AEP-NRC-2012-54, Application to Revise Technical Specifications to Adopt Technical Specification Task Force (TSTF)-510, Revision 2, Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection

Application to Revise Technical Specifications to Adopt Technical Specification Task Force (TSTF)-510, Revision 2, Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection

ML12268A047
Person / Time
Site:	Cook
Issue date:	09/12/2012
From:	Gebbie J Indiana Michigan Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP-NRC-2012-54 TSTF-510, Revision 2
Download: ML12268A047 (30)

Text

INDIANA MICHIGAN Indiana Michigan Power P

V'ffR One Cook Place Bridgman, MI 49106 A unit ofAmerican Electric Power IndianaMichiganPower.com September 12, 2012 AEP-NRC-2012-54 10 CFR 50.90 U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC. 20555-0001 Docket Nos.: 50-315 50-316 Donald C. Cook Nuclear Plant, Units 1 and 2 APPLICATION TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT TECHNICAL SPECIFICATION TASK FORCE (TSTF)-510, REVISION 2, "REVISION TO STEAM GENERATOR PROGRAM INSPECTION FREQUENCIES AND TUBE SAMPLE SELECTION," USING THE CONSOLIDATED LINE ITEM IMPROVEMENT PROCESS Pursuant to 10 CFR 50.90, Indiana Michigan Power Company (I&M), is submitting a request for an amendment to the Technical Specifications (TS) for Donald C. Cook Nuclear Plant (CNP), Units 1 and 2.

The proposed amendment would modify TS requirements regarding steam generator tube inspections and reporting as described in TSTF-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection." provides an affirmation. Enclosure 2 provides a description and assessment of the proposed changes, the requested confirmation of applicability, and plant-specific verifications. provides the existing Unit 1 TS pages marked up to show the proposed changes. provides the existing Unit 2 TS pages marked up to show the proposed changes. provides existing Unit 1 TS Bases pages marked up to show the proposed changes. provides existing Unit 2 TS Bases pages marked up to show the proposed changes.

Revised (clean) TS pages with proposed changes incorporated will be provided to the Nuclear Regulatory Commission (NRC) Licensing Project Manager when requested. The TS Bases are provided for information only. TS Bases changes will be processed in accordance with the CNP TS Bases Control Program.

I&M requests approval of the proposed license amendment in accordance with the normal NRC review schedule for such changes, with the amendment being implemented within 180 days of NRC approval.

There are no new regulatory commitments made in this letter.

1Aoo

U. S. Nuclear Regulatory Commission AEP-NRC-2012-54 Page 2 In accordance with 10 CFR 50.91, a copy of this application, with enclosures and attachments, is being provided to the designated Michigan state officials. Should you have any questions, please contact Mr. Michael K. Scarpello, Regulatory Affairs Manager, at (269) 466-2649.

Sincerely, Joel P. Gebbie Site Vice President DMB/jmr c:

C. A. Casto, NRC Region III J. T. King - MPSC S. M. Krawec, AEP Ft. Wayne, w/o attachments MDEQ - RMD/RPS NRC Resident Inspector T. J. Wengert - NRC Washington DC

Enclosures:

1. Affirmation

2. Description and Assessment

3. CNP Unit 1 TS Pages Marked To Show Proposed Changes

4. CNP Unit 2 TS Pages Marked To Show Proposed Changes

5. CNP Unit 1 TS Bases Pages Marked To Show Proposed Changes [Information Only]

6. CNP Unit 2 TS Bases Pages Marked To Show Proposed Changes [Information Only]

Enclosurel to AEP-NRC-2012-54 AFFIRMATION I, Joel P. Gebbie, 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 Joel P. Gebbie Site Vice President SWORN TO AND SUBSCRIBED BEFORE ME THIS Q DAY OF

,2012 My Commission Expires £'M -*-'-

ŽA <

DANIELLE BURGOYNE Notary Public, State of Michigan County of Berrien My Commission Expires 0-44-2018 Acting in the C

°untys0T-, i-,- \\

to AEP-NRC-2012-54 Page 1 DESCRIPTION AND ASSESSMENT

1.0 DESCRIPTION

The proposed change revises Specification 5.5.9, "Steam Generator (SG) Program" and 5.6.7, "Steam Generator Tube Inspection Report." The proposed changes are needed to address implementation issues associated with the inspection periods, and address other administrative changes and clarifications. The SGs in both Cook Nuclear Plant (CNP) units are equipped with alloy 690 thermally treated tubing; therefore, the TSTF-510 material specific requirements for that material type are reflected in the proposed change.

The proposed amendment is consistent with TSTF-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection."

2.0 ASSESSMENT

2.1 Applicability of Published Safety Evaluation Indiana Michigan Power Company (I&M) has reviewed TSTF-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection" (TSTF-510)

(ADAMS Accession No. ML110610350) and the model safety evaluation dated October 19, 2011 (ADAMS Accession No. ML112101513) as identified in the Federal Register Notice of Availability dated October 27, 2011 (76 FR 66763).

As described in the subsequent paragraphs, I&M has concluded that the justifications presented in TSTF-510 and the model safety evaluation prepared by the NRC staff are applicable to CNP, Units 1 and 2, and justify this amendment for the incorporation of the changes to the CNP TS.

2.2 Optional Changes and Variations I&M is proposing the following variations from the TS changes described in TSTF-510, Revision 2, or the applicable parts of the NRC staff's model safety evaluation dated October 19, 2011.

CNP TSs utilize different numbering than the Standard Technical Specifications on which TSTF-510 was based. The specific numbering differences are:

TSTF-510 Revision 2 CNP TS's 3.4.20, "Steam Generator Tube Integrity" 3.4.17 5.5.9, "Steam Generator (SG) Program" 5.5.7 During the February 15, 2012, TSTF meeting with the NRC, which was documented by a letter from the TSTF to the NRC dated March 28, 2012 (ML12088A082), it was identified that the proposed wording for Paragraph d.2 contained an administrative error. Paragraph d.2 contains the following statement:

If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of to AEP-NRC-2012-54 Page 2 degradation at this location and that may satisfy the applicable tube repair criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. (emphasis added)

The underlined phrase should state "tube plugging [or repair] criteria," consistent with the other changes made in TSTF-510.

I&M is changing the phrase to "tube plugging criteria."

This change is an administrative change and should not result in this application being removed from the Consolidated Line Item Improvement Process.

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Determination CNP Units 1 and 2 request adoption of an approved change to the standard technical specifications (STS) into the CNP plant specific technical specifications (TS), to revise Specification 5.5.7, "Steam Generator (SG) Program," and Specification 5.6.7, "Steam Generator Tube Inspection Report," and Limiting Condition of Operation (LCO) 3.4.17, "Steam Generator Tube Integrity," to address inspection periods and other administrative changes and clarifications.

As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below:

1.

Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change revises the Steam Generator (SG) Program to modify the frequency of verification of SG tube integrity and SG tube sample selection. A steam generator tube rupture (SGTR) event is one of the design basis accidents that are analyzed as part of a plant's licensing basis. The proposed SG tube inspection frequency and sample selection criteria will continue to ensure that the SG tubes are inspected such that the probability of an SGTR is not increased. Section 4.0, Technical Analysis, of the TSTF demonstrates that the change in frequencies will not increase the probability of an SGTR.

The consequences of an SGTR are bounded by the conservative assumptions in the design basis accident analysis. The proposed change will not cause the consequences of an SGTR to exceed those assumptions. Therefore, it is concluded that this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2.

Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed changes to the Steam Generator Program will not introduce any adverse changes to the plant design basis or postulated accidents resulting from potential tube to AEP-NRC-2012-54 Page 3 degradation. The proposed change does not affect the design of the SGs or their method of operation. In addition, the proposed change does not impact any other plant system or component.

Therefore, it is concluded that this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3.

Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The SG tubes in pressurized water reactors are an integral part of the reactor coolant pressure boundary and, as such, are relied upon to maintain the primary system's pressure and inventory. As part of the reactor coolant pressure boundary, the SG tubes are unique in that they are also relied upon as a heat transfer surface between the primary and secondary systems such that residual heat can be removed from the primary system. In addition, the SG tubes also isolate the radioactive fission products in the primary coolant from the secondary system. In summary, the safety function of an SG is maintained by ensuring the integrity of its tubes.

Steam generator tube integrity is a function of the design, environment, and the physical condition of the tube. The proposed change does not affect tube design or operating environment. The proposed change will continue to require monitoring of the physical condition of the SG tubes such that there will not be a reduction in the margin of safety compared to the current requirements.

Therefore, it is concluded that the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, I&M concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

3.2 Applicable Regulatory Requirements/Criteria The Traveler and model Safety Evaluation for TSTF-510 discuss the applicable regulatory requirements and guidance, including the 10 CFR 50, Appendix A, General Design Criteria (GDC). The construction permits for CNP were issued and the majority of construction was completed prior to issuance of 10 CFR 50, Appendix A, General Design Criteria, in 1971 by the Atomic Energy Commission (AEC); therefore, CNP was not licensed to the 10 CFR 50 Appendix A GDC. CNP was designed and constructed to comply with the AEC GDC as proposed on July 10, 1967. The application of the AEC proposed GDC to the CNP is contained in the CNP UFSAR, Section 1.4, as the Plant Specific Design Criteria (PSDC). Appendix A of 10 CFR Part

.50 GDC differ both in numbering and content from the PSDC for CNP. However, the following information demonstrates that CNP's PSDC meets the intent of GDC 14, 30, and 32 of 10 CFR 50, Appendix A.

to AEP-NRC-2012-54 Page 4 CNP's PSDC 9 is the equivalent of GDC 14 as can be seen in the PSDC language below. With the exception of the testing requirement which is also prescribed within GDC 32, which is discussed further below in CNP's PSDC equivalent to GDC 32. These differences do not alter the conclusion that the proposed change is applicable to CNP.

PSDC 9 Reactor Coolant Pressure Boundary The reactor coolant pressure boundary shall be designed, fabricated and constructed so as to have an exceedingly low probabi.lity of gross rupture or significant uncontrolled leakage throughout its design lifetime.

There are no changes to the SG design that impact CNP's PSDC 9 which is the equivalent of GDC 14. The evaluation performed in Section 4.0 of TSTF-510 concludes that the proposed change will continue to comply with CNP's PSDC.

CNP's PSDC 34 and PSDC 16 are the equivalent of GDC 30 as can be seen in the PSDC language below. These differences do not alter the conclusion that the proposed change is applicable to CNP.

PSDC 34 Reactor Coolant Pressure Boundary Propagation Failure Prevention The reactor coolant pressure boundary shall be designed and operated to reduce to an acceptable level the probability of rapidly propagating type failure.

Consideration is given (a) to the provisions for control over service temperature and irradiation effects which may require operational restrictions, (b) to the design and construction of the reactor pressure vessel in accordance with applicable codes, including those which establish requirements for absorption of energy within the elastic strain energy range and for absorption of energy by plastic deformation and (c) to the design and construction of reactor coolant pressure boundary piping and equipment in accordance with applicable codes.

PSDC 16 Monitoring Reactor Coolant Leakage Means shall be provided to detect significant uncontrolled leakage from the reactor coolant pressure boundary.

There are no changes to the SG design that impact CNP's PSDC 34 and 16 which is the equivalent of GDC 30. The evaluation performed in Section 4.0 of TSTF-510 concludes that the proposed change will continue to comply with CNP's PSDC.

CNP's PSDC 36 is the equivalent of GDC 32 as can be seen in the PSDC language below.

This difference does not alter the conclusion that the proposed change is applicable to CNP.

PSDC 36 Reactor Coolant Pressure Boundary Surveillance Reactor coolant pressure boundary components shall have provisions for inspection, testing, and surveillance of critical areas by appropriate means to assess the structural and leaktight integrity of the boundary components during their service lifetime. For the to AEP-NRC-2012-54 Page 5 reactor vessel, a material surveillance program conforming with current applicable codes shall be provided.

There are no changes to the SG design that impact CNP's PSDC 36 which is the equivalent of.

GDC 32. The evaluation performed in Section 4.0 of TSTF-510 concludes that the proposed change will continue to comply with CNP's PSDC.

4.0 ENVIRONMENTAL EVALUATION The proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change. -

to AEP-NRC-2012-54 DONALD C. COOK NUCLEAR PLANT UNIT 1 TECHNICAL SPECIFICATION PAGES MARKED TO SHOW CHANGES

SG Tube Integrity 3.4.17 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.17 Steam Generator (SG) Tube Integrity LCO 3.4.17 SG tube integrity shall be maintained.

AND All SG tubes satisfying the tube inr-epW criteria shall be plugged in accordance with the Steam Generator Program.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS NOTE----

Separate Condition entry is allowed for each SG tube.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more SG tubes A.1 Verify tube integrity of the 7 days satisfying the tube affected tube(s) is 1plugging FepaiF criteria maintained until the next and not plugged in refueling outage or SG accordance with the tube inspection.

Steam Generator Program.

AND A.2 Plug the affected tube(s) in Prior to entering accordance with the Steam MODE 4 following the Generator Program.

next refueling outage or SG tube inspection B. Required Action and B.1 Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A not AND met.

B.2 Be in MODE 5.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR SG tube integrity not maintained.

Cook Nuclear Plant Unit 1 3.4.17-1 Amendment No. 298

SG Tube Integrity 3.4.17 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.17.1 Verify SG tube integrity in accordance with the In accordance Steam Generator Program.

with the Steam Generator Program SR 3.4.17.2 Verify that each inspected SG tube that satisfies the Prior to entering tube 1plugging Fepl criteria is plugged in MODE 4 following accordance with the Steam Generator Program.

a SG tube inspection Cook Nuclear Plant Unit 1 3.4.17-2 Amendment No. 298

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Steam Generator (SG) Progqram A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following previeiene:

a.

Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes.

Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.

b.

Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.

1. Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool downs and all anticipated transients included in the design specificationD and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

2. Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed 1 gpm for all SGs.

Cook Nuclear Plant Unit I 5.5-5 Amendment No. 2&~, 298 Cook Nuclear Plant Unit 1 5.5-5 Amendment No. 297, 298

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Steam Generator (SG) Program (continued)

3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS Operational LEAKAGE."

c.

Provisions for SG tube infrpir criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.

d.

Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube ijngeaif criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assesspment o degradation

[assessmentl shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

1. Inspect 100% of the tubes in each SG during the first refueling outage following SG isllationnlre.

2. Inspect 100% of the tubes at sequential periods of 144, 108, 72, and, thereafter, 60 effective full power month.

first sequential period shall be considered to begin after the first inserVice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the Fremaining 50% by the FeuliR l

ua neares th end of th ld No Sshll nat

,,h.h.....

less) wit,,,h,.u,,bein the first refuelinal outage following SG installation, inspect each SG at least every 721 effective full power months or at least every third refueling outag (whichever results in more frequent inspections). In addition, thel minimum number of tubes inspected at each scheduled inspectioni shall be the number of tubes in all SGs divided by the number of SGI inspection outages scheduled in each inspection period as defined inI a, b, c and d below. If a degradation assessment indicates thel potential for a type of degradation to occur at a location not reviouslyI inspected with a technique capable of detecting this type o degradation at this location and that may satisfy the applicable tube]

DluQaina criteria, the minimum number of locations inspected with such]

Cook Nuclear Plant Unit 1 5.5-6 Amendment No. 2g~, 298 Cook Nuclear Plant Unit 1 5.5-6 Amendment No. 2-87, 298

Programs and Manuals 5.5 a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations.to be inspected fod this potential type of degradation at this location at the end of thel inspection period shall be no less than the ratio of the number of timesl the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG isl scheduled to be inspected in the inspection period. Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period andl the subsequent inspection period begins at the conclusion of thel included SG inspection outage.1 A) After the first refueling outage following SG installation, inspect 100% of the tubes during the next 144 effective full power months.

This constitutes the first inspection period; B) During the next 120 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period; C) During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the third inspection period; and D) During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the fourth and subsequent inspection periods.

3.

If crack indications are found in any SG tube, then the next inspection for each Faffected and potentially affected JSG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever lin more frequent inspections is-less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

e.

Provisions for monitoring operational primary to secondary LEAKAGE.

Cook Nuclear Plant Unit I 5.5-7 Amendment No. 2&7, 298 Cook Nuclear Plant Unit 1 5.5-7 Amendment No. 287, 298

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

8. WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLOTM, July 2006 (Westinghouse Proprietary).

c.

The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d.

The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Post Accident Monitoring Report When a report is required by Condition B or G of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.7 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.7, Steam Generator (SG) Program. The report shall include:

a.

The scope of inspections performed on each SG,

b.

AGtie-Jegradation mechanisms found,

c.

Nondestructive examination techniques utilized for each degradation mechanism,

d.

Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e.

Number of tubes plugged during the inspection outage for each aotive degradation mechanism,

f.

Total Rube...

,4~

ntg of.. tue lugged to,4 t.rThe number andI 1percentage of tubes plu~ged to date, and the effective plugging percentage]

ln each steam generatod1, and

g.

The results of condition monitoring, including the results of tube pulls and in-situ testing.

Cook Nuclear Plant Unit 1 5.6-4 Amendment No. 287, 288, 2-98, 3-1-3, 316 to AEP-NRC-2012-54 DONALD C. COOK NUCLEAR PLANT UNIT 2 TECHNICAL SPECIFICATION PAGES MARKED TO SHOW CHANGES

SG Tube Integrity 3.4.17 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.17 Steam Generator (SG) Tube Integrity LCO 3.4.17 SG tube integrity shall be maintained.

AND All SG tubes satisfying the tube i*n Ipir criteria shall be plugged in accordance with the Steam Generator Program.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS I. i #"'1=

II--.

Separate Condition entry is allowed for each SG tube.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more SG tubes A.1 Verify tube integrity of the 7 days satisfying the tube affected tube(s) is in

. criteria maintained until the next and not plugged in refueling outage or SG accordance with the tube inspection.

Steam Generator Program.

AND A.2 Plug the affected tube(s) in Prior to entering accordance with the Steam MODE 4 following the Generator Program.

next refueling outage or SG tube inspection B. Required Action and B.1 Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A not AND met.

B.2 Be in MODE 5.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR SG tube integrity not maintained.

Cook Nuclear Plant Unit 2 3.4.17-1 Amendment No. 279

SG Tube Integrity 3.4.17 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.17.1 Verify SG tube integrity in accordance with the In accordance Steam Generator Program.

with the Steam Generator Program SR 3.4.17.2 Verify that each inspected SG tube that satisfies the Prior to entering tube p ingQiý criteria is plugged in MODE 4 following accordance with the Steam Generator Program.

a SG tube inspection Cook Nuclear Plant Unit 2 3.4.17-2 Amendment No. 279

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following previsiens:

a.

Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes.

Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.

b.

Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.

1. Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool downs and all anticipated transients included in the design specificationLj) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-Secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

2. Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed 1 gpm for all SGs.

Cook Nuclear Plant Unit 2 5.5-5 Amendment No. 2~9, 279 Cook Nuclear Plant Unit 2 5.5-5 Amendment No. 269, 279

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Steam Generator (SG) Progqram (continued)

3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS Operational LEAKAGE."

c.

Provisions for SG tube i jgeail criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.

d.

Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube ingea criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation lassessment shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

1. Inspect 100% of the tubes in each SG during the first refueling outage following SG installation.:h:

2. Inspect 100% of the tubes at sequential periods of 144, 108, 72, and, thereafter, 60 effctive full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the Femaining 50% by the rp 4

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/:hnOl r,rrI n

+ka nnl in.

I n, QI % ekr, I nl n

a,,*

,..r, refulolll eutao neaes them ev, of the

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No SG shl el nm v

ra t.=

(whichever iS less) without being inspected.

fter the first refuelin outage following SG installation, inspect each SG at least every 72 effective full power months or at least every third refueling outag (whichever results in more frequent inspections). In addition, the minimum number of tubes inspected at each scheduled inspectionj shall be the number of tubes in all SGs divided by the number of SG]

inspection outages scheduled in each inspection period as defined ini a, b, c and d below. If a degradation assessment indicates theI potential for a type of degradation to occur at a location not previousl inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tubeý Dluaaina criteria, the minimum number of locations inspected with such]

Cook Nuclear Plant Unit 2 5.5-6 Amendment No. 269, 279 Cook Nuclear Plant Unit 2 5.5-6 Amendment No. 269, 279

Programs and Manuals 5.5 a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of theý inspection period shall be no less than the ratio of the number of timesl the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially bel occurring at this location divided by the total number of times the SG isl scheduled to be inspected in the inspection period. Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period andl the subsequent inspection period begins at the conclusion of thel included SG inspection outage.

a) After the first refueling outage following SG installation, inspect 100% of the tubes during the next 144 effective full power months.

This constitutes the first inspection period; b) During the next 120 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period; c) During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the third inspection period; and d) During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the fourth and subsequent inspection periods.

3. If crack indications are found in any SG tube, then the next inspection for each Faffected and potentially affected SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever results jin more frequent inspections is-less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

e.

Provisions for monitoring operational primary to secondary LEAKAGE.

Cook Nuclear Plant Unit 2 5.5-7 Amendment No. 2-W 279

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

7.

WCAP-13749-P-A, "Safety Evaluation Supporting the Conditional Exemption of the Most Negative EOL Moderator Temperature Coefficient Measurement," (Westinghouse Proprietary).

c.

The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d.

The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Post Accident Monitoring Report When a report is required by Condition B or G of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.7 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.7, Steam Generator (SG) Program. The report shall include:

a.

The scope of inspections performed on each SG,

b.

AGtiveegradation mechanisms found,

c.

Nondestructive examination techniques utilized for each degradation mechanism,

d.

Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e.

Number of tubes plugged during the inspection outage for each aeti-ve degradation mechanism,

f.

Total number and percentagc of tubes plugged to date ýhe number and 1percentage of tubes plugged to date, and the effective plugging percentagel

.lin each steam generator*, and

g.

The results of condition monitoring, including the results of tube pulls and in-situ testing.

Cook Nuclear Plant Unit 2 5.6-4 Amendment No. 269, 270, 279, 296, 297 to AEP-NRC-2012-54 DONALD C. COOK NUCLEAR PLANT UNIT 1 TECHNICAL SPECIFICATION BASES PAGES MARKED TO SHOW CHANGES

[INFORMATION ONLY]

SG Tube Integrity B 3.4.17 BASES APPLICABLE The steam generator tube rupture (SGTR) accident is the limiting design SAFETY basis event for SG tubes and avoiding an SGTR is the basis for this ANALYSES Specification. The analysis of an SGTR event assumes a bounding primary to secondary LEAKAGE rate equal to the operational LEAKAGE rate limits in LCO 3.4.13, "RCS Operational LEAKAGE," plus the leakage rate associated with a double-ended rupture of a single tube. The accident analysis for an SGTR assumes the contaminated secondary fluid is only briefly released to the atmosphere via the SG power operated relief valves.

The analysis for design basis accidents and transients other than an SGTR assumes the SG tubes retain their structural integrity (i.e., they are assumed not to rupture.) In these analyses, the steam discharge to the atmosphere is based on 150 gpd per SG primary to secondary LEAKAGE as an initial condition. For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT 1-131 is assumed to be equal to the LCO 3.4.16, "RCS Specific Activity," limits. For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaged fuel. The dose consequences of these events are within the limits of GDC 19 (Ref. 2), 10 CFR 100 (Ref. 3) or the NRC approved licensing basis (e.g., a small fraction of these limits).

Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The LCO requires that SG tube integrity be maintained. The LCO also requires that all SG tubes that satisfy the uinrpi criteria be plugged in accordance with the Steam Generator Program.

During an SG inspection, any inspected tube that satisfies the Steam Generator Program jplugginger

-criteria is removed from service by plugging. If a tube was determined to satisfy the i ngip criteria but was not plugged, the tube may still have tube integrity.

In the context of this Specification, an SG tube is defined as the entire length of the tube, including the tube wall, between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet.

The tube-to-tubesheet weld is not considered part of the tube.

An SG tube has tube integrity when it satisfies the SG performance criteria. The SG performance criteria are defined in Specification 5.5.7, "Steam Generator Program," and describe acceptable SG tube performance. The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria.

Cook Nuclear Plant Unit 1 B 3.4.17-2 Revision No. 26

SG Tube Integrity B 3.4.17 BASES LCO (continued)

The operational LEAKAGE performance criterion provides an observable indication of SG tube conditions during plant operation. The limit on operational LEAKAGE is contained in LCO 3.4.13, "RCS Operational LEAKAGE," and limits primary to secondary LEAKAGE through any one SG to 150 gallons per day. This limit is based on the assumption that a single crack leaking this amount would not propagate to an SGTR under the stress conditions of a LOCA or a main steam line break. If this amount of LEAKAGE is due to more than one crack, the cracks are very small, and the above assumption is conservative.

APPLICABILITY Steam generator tube integrity is challenged when the pressure differential across the tubes is large. Large differential pressures across SG tubes can only be experienced in MODE 1, 2, 3, or 4.

RCS conditions are far less challenging in MODES 5 and 6 than during MODES 1, 2, 3, and 4. In MODES 5 and 6, primary to secondary differential pressure is low, resulting in lower stresses and reduced potential for LEAKAGE.

ACTIONS The ACTIONS are modified by a Note clarifying that the Conditions may be entered independently for each SG tube. This is acceptable because the Required Actions provide appropriate compensatory actions for each affected SG tube. Complying with the Required Actions may allow for continued operation, and subsequent affected SG tubes are governed by subsequent Condition entry and application of associated Required Actions.

A.1 and A.2 Condition A applies if it is discovered that one or more SG tubes examined in an inservice inspection satisfy the tube ijngrea criteria but were not plugged in accordance with the Steam Generator Program as required by SR 3.4.17.2. An evaluation of SG tube integrity of the affected tube(s) must be made. Steam generator tube integrity is based on meeting the SG performance criteria described in the Steam Generator Program. The SG ui*n6ýp criteria define limits on SG tube degradation that allow for flaw growth between inspections while still providing assurance that the SG performance criteria will continue to be met. In order to determine if an SG tube that should have been plugged has tube integrity, an evaluation must be completed that demonstrates that the SG performance criteria will continue to be met until the next refueling outage or SG tube inspection. The tube integrity determination is based on the estimated condition of the tube at the time the situation is discovered and the estimated growth of the degradation prior to the next SG tube inspection. If it is determined that tube integrity is not being maintained, Condition B applies.

Cook Nuclear Plant Unit 1 B 3.4.17-4 Revision No. 12 Cook Nuclear Plant Unit 1 B 3.4.17-4 Revision No. 12

SG Tube Integrity B 3.4.17 BASES SURVEILLANCE-REQUIREMENTS (continued)

The Steam Generator Program determines the scope of the inspection and the methods used to determine whether the tubes contain flaws satisfying the tube i nfealF criteria. Inspection scope (i.e., which tubes or areas of tubing within the SG are to be inspected) is a function of existing and potential degradation locations. The Steam Generator Program also specifies the inspection methods to be used to find potential degradation. Inspection methods are a function of degradation morphology, non-destructive examination (NDE) technique capabilities, and inspection locations.

The Steam Generator Program defines the Frequency of SR 3.4.17.1.

The Frequency is determined by the operational assessment and other limits in the SG examination guidelines (Ref. 6). The Steam Generator Program uses information on existing degradations and growth rates to determine an inspection Frequency that provides reasonable assurance that the tubing will meet the SG performance criteria at the next scheduled inspection. In addition, Specification 5.5.7 contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections.i If crack indications are found in any SG tube, thel maximum interval for all affected and potentially affected SGs is restrictedi by Specification 5.5.7 until subsequent inspections support extending thel inspection interval.

SR 3.4.17.2 During an SG inspection, any inspected tube that satisfies the Steam Generator Program lu ing eaif criteria is removed from service by plugging. The tube repair criteria delineated in Specification 5.5.7 are intended to ensure that tubes accepted for continued service satisfy the SG performance criteria with allowance for error in the flaw size measurement and for future flaw growth. In addition, the tube in criteria, in conjunction with other elements of the Steam Generator Program, ensure that the SG performance criteria will continue to be met until the next inspection of the subject tube(s). Reference 1 provides guidance for performing operational assessments to verify that the tubes remaining in service will continue to meet the SG performance criteria.

The Frequency of prior to entering MODE 4 following an SG inspection ensures that the Surveillance has been completed and all tubes meeting the 71_1gpir criteria are plugged prior to subjecting the SG tubes to significant primary to secondary pressure differential.

Cook Nuclear Plant Unit 1 B 3.4.17-6 Revision No. 12 Cook Nuclear Plant Unit 1 B 3.4.17-6 Revision No. 12 to AEP-NRC-2012-54 DONALD C. COOK NUCLEAR PLANT UNIT 2 TECHNICAL SPECIFICATION BASES PAGES MARKED TO SHOW CHANGES

[INFORMATION ONLY]

SG Tube Integrity B 3.4.17 BASES APPLICABLE The steam generator tube rupture (SGTR) accident is the limiting design SAFETY basis event for SG tubes and avoiding an SGTR is the basis for this ANALYSES Specification. The analysis of an SGTR event assumes a bounding primary to secondary LEAKAGE rate equal to the operational LEAKAGE rate limits in LCO 3.4.13, "RCS Operational LEAKAGE," plus the leakage rate associated with a double-ended rupture of a single tube. The accident analysis for an SGTR assumes the contaminated secondary fluid is only briefly released to the atmosphere via the SG power operated relief valves.

The analysis for design basis accidents and transients other than an SGTR assumes the SG tubes retain their structural integrity (i.e., they are assumed not to rupture.) In these analyses, the steam discharge to the atmosphere is based on the total primary to secondary LEAKAGE from all SGs of 1 gallon per minute (gpm) or is assumed to increase to 1 gpm as a result of accident induced conditions. For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT I-131 is assumed to be equal to the LCO 3.4.16, "RCS Specific Activity,"

limits. For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaged fuel. The dose consequences of these events are within the limits of GDC 19 (Ref. 2), 10 G FR 100 (Ref. 3) or the NRC approved licensing basis (e.g., a small fraction of these limits).

Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The LCO requires that SG tube integrity be maintained. The LCO also requires that all SG tubes that satisfy the i

criteria be plugged in accordance with the Steam Generator Program.

During an SG inspection, any inspected tube that satisfies the Steam Generator Program lu ing repah criteria is removed from service by plugging. If a tube was determined to satisfy the iinreai criteria but was not plugged, the tube may still have tube integrity.

In the context of this Specification, an SG tube is defined as the entire length of the tube, including the tube wall, between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet.

The tube-to-tubesheet weld is not considered part of the tube.

An SG tube has tube integrity when it satisfies the SG performance criteria. The SG performance criteria are defined in Specification 5.5.7, "Steam Generator Program," and describe acceptable SG tube performance. The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria.

Cook Nuclear Plant Unit 2 B 3.4.17-2 Revision No. 12

SG Tube Integrity B 3.4.17 BASES LCO (continued)

The operational LEAKAGE performance criterion provides an observable indication of SG tube conditions during plant operation. The limit on operational LEAKAGE is contained in LCO 3.4.13, "RCS Operational LEAKAGE," and limits primary to secondary LEAKAGE through any one SG to 150 gallons per day. This limit is based on the assumption that a single crack leaking this amount would not propagate to an SGTR under the stress conditions of a LOCA or a main steam line break. If this amount of LEAKAGE is due to more than one crack, the cracks are very small, and the above assumption is conservative.

APPLICABILITY Steam generator tube integrity is challenged when the pressure differential across the tubes is large. Large differential pressures across SG tubes can only be experienced in MODE 1, 2, 3, or 4.

RCS conditions are far less challenging in MODES 5 and 6 than during MODES 1, 2, 3, and 4. In MODES 5 and 6, primary to secondary differential pressure is low, resulting in lower stresses and reduced potential for LEAKAGE.

ACTIONS The ACTIONS are modified by a Note clarifying that the Conditions may be entered independently for each SG tube. This is acceptable because the Required Actions provide appropriate compensatory actions for each affected SG tube. Complying with the Required Actions may allow for continued operation, and subsequent affected SG tubes are governed by subsequent Condition entry and application of associated Required Actions.

A.1 and A.2 Condition A applies if it is discovered that one or more SG tubes examined in an inservice inspection satisfy the tube i7jfreai criteria but were not plugged in accordance with the Steam Generator Program as required by SR 3.4.17.2. An evaluation of SG tube integrity of the affected tube(s) must be made. Steam generator tube integrity is based on meeting the SG performance criteria described in the Steam Generator Program. The SG WngFpiI criteria define limits on SG tube degradation that allow for flaw growth between inspections while still providing assurance that the SG performance criteria will continue to be met. In order to determine if an SG tube that should have been plugged has tube integrity, an evaluation must be completed that demonstrates that the SG performance criteria will continue to be met until the next refueling outage or SG tube inspection. The tube integrity determination is based on the estimated condition of the tube at the time the situation is discovered and the estimated growth of the degradation prior to the next SG tube inspection. If it is determined that tube integrity is not being maintained, Condition B applies.

Cook Nuclear Plant Unit 2 B 3.4.17-4 Revision No. 12

SG Tube Integrity B 3.4.17 BASES SURVEILLANCE REQUIREMENTS (continued)

The Steam Generator Program determines the scope of the inspection and the methods used to determine whether the tubes contain flaws satisfying the tube ipinegph criteria. Inspection scope (i.e., which tubes or areas of tubing within the SG are to be inspected) is a function of existing and potential degradation locations. The Steam Generator Program also specifies the inspection methods to be used to find potential degradation. Inspection methods are a function of degradation morphology, non-destructive examination (NDE) technique capabilities, and inspection locations.

The Steam Generator Program defines the Frequency of SR 3.4.17.1.

The Frequency is determined by the operational assessment and other limits in the SG examination guidelines (Ref. 6). The Steam Generator Program uses information on existing degradations and growth rates to determine an inspection Frequency that provides reasonable assurance that the tubing will meet the SG performance criteria at the next scheduled inspection. In addition, Specification 5.5.7 contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections. Ilf crack indications are found in any SG tube, theI maximum inspection interval for all affected and potentially affected SGsI is restricted by Specification 5.5.7 until subsequent inspections supporT extending the inspection interval.

SR 3.4.17.2 During an SG inspection, any inspected tube that satisfies the Steam Generator Program pluginpaif criteria is removed from service by plugging. The tube plu inrep criteria delineated in Specification 5.5.7 are intended to ensure that tubes accepted for continued service satisfy the SG performance criteria with allowance for error in the flaw size measurement and for future flaw growth. In addition, the tube

~ingepai criteria, in conjunction with other elements of the Steam Generator Program, ensure that the SG performance criteria will continue to be met until the next inspection of the subject tube(s). Reference 1 provides guidance for performing operational assessments to verify that the tubes remaining in service will continue to meet the SG performance criteria.

The Frequency of prior to entering MODE 4 following an SG inspection ensures that the Surveillance has been completed and all tubes meeting the repair criteria are plugged prior to subjecting the SG tubes to significant primary to secondary pressure differential.

Cook Nuclear Plant Unit 2 B 3.4.17-6 Revision No. 12