ML041670523

License Renewal Application - Response to Requests for Additional Information on Structures and Component Supports

ML041670523
Person / Time
Site:	Cook
Issue date:	06/08/2004
From:	Nazar M Indiana Michigan Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP:NRC:4034-07, TAC MC1202, TAC MC1203
Download: ML041670523 (15)

Text

Indiana Michigan Power Company 500 Circle Drive Buchanan, Ml 49107 1373 z

INDIANA MICHIGAN POWER June 8, 2004 AEP:NRC:4034-07 10 CFR 54 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop O-P1-17 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 on Structures and Component Supports (TAC Nos. MC 1202 and MC 1203)

REFERENCE:

1. Letter from J. Rowley, Nuclear Regulatory Commission (NRC), to M. K. Nazar, Indiana Michigan Power Company (I&M), "Request for Additional Information for the Review of the Donald C. Cook Nuclear Plant, Unit I and 2 License Renewal Application," dated May 7, 2004.

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.

Dear Sir or Madam:

By letter dated October 31, 2003, I&M submitted an application to renew the operating licenses for Donald C. Cook Nuclear Plant, Units I and 2.

During the conduct of its review, the NRC Mechanical and Civil Engineering Branch Staff identified areas where additional information was needed to complete its review of the license renewal application (LRA).

This letter responds to the LRA Section 3.5 requests for additional information (RAls),

which were documented in an NRC letter dated May 7, 2004 (Reference 1).

RAIs in Reference I pertaining to LRA Section 2.4 were addressed in an I&M letter dated May 7, 2004 (Reference 2).

,qjA 0

U. S. Nuclear Regulatory Commission Page 2 AEP:NRC:4034-07 The enclosure to this letter provides an affirmation pertaining to the statements made in this letter.

The attachment to this letter provides the additional information requested from the NRC Staff. 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, M. K. Nazar Senior Vice Prey NH/rdw

/

Enclosure:

Afl lent and Chief Nuclear Officer irmation

Attachment:

Response to Requests for Additional Information for the Donald C. Cook Nuclear Plant License Renewal Application -

Structures and Component Supports Aging Management Review Results c:

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

Enclosure to AEP:NRC:4034-07 AFFIRMATION I, Mano K. Nazar, being duly sworn, state that I am Senior Vice President and Chief Nuclear Officer of American Electric Power Service Corporation and 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.

American Electric Power Service Corporation M. K. Nazar Senior Vice P sident and Chief Nuclear Officer SWORN TO AND SUBSCRIBED BEFORE ME THIS 6 DAY OF 2004 v - a \\-Ntar.tYPublic My Commission Expires

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Attachment to AEP:NRC:4034-07 Page I Response to Requests for Additional Information for the Donald C. Cook Nuclear Plant License Renewal Application Structures and Component Supports Aging Management Review Results This attachment provides Indiana Michigan Power Company's (I&M's) responses to the Donald C. Cook Nuclear Plant (CNP) License Renewal Application (LRA) Requests for Additional Information (RAIs) provided in a Nuclear Regulatory Commission (NRC) letter dated May 7, 2004 (Reference 1). The RAIs addressed in this attachment pertain to LRA Section 3.5 -

Aging Management Review Results: Structures and Component Supports. RAIs in Reference 1 pertaining to LRA Section 2.4 were addressed in an I&M letter dated May 7, 2004 (Reference 2).

References

1. Letter from J. Rowley, NRC, to M. K. Nazar, I&M, "Request for Additional Information for the Review of the Donald C. Cook Nuclear Plant, Unit 1 and 2 License Renewal Application," dated May 7, 2004.

2. Letter from M. K. Nazar, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant, Units 1 and 2, Docket Nos. 50-315 and 50-316, License Renewal Application -

Response to Requests for Additional Information on Scoping and Screening Results (TAC Nos. MC1202 and MC1203)," AEP:NRC:4034-01, dated May 7, 2004.

RAI 3.5-1:

In line item 3.5.1-3 of Table 3.5.1 of the LRA, the applicant indicates that the aging effects related to loss of material due to corrosion of bellows, and dissimilar metal wields are managed consistent with NUREG-1801.

NUREG-1801 recommends the examination of penetration bellows and the associated dissimilar welds based on the operating experience with the stress corrosion cracking of bellows as documented in NRC Information Notice 92-20. The applicant is requested to provide the following information related to the examination and/or testing of containment penetration bellows:

How many penetration bellows are in the Cook containments? Please summarize the operating experience related to the examination of these bellows. If provisions are made to assess their leaktightness (as they are not accessible for visual examination), please provide a summary of these provisions (including frequency of tests), and indicate if such leaktightness assessment of the bellows is part of the LRA AMP [aging management program] B.1.15 or B.1.8.

Attachment to AEP:NRC:4034-07 Page 2 I&M Response to RAI 3.5-1:

As stated in Updated Final Safety Analysis Report (UFSAR) Section 1.4.7, Criterion 56. and Section 5.2.4.2, Fuel Transfer Penetration, penetration expansion bellows are not used to maintain containment integrity and do not serve as part of the containment pressure boundary.

Therefore, CNP expansion bellows do not serve an intended function, and are not subject to aging management review. No provisions are made to assess their leaktightness.

RAI 3.5-2:

For seals and gaskets related to containment penetrations, in Item Number 3.5.1-6 of the LRA, containment ISI [inservice inspection] and containment leak rate testing have been stated as the aging management programs. For equipment hatches and air-locks at Cook, the staff agrees with the applicant's assertion that the leak rate testing program wvill monitor aging degradation of their seals and gaskets, as they are leak rate tested after each closing. For other penetrations with seals and gaskets, the applicant is requested to provide information regarding the adequacy of Type B leak rate testingfrequency to monitor aging degradation of seals and gaskets at Cook I&M Response to RAI 3.5-2:

The line item for "Air lock seals" in LRA Table 3.5.2-1 (Page 3.5-39) is the only line item for seals or gaskets in which I&M credits the Containment Leakage Rate Testing Program. The air lock seals line item is also the only line item that refers to Item 3.5.1-6, "Seals, gaskets, and moisture barriers," of LRA Table 3.5.1.

RAI 3.5-3:

In the discussion of Item 3.5.1-12 in Section 3.5.2.2.1.4, the applicant notes that the moisture barrier is monitored under IJWE for aging degradation. The industry experience indicates that the moisture barrier degrades ivith time, and any moisture accumulation in the degraded barrier corrodes the steel liner.

The applicant is requested to provide information regarding the operating experience related to the degradation of moisture barrier and the containment liner plate at Cook Please include a discussion of acceptable liner plate corrosion before it is reinstated to the nominal thickness.

I&M Response to RAI 3.5-3:

In the past, instances of containment liner degradation in the vicinity of the moisture barrier have been identified on both CNP units. However, in only one case was the minimum containment liner thickness found to be less than the acceptance criterion (0.250 inches). In March 1998, an inspection of the Unit I containment liner plate identified aging degradation of moisture barrier

Attachment to AEP:NRC:4034-07 Page 3 seal due to poor maintenance and the consequent pitting. The inspection reported the thickness of the steel containment liner to be less than 0.250 inches.

The cause of the pitting was determined to have been inadequate installation practices at the time of original construction and a lack of proper maintenance of the seal located between the concrete floor slab and the steel liner. As documented in Unit 1 LER 98-011-02, an analysis of this event determined that the identified steel containment liner pitting would be of no safety significance, as the leak-tight integrity of the containment would not be impaired and the as-found liner will continue to fulfill its function as an effective leaktight membrane. The existing seal was removed and the surface on the containment liner plate prepared and coated with new seals applied. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, Subsection IWE, provides the requirements for inservice inspection of containment structures. The requirements include examination, evaluation, repair, and replacement of concrete containment liner plate in accordance with 10 CFR 50.55a. The acceptance criteria for CNP liner plate are in accordance with ASME Boiler and Pressure Vessel Code,Section XI, Subsection IWE.

RAI 3.5-4:

In addressing item 3.5.1-27 in table 3.5.2-1, for the reinforced concrete structures subjected to elevated temperatures and. high humidity (e.g., primary shield walls, pressurizer and steam generator enclosures, reactor vessel supports) the environment column should read "elevated temperature." For these structures, the applicant is requested to provide the following information:

(a) The method(s) of monitoring temperatures within the primary shield wvall concrete, and around the reactor vessel, and in the reactor cavity.

(b) If the primary shield wvall concrete (or any other structure wvithin Cook containment is kept below the threshold temperature (i.e., 150F) by means of air cooling, provide the operating experience related to the performance of the cooling system.

(c) The results of the latest inspection of these structures, in terms of cracking, spalling, and condition of reactor vessel support structures, etc.

I&M Response to RAI 3.5-4:

More appropriate pressurizer enclosure component type and environment entries listed in LRA Table 3.5.2-1 on Page 3.5-38 are "Pressurizer enclosure (Unit 1)" and "Protected from weather with elevated temperature," respectively.

The environment is defined in the LRA Table 3.0-2 on Page 3.0-10.

(a) The primary containment upper and lower compartment average temperature is monitored at various containment building elevations in accordance with applicable technical

Attachment to AEP:NRC:4034-07 Page 4 specifications. However, the temperatures within the primary shield wall concrete, around the reactor vessel, and in the reactor cavity are not directly monitored.

(b) The containment ventilation system is designed to maintain a maximum temperature of 1000F in the containment upper compartment during plant operation, and a maximum temperature of 120'F in the lower compartment (1350F inside the primary concrete shield) during plant operation. A search of the corrective action program database for condition reports (CRs) generated over the past five years discovered two CRs that relate to the performance of containment cooling ventilation systems. Both of these CRs discuss repair of back draft dampers.

• A Unit I pressurizer enclosure ventilation exhaust fan had significantly lower output than its alternate train fan.

A job order adjusted the back draft damper to correct fan performance.

• A pressurizer enclosure ventilation exhaust fan discharge backdraft damper was found to be stuck open with broken linkage. Ajob order replaced the back draft damper.

Neither CR indicated that pressurizer compartment temperatures had exceeded 1 50'F.

(c) The results of the latest inspections of the structures are provided below.

• Concrete examinations November 1999 The concrete surfaces below the nozzles of the reactor vessel in CNP Unit 2 were examined on November 12, 1999. The methods of examination were visual observation and sounding of the concrete surfaces with a hammer.

The scope of the examination was the horizontal concrete surfaces located at elevation 609'-2" below the Number 2 inlet, and the Number ] outlet nozzles.

Steel reactor supports are located at each nozzle. The junction of the two flanges between the upper and lower portion of the support is approximately 6" above the concrete surfaces observed.

In each location, there was a thin crack parallel to the steel support, approximately 8" to 10" away from the outside of the support. A similar width crack, normal to the support centerline, extended from the center stiffener plate to the parallel crack. The cracks observed were between 1/32" and 1/8" in width at the surface and the width tapered to hairline width within 1/16" of the surface. Solid ringing sounds were produced in response to hammer strikes on all of the accessible surface area. There was no evidence of moisture movement through the cracks or any structural distress or material deterioration of the concrete.

,1i I Attachment to AEP:NRC:4034-07 Page 5 The concrete in question was placed in four 450 sectors, at each support location, between elevations 607'-7" and 609'-2", after the supports were installed. It appeared that the concrete did not support mechanical loads and that its primary function is to fill the space left in the pedestal for installation of the reactor supports. The observed cracks were probably caused by drying shrinkage and were not considered unusual. It did not appear that there has been significant moisture movement through the cracks and it is doubtful that serious corrosion has resulted in the embedded portion of the support. Therefore, no remedial action was considered necessary for the concrete examined.

Concrete examinations April 2003 and November 2003 A summary of the inspection results of the Unit I and Unit 2 biological shield walls in the Containment Buildings is provided below. These inspections were performed during the Ul Cl9 (November 2003) and U2C14 (April 2003) refueling outages.

Starting at elevation 598' (which is about 12 feet below the vessel supports) and going upward, both wall exterior faces are in very good condition. The walls exhibited a few, limited, scattered, hairline cracks. The longest hairline cracks were no more than ten feet long. Almost all of the observed cracks were between elevation 598' and about 608' (between the floor and a horizontal construction joint). There was very little indication of efflorescence (calcium deposits) in these cracks.

There were no indications of any spalling. The condition of the *walls (very limited cracking) can be attributed to their extremely robust design. The walls at this location are approximately seven feet thick, with several layers of reinforcing steel bars.

RAI 3.5-5:

Section 3.5.2.2.2.2, "Aging Management of Inaccessible Areas, " of the LRA (Page 3.5-13) states that inspection of accessible concrete have not revealed degradation related to corrosion of embedded steel. The Cook belowv-grade environment is not aggressive. Therefore, corrosion of embedded steel is not an applicable aging mechanism for Cook concrete. The staff agrees with this statement only for the case of uncracked reinforced concrete elements.

Howstever, the embedded structuralfoundations may crack due to settlement and corrosion of reinforcing steel may be expected. The applicant is requested to provide additional information to justify the validity of the LRA statement.

I&M Response to RAI 3.5-5:

As discussed in LRA Section 3.5.2.2.1.2, cracking due to settlement is not applicable to CNP concrete structures. Settlement was monitored at CNP until discontinued after confirmation that significant settlement was not occurring.

In addition, NUREG-1 800, Section 3.5.2.2.2.2, recommends further evaluation to manage aging effects only if specific criteria defined in

Attachment to AEP:NRC:4034-07 Page 6 NUREG-1801 cannot be met. Those criteria are the criteria for an aggressive environment as stated in NUREG-1801, Volume 2, Item III.Al.1-e. Since the CNP below-grade environment is not aggressive, corrosion of embedded steel is not an applicable aging effect/mechanism RAI 3.5-6:

Section B.].4, "Boric Acid Corrosion Prevention, " of the LRA (Page B-26) states that, the Boric Acid Corrosion Prevention Program is an existing Cook program which is comparable to the program Section XI.Ml 0. This existing program vill be revised to include electrical components in addition to ferritic steel. As stated in the GALL (Section XI.MJO, "Boric Acid Corrosion"),

the program scope covers any carbon steel and lowv-alloy structures or components, and electrical components, on uvhich borated reactor water may leak. It is the staffs Understanding, based on the conversation betveen NRC project manager and the applicant, that this program also covers structures and structural components related to or adjacent to the boronic injection system (portions of this system are located in the auxiliary building). Howvever, some structural components, such as anchor bolts (includes sivitchyard structures and tank anchors), (LRA, Page 3.5-62) are obviously not located in the containment nor in the auxiliary building.

Therefore, the applicant is requested to:

(a) Clearly state that the scope of the boric acid corrosion program will cover these structural components, and (b) Clarify that are there any other structures and structural components not located in the containment nor auxiliary building and to be covered by this AMP.

I&M Response to RAI 3.5-6:

(a)

The Boric Acid Corrosion Prevention Program is implemented by a plant procedure. The procedure is applicable where the potential exists to degrade ferritic steel components within the reactor coolant pressure boundary and other plant systems due to contact with borated water.

The scope of the Boric Acid Corrosion Prevention Program includes structures and structural components potentially exposed to borated water leakage, whether they are included in the containment, auxiliary building, or other locations on the plant site (including yard structures).

Switchyard structures are not exposed to borated water leakage, but some tank anchors are, as discussed in response to sub-part (b), below.

(b)

The Boric Acid Corrosion Prevention Program listed for anchor bolts in LRA Table 3.5.2-5 applies only to anchor bolts with the potential for exposure to borated water leakage.

Structural components located outside the containment and auxiliary buildings that are covered by the Boric Acid Prevention Corrosion Program are the refueling water storage tank anchorage in the yard.

Attachment to AEP:NRC:4034-07 Page 7 RAI 3.5-7:

As described in Table 3.5.2-3 and B.1.32 of the LRA, the Structure Monitoring Program (SMP) is consistent with GALL and is to be usedfor the aging management of wtater control structures, but the applicant has not compared the SMP wvith the GALL RG 1.12 7 Program, as specified in the GALL.

With the concern stated above, the applicant is requested to provide a comparison of the SMP with the GALL RG 1.127 Program and demonstrate that the SMP is suitable for managing the aging effects of water control structures.

I&M Response to RAI 3.5-7:

Regulatory Guide (RG) 1.127, Inspection of Water-Control Structures Associated with Nuclear Powver Plants, is identified as an acceptable basis for developing an inservice inspection and surveillance program for water-control structures in NUREG-1801,Section XI.S7. For plants not committed to RG 1.127, such as CNP, managing aging effects associated with structures and structural components may be included in the Structures Monitoring Program. The major water control structures at CNP are the screenhouse (intake structure) and the roadway wvest of the screenhouse. CNP uses the Structures Monitoring Program described in LRA Section B. 1.32 to mange the effects of aging on water control structures. The Structures Monitoring Program, with enhancements, is consistent with NUREG-1 801,Section XI.S6.

The attributes that are in NUREG-1 801,Section XI.S7, aging management program, but not in the CNP Structures Monitoring Program, are attributes dealing with earthen embankments water control structures. NUREG-1801,Section XI.S7, refers to RG 1.127 which proposes inspection parameters, including settlement, depressions, sink holes, slope stability (e.g., irregularities in alignment and variances from originally constructed slopes), seepage, proper functioning of drainage systems, and degradation of slope protection features, and frequency (not to exceed five years) for earthen embankment water control structures. During the CNP aging management review, the aging effects requiring management for earthen structures (roadway) were determined to be loss of material, loss of form and change in material properties. As indicated in LRA Section B.1.32, the Structures Monitoring Program will be enhanced to include visual inspections to manage aging effects for the roadway xvest of the screenhouse.

The visual inspections will detect degradation of the roadway due to the identified aging effects.

The remaining water control structure (screenhouse) is similar to other CNP structures that are addressed by the Structures Monitoring Program. The Structures Monitoring Program will be effective in managing the effects of aging for water control structures.

Attachment to AEP:NRC:4034-07 Page 8 A comparison of the CNP Structures Monitoring Program with the NUREG-1801,Section XI.S7, RG 1.127 Program for water control structures is provided in the following paragraphs.

Aging Management Program Elements

1.

Scope of Program

a.

NUREG-1801 Section XI.S7 Scope "RG 1.127 applies to water-control structures associated with emergency cooling water systems or flood protection of nuclear power plants.

The water-control structures included in the RG 1.127 program are concrete structures; embankment structures; spillway structures and outlet works; reservoirs; cooling water channels and canals, and intake and discharge structures; and safety and performance instrumentation."

b.

Comparison Water-control structures are included in CNP Structures Monitoring Program.

The Structures Monitoring Program scope includes the screenhouse and, as indicated in LRA Section B.1.32, the program will be enhanced to include the roadway west of the screenhouse.

2.

Preventive Action

a.

NUREG-1 801 Section XI.S7. Preventive Action "No preventive actions are specified; RG 1.127 is a monitoring program."

b.

Comparison No preventive actions are included in the Structures Monitoring Program and none are required to address water control structures. The CNP preventive actions are consistent with NUREG-1 801,Section XI.S7.

3.

Parameters Monitored or Inspected

a.

NUREG-1801 Section XI.S7 Parameters Monitored or Inspected "RG 1.127 identifies the parameters to be monitored and inspected for water-control structures. The parameters vary depending on the particular structure. Parameters to be monitored and inspected for concrete structures include cracking, movements (e.g., settlement, heaving, deflection), conditions at junctions with abutments and embankments, erosion, cavitation, seepage, and leakage. Parameters to be monitored and inspected for earthen embankment structures include settlement, depressions, sink holes, slope stability (e.g., irregularities in alignment and variances from originally constructed slopes), seepage, proper functioning of drainage systems, and degradation of slope

Attachment to AEP:NRC:4034-07 Page 9 protection features. Further details of parameters to be monitored and inspected for these and other water-control structures are specified in Section C.2 of RG 1.127."

b.

Comparison For concrete water-control structures at CNP, the specific parameters monitored or inspected were selected to ensure that aging degradation leading to loss of intended functions is detected and the extent of degradation is determined.

The parameters monitored or inspected (such as cracking, settlement, leakage, and water infiltration) were selected considering industry codes, standards and guidelines, and also consider industry and plant-specific operating experience.

Settlement and erosion of porous concrete subfoundations are not problems at CNP, so a site de-watering system is not necessary.

Consistent with NUREG-1 801,Section XI.S6, the Structures Monitoring Program adequately addresses concrete and steel structure parameters to 'be monitored that are applicable to water-control structures.

The Structures Monitoring Program will be enhanced to include parameters monitored for earthen structures (roadway west of the screenhouse).

In accordance with NUREG-1801,Section XI.S6, parameters to be monitored and inspected for earthen embankment structures will consider industry codes, standards and guidelines, RG 1.127, and also consider industry and plant-specific operating experience.

4.

Detection of Aging Effects

a.

NUREG-1 801. Section XL.S7. Detection of Aging Effects "Visual inspections are primarily used to detect degradation of water-control structures.

In some cases, instruments have been installed to measure the behavior of water-control structures.

RG 1.127 indicates that the available records and readings of installed instruments are to be reviewed to detect any unusual performance or distress that may be indicative of degradation. RG 1.127 describes periodic inspections, to be performed at least once every five years. Similar intervals of five years are specified in ACI 349.3R for inspection of structures continually exposed to fluids or retaining fluids.

Such intervals have been shown to be adequate to detect degradation of water-control structures before they have a significant effect on plant safety. RG 1.127 also describes special inspections immediately following the occurrence of significant natural phenomena, such as large floods, earthquakes, hurricanes, tornadoes, and intense local rainfalls."

b.

Comparison As indicated in LRA Section B. 1.32, the Structures Monitoring Program will be enhanced to include detection of aging effects for the roadway west of the screenhouse.

This enhancement will include inspecting the roadway for degradation or damage following a significant natural phenomenon, such as large floods, earthquakes, hurricanes, tornadoes, and intense local rainfalls.

When enhanced, the examination

Attachment to AEP:NRC:4034-07 Page 1 0 criteria for the roadway will detect degradation of the roadway due to weather-related damage.

The screenhouse is designed as a seismic Class I structure providing protection to safety-related equipment from seismic events, tornado-velocity wind effects, tornado-borne missiles and flood conditions anticipated due to a seiche or surge phenomenon.

Consistent with NUREG-1 801,Section XI.S6, the detection of aging effects for water control structures concrete and steel elements is adequately addressed in the Structures Monitoring Program.

Inspection methods, inspection schedule, and inspector qualifications are commensurate with industry codes, standards, guidelines, and also consider industry and plant-specific operating experience.

Special inspections are performed following the occurrence of significant natural phenomena, such as earthquakes, floods, seiche, severe weather, and fires.

5.

Monitoring and Trending

a.

NUREG-1801. Section XL.S7. Monitoring and Trending "Water-control structures are monitored by periodic inspection as described in RG 1.127.

In addition to monitoring the aging effects identified in Attribute (3) above, inspections also monitor the adequacy and quality of maintenance and operating procedures.

RG 1.127 does not discuss trending."

b.

Comparison Consistent with NUREG-1801, Section XL.S6, CNP structures are monitored in accordance with 10 CFR 50.65. This approach is adequate for managing aging effects associated with water control structures included in the Structures Monitoring Program.

6.

Acceptance Criteria

a.

NUREG-1801. Section XL.S7. RG 1.127. Acceptance Criteria "Acceptance criteria to evaluate the need for corrective actions are not specified in RG 1.127. However, the 'Evaluation Criteria' provided in Chapter 5 of ACI 349.3R-96 provides acceptance criteria (including quantitative criteria) for determining the adequacy of observed aging effects and specifies criteria for further evaluation.

Although not required, plant-specific acceptance criteria based on Chapter 5 of ACI 349.3R-96 are acceptable.

Acceptance criteria for earthen structures such as dams, canals, and embankments are to be consistent with programs falling within the regulatory jurisdiction of the Federal Energy Regulatory Commission (FERC) or the.U.S. Army Corps of Engineers."

b.

Comparison The Structures Monitoring Program acceptance criteria are consistent with NUREG-1801, Section XL.S6, for concrete and steel components of water control

Attachment to AEP:NRC:4034-07 Page I I structures. Including the roadway west of the screenhouse in the Structures Monitoring Program wvill result in developing acceptance criteria for visual inspections.

When enhanced, acceptance criteria will be selected to ensure that the need for corrective actions will be identified before loss of intended functions. Acceptance criteria will be commensurate with industry codes, standards and guidelines, and will also consider industry and plant-specific operating experience.

7, 8, 9. Corrective Actions, Confirmation Process, and Administrative Controls CNP applies the requirements of 10 CFR Part 50, Appendix B, to the Structures Monitoring Program through the use of the Corrective Action Program. The Structures Monitoring Program Corrective Actions, Confirmation Process, and Administrative Controls attributes are applicable to water-control structures without enhancement.

10.

Operating Experience Operating experience discussed in LRA Section B. 1.32 applies to water-control structures with the exception of the roadway west of the screenhouse. Operating experience with the roadway was not discussed in LRA Section B.1.32 since the Structures Monitoring Program does not yet include the roadway; however, the operating experience review identified no significant degradation of the roadway.

The Structures Monitoring Program will effectively manage the aging effects requiring management for water-control structures at CNP with enhancements identified in LRA Section B.1.32.

In summary, with enhancements, the Structures Monitoring Program wvill effectively manage the aging effects requiring management for water control structures at CNP.

RAI 3.5-8:

Section 3.5 of the LRA states that at Cook, the concrete is not exposed to flowving water and the belowv-grade environment is not aggressive (pH is greater than 5.5, chlorides is less than 50ppm, and sulfates is less than 1,50ppm). Therefore, the LRA concludes that increase in porosity and permeability and loss of strength of due to leaching of calcium hydroxide are not applicable aging effects for Cook concrete structures. However, the applicant did not commit, in the SMP to periodically monitor the ground water chemistry, as specified in the GALL.

The applicant is requested to:

(a) Either augment its SMP to include the monitoring program and to ensure that the ground water will continuously be non-aggressive, or

Attachment to AEP:NRC:4034-07 Page 12 (b) Provide technical basis and justify that there is no need to continuously monitor the ground wvater chemistry.

I&M Response to RAI 3.5-8:

Sample data tabulated below indicates the limiting chemistry parameters have shown no significant increase and are still far below established limits. Because existing data show no significant change over a period of approximately 25 years, groundwater chemistry is not anticipated to significantly change in the future. Therefore, periodic monitoring of groundwater chemistry is not required to assure the non-aggressiveness of the below-grade environment.

Sample Sample Date Sample well IA Sample well 12 pH 3/4/1976 6.4 7.8 1/15/2002 7.1 7.4 Chloride (ppm) 3/4/1976 20.3 9.7 1/15/2002 10 12 Sulfate (ppm) 3/4/1976 18.1 310.3 1/15/2002 134 67