

RAI B.1.4-2:

LRA Section B.1.4 states that Boric Acid Corrosion Prevention Program continues to beimproved based on operating experience, and program revisions have incorporated lessonslearned from condition reports and industry guidance. Provide information about theseimprovements as related to lessons learnedfrom the Davis-Besse vessel head degradation andthe 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 correctiveactions in the program to prevent the recurrence of degradation caused by boric acid leakage, asrequired by GL [Generic Letter] 88-05.



Boric acid corrosion control is an issue that has received, and continues to receive significantregulatory and industry attention. In accordance with the Statements of Consideration for theFinal Part 54 Rule, the existing regulatory process, which includes consideration of industryoperating experience, will ensure that the Boric Acid Corrosion Prevention Program caneffectively 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 inaccordance with the existing regulatory process to address the issue of boric acid corrosion. Ifadditional guidance or requirements are promulgated to address this issue prior to the end of thecurrent licensing period, any regulatory obligations or on-going commitments made in responseto those requirements will also apply to the period of extended operation.

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

The Boric Acid Corrosion Prevention Program includes corrective actions to prevent recurrenceof degradation caused by boric acid leakage, as required by GL 88-05. An example wasprovided in LRA Section B. 1.4, which discussed a recent condition report documenting anaccumulation of boric acid crystals on a heat exchanger flange. A brown stain was noted in theacid crystals, indicating corrosion of the carbon steel bolts. The bolts were replaced withstainless 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 thesecommitments was to develop a program for replacement of carbon steel packing follow studs onvalves within the reactor coolant pressure boundary. The program was developed, and the studswere replaced on the valves that were identified as requiring replacement. The secondcommitment was to ensure proper consideration is given to (a) reducing the probability of reactorcoolant leaks at locations where they may cause corrosion damage and (b) the use of suitablecorrosion resistant materials or the application of protective coatings/claddings. Thiscommitment was satisfied based on guidance that was incorporated into the boric acid corrosioncontrol procedure. The third corrective action commitment was to review the training programsand procedures to ensure they contain adequate guidance on issues relevant to reactor coolantpressure 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 theimplementing documents (i.e., training modules) do not properly reference the commitment. Anaction has been entered into the Corrective Action Program to ensure this commitment isreferenced properly in the appropriate training modules. Generic Letter 88-05 actions wereaudited in July 1989 and found by the NRC to be adequately implemented, as documented in anaudit 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 NuclearPlant, Units 1 and 2, License Renewal Application - Response to Requests for AdditionalInformation 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 NuclearPlant, Units 1 and 2, NRC Generic Letter 88-05: Boric Acid Corrosion of Carbon Steel ReactorPressure 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 atD. C. Cook Units 1 and 2 (Generic Letter 88-05) (TAC Nos. MC1202 and MC1203)," datedFebruary 22, 1990.

RAI B.1.5-1:

LRA Section B.1.5, "Bottom-Mounted Instrumentation Thimble Tube Inspection, " was designedfor 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 agingmanagement for BMI thimble tubes, explain how your proposed program is adequate to detectSCC or modify the thimble tube inspection program to include inspections for thimble tubecracking 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 LRATable 3.1.2-1 should be revised accordingly.



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 SecondaryWater Chemistry Control Program, which mitigates loss of material for stainless steel exposed totreated borated water, as discussed in the referenced response to RAI 3.1-3. Cracking due tostress corrosion cracking (SCC) is managed by the Primary and Secondary Water ChemistryControl Program and the Inservice Inspection Program (i.e., leak detection as part ofExamination Category B-P). The Bottom-Mounted Instrumentation Thimble Tube InspectionProgram eddy current testing (ECT) is not credited for detection and sizing of SCC. No changesto 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 NuclearPlant, Units I and 2, License Renewal Application - Response to Requests for AdditionalInformation 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 orisolation of a thimble tube with 80 percent through-wall wear, (2) reposition of a thimble tubewith more than 40 percent through-wall wear, provided that it is projected to remain under80 percent until the next inspection, and (3) replacement, isolation, or reposition of a thimbletube with more than 40 percent through-wall wear if it is projected to exceed 80 percent by thenext inspection. Using reposition as an option for Criterion 3 for a tube which is projected toexceed 80-percent wear by the next inspection is inadequate because the uncertainty of the tubewear rate at the selected location for the tube reposition in a certain time period might make thereposition ineffective. Provide a revision of the AMP by incorporating ET uncertainty in futurewear measurements and by considering only replacement and isolation of tubes as options forCriterion 3 of the acceptance criteria.


The BMI inspection is based on recommendations provided in WCAP-12866, Bottom MountedInstrument Flux Thimble Wear. The WCAP demonstrates that thimble tube percent wall lossvaries at different core locations over several operating cycles. The current inspection procedurepermits relocation of a BMI thimble tube from a location with wear predicted to equal or exceed80% through-wall by the next inspection to a location that would not result in 80% wear by the


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

Reference for RAI B. 1.5-2Letter from E. E. Fitzpatrick, I&M, to NRC Document Control Desk, "Donald C. Cook NuclearPlant, Units 1 and 2, Response to Confirmatory Action Letter No. RIII 97-011 NRC ArchitectEngineer (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 ETinspection data and considering the anticipated operating conditions during the period ofextended operation. It should be noted that the proposed thimble inspection every third outage isonly acceptable if no wear has been discovered in the past three refueling outages for all thimbletubes. When wear appears, the inspection interval must be reevaluated based on the observedthimble tube-specific wear rates. Please provide a revised inspection schedule, anticipatingwear and based on severity of wear. The UFSAR Supplement should be revised to include adescription of this inspection schedule. In addition, discuss any mitigative measures, such asflushing of the tubes, taken during refueling outages. If SCC is determined to be a potentialdegradation mechanism for thimble tubes in your response to RAI 3.0.3.3-1, provide justificationfor the inspectionfrequencyfor detecting the SCCflaws.


In accordance with NRC Bulletin 88-09, "Thimble Tube Thinning in Westinghouse Reactors,"all holders of operating licenses or construction permits for Westinghouse-designed nuclearpower reactors that utilize BMI were required to establish and implement an inspection programto periodically confirm incore neutron monitoring system thimble tube integrity. In response tothis bulletin, I&M established an incore thimble tube eddy current inspection program with ECTperformed at the next refueling outages of both units. Inspections continued to be performedeach refueling outage based on observed ECT wear results until changed as discussed below.


During the 1992 refueling outages, 15 thimble tubes were replaced in Unit 1 and 22 thimbletubes were replaced in Unit 2; replacement tubes were chrome-plated at axial locationscorresponding 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 noindications of wear and provided the NRC with an update of plans to replace the remainingdesign thimble tubes with chrome-plated tubes.

In 2000, I&M completed the replacement of thimble tubes with tubes that are chrome-platedfrom approximately twelve feet below the tip for a length of approximately fourteen feet. Thisencompasses the entire lower reactor internal region from immediately above the lower fuelnozzle into conduits below, which covers the most active wear location. During the 2002refueling outages, after one cycle of operation, ECT results showed no indication of wear on anyof the thimble tubes, at any axial location (from the bullet tip of the thimble tube to the sealtable).

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

Projections of thimble tube wear are based on the applicable inspection interval and are verifiedthrough eddy current tests. On-going program requirements and commitments to NRC Bulletin88-09 will be carried forward through the period of extended operation. This is consistent withthe second principle of license renewal, as discussed in the Statements of Consideration for theFinal Part 54 Rule, which states that the plant-specific licensing basis must be maintained duringthe 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 currentterm of operation or the period of extended operation, the ECT frequency will be revised inaccordance with the Corrective Action Program. For clarification of the inspection schedulebasis, the UFSAR Supplement for this program in LRA Section A.2.1.5 is revised to add thefollowing sentence to the Bottom-Mounted Instrumentation Thimble Tube Inspection Programdescription:

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

Mitigative measures include flushing, drying, and lubricating the thimble tubes each outage. Inaddition, foreign material exclusion and cleanliness controls are required during maintenanceactivities.

Attachment I to AEP:NRC:4034-13 Page IO

For consistency with other stainless steel components exposed to a treated (borated) waterenvironment, LRA Table 3.1.2-1 lists cracking (including cracking due to SCC) as an agingeffect for the component type "BMI thimble tubes and bullet plugs." This aging effect ismanaged by the Primary and Secondary Water Chemistry Control Program and the InserviceInspection Program. A review of operating experience indicated no instances of cracking bySCC of BMI thimble tubes at CNP.

RAI B.1.24-1:

LRA Section B. 1.24, "Pressurizer Examinations, " assesses the cladding and attachment welds tothe cladding of the pressurizer. Identify all nickel-alloy welds which were used to attach variouspenetrations to the pressurizer; confirm that these welds are managed by this AMP and justifythat 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.


There are no nozzles attached to the pressurizer with nickel-based alloy welds. All pressurizercladding is stainless steel. The surge, spray, relief, and safety nozzle-to-piping safe endconnections are buttered with nickel-alloy weld material prior to attachment of the stainless steelsafe ends with nickel-alloy weld material. As described in LRA Section B. 1. 1, aging effects forthese welds will be managed by the Alloy 600 Aging Management Program, not by thePressurizer Examinations Program. Justification, including codes and standards referenced, thatthe technique and frequency used in the Alloy 600 Aging Management Program will be adequatefor managing the effects of aging effects on nickel-alloy welds is provided in I&M's response toRAI B. 1. 1.2-2 in the referenced letter.


Letter from J. N. Jensen, I&M, to NRC Document Control Desk, "Donald C. Cook NuclearPlant, Units I and 2, License Renewal Application - Response to Requests for AdditionalInformation 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 tosevere thermal cycling. Inadequate justification was provided to demonstrate that a VT-3examination is sufficient to detect a potential flaw in the spray head which could lead to failureof the component. Provide justification for using VT-3 examination instead of VT-I examinationfor the one-time inspection of these components in either Unit I or Unit 2. In addition, provideinformation regarding acceptance criteria; the evaluation methodology for disposition ofindications; 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 commitmentlist and in the UFSAR Supplement.


The pressurizer spray head and associated components are not pressure-retaining items. Theprimary aging effect of concern is cracking. Reduction of fracture toughness of the castaustenitic stainless steel (CASS) spray head may contribute to accelerated crack growth. Theone-time visual inspection (VT-3) of the spray head will detect cracking. If cracks are detectedin the spray head, engineering analysis will determine corrective actions, which could includefollow-up examinations or replacement of the spray head. The acceptance standards for thevisual examinations will be in accordance with American Society of Mechanical Engineers(ASME) Section XI VT-3 examinations. This approach is consistent with the Oconee NuclearStation (ONS) Pressurizer Examinations Program for CASS spray heads, as accepted by the Staffin NUREG-1723, Safety Evaluation Report Related to the License Renewal of Oconee NuclearStation, Units 1, 2, and 3, in Section 3.4.3.3 on page 3-115. As summarized in the ONS SafetyEvaluation Report (SER), the Staff expects cracking of the spray head to be a slow acting agingeffect 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 similarin design and function to the CNP pressurizer spray heads.

The acceptance criteria and corrective actions to disposition identified flaws are currently statedin LRA Section B.1.24, and the related commitments are listed in Attachment 1 to the referencedLRA submittal letter. The LRA Updated Final Safety Analysis Report Supplement also includesthe 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 internalspray head, spray head locking bar, and coupling by a one-time visual examination of thesecomponents in one CNP unit. This program requires enhancements that will be implementedprior to the period of extended operation." This description is consistent with the level of detailin other LRA Appendix A program descriptions. Because LRA Attachments A and B providethe 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 NuclearPlant, Units 1 and 2, Application for Renewed Operating Licenses," AEP:NRC:3034, datedOctober 31, 2003 [Accession No. ML033070177].

RAI B.1.24-3:

LRA Section B. 1.24 states that the volumetric inspections have been performed with inserviceinspection techniques that have been proven effective within the industry at detecting cracking.Provide plant-specific and industry operating experience regarding detection, sizing, and


disposition of cracking in the pressurizer cladding using volumetric examinations and crackingand loss of parts in spray head components using visual examinations consistent with theinspection discussed in LRA BJ.24.


The Pressurizer Examinations Program described in LRA Section B. 1.24 includes volumetricexaminations of pressurizer items having the highest fatigue usage factors. The stainless steelclad item with the highest fatigue cumulative usage factor is the circumferential weld at the headto shell junction. In accordance with ASME Section XI, Examination Category B-B, volumetricexamination of essentially 100 percent of the circumferential shell-to-head weld is performedeach inspection interval. In addition, the weld metal between the surge nozzle and the vessellower head is subjected to high stress cycles. Periodic monitoring of this area in accordance withASME Section XI, Examination Category B-D, provides monitoring for cracking of the claddingthat may extend into the underlying base metal. The volumetric inspections included as part ofthe Pressurizer Examination Program are performed in accordance with ASME Section XI usingtechniques that have been proven effective throughout the industry in detecting cracking. Theuse of ASME Section XI volumetric inspections to manage cracking of cladding that may extendinto 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 ReportRelated to the License Renewal of Arkansas Nuclear One, Unit 1, May 2001 inSection 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 associatedSER, non-destructive examination has been used to support the acceptable disposition ofpressurizer cladding cracking at Haddam Neck Plant.

The plant-specific operating experience for the pressurizer volumetric inspections performedthrough the Inservice Inspection Program is detailed in LRA Section B.1.14. The proposedVT-3 examination of the pressurizer spray head and associated components is a new inspectionthat will be in accordance with ASME Section XI, Paragraph IWA-2213. At present, ASMESection XI does not require visual inspection of the pressurizer spray head and associatedcomponents and CNP has no operating experience regarding visual examination of thesecomponents.

RAI B.1.26-1:

The staff reviewed documents supporting LRA Section B.1.26, "Reactor Vessel Integrity, " andfoundfrom 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. CookUnit 1 Reactor Vessel Radiation Surveillance Program, " that Capsule Wwas formerly located atthe 4° position and known as Capsule S and Capsule S was formerly known as Capsule WPlease confirm that the LRA has reported the most recent information regarding capsule


identification. In addition, please provide the projectedfluence in n/cm2 and in EFPY relative tothe fluence at the peak reactor pressure vessel (RPJ9 fluence location for Capsule Wfor Unit Iand Capsule Sfor Unit 2 at the proposed time of their next withdrawal.


As specified in WCAP-12483, Revision 1, Table 7-1, footnote (d), current Capsule W wasformerly located at the 40 position and known as Capsule S. Current Capsule S was formerlyknown 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. Thisrepresents 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 Table4.4-5 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 andTrending, identifies the Unit I capsule with the outdated "Capsule W' designation. This tableentry should be changed to read as follows:

"I&M will pull and test one additional standby capsule for each unit between 32 EFPYand 48 EFPY to address the peak fluence expected at 60 years. A fluence update will beperformed 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 andtested between 32 EFPY and 48 EFPY."

The projected fluence and removal time for Unit 1 Capsule S (i.e., the original Capsule W thatwas relocated and re-designated as Capsule S) are estimated as 2.018 x 1019 n/cm2 and 32effective full power years (EFPY), respectively. The projected fluence and removal time forUnit 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 InternalsPlates, Forgings, Welds, and Bolting, " the staff could not verify that this program is consistentwith GALL for most of the 10 elements. For example, the LRA does not mention theidentification of the most susceptible items, an Attribute I concern; the speciflc water chemistryguidelines used, an Attribute 2 concern; and whether enhanced visual VT-I examinations orultrasonic 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 programare in accordance with GALL Program XI.M16, "PWR Vessel Internals, " and whether yourprogram contains any exceptions or enhancements.



As stated in LRA Section B. 1.27, the Reactor Vessel Internals Plates, Forging, Weld, andBolting Program will be consistent with the program described in NUREG-1801,Section XI.M16, "PWR Vessel Internals." In accordance with the standard license renewalapplication format, the information provided in LRA Section B. 1.27 is consistent with the levelof detail provided for all programs that are consistent with NUREG-1801. There are noexceptions to the NUREG-1801 program. As identified in LRA Section B.1.27, and discussedbelow 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 tointernal reactor vessel stainless steel and nickel-based alloy components, as listed in LRATable 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 acondition monitoring program, no actions to prevent or mitigate aging effects are applicable.However, the Primary and Secondary Water Chemistry Control Program is an effectivepreventive program to deter SCC and localized corrosion of the stainless steel and nickel-basedalloy reactor vessel internal components. The Primary and Secondary Water Chemistry ControlProgram includes periodic monitoring and control of contaminants in accordance with theguidelines in the Electric Power Research Institute (EPRI) document TR-105714 for primarywater 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 thefollowing parameters through inspections for items comprised of plates, forgings, welds, andmiscellaneous bolting:

* Detection and sizing of cracks which could be initiated by SCC or irradiation-assistedstress corrosion cracking (IASCC),


* 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 whichis not included in NUREG-1801, Section XI.M16. (This is included as an enhancement in LRASection 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 thereactor vessel internals intended function(s) utilizing the following activities:

* A visual inspection will be performed on plates, forgings, and welds to detect crackingcaused by IASCC enhanced by reduction of fracture toughness by irradiationembrittlement and distortion due to void swelling. Other demonstrated acceptableinspection methods will be utilized for bolted joints (core barrel bolts and thermal shieldbolts), if deemed necessary.

* For baffle bolts, a volumetric inspection of critical locations will be performed to assesscracking.

This program will supplement the normal inservice inspections conducted in accordance with theinterval 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 establishedacceptance standards. The engineering evaluations will consider the extent of degradation toreasonably assure that timely corrective or mitigative actions are taken. The Unit 1 and Unit 2reactor 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 FluenceUnit (MWt) Material Material at Vessel Inside

1 3304 Type 304 stainless Type 347 stainless 2.83E19steel steel

2 3468 Type 304 stainless Type 347 stainless 2.46E 19__ _ __ _ __ _ _ _ __ _ _ steel I steel _ _ _ _ _ _ _ _ _ _


Therefore, inspections on one unit will be representative of the other and inspections of bothunits will not be necessary in each inspection interval. Unit 1 will be inspected in the fifthinspection interval while Unit 2 will be inspected in the sixth interval, prior to the last year of thefirst license renewal period. Should industry data or evaluations of the Unit 1 inspection dataindicate that the inspection interval for Unit 2 should be modified (or re-inspection is requiredfor 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 fordimensional 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. Thecritical number of baffle bolts needed to be intact and their locations will be determined byanalysis 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 ActionProgram. Required repairs and replacements will be completed in accordance with ASMESection 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 CNPsince initial operation. In general, visual examinations have proven effective to detect cracking.CNP also participates in the Westinghouse Owners Group (WOG) program for baffle/formerbolting. Most of the current industry activities addressing aging effects on reactor vessel


internals are conducted under the EPRI Materials Reliability Project (MRP). The MRP strategyis to evaluate potential aging mechanisms and their effects on specific reactor vessel internalsparts 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 integratedindustry, 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 willparticipate in industry-wide programs designed by the PWR Materials Reliability Project IssuesTask 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 programbased on the above program elements.


Letter from M. K. Nazar, I&M, to NRC Document Control Desk, "Donald C. Cook NuclearPlant, Units 1 and 2, Application for Renewed Operating Licenses," AEP:NRC:3034, datedOctober 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 whetherthe 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 distortiondue to void swelling. Provide a description of all the tasks under the MRP program and theirgoals and an assessment of the relevance of these tasks to the three aging effects mentionedabove. Provide the same for the WOG program for baffle and former bolting. Further, yourparticipation in the MRP program should be included as a commitment in your LRA commitmentlist and in the UFSAR Supplement to be submitted to the NRC. Also, please provide acommitment that the program to manage void swelling will be submitted for staff review andapproval three years prior to the period of extended operation.


The status of EPRI MRP Reactor Internals (RI) Issues Task Group (ITG) (RI-ITG) initiativeswas 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 managedby the RI-ITG , which also supports license renewal aging management programs referenced byvarious owners groups and utilities. The goal of the RI-ITG is to establish aging management


programs that will assure reactor vessel internals integrity through plant life, including 60+ yearsof operation. The group serves as an industry focal point for resolution of issues related topressurized water reactor (PWR) internals materials degradation, performs research to identifyaging mechanisms and their effects on reactor internals, and provides a focal point to supportcommunication with the agency. The RI-ITG, which is coordinated by the EPRI MRP, includesthe WOG Subcommittee, the Babcock and Wilcox Owners Group Subcommittee, theInternational IASCC Program, the Electricite de France RI Materials Reliability Program, theBoiling 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 mechanismslisted in this RAI and to address management of baffle and former bolting aging effects isprovided in Slide 13 of the October 23, 2003, meeting handout (Reference 2). As presented inSlide 15 of the handout, the associated timeline for completion of program tasks includesobtaining material aging data through 2005, developing aging management guidelines andanalysis between 2005 and 2008, and developing inspection guidelines after 2008. Developmentof these inspection guidelines is intended to support extended operation for those plants pursuinglicense renewal. Subsequent slides in the handout provide completed products status andresearch program results.

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

Participation in the EPRI MRP and management of void swelling are currently addressed inLRA 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 examinationsof the reactor vessel internals. I&M will participate in industry-wide programsdesigned by the PWR Materials Reliability Project Reactor Internals Issues TaskGroup for investigating the impacts of aging on PWR vessel internalsubcomponents. The Reactor Vessel Internals Plates, Forgings, Welds, andBolting 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.]


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, ProgramDirector, PRP/DRIP/NRR, "Summary of October 23, 2003, Meeting with Nuclear EnergyInstitute (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 NuclearPlant, Units I and 2, Application for Renewed Operating Licenses," AEP:NRC:3034, datedOctober 31, 2003 [Accession No. ML033070177].

RAI B.1.31-1:

This aging program is called "Steam Generator Integrity," but the program descriptionaddresses only tubes. It is therefore consistent with NUREG-1801, Xl.M19, "Steam GeneratorTube 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


The staff requests that the applicant discuss, according to the ten Aging Management ProgramElements (NUREG-1800 Section A.1.2.3), how the Steam Generator Integrity Program managesaging of components other than tubes.


NUREG-1801, Section XI.M19, "Steam Generator Tube Integrity," is not specific only to steamgenerator tubes. As described in NUREG-1800 XI.M19, Program Element 3, ParametersMonitored/Inspected, "The inspection activities in this program detect flaws in tubing ordegradation of secondary side internals needed to maintain tubing integrity ... Degradation ofsteam generator internals is evaluated for corrective actions." The components discussed in thisRAI form the steam generator secondary side tube support structure. As noted inLRA Table 3.1.2-5, these components perform the intended function of providing structuraland/or functional support for in-scope components (i.e., the steam generator tubes), and aretherefore subject to aging management review. The CNP Steam Generator Integrity Programincludes 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 integrityof the steam generator secondary side internals. These areas are visually inspected for evidenceof 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 correctiveactions such as metallurgical testing of the part; categorization of probable causes, origin, andmitigation; and determination of the need to expand inspections. If degraded conditions orforeign objects are found, the condition is documented using the Corrective Action Program andthe 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 theintegrity only of tubes. However, the Steam Generator Integrity Program is credited withmanaging aging of other components. Please change the UFSAR Supplement to reflect the fullscope of the Steam Generator Integrity program and reference the NEI 97-06 Steam GeneratorProgram Guidelines.


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

"The Steam Generator Integrity Program, which is based on guidance provided inNEI 97-06, Steam Generator Program Guidelines, uses nondestructive examinationtechniques to identify tubes that are defective and need to be removed from service orrepaired in accordance with the Technical Specifications. In addition, the Steam


Generator Integrity Program uses visual inspections to manage the effects of aging onsecondary 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.]

Attachment 2 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 orplanned actions by I&M. They are described to the Nuclear Regulatory Commission (NRC) forinformation 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, 2011manage void swelling will be submitted for staff review and approvalthree years prior to the period of extended operation. Unit 2:

December 23, 2014

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