ML042530551

License Renewal Application - Response to Requests for Additional Information

ML042530551
Person / Time
Site:	Cook
Issue date:	09/02/2004
From:	Jensen J Indiana Michigan Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP:NRC:4034-15, TAC MC1202, TAC MC1203
Download: ML042530551 (23)

Text

Indiana Michigan Power Company 500 Circle Drive Buchanan, MI 49107 1395 INDIANA MICHIGAN POWER September 2, 2004 AEP:NRC:4034-15 10 CFR 54 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop O-PI-17 Washington, DC 20555-0001

SUBJECT:

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

The Nuclear Regulatory Commission (NRC) review process includes audits of the aging management programs (AMPs) credited in the CNP license renewal application (LRA).

During the conduct of these audits and subsequent to conversations between I&M and NRC Staff, the NRC Staff identified areas where additional information was needed to complete its review of the credited AMPs.

This letter provides the information requested by the NRC Staff subsequent to the completion of the AMP audits pertaining to the following 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 requests for additional information (RAls) addressed in this letter were received in an NRC letter dated August 20, 2004 (Reference 2).

U S. Nuclear Regulatory Commission AEP:NRC:4034-15 Page 2 In addition, this letter provides supplemental responses to RAls requested by the NRC Staff and response to information requested in a draft RAI pertaining to the following 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 statements made in this letter. Attachment I to this letter provides I&M's responses to the NRC Staff's RAls. Attachment 2 to this letter provides I&M's supplemental responses to RAls and a response to the NRC Staff's draft RAI. There are no new commitments contained in this submittal.

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

Sincerely, J. N. Jensen Site Vice President NH/rdw

Enclosure:

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

2. Supplemental Responses to Requests for Additional Information and Response to Draft Request for Additional Information for the Donald C. Cook Nuclear Plant License Renewal Application

U S. Nuclear Regulatory Commission AEP:NRC:4034-15 Page 3

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 Operating Licenses,"

AEP:NRC:3034, dated October 31, 2003

[Accession No. ML033070177].

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

dated August 20, 2004

[Accession No. ML042330355].

c:

J. L. Caldwell, NRC Region III K. D. Curry, AEP Ft. Wayne, w/o attachments J. T. King, MPSC, w/o attachments J. G. Lamb, NRC Washington DC MDEQ - WHMD/HWRPS, w/o attachments NRC Resident Inspector J. 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 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 ph N. Jensen Site 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 the Donald 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 additional information (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 dated August 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 I and 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 material properties of elastomers, specifically, the pressure seals (divider barrier).

The SRP-LR Appendix A. 1.2.3.3 states that "parameters to be monitored or inspected should be identified and linked 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 "change in material properties" was intended to convey a visual inspection to ensure the absence of apparent deterioration (i.e., cracks or defects in the sealing surfaces) as discussed in the implementing procedures.

Please provide the basis for concluding that the elastomeric divider barrier will continue to perform 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 and the 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 also includes personnel access doors between the containment's upper and lower compartments.

• Divider barrier penetration seals installed around penetrations and openings through the divider barrier.

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

The divider barrier between the lower and upper containment limits steam bypassing the ice condenser in the event of a loss of coolant accident or postulated pipe break.

Penetrations through the divider barrier are sealed by elastomeric materials.

As indicated in LRA Table 3.5.2-1, under the line item for Removable gate (bulkhead) seals, and Table 3.5.2-5, under the line item for Divider barrier penetration seals, the aging effects applicable to elastomeric seals are cracking and change in material properties. These aging effects result from two general aging mechanisms (thermal exposure and ionizing radiation). The noteworthy effects of thermal exposure on these seals are elongation and cracking, whereas those applicable to ionization radiation are cracking, swelling, and melting. Abnormalities, such as swelling, surface cracking, discoloration, surface peeling, separation, and melting are readily identifiable by visual inspection.

Consequently, during the performance of the Divider Barrier Seal Inspection Program visual inspections, these aging effects would be observed as obvious abnormalities indicative of material degradation prior to having a detrimental effect on the intended function of the seals. Therefore, seal degradation due to cracking and change in material properties will be detected by the visual inspections before the intended function of the seals is challenged.

Attachment I to AEP NRC:4034-15 Page 3 Reference for RAI B.1.34-1 Letter from M. K. Nazar, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant, Units I and 2, License Renewal Application - Response to Requests for Additional Information on 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 Inspection Program manages cracking and change in material property of elastomeric seals. Please clarify the acceptance criteria for evaluating changes in material properties of elastomeric components, specifically, the pressure seals (divider barrier). Implementing procedures mention evidence of chemical attack, radiation damage, or changes in physical appearance. Please clarify hows these will be evaluated (acceptance criteria) and confirm that visual evidence of degradation will precede 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 pressure seals to be addressed by this response are also the divider barrier penetration seals installed around 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, and Table 3.5.2-5, under the line item for Divider barrier penetration seals, the aging effects applicable to elastomeric seals are cracking and change in material properties.

During the performance of visual inspections in accordance with the Structures Monitoring Program -

Divider Barrier Seal Inspection Program, any abnormalities that would be indicative of material degradation that could affect intended function of the seals would be identified. The acceptance criteria for these inspections is the absence of elastomeric seal material abnormalities, such as swelling, surface cracking, discoloration, surface peeling, separation, melting, holes, ruptures, abrasions, or other changes in appearance. If the elastomeric seal material demonstrates obvious signs 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 seal is not affected.

Any abnormality and degradation (no matter how minor) is evaluated for acceptability and possible repair or replacement. Therefore, visual evidence of degradation of the containment divider barrier penetration seals will precede loss of intended function of these components.

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'ay cover bolts/studs in ambient air, and in LRA Table 3.1.2-5, to manage cracking carbon steel bolting of the secondary manway, handhold, recirculation port (Unit 1), and inspection port closure in ambient air of the steam generators, the applicant proposes to use the Inservice Inspection Program. Although a precedent was cited, the staff was not able to confirm its applicability. For the components referenced, the Generic Aging Lessons Learned (GALL)

Report recommends a program consistent with "Bolting Integrity" (GALL AMP XI.M18) which references the guidelines of NUREG-1339 to prevent and mitigate bolting degradation. Please explain the rationale for excluding this bolting material from the scope of CNP LRA Aging Management Program (AMP) B.1.2, "Bolting and Torquing Activities, " or confirm that it is managed 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 of mechanical closure integrity may be attributed to one or more aging effects applicable to the bolted 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 may result in joint failure and loss of intended function. Loss of mechanical closure integrity is managed using a combination of programs that include the Bolting and Torquing Activities Program, the Inservice Inspection Program, and the Boric Acid Corrosion Prevention Program.

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

Cracking is listed separately for the referenced components because the Inservice Inspection Program more appropriately manages cracking of bolted closures in Class I systems than the Bolting and Torquing Activities Program.

Both the Inservice Inspection - ASME Section XI, Subsection IWB, IWC, and IWD Program and the Inservice Inspection - ASME Section 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 Acid Corrosion Prevention Program, which is described in LRA Section B.1.2, more appropriately manages loss of material for the referenced closure bolting exposed to boric acid than the Bolting and 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 of loss of intended function.

Attachment I to AEP NRC:4034-15 Page 5 RAI 3.1.3-2:

In LRA Table 3.1.2-4, the applicant proposes to manage cracking of heater support plates, their brackets, and the bracket bolts using the Water Chemistry Control Program. Although a precedent it'as cited, the staff isas not able to confirm that it is applicable to CANP. For the components referenced, the GALL Report recommends the use of Inservice Inspection in addition to the wvater chemistry control program. Please justify the absence of an inspection or monitoring program to manage this aging effect, or identify the program used I&M Response to RAI 3.1.3-2:

CNP credits the Water Chemistry Control Program and Inservice Inspection Program to manage cracking of the pressurizer immersion heater sheaths. These pressure boundary components are comparable to the NUREG-1801, Volume 2, pressurizer heater sheaths and sleeves, which also credit the Water Chemistry Control Program and Inservice Inspection Program to manage cracking, as indicated in LRA Table 3.1.2-4 on Page 3.1-74. The pressurizer heater support plates, brackets and bracket bolts, also listed in LRA Table 3.1.2-4 on Page 3.1-74, are internal to the pressurizer and not accessible for inspection. These components provide lateral support for the heaters and serve no pressure boundary function.

As these components do not serve a pressure boundary function, it is acceptable to credit a mitigative program, such the Water Chemistry Control Program, to manage cracking of the components. This is consistent with the NRC Staffs review of the pressurizer spray and surge nozzle thermal sleeves discussed on Page 3-110 of NUREG-1772, Safety Evaluation Report Related to the License Renewal of McGuire 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, secondary blowvdowvn and instrumentation connections, recirculation connections (Unit 1), and secondary shell drain connections, secondary handhole and inspection ports, and carbon steel secondary manivay and feedwvater elbow thermal liner (Unit 2) component types in treated wvater is managed 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, in the case cited, a Water Chemistry Control Program had been credited as wvell.

No water chemistry control program wvas identified for managing of this aging effect at CNP. Please provide the basis for concluding that water chemistry control is not required or identify the Water Chemistry Control Program that vill be used.

Attachment I to AEP NRC:4034-15 Page 6 I&M Response to RAI 3.1.3-4:

The referenced LRA Table 3.1.2-5 components are fabricated of carbon or low alloy steel that are exposed to a treated water environment.

Carbon steel and low alloy steel items in this environment are potentially susceptible to loss of material by wear and general, pitting, and crevice corrosion, and cracking due to metal fatigue and growth of pre-service flaws at the welded joints due to service loadings.

Loss of material for the referenced components is managed by the Water Chemistry Control Program.

Growth of pre-service flaws at welded joints is managed by the Inservice Inspection Program and metal fatigue is a time-limited aging analysis. Cracking of carbon and low alloy steel components that are exposed to a treated water environment is an aging effect that is not mitigated by Water Chemistry Control. Management of these aging effects is consistent with the treatment of other steam and power conversion systems 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 for ventilation systems with the Preventive Maintenance Program. The staff requests clarification on the method(s) used to monitor a change in material properties of elastomers in the ventilation systems. Material properties that could affect the performance of elastomers (e.g., hardness, flexibility) are not directly measured. Please provide a basis for concluding that degradation will be identified before the intendedfiunction is compromised. Othernvise, provide a technical basis for the conclusion that the elastomers in question are not subject to these effects or that these effects itill not interfere nit/h the intendedfiunction of the component.

I&M Response to RAI 3.3.2-2:

The intended function of the elastomeric components in the ventilation systems is to maintain pressure boundary. These components will be inspected visually to detect cracking and changes in material properties. Visual inspections can detect cracking, discoloration, or change in surface condition in elastomer materials, all of which would be indicative of degradation that could lead to loss of pressure boundary. Hardness and flexibility are not critical properties for maintaining the pressure boundary intended function. However, if the material should become excessively hard or brittle, cracking would result which would be visible during these inspections. Visual examinations of the ventilation system elastomeric components will be performed at an inspection interval that provides assurance that any visually detectable abnormalities would be detected and corrected prior to degradation from properties that are not visually detectable, such as hardness and flexibility, before a loss of intended function would occur.

Therefore, the Preventive Maintenance Program will use appropriate examination methods to ensure that degradation of elastomeric components in the ventilation systems will be identified before the

Attachment I to AEP NRC:4034-15 Page 7 material properties that are not directly measured would compromise the pressure boundary function 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 Table 3.3.2-8, 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, and in LRA Table 3.3.2-10, for elastomers in the flex hoses associated with the containment hydrogen monitoring system, the applicant proposes to manage change in material properties with the Preventive Maintenance Program. The staff requests clarification on the method(s) used to monitor a change in material properties of elastomers in these flex hoses associated wt'ith these systems. Material properties that could affect the performance of elastomers (e.g., hardness, flexibility) are not directly measured Please provide a basis for concluding that degradation it-ill be identified before the intendedfiunction is compromised. Otherwise, provide a technical basis for the conclusion that the elastomers in question are not subject to these effects or that these 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 to manage the effects of aging on flexible hoses in the compressed air, EDG, security diesel, and containment hydrogen monitoring systems through visual examination and replacement as needed. This visual examination of external and internal surfaces will look for cracking as evidence of changes in material properties such as loss of flexibility and embrittlement at an inspection interval that will provide assurance that degradation is identified before loss of intended function. The flexibility of the hoses will be verified through physical manipulation of the hose during the visual inspection, thereby enhancing the inspector's ability to sense (both visually and through touch) a change in material properties that could affect the performance of the elastomers.

Therefore, the Preventive Maintenance Program will use appropriate examination methods to ensure that degradation of flexible hoses in the compressed air, EDG, security diesel, and containment hydrogen monitoring system will be identified before the intended 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 support slab and ice condenser wear slab using the Structures Monitoring Program. Please clarify whether these component tjpes are accessible for direct monitoring and if not, describe specifically hoit' the associated aging effects will be monitored.

Attachment I to AEP NRC:4034-15 Page 8 I&M Response to RAI 3.5.3-1:

The ice condenser wear slab is accessible for inspection (direct monitoring) from inside the ice condenser 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 using the Fire Protection Program. Separation, cracking, and loss of material are considered to be applicable aging effects for pyrocrete materials. The staff requests the applicant to identify how the aging effects of separation, cracking, and loss of material are managed by the Fire Protection Program or justify i/

why these aging effects are not applicable.

I&M Response to RAI 3.5.3-2:

Pyrocrete fire-proofing material is not identified in NUREG-1801.

As indicated in LRA Table 3.5.2-5, I&M's aging management review did not identify any aging effects requiring management for this material and environment combination. Review of operating experience did not indicate any aging effects requiring management for this material. However, during the conduct of Fire Protection Program inspections, pyrocrete material is typically monitored by visual inspection for obvious degradation such as flaking, cracking, separation, and loss of material.

to AEP:NRC:4034-15 Page I Supplemental Responses to Requests for Additional Information and Response to Draft Request for Additional Information for the Donald 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 Regulatory Commission (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 intended fiunction of the spent fiuel pool system is to maintain adequate water inventory for shielding and to prevent criticality of the storedfiuel. " In a letter dated February 4, 1992, in response to the staffs 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 water storage tank]... W[ith these diverse sources, make-up water wvill be readily available in the event of loss of spent fiel pool cooling.

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

In the safety evaluation issued pursuant to Amendment No. 32 to Facility Operating License 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 pool meets the design criteria of Regulatory Guide 1.13 which requires a diversity of make lip wloater sources to the spent fiuel pool. The SE states that in a previous SE for Amendment No. 32 and 13 to licenses DPR-58 and DPR-74, the staff accepted the chemical and volume control system hold-lip tanks as the Seismic Category I source of make lop 'water to the SFP. The hold-lip tank recirculation pump, which is rated at 500 gpin, can be used to plump water from the hold lup tank to the SFP.

to AEP NRC:4034-15 Page 2 However, the license renewal drawing of the SFP. LRA-12-5136, does not showv the source of make-up waterfrom the chemical and voline control system (CVCS) hold-up tanks to the SEP as being subject to an AA1MR [Aging Management Reviewj. Justify the exclusion ofthe piping and components linking the make-uip water source from the CVCS hold-up tanks, and at least one other make-up water source to the SFP fromn being subject to an AMR in accordance with the requirements 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 maintain adequate water inventory for shielding and to prevent criticality of the storedfiuel. Therefore, a source of makeup water is required to be wt'ithin the scope of license renewralfor meeting criteria IO 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 hose reel stations, as the in-scope source of makeup water to the SFP. The capacity of this makeup source 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 makeup water to the SFP are currently included in the scope of license renewal. The in-scope fire water hose 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 location H6.

The "B" and "C" suffix describes the FHC arrangement shown on LRA-12-5152D at location C2.

As discussed in the LRA Section 2.3.3.7, the fire protection system, which includes the fire water system, is included in the scope of license renewal based on the criteria of 10 CFR 54.4(a)(2) and 10 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 is within the scope of license renewal and will functionally support the SFP system's intended function.

RAI 2.3.3.11-2:

LRA Table 2.3.3-11 identifies component types and intended functions as a group for these 17 systems. The staff is unable to identify iWhich component types and intendedfiunctions in the table correlate to which of the 17 systems described in LRA Section 2.3.3. 11. License renewal drawings have not been providedfor these systems, nor does the UFSAR [Updated Final Safety Analysis Report] provide sufficient descriptive information. Therefore, the staff is unable to conclude, iith reasonable assurance, that the applicant has identified the mechanical system to AlP NRC:4034-15 Page 3 components for these systems that are it ilhin the scope of license reneival and subject to an AMvfR in accordance with the requirements of 10 CFR 54.4(a)(2) and 10 CFR 54.21('a)(1). In orderfor the staff to make this determination, provide drawings or text information wthich identifies the components by system that are subject to an ASMR because they meet the intendedfiunction of 10 CFR 54.4(a) (2) and 10 CFR 54.21(a) (1).

If any of these components are not included in LRA 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-related equipment is in the "area. " Describe what is meant by the term "area. "

Identify the components excludedfronm the scope of license renetial because protection of the safety-related equipment is provided by "design features. " Identify the "design features" and discuss 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 dated May 7, 2004, all nonsafety-related components containing liquid or steam located in the containment building, auxiliary building, screenhouse, and the portion of the turbine building that contains the auxiliary feedwater pumps were considered subject to aging management review unless no safety-related equipment was in the area. For the purpose of this review, an area is defined as a plant space that is on the same floor (elevation) in a building with no barrier walls between the nonsafety-related fluid-filled system and the safety-related component(s). At the Donald C. Cook Nuclear Plant (CNP), areas are identified with room numbers. Structural walls form the boundary of a room on the same elevation of a major building and separate safety-related components from a spray or a leak from a nonsafety-related component. These walls 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 on safety-related systems, the evaluation considered that if there were no safety-related components installed in the same area as the nonsafety-related fluid-filled components, then these nonsafety-related components were not included in the scope of 10 CFR 54.4(a)(2).

Some typical components that were excluded from the scope of 10 CFR 54.4(a)(2) based on this criterion include valves, piping, pump casings, tubing, thermowells, and strainers in the circulating 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 contain safety-related equipment in the auxiliary building, the screenhouse, or the portion of the turbine building that contains the auxiliary feedwater pumps.

Additional reviews were performed to exclude specific nonsafety-related components where design features, such as panels or enclosures, would protect safety-related equipment from to AEP NRC:4034-15 Page 4 leakage or spray. Some typical components and portions of systems that were excluded from the scope 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 that protect safety-related equipment from leakage or spray are ventilation equipment housings and sample panels that contain the fluid-filled nonsafety-related components. These design features prevent failures of the nonsafety-related equipment from spraying or leaking onto safety-related equipment in the area. These enclosures are within the scope of license renewal and subject to aging 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 Nuclear Plant, Units I and 2, Docket Nos. 50-315 and 50-316, License Renewal Application - Response to Requests for Additional Information on Scoping and Screening Results (TAC Nos. MC1202 and 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 the non-EQ [environmental qualificationI cable AMqP does not provide an adequate description of the program as required by 1 CFR 54.21(d). The description of FSAR supplement for aging management of electrical and instrumentation and controls system should be consistent with Table 3.6-2 of NUREG-1800. Please submit a revised FSAR supplement that is consistent with NUREG-1800 to satisfy 10 CFR 54.21(d).

Clarification Requested by the Staff:

Although not explicitly stated in the RAI, the staff indicated that the intent of the RAI was to have all three Non-EQ programs provide an adequate description in the revised FSAR supplement in accordance with Table 3.6-2 of NUREG-1800.

The applicant acknowledged the question's intent and agreed to revise the response to include all three Non-EQ programs in the FSAR supplement.

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, the program descriptions for Updated Final Safety Analysis Report Sections A.2.1.23 and A.2.1.24 are revised as follows:

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

to AEP NRC:4034-15 Page 5 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 license renewal that are exposed to significant moisture simultaneously with significant voltage. 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 as being subjected to system voltage for more than 25 percent of the time. Under this program, in-scope medium-voltage cables that are exposed to significant moisture and significant voltage will be tested at least once every 10 years to provide an indication of the condition of the conductor insulation. The specific type of test performed will be determined prior to the initial test, and itiill be based on technology that is state-of-the-art at the time the test is performed. The Non-EQ Inaccessible Medium-Voltage Cable Program will be implemented prior to the period of extended operation.

A.2.1.24 NON-EQ INSTRUMENTATION CIRCUITS TEST REVIEW PROGRAM The Non-EQ Instrumentation Circuits Test Review Program will manage aging effects 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-level signals, such as radiation monitoring and nuclear instrumentation, which are exposed to adverse localized environments caused by heat, radiation, or moisture.

An adverse localized environment is defined as being significantly more severe than the specified service environment for the cable.

This program will detect aging effects by reviewing calibration or surveillance results for components within the program scope at afrequency not to exceed 10 years or as part of corrective actions then acceptance criteria are exceeded at the normal calibration frequency. The Non-EQ Instrumentation Circuits Test Review Program will be implemented prior to the period of extended operation.

to AEP NRC:4034-15 Page 6 RAI 4.4-1:

The environmental qualification [EQ] of electrical equipmnent resuhlts in Section 4.4 indicate that the 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 to extend the qualification life of electrical equipment identified in the TLAA.

The important attributes of a re-analysis include analytical methods, data collection and reduction methods, underlying assumptions, acceptance criteria and corrective actions. Provide information on the important attributes for reanalysis of an aging evaluation of electrical equipment identified in the TLAA to extend the qualification under IO CFR 50.49(e).

Clarification Requested by the Staff:

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 previous applicants simply "cut and pasted" the re-analysis attributes fromn GALL [Generic Aging Lessons 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 also indicated that the EQ Program is available onsite, and av'ailable for JVRC audit, if so desired This 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 of GALL 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 Components Program is an existing program that was established to meet commitments associated with IO CFR 50.49.

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

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

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

to AEP NRC:4034-15 Page 7 Program Description The CNP Environmental Qualification of Electric Components Program manages component thermal, radiation, and cyclical aging through the use of aging evaluations based on 10 CFR 50.49(f) qualification methods.

As required by 10 CFR 50.49, EQ components not qualified for the current license term are to be refurbished, replaced, or have their qualification extended prior to reaching the aging limits established in the evaluation. Aging evaluations for EQ components that specify a qualification of at least 40 years are considered TLAAs for license renewal.

Aging Management Program Elements Scope The CNP Environmental Qualification of Electric Components Program provides the requirements for the environmental qualification of electrical equipment important to safety as required 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 Qualification of 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 effects within 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 related surveillance and maintenance requirements for EQ equipment, but does not require condition or performance monitoring.

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

to AEP NRC:4034-15 Page 8 Detection of Aging Effects The detection of aging effects is not required for compliance with 10 CFR 50.49 at CNP.

Monitoring or inspection of certain environmental conditions or component parameters may be used to ensure that the component is within the bounds of its qualification basis, or as a means to modify 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 require monitoring and trending of EQ equipment. The program provides surveillance and maintenance requirements 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 with NUREG-1801.

Acceptance Criteria The CNP Environmental Qualification of Electric Components Program provides surveillance and maintenance activities to assure that the acceptance criteria for EQ equipment is maintained within its qualification basis and qualified life. The program provides that any EQ equipment shall be replaced, refurbished, or requalified prior to exceeding the qualified life of the equipment.

If monitoring is used to modify a component qualified life, appropriate plant-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, unexpected adverse conditions are identified during operational or maintenance activities that affect the environment of a qualified component, or an emerging industry aging issue is identified that affects the qualification of an EQ component, the affected EQ component is evaluated and appropriate corrective actions are taken, which may include changes to the qualification bases and conclusions. The adverse condition can be identified in a condition report (CR), industry notification such as NRC Information Notice, NRC Generic Letter, 10 CFR Part 21 Notice or design change.

The requirements of 10 CFR Part 50, Appendix B, are applied through implementation of the CNP Corrective Action Program.

to AEP NRC:4034-15 Page 9 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 Components Program includes CRs, NRC Inspection Reports, and documentation of the results of internal program assessments.

The operating experience discussion in NUREG-1801,Section X.EI, states, "EQ programs include consideration of operating experience to modify qualification bases and conclusions, including qualified life.

Compliance with 10 CFR 50.49 provides reasonable assurance that components can perform their intended functions during accident conditions after experiencing the effects of inservice aging."

The I&M report titled "Engineering Functional Area Attribute Restart Readiness, Affirmation Report for Environmental Qualification (EQ) Program," dated December 22, 1999, was reviewed. This report concludes "...there is reasonable assurance that this program will function satisfactorily to support restart and continuous operation of CNP."

Therefore based on the operating experience discussion of NUREG-1801, the CNP program provides reasonable assurance that components can perform their intended functions during accident conditions after experiencing the effects of inservice aging.

For further confirmation of the program effectiveness, CRs identified in the program procedure revisions and restart assessment report were reviewed. The vast majority address administrative issues associated with the program, records, and auditability. Program revisions have been made as part of corrective actions for these CRs. The only equipment failures identified were failures of flood-up tubes, which were determined to have been related to installation techniques and not aging. No CRs identified aging effects for which the program is intended to prevent. This operating experience is consistent with the conclusion that the program is effective in preventing the effects of aging.

A CR initiated in 1998 identified programmatic problems with the Environmental Qualification of Electric Components Program.

The root cause analysis for this CR determined that the underlying causes were programmatic issues, including lack of management support/attention.

The resultant corrective actions included numerous action plans and programmatic changes to to AEP NRC:4034-15 Page I10 correct these deficiencies.

The result of this operating experience was a much stronger Environmental Qualification of Electric Components Program.

The Environmental Qualification of Electric Components Program has been effective at managing aging effects.

Operating experience has identified no aging effects for which the program is intended to prevent. The program is continuing to be improved as a result of ongoing program assessments.

The continued implementation of this program provides reasonable assurance that the aging effects will be managed so that the applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.

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

The overall effectiveness of the Environmental Qualification of Electric Components Program is demonstrated by the excellent operating experience for systems and components in the program.

The program has been subject to periodic internal and external assessments that facilitate continuous improvement.

Continued implementation of this program provides reasonable assurance that components within the scope of license renewal will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.

RAI 4.7.4-1:

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

I&M Response to RAI 4.7.4-1:

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 years of operation. For CNP, the types and numbers of RCS design transients, with the exception of the feedwater cycling at hot shutdown, were verified to be bounded by the design transients assumed in WCAP-15338-A, thereby satisfying Renewal Applicant Action Item (1) of the Revised Safety Evaluation Report of WCAP-15338, dated September 25, 2002 [Accession No. ML022690375]. The feedwater cycling at hot shutdown transient is associated with a feedwater nozzle cracking concern and is not monitored at CNP due to design and operating modifications

- to AEP NRC:4034-15 Page 11 to preclude feedwater nozzle cracking. This transient is not anticipated to have a significant impact on crack growth beneath the reactor vessel cladding. CNP's projected number of RCS design transients for 60 years, as shown in LRA Table 4.3-1, does not exceed applicable design assumptions assumed in WCAP-15338-A.

Therefore, the WCAP-15338-A RCS transients bound the CNP RCS transients, and WCAP-15338-A remains applicable to CNP for the period of extended operation.