michigan power september 2, 2004 aep:nrc:4034-15 mail … · j. n. jensen site vice president...

Indiana MichiganPower Company500 Circle DriveBuchanan, MI 49107 1395

INDIANAMICHIGANPOWER

September 2, 2004 AEP:NRC:4034-1510 CFR 54

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

SUBJECT: Donald C. Cook Nuclear Plant, Units I and 2Docket Nos. 50-315 and 50-316License Renewal Application - Response to Requests forAdditional Information(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 (CNP), Units 1 and 2 (Reference 1).

The Nuclear Regulatory Commission (NRC) review process includes audits ofthe aging management programs (AMPs) credited in the CNP license renewalapplication (LRA). During the conduct of these audits and subsequent toconversations between I&M and NRC Staff, the NRC Staff identified areaswhere additional information was needed to complete its review of the creditedAMPs. This letter provides the information requested by the NRC Staffsubsequent to the completion of the AMP audits pertaining to the following LRAsections:

* B.1.34 - Structure Monitoring - Divider Barrier Seal Inspection

* 3.1 - Reactor Vessel, Internals and Reactor Coolant System

* 3.3 -Auxiliary Systems

* 3.5 - Structures and Component Supports

The requests for additional information (RAls) addressed in this letter werereceived in an NRC letter dated August 20, 2004 (Reference 2).

U S. Nuclear Regulatory Commission AEP:NRC:4034-15

In addition, this letter provides supplemental responses to RAls requested by theNRC Staff and response to information requested in a draft RAI pertaining to thefollowing LRA sections:

* 2.3.3 - Auxiliary Systems

* 3.6 - Electrical and Instrumentation Controls

* 4.4 - Environmental Qualification of Electrical Components

* 4.7 - Other Plant-Specific Time-Limited Aging Analyses

The enclosure to this letter provides an affirmation pertaining to the statementsmade in this letter. Attachment I to this letter provides I&M's responses to theNRC Staff's RAls. Attachment 2 to this letter provides I&M's supplementalresponses to RAls and a response to the NRC Staff's draft RAI. There are nonew commitments contained in this submittal.

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

Sincerely,

J. N. JensenSite Vice President

NH/rdw

Enclosure: Affirmation

Attachments: 1. Response to Requests for Additional Information for theDonald C. Cook Nuclear Plant License Renewal Application

2. Supplemental Responses to Requests for AdditionalInformation and Response to Draft Request for AdditionalInformation for the Donald C. Cook Nuclear Plant LicenseRenewal Application

U S. Nuclear Regulatory Commission AEP:NRC:4034-15

References:

I. Letter from M. K. Nazar, l&M, to NRC Document Control Desk, "Donald C.Cook Nuclear Plant Units I 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 I and 2License Renewal Application," dated August 20, 2004 [AccessionNo. ML042330355].

c: J. L. Caldwell, NRC Region IIIK. D. Curry, AEP Ft. Wayne, w/o attachmentsJ. T. King, MPSC, w/o attachmentsJ. G. Lamb, NRC Washington DCMDEQ - WHMD/HWRPS, w/o attachmentsNRC Resident InspectorJ. G. Rowley, NRC Washington DC

Enclosure to AEP:NRC:4034-15

AFFIRMATION

I, Joseph N. Jensen, being duly sworn, state that I am Site Vice President of Indiana MichiganPower Company (I&M), that I am authorized to sign and file this request with the NuclearRegulatory Commission on behalf of I&M, and that the statements made and the matters setforth herein pertaining to I&M are true and correct to the best of my knowledge, information,and belief.

Indiana Michigan Power Company

ph N. JensenSite Vice President

SWORN TO AND SUBSCRIBED BEFORE ME

THIS PLAY OF IC 2004

__ __ ,ty Publ

My Commission Expires __a__l_________

Attachment I to AEP:NRC:4034-15 Page I

Response to Requests for Additional Information for theDonald C. Cook Nuclear Plant License Renewal Application

This attachment provides the information requested by the Nuclear Regulatory Commission(NRC) Staff to complete the aging management program audit report in requests for additionalinformation (RAIs) pertaining to the following license renewal application (LRA) sections:

* B.1.34 - Structure Monitoring - Divider Barrier Seal Inspection

* 3.1 - Reactor Vessel, Internals and Reactor Coolant System

* 3.3 - Auxiliary Systems

* 3.5 - Structures and Component Supports.

The RAls addressed in this attachment were received in the referenced NRC letter datedAugust 20, 2004.

Reference

Letter from J. Rowley, NRC, to M. K. Nazar, Indiana Michigan Power Company (I&M),"Request for Additional Information for the Review of Donald C. Cook Nuclear Plant, Units Iand 2 License Renewal Application," dated August 20, 2004 [Accession No. ML042330355].

RAI B.l.34-1:

The project team requested clarification on the method(s) used to monitor a change in materialproperties of elastomers, specifically, the pressure seals (divider barrier). The SRP-LRAppendix A. 1.2.3.3 states that "parameters to be monitored or inspected should be identified andlinked to the degradation of the particular structure and component intendedfiunction(s) " and"should detect the presence and extent of aging effects."

By letter dated April 23, 2004 (ML041270484), the applicant responded that the phrase "changein material properties" was intended to convey a visual inspection to ensure the absence ofapparent deterioration (i.e., cracks or defects in the sealing surfaces) as discussed in theimplementing procedures.

Please provide the basis for concluding that the elastomeric divider barrier will continue toperform its designffunction despite changes in material properties that may not be visible.

I&M Response to RAI B.1.34-1:

As noted in I&M's response to RAI 2.4-2, which was provided in the referenced May 7, 2004,RAI response letter, the seals that provide a boundary between the lower and upper containment

Attachment I to AEP NRC:4034-15 Page 2

compartments are of three types.

* Main divider barrier seals between the bottom of the ice condenser compartment slab andthe containment wall and up the sides of the ice condenser end walls.

* Divider barrier hatch seals provided on the hatches in the operating deck. This alsoincludes personnel access doors between the containment's upper and lowercompartments.

* Divider barrier penetration seals installed around penetrations and openings through thedivider barrier.

Subsequent to issuing this RAI, the NRC Staff provided clarification that the elastomericpressure seals that are the subject of this RAI are the penetration seals installed aroundpenetrations and openings through the divider barrier. The basis for excluding the main dividerbarrier seals and the divider barrier hatch seals from this RAI is that these seals are addressed inthe Donald C. Cook Nuclear Plant (CNP) Technical Specifications. The main divider barrierseals are inspected and replaced based on their condition in accordance with CNP TechnicalSpecification Surveillance Requirement 4.6.5.9, and as such are short-lived and not subject toaging management review. The divider barrier personnel access door and equipment hatch sealsare visually inspected before containment closure each outage, in accordance with CNPTechnical Specification Surveillance Requirement 4.6.5.5, and replaced as needed. Therefore,the divider barrier personnel access doors and equipment hatch seals are also short-lived and arenot subject to aging management review.

The divider barrier between the lower and upper containment limits steam bypassing the icecondenser in the event of a loss of coolant accident or postulated pipe break. Penetrationsthrough the divider barrier are sealed by elastomeric materials. As indicated in LRATable 3.5.2-1, under the line item for Removable gate (bulkhead) seals, and Table 3.5.2-5, underthe line item for Divider barrier penetration seals, the aging effects applicable to elastomericseals are cracking and change in material properties. These aging effects result from two generalaging mechanisms (thermal exposure and ionizing radiation). The noteworthy effects of thermalexposure on these seals are elongation and cracking, whereas those applicable to ionizationradiation are cracking, swelling, and melting. Abnormalities, such as swelling, surface cracking,discoloration, surface peeling, separation, and melting are readily identifiable by visualinspection. Consequently, during the performance of the Divider Barrier Seal InspectionProgram visual inspections, these aging effects would be observed as obvious abnormalitiesindicative of material degradation prior to having a detrimental effect on the intended function ofthe seals. Therefore, seal degradation due to cracking and change in material properties will bedetected by the visual inspections before the intended function of the seals is challenged.


Reference for RAI B.1.34-1

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 Scoping and Screening Results," AEP:NRC:4034-01, dated May 7, 2004[Accession No. ML041390360].

RAI B.1.34-2:

In license renewal application (LRA) Section B.1.34, the Divider Barrier Seal InspectionProgram manages cracking and change in material property of elastomeric seals. Please clarifythe acceptance criteria for evaluating changes in material properties of elastomeric components,specifically, the pressure seals (divider barrier). Implementing procedures mention evidence ofchemical attack, radiation damage, or changes in physical appearance. Please clarify hows thesewill be evaluated (acceptance criteria) and confirm that visual evidence of degradation willprecede loss offunction.

I&M Response to RAI B.1.34-2:

Similar to the response clarification provided to RAI B.1.34-1 above, the elastomeric pressureseals to be addressed by this response are also the divider barrier penetration seals installedaround penetrations and openings through the divider barrier.

As indicated in LRA Table 3.5.2-1, under the line item for Removable gate (bulkhead) seals, andTable 3.5.2-5, under the line item for Divider barrier penetration seals, the aging effectsapplicable to elastomeric seals are cracking and change in material properties. During theperformance of visual inspections in accordance with the Structures Monitoring Program -Divider Barrier Seal Inspection Program, any abnormalities that would be indicative of materialdegradation that could affect intended function of the seals would be identified. The acceptancecriteria for these inspections is the absence of elastomeric seal material abnormalities, such asswelling, surface cracking, discoloration, surface peeling, separation, melting, holes, ruptures,abrasions, or other changes in appearance. If the elastomeric seal material demonstrates obvioussigns of degradation, the seal's condition will be identified in the Corrective Action Program,and the condition will be evaluated to ensure the pressure-retaining function of the degraded sealis not affected. Any abnormality and degradation (no matter how minor) is evaluated foracceptability and possible repair or replacement. Therefore, visual evidence of degradation ofthe containment divider barrier penetration seals will precede loss of intended function of thesecomponents.

Attachment 1 to AEP NRC:4034-15 Page 4

RAI 3.1.3-1:

In LRA Table 3.1.2-3, to manage cracking of the bolting material for valves and blindflanges,and main flange bolts, in LRA Table 3.1.2-4, to manage cracking of low-alloy steel manis'aycover bolts/studs in ambient air, and in LRA Table 3.1.2-5, to manage cracking carbon steelbolting of the secondary manway, handhold, recirculation port (Unit 1), and inspection portclosure in ambient air of the steam generators, the applicant proposes to use the InserviceInspection Program. Although a precedent was cited, the staff was not able to confirm itsapplicability. For the components referenced, the Generic Aging Lessons Learned (GALL)Report recommends a program consistent with "Bolting Integrity" (GALL AMP XI.M18) whichreferences the guidelines of NUREG-1339 to prevent and mitigate bolting degradation. Pleaseexplain the rationale for excluding this bolting material from the scope of CNP LRA AgingManagement Program (AMP) B.1.2, "Bolting and Torquing Activities, " or confirm that it ismanaged using this program.

I&M Response to RAI 3.1.3-1:

The governing aging effect for bolted closures is loss of mechanical closure integrity. Loss ofmechanical closure integrity may be attributed to one or more aging effects applicable to thebolted joint, e.g., loss of pre-load, cracking of bolting material, and loss of material.Combinations of these aging effects may lead to loss of closure integrity of the joint, which mayresult in joint failure and loss of intended function. Loss of mechanical closure integrity ismanaged using a combination of programs that include the Bolting and Torquing ActivitiesProgram, the Inservice Inspection Program, and the Boric Acid Corrosion Prevention Program.

The Bolting and Torquing Activities Program is credited for managing loss of mechanicalclosure integrity for valve and pump bolting in LRA Table 3.1.2-3, for pressurizer manway coverbolts/studs in LRA Table 3.1.2-4, and for the steam generator secondary manway closure boltingin LRA Table 3.1.2-5.

Cracking is listed separately for the referenced components because the InserviceInspection Program more appropriately manages cracking of bolted closures in Class I systemsthan the Bolting and Torquing Activities Program. Both the Inservice Inspection - ASMESection XI, Subsection IWB, IWC, and IWD Program and the Inservice Inspection - ASMESection XI, Subsection IWE Program, which are described in LRA Sections B.1.14 and B.1.15,respectively, provide for ASME Section XI inservice inspections of Class I bolted closures.

Loss of material is listed separately for the referenced components because the Boric AcidCorrosion Prevention Program, which is described in LRA Section B.1.2, more appropriatelymanages loss of material for the referenced closure bolting exposed to boric acid than the Boltingand Torquing Activities Program. Loss of material for external surfaces, such as closure bolting,is a long-term aging effect that would be observed well before aging progressed to the point ofloss of intended function.


RAI 3.1.3-2:

In LRA Table 3.1.2-4, the applicant proposes to manage cracking of heater support plates, theirbrackets, and the bracket bolts using the Water Chemistry Control Program. Although aprecedent it'as cited, the staff isas not able to confirm that it is applicable to CANP. For thecomponents referenced, the GALL Report recommends the use of Inservice Inspection inaddition to the wvater chemistry control program. Please justify the absence of an inspection ormonitoring program to manage this aging effect, or identify the program used


CNP credits the Water Chemistry Control Program and Inservice Inspection Program to managecracking of the pressurizer immersion heater sheaths. These pressure boundary components arecomparable to the NUREG- 1801, Volume 2, pressurizer heater sheaths and sleeves, which alsocredit the Water Chemistry Control Program and Inservice Inspection Program to managecracking, as indicated in LRA Table 3.1.2-4 on Page 3.1-74. The pressurizer heater supportplates, brackets and bracket bolts, also listed in LRA Table 3.1.2-4 on Page 3.1-74, are internal tothe pressurizer and not accessible for inspection. These components provide lateral support forthe heaters and serve no pressure boundary function. As these components do not serve apressure boundary function, it is acceptable to credit a mitigative program, such the WaterChemistry Control Program, to manage cracking of the components. This is consistent with theNRC Staffs review of the pressurizer spray and surge nozzle thermal sleeves discussed onPage 3-110 of NUREG-1772, Safety Evaluation Report Related to the License Renewal ofMcGuire Nuclear Station, Units I and 2, and Catawtba Nuclear Station, Units I and 2.

RAI 3.1.3-4:

In LRA Table 3.1.2-5, cracking of the lowt-alloy steel lowver shell, uipper shell, transition cone,steam drum, elliptical zipper head, feedwvater nozzle and main steam nozzle, secondaryblowvdowvn and instrumentation connections, recirculation connections (Unit 1), and secondaryshell drain connections, secondary handhole and inspection ports, and carbon steel secondarymanivay and feedwvater elbow thermal liner (Unit 2) component types in treated wvater ismanaged by CNP AMP B. 1. 14, "Inservice Inspection - ASME Section XI, Subsection IWYB, IDVC,and IJW'D. " The applicant made reference to a previously approved staff position, howvever, inthe case cited, a Water Chemistry Control Program had been credited as wvell. No waterchemistry control program wvas identified for managing of this aging effect at CNP. Pleaseprovide the basis for concluding that water chemistry control is not required or identify theWater Chemistry Control Program that vill be used.



The referenced LRA Table 3.1.2-5 components are fabricated of carbon or low alloy steel thatare exposed to a treated water environment. Carbon steel and low alloy steel items in thisenvironment are potentially susceptible to loss of material by wear and general, pitting, andcrevice corrosion, and cracking due to metal fatigue and growth of pre-service flaws at thewelded joints due to service loadings. Loss of material for the referenced components ismanaged by the Water Chemistry Control Program. Growth of pre-service flaws at weldedjoints is managed by the Inservice Inspection Program and metal fatigue is a time-limited aginganalysis. Cracking of carbon and low alloy steel components that are exposed to a treated waterenvironment is an aging effect that is not mitigated by Water Chemistry Control. Managementof these aging effects is consistent with the treatment of other steam and power conversionsystems components, such as main feedwater system carbon steel component types "Piping" and"Valve" listed in LRA Table 3.4.2-1.

RAI 3.3.2-2:

In LRA Section 3.3.2.2.2, the applicant proposes to manage degradation of elastomers forventilation systems with the Preventive Maintenance Program. The staff requests clarificationon the method(s) used to monitor a change in material properties of elastomers in the ventilationsystems. Material properties that could affect the performance of elastomers (e.g., hardness,flexibility) are not directly measured. Please provide a basis for concluding that degradationwill be identified before the intendedfiunction is compromised. Othernvise, provide a technicalbasis for the conclusion that the elastomers in question are not subject to these effects or thatthese effects itill not interfere nit/h the intendedfiunction of the component.


The intended function of the elastomeric components in the ventilation systems is to maintainpressure boundary. These components will be inspected visually to detect cracking and changesin material properties. Visual inspections can detect cracking, discoloration, or change in surfacecondition in elastomer materials, all of which would be indicative of degradation that could leadto loss of pressure boundary. Hardness and flexibility are not critical properties for maintainingthe pressure boundary intended function. However, if the material should become excessivelyhard or brittle, cracking would result which would be visible during these inspections. Visualexaminations of the ventilation system elastomeric components will be performed at aninspection interval that provides assurance that any visually detectable abnormalities would bedetected and corrected prior to degradation from properties that are not visually detectable, suchas hardness and flexibility, before a loss of intended function would occur. Therefore, thePreventive Maintenance Program will use appropriate examination methods to ensure thatdegradation of elastomeric components in the ventilation systems will be identified before the


material properties that are not directly measured would compromise the pressure boundaryfunction of these components.

RAI 3.3.3-2:

In LRA Table 3.3.2-4, for elastomers in the compressed air system flex hoses, and in LRA Table3.3.2-8, for elastomers in the flex hoses associated with the emergency diesel generator (EDG),and in LRA Table 3.3.2-9, for elastomers in the flex hoses associated with the security diesel, andin LRA Table 3.3.2-10, for elastomers in the flex hoses associated with the containment hydrogenmonitoring system, the applicant proposes to manage change in material properties with thePreventive Maintenance Program. The staff requests clarification on the method(s) used tomonitor a change in material properties of elastomers in these flex hoses associated wt'ith thesesystems. Material properties that could affect the performance of elastomers (e.g., hardness,flexibility) are not directly measured Please provide a basis for concluding that degradationit-ill be identified before the intendedfiunction is compromised. Otherwise, provide a technicalbasis for the conclusion that the elastomers in question are not subject to these effects or thatthese effects wt'ill not interfere with the intendedfunction ofthe component.

I&M Rcsponsc to RAI 3.3.3-2:

As discussed in LRA Section B.1.25, the Preventive Maintenance Program will be enhanced tomanage the effects of aging on flexible hoses in the compressed air, EDG, security diesel, andcontainment hydrogen monitoring systems through visual examination and replacement asneeded. This visual examination of external and internal surfaces will look for cracking asevidence of changes in material properties such as loss of flexibility and embrittlement at aninspection interval that will provide assurance that degradation is identified before loss ofintended function. The flexibility of the hoses will be verified through physical manipulation ofthe hose during the visual inspection, thereby enhancing the inspector's ability to sense (bothvisually and through touch) a change in material properties that could affect the performance ofthe elastomers. Therefore, the Preventive Maintenance Program will use appropriateexamination methods to ensure that degradation of flexible hoses in the compressed air, EDG,security diesel, and containment hydrogen monitoring system will be identified before theintended function is compromised.

RAI 3.5.3-1:

In LRA Table 3.5.2-1, page 3.5-37, the applicant proposes to manage loss of material, cracking,and change of material properties of concrete exposed to borated ice for ice condenser supportslab and ice condenser wear slab using the Structures Monitoring Program. Please clarifywhether these component tjpes are accessible for direct monitoring and if not, describespecifically hoit' the associated aging effects will be monitored.



The ice condenser wear slab is accessible for inspection (direct monitoring) from inside the icecondenser during refueling outages. The ice condenser support slab is accessible for inspection(direct monitoring) from below in various rooms within the Containment annulus area(i.e., outside the crane wall).

RAI 3.5.3-2:

In LRA Table 3.5.2-5, the applicant proposes to manage fire proofing pyrocrete materials usingthe Fire Protection Program. Separation, cracking, and loss of material are considered to beapplicable aging effects for pyrocrete materials. The staff requests the applicant to identify howthe aging effects of separation, cracking, and loss of material are managed by the FireProtection Program or justify i/ why these aging effects are not applicable.


Pyrocrete fire-proofing material is not identified in NUREG-1801. As indicated in LRATable 3.5.2-5, I&M's aging management review did not identify any aging effects requiringmanagement for this material and environment combination. Review of operating experiencedid not indicate any aging effects requiring management for this material. However, during theconduct of Fire Protection Program inspections, pyrocrete material is typically monitored byvisual inspection for obvious degradation such as flaking, cracking, separation, and loss ofmaterial.

Attachment 2 to AEP:NRC:4034-15 Page I

Supplemental Responses to Requests for Additional Information andResponse to Draft Request for Additional Information for theDonald C. Cook Nuclear Plant License Renewal Application

This attachment provides information requested in a draft request for additional information(RAI) and other supplemental responses to RAIs requested by the Nuclear RegulatoryCommission (NRC) Staff pertaining to the following license renewal application (LRA) sections:

* 2.3.3 - Auxiliary Systems

* 3.6 - Electrical and Instrumentation Controls

* 4.4 - Environmental Qualification of Electrical Components

* 4.7 - Other Plant-Specific Time-Limited Aging Analyses

RAI 2.3.3.1-1:

Section 2.3.3.1, "Spent Fuel Pool" (SFP) of the LRA states that "The primary safety intendedfiunction of the spent fiuel pool system is to maintain adequate water inventory for shielding andto prevent criticality of the storedfiuel. " In a letter dated February 4, 1992, in response to thestaffs request for additional information on a license amendment request for D. C Cook,Units I and 2, the Indiana Michigan Power Company stated the follows'ing:

Make-lop water to the [spent fiuel] pool can be obtained from several reliable,permanently installed sources, including the [chemical and volume control system]hold-up tank recirculation pump, demineralized water supply, and [reactor waterstorage tank]... W[ith these diverse sources, make-up water wvill be readily availablein the event of loss of spent fiel pool cooling.

In the safety evaluation issued pursuant to the above amendments (Amendment Nos. 169 and 152to licenses DPR-58 and DPR-74 for D. C. Cook, Units I and 2) dated January 14, 1993, the staffstated the following:

In the safety evaluation issued pursuant to Amendment No. 32 to Facility OperatingLicense No. DPR-58 and Amendment No. 13 to Facility Operating License No.DPR- 74 for D. C. Cook, Units I and 2, respectively state that the spent fiuel poolmeets the design criteria of Regulatory Guide 1.13 which requires a diversity of makelip wloater sources to the spent fiuel pool. The SE states that in a previous SE forAmendment No. 32 and 13 to licenses DPR-58 and DPR-74, the staff accepted thechemical and volume control system hold-lip tanks as the Seismic Category I sourceof make lop 'water to the SFP. The hold-lip tank recirculation pump, which is ratedat 500 gpin, can be used to plump water from the hold lup tank to the SFP.


However, the license renewal drawing of the SFP. LRA-12-5136, does not showv the source ofmake-up waterfrom the chemical and voline control system (CVCS) hold-up tanks to the SEPas being subject to an AA1MR [Aging Management Reviewj. Justify the exclusion ofthe piping andcomponents linking the make-uip water source from the CVCS hold-up tanks, and at least oneother make-up water source to the SFP fromn being subject to an AMR in accordance with therequirements of 10 CFR 54.4(a)(1)(1ii) and 10 CFR 54.21(a)(1).

Clarification requested for RAI 2.3.3.1-1:

LRA 2.3.3.1 states that the primary safety intended function of the SFP system is to maintainadequate water inventory for shielding and to prevent criticality of the storedfiuel. Therefore, asource of makeup water is required to be wt'ithin the scope of license renewralfor meeting criteriaIO CFR 54.4(a) (2) because makeup finctionally supports the SFP system 's intendedfunction.

I&M's Supplemental Response to RAI 2.3.3.1-1:

Indiana Michigan Power Company (I&M) will credit fire water, 'which is supplied via the hosereel stations, as the in-scope source of makeup water to the SFP. The capacity of this makeupsource has been evaluated and determined to exceed the maximum calculated SFP boil-off rate.The fire water hose stations and associated supply piping that are capable of supplying makeupwater to the SFP are currently included in the scope of license renewal. The in-scope fire waterhose stations are depicted on license renewal drawing LRA-12-5152D as fire hose connection(FHC) stations FHC-83B at location H4, FHC-209C at location F6, and FHC-210C at locationH6. The "B" and "C" suffix describes the FHC arrangement shown on LRA-12-5152D atlocation C2.

As discussed in the LRA Section 2.3.3.7, the fire protection system, which includes the fire watersystem, is included in the scope of license renewal based on the criteria of 10 CFR 54.4(a)(2) and10 CFR 54.4(a)(3). Fire hoses are consumables, as discussed in LRA Section 2.1.2.4.4.

Therefore, fire water supplied via hose reel stations provide a source of makeup water that iswithin the scope of license renewal and will functionally support the SFP system's intendedfunction.

RAI 2.3.3.11-2:

LRA Table 2.3.3-11 identifies component types and intended functions as a group for these 17systems. The staff is unable to identify iWhich component types and intendedfiunctions in thetable correlate to which of the 17 systems described in LRA Section 2.3.3. 11. License renewaldrawings have not been providedfor these systems, nor does the UFSAR [Updated Final SafetyAnalysis Report] provide sufficient descriptive information. Therefore, the staff is unable toconclude, iith reasonable assurance, that the applicant has identified the mechanical system

Attachment 2 to AlP NRC:4034-15 Page 3

components for these systems that are it ilhin the scope of license reneival and subject to an AMvfRin accordance with the requirements of 10 CFR 54.4(a)(2) and 10 CFR 54.21('a)(1). In orderforthe staff to make this determination, provide drawings or text information wthich identifies thecomponents by system that are subject to an ASMR because they meet the intendedfiunction of10 CFR 54.4(a) (2) and 10 CFR 54.21(a) (1). If any of these components are not included inLRA Table 2.3.3- 1, revise the table.

Clarification Requested by the Staff:

Identify the components excludedfrom the scope of license renewtal because no safety-relatedequipment is in the "area. " Describe what is meant by the term "area. "

Identify the components excludedfronm the scope of license renetial because protection of thesafety-related equipment is provided by "design features. " Identify the "design features" anddiscuss wvhether they are it'ithin the scope of license reneit'al and subject to an A MR.

I&M's Supplemental Response to RAI 2.3.3.11-2:

As stated in I&M's response to RAI 2.3.3.11-2 provided in the referenced letter datedMay 7, 2004, all nonsafety-related components containing liquid or steam located in thecontainment building, auxiliary building, screenhouse, and the portion of the turbine buildingthat contains the auxiliary feedwater pumps were considered subject to aging managementreview unless no safety-related equipment was in the area. For the purpose of this review, anarea is defined as a plant space that is on the same floor (elevation) in a building with no barrierwalls between the nonsafety-related fluid-filled system and the safety-related component(s). Atthe Donald C. Cook Nuclear Plant (CNP), areas are identified with room numbers. Structuralwalls form the boundary of a room on the same elevation of a major building and separatesafety-related components from a spray or a leak from a nonsafety-related component. Thesewalls are within the scope of license renewal and subject to aging management review.

When performing the evaluation of nonsafety-related components for potential spatial impact onsafety-related systems, the evaluation considered that if there were no safety-related componentsinstalled in the same area as the nonsafety-related fluid-filled components, then thesenonsafety-related components were not included in the scope of 10 CFR 54.4(a)(2). Sometypical components that were excluded from the scope of 1O CFR 54.4(a)(2) based on thiscriterion include valves, piping, pump casings, tubing, thermowells, and strainers in thecirculating water, plant heating boiler, main generator, turbine auxiliary cooling water systems,and other fluid-filled systems. Such components are located in areas that do not containsafety-related equipment in the auxiliary building, the screenhouse, or the portion of the turbinebuilding that contains the auxiliary feedwater pumps.

Additional reviews were performed to exclude specific nonsafety-related components wheredesign features, such as panels or enclosures, would protect safety-related equipment from


leakage or spray. Some typical components and portions of systems that were excluded from thescope of 10 CFR 54.4(a)(2) using this criterion include valves, piping, pump casings, tubing,sample coolers, cooling coils and thermowells in the nuclear sampling, post accident sampling,auxiliary building ventilation, and miscellaneous ventilation systems. The design features thatprotect safety-related equipment from leakage or spray are ventilation equipment housings andsample panels that contain the fluid-filled nonsafety-related components. These design featuresprevent failures of the nonsafety-related equipment from spraying or leaking onto safety-relatedequipment in the area. These enclosures are within the scope of license renewal and subject toaging management review.

Reference for Supplemental Response to RAI 2.3.3.11-2

Letter from M. K. Nazar, I&M, to NRC Document Control Desk, "Donald C. Cook NuclearPlant, Units I and 2, Docket Nos. 50-315 and 50-316, License Renewal Application - Responseto Requests for Additional Information on Scoping and Screening Results (TAC Nos. MC1202and MC1203)," AEP:NRC:4034-01, dated May 7, 2004 [Accession No. ML041390360].

RAI 3.6-5:

The updated FSAR [final safety analysis report] supplement description in the LRA for thenon-EQ [environmental qualificationI cable AMqP does not provide an adequate description ofthe program as required by 1 CFR 54.21(d). The description of FSAR supplement for agingmanagement of electrical and instrumentation and controls system should be consistent withTable 3.6-2 of NUREG-1800. Please submit a revised FSAR supplement that is consistent withNUREG-1800 to satisfy 10 CFR 54.21(d).


Although not explicitly stated in the RAI, the staff indicated that the intent of the RAI was to haveall three Non-EQ programs provide an adequate description in the revised FSAR supplement inaccordance with Table 3.6-2 of NUREG-1800. The applicant acknowledged the question'sintent and agreed to revise the response to include all three Non-EQ programs in the FSARsupplement.

I&M's Supplemental Response to RAI 3.6-5:

Based on a review of NUREG-1800, Table 3.6-2, and NUREG-1801, Section XI.E2 and E3, theprogram descriptions for Updated Final Safety Analysis Report Sections A.2.1.23 and A.2.1.24are revised as follows:

(NOTE: The text added for clarification is in italics.)


A.2.1.23 NON-EQ INACCESSIBLE MEDIUMI-VOLTAGE CABLE PROGRAM

The Non-EQ Inaccessible Medium-Voltage Cable Program will apply to inaccessible(e.g., in conduit or direct-buried) medium-voltage cables within the scope of licenserenewal that are exposed to significant moisture simultaneously with significantvoltage. Significant moisture is defined as periodic exposures that last more than a

felt} days (e.g., cable in standing it'ater). Significant voltage exposure is defined asbeing subjected to system voltage for more than 25 percent of the time. Under thisprogram, in-scope medium-voltage cables that are exposed to significant moistureand significant voltage will be tested at least once every 10 years to provide anindication of the condition of the conductor insulation. The specific type of testperformed will be determined prior to the initial test, and itiill be based on technologythat is state-of-the-art at the time the test is performed. The Non-EQ InaccessibleMedium-Voltage Cable Program will be implemented prior to the period of extendedoperation.

A.2.1.24 NON-EQ INSTRUMENTATION CIRCUITS TEST REVIEW PROGRAM

The Non-EQ Instrumentation Circuits Test Review Program will manage agingeffects for electrical cables that:

* Are not subject to the environmental qualification requirements of 10 CFR 50.49,and

* Are used in instrumentation circuits with sensitive, high-voltage, low-levelsignals, such as radiation monitoring and nuclear instrumentation, which areexposed to adverse localized environments caused by heat, radiation, or moisture.

An adverse localized environment is defined as being significantly more severe thanthe specified service environment for the cable. This program will detect agingeffects by reviewing calibration or surveillance results for components within theprogram scope at afrequency not to exceed 10 years or as part of corrective actionsthen acceptance criteria are exceeded at the normal calibration frequency. TheNon-EQ Instrumentation Circuits Test Review Program will be implemented prior tothe period of extended operation.


RAI 4.4-1:

The environmental qualification [EQ] of electrical equipmnent resuhlts in Section 4.4 indicate thatthe aging effects of the EQ of electrical equipment identified in the Time Limited Aging Analysis(TLAA) will be managed (lhring the extended period of operation under I0 CFR 54.21 (c))(J) (iii).Howt-ever, no information is provided on the attributes for re-analysis of an aging evaluation toextend the qualification life of electrical equipment identified in the TLAA. The importantattributes of a re-analysis include analytical methods, data collection and reduction methods,underlying assumptions, acceptance criteria and corrective actions. Provide information on theimportant attributes for reanalysis of an aging evaluation of electrical equipment identified inthe TLAA to extend the qualification under IO CFR 50.49(e).


The staff stated that previous applicants hav e stated hows they applied the re-analysis attributes.The applicant stated that based on a review of RAI responses, it appeared that previousapplicants simply "cut and pasted" the re-analysis attributes fromn GALL [Generic AgingLessons Learned], and that I&M opted to commit to a program that is consistent wtith GALL,Section X.EI, rather than duplicating this section in the RAI response. The applicant alsoindicated that the EQ Program is available onsite, and av'ailable for JVRC audit, if so desiredThis program wsas not part of the audit agenda and w-as not reviewed during the audit.Therefore, the licensee is requested to submit a discussion of holw they met the ten elements ofGALL regarding environmental qualification of electrical components.

I&M's Supplemental Response to RAI 4.4-1:

As noted in LRA Section 4.4, the CNP Environmental Qualification of Electric ComponentsProgram is an existing program that was established to meet commitments associated withIO CFR 50.49. The CNP Environmental Qualification of Electric Components Program isconsistent with the program described in NUREG-1801, Section X.EI, "EnvironmentalQualification (EQ) of Electric Components."

Based upon a review of the existing program and operating experience, continuedimplementation of the Environmental Qualification of Electric Components Program providesreasonable assurance that the aging effects will be managed and that in-scope EQ componentswill continue to perform their intended function(s) for the period of extended operation. Theeffects of aging related to EQ evaluations will be managed for the period of extended operationin accordance with 10 CFR 54.21 (c)(1)(iii).

The comparison of the CNP Environmental Qualification of Electric Components Program to theten elements of NUREG-1801. Section X.El, is provided in the following paragraphs.


Program Description

The CNP Environmental Qualification of Electric Components Program manages componentthermal, radiation, and cyclical aging through the use of aging evaluations based on1O CFR 50.49(f) qualification methods. As required by 1O CFR 50.49, EQ components notqualified for the current license term are to be refurbished, replaced, or have their qualificationextended prior to reaching the aging limits established in the evaluation. Aging evaluations forEQ components that specify a qualification of at least 40 years are considered TLAAs for licenserenewal.

Aging Management Program Elements

Scope

The CNP Environmental Qualification of Electric Components Program provides therequirements for the environmental qualification of electrical equipment important to safety asrequired by 10 CFR 50.49.

The scope of the CNP program is consistent with NUREG-1 801.

Preventive Actions

10 CFR 50.49 does not require actions that prevent aging effects. Environmental Qualificationof Electric Components Program actions that could be viewed as preventive actions include(a) establishing the component service condition tolerance and aging limits (for example,qualified life or condition limit), and (b) where applicable, requiring specific installation,inspection, monitoring or periodic maintenance actions to maintain component aging effectswithin the bounds of the qualification basis.

The preventive actions of the CNP program are consistent with NUREG- 1801.

Parameters Monitored or Inspected

The CNP Environmental Qualification of Electric Components Program provides EQ relatedsurveillance and maintenance requirements for EQ equipment, but does not require condition orperformance monitoring.

The parameters monitored and inspected by the CNP program are consistent with NUREG-1801.


Detection of Aging Effects

The detection of aging effects is not required for compliance with 1O CFR 50.49 at CNP.Monitoring or inspection of certain environmental conditions or component parameters may beused to ensure that the component is within the bounds of its qualification basis, or as a means tomodify the qualified life.

The detection of aging effects of the CNP program is consistent with NUREG-1 801.

Monitoring and Trending

The CNP Environmental Qualification of Electric Components Program does not requiremonitoring and trending of EQ equipment. The program provides surveillance and maintenancerequirements for EQ equipment and verifies that the required activities are performed.Monitoring the service life of qualified components is part of the CNP program.

The monitoring and trending requirements of the CNP program are consistent withNUREG-1801.

Acceptance Criteria

The CNP Environmental Qualification of Electric Components Program provides surveillanceand maintenance activities to assure that the acceptance criteria for EQ equipment is maintainedwithin its qualification basis and qualified life. The program provides that any EQ equipmentshall be replaced, refurbished, or requalified prior to exceeding the qualified life of theequipment. If monitoring is used to modify a component qualified life, appropriateplant-specific acceptance criteria will be established based on applicable 10 CFR 50.49(f)qualification methods.

The CNP acceptance criteria are consistent with NUREG-1801.

Corrective Actions

If an EQ component is found to be outside the bounds of its qualification basis, unexpectedadverse conditions are identified during operational or maintenance activities that affect theenvironment of a qualified component, or an emerging industry aging issue is identified thataffects the qualification of an EQ component, the affected EQ component is evaluated andappropriate corrective actions are taken, which may include changes to the qualification basesand conclusions. The adverse condition can be identified in a condition report (CR), industrynotification such as NRC Information Notice, NRC Generic Letter, 10 CFR Part 21 Notice ordesign change. The requirements of 10 CFR Part 50, Appendix B, are applied throughimplementation of the CNP Corrective Action Program.


The corrective actions for the CNP program are consistent with NUREG-1 801.

Confirmation Process

The confirmation process is discussed in LRA Section B.O.3.

Administrative Controls

Administrative controls are discussed in LRA Section B.0.3.

Operating Experience

CNP operating experience relative to the Environmental Qualification of Electric ComponentsProgram includes CRs, NRC Inspection Reports, and documentation of the results of internalprogram assessments.

The operating experience discussion in NUREG-1801, Section X.EI, states, "EQ programsinclude consideration of operating experience to modify qualification bases and conclusions,including qualified life. Compliance with 10 CFR 50.49 provides reasonable assurance thatcomponents can perform their intended functions during accident conditions after experiencingthe effects of inservice aging."

The I&M report titled "Engineering Functional Area Attribute Restart Readiness, AffirmationReport for Environmental Qualification (EQ) Program," dated December 22, 1999, wasreviewed. This report concludes "...there is reasonable assurance that this program will functionsatisfactorily to support restart and continuous operation of CNP." Therefore based on theoperating experience discussion of NUREG-1801, the CNP program provides reasonableassurance that components can perform their intended functions during accident conditions afterexperiencing the effects of inservice aging.

For further confirmation of the program effectiveness, CRs identified in the program procedurerevisions and restart assessment report were reviewed. The vast majority address administrativeissues associated with the program, records, and auditability. Program revisions have been madeas part of corrective actions for these CRs. The only equipment failures identified were failuresof flood-up tubes, which were determined to have been related to installation techniques and notaging. No CRs identified aging effects for which the program is intended to prevent. Thisoperating experience is consistent with the conclusion that the program is effective in preventingthe effects of aging.

A CR initiated in 1998 identified programmatic problems with the Environmental Qualificationof Electric Components Program. The root cause analysis for this CR determined that theunderlying causes were programmatic issues, including lack of management support/attention.The resultant corrective actions included numerous action plans and programmatic changes to

Attachment 2 to AEP NRC:4034-15 Page I10

correct these deficiencies. The result of this operating experience was a much strongerEnvironmental Qualification of Electric Components Program.

The Environmental Qualification of Electric Components Program has been effective atmanaging aging effects. Operating experience has identified no aging effects for which theprogram is intended to prevent. The program is continuing to be improved as a result of ongoingprogram assessments. The continued implementation of this program provides reasonableassurance that the aging effects will be managed so that the applicable components will continueto perform their intended functions consistent with the current licensing basis for the period ofextended operation.

Conclusion

The CNP Environmental Qualification of Electric Components Program is consistent with theprogram described in NUREG-1801, Section X.EI, "Environmental Qualification (EQ) ofElectric Components."

The overall effectiveness of the Environmental Qualification of Electric Components Program isdemonstrated by the excellent operating experience for systems and components in the program.The program has been subject to periodic internal and external assessments that facilitatecontinuous improvement. Continued implementation of this program provides reasonableassurance that components within the scope of license renewal will continue to perform theirintended functions consistent with the current licensing basis for the period of extendedoperation.

RAI 4.7.4-1:

The LRA Section 4.7.4, "Reactor Vessel Underclad Cracking, " states, "The numbers of designcycles and transients assumed in the T'CAP-15338 analysis bound the number of design cyclesand transients projectedfor 60 years of operation. " Please provide information regarding howyou arrived at this conclusion.


WCAP-15338-A, dated October 2002, includes the types and numbers of reactor coolant system(RCS) design transients utilized for evaluation of underclad cracking flaw growth over 60 yearsof operation. For CNP, the types and numbers of RCS design transients, with the exception ofthe feedwater cycling at hot shutdown, were verified to be bounded by the design transientsassumed in WCAP-15338-A, thereby satisfying Renewal Applicant Action Item (1) of theRevised Safety Evaluation Report of WCAP-15338, dated September 25, 2002 [Accession No.ML022690375]. The feedwater cycling at hot shutdown transient is associated with a feedwaternozzle cracking concern and is not monitored at CNP due to design and operating modifications

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Attachment 2 to AEP NRC:4034- 15 Page 11

to preclude feedwater nozzle cracking. This transient is not anticipated to have a significantimpact on crack growth beneath the reactor vessel cladding. CNP's projected number of RCSdesign transients for 60 years, as shown in LRA Table 4.3-1, does not exceed applicable designassumptions assumed in WCAP-15338-A. Therefore, the WCAP-15338-A RCS transientsbound the CNP RCS transients, and WCAP-15338-A remains applicable to CNP for the periodof extended operation.

michigan power september 2, 2004 aep:nrc:4034-15 mail … · j. n. jensen site vice president...

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