Text

Application to Modify Technical Specifications for Replacement Steam Generators (TS-SQN-2011-01)
	ML11199A212
Person / Time
Site:	Sequoyah ; (DPR-079)
Issue date:	07/15/2011
From:	Krich R; Tennessee Valley Authority
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	TAC MD0145
	Download: ML11199A212 (43)
	v • d • e

Tennessee Valley Authority 1101 Market Street, LP 3R Chattanooga, Tennessee 37402-2801 R. M. Krich Vice President Nuclear Licensing July 15, 2011 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Sequoyah Nuclear Plant, Unit 2 Facility Operating License No. DPR-79 NRC Docket No. 50-328

Subject:

Application to Modify Technical Specifications for Replacement Steam Generators (TS-SQN-2011-01)

Reference:

Letter from NRC to TVA, "Sequoyah Nuclear Plant, Unit 2 - Issuance of Amendment Regarding Steam Generator Tube Integrity (TAC No.

MD0145)," dated May 22, 2007 In accordance with the provisions of 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," the Tennessee Valley Authority (TVA) is submitting a request for an amendment to Facility Operating License DPR-79 for Sequoyah Nuclear Plant (SQN), Unit 2.

The license amendment request proposes to revise the SQN, Unit 2, Technical Specifications (TS) requirements for steam generator tube inspections to reflect the replacement steam generators to be installed during SQN, Unit 2, refueling outage 18 (U2R18) presently scheduled for the fall of 2012. Previous changes to the SQN, Unit 2, TS to reflect Technical Specification Task Force (TSTF) Standard Technical Specification Traveler, TSTF-449, "Steam Generator Tube Integrity," Revision 4 were approved as Amendment No. 305 by the reference letter dated May 22, 2007. The changes proposed in this license amendment request reflect the inspection requirements of TSTF-449, Revision 4.

The enclosure to this letter provides a description of the proposed changes and confirmation of applicability. Attachments 1 and 2 to the enclosure provide the existing TS and Bases pages marked-up to show the proposed changes. Attachments 3 and 4 to the enclosure provide the existing TS and Bases pages retyped to show the proposed changes.

printed on recycled paper

U.S. Nuclear Regulatory Commission Page 2 July 15, 2011 The change is proposed for implementation following SQN, Unit 2 refueling outage U2R18, when the replacement steam generators will have been installed. Accordingly, TVA requests NRC approval on a schedule to allow implementation of this TS change to coincide with the SQN, Unit 2, refueling outage U2R18.

TVA has determined that there are no significant hazards considerations associated with the proposed change and that the TS change qualifies for categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22(c)(9).

The SQN Plant Operations Review Committee and the SQN Nuclear Safety Review Board have reviewed this proposed change and determined that operation of SQN in accordance with the proposed change will not endanger the health and safety of the public.

Additionally, in accordance with 10 CFR 50.91(b)(1), TVA is sending a copy of this letter and enclosure to the Tennessee Department of Environment and Conservation.

There are no new regulatory commitments associated with this submittal. Please address any questions regarding this request to Dan Green at 423-751-8423.

I declare under penalty of perjury that the foregoing is true and correct. Executed on this 15th day of July 2011.

Respectfully, R. M. Krich

Enclosure:

Evaluation of the Proposed Change cc (Enclosure):

NRC Regional Administrator - Region II NRC Senior Resident Inspector - Sequoyah Nuclear Plant Director, Division of Radiological Health - Tennessee Department of Environment and Conservation

ENCLOSURE TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT UNIT 2 EVALUATION OF THE PROPOSED CHANGE

Subject:

Request for Change to Technical Specification Section 6 for Replacement Steam Generators (TS-SQN-201 1-01)

1.

SUMMARY

DESCRIPTION

2.

DETAILED DESCRIPTION

3.

TECHNICAL EVALUATION

4.

REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 Significant Hazards Consideration 4.4 Conclusion

5.

ENVIRONMENTAL CONSIDERATION

6.

REFERENCES ATTACHMENTS

1.

Proposed TS Changes (Mark-Ups)

2.

Proposed TS Bases Changes (Mark-Ups)

3.

Proposed TS Changes (Final Typed)

4.

Proposed TS Bases Changes (Final Typed)

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1.0

SUMMARY

DESCRIPTION By letter dated May 22, 2007 (Reference 1), License Amendment No. 305 was approved for the Sequoyah Nuclear Plant (SQN), Unit 2. That amendment implemented Revision 4 to Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler 449 (TSTF-449) for the existing SQN, Unit 2, steam generators (SGs). This License Amendment Request (LAR) seeks implementation of the TSTF-449 inspection requirements for the replacement steam generators (RSGs),

that are being installed during the SQN, Unit 2, refueling outage 18 (U2R18) scheduled to start in the fall of 2012. The RSGs differ from the existing SGs in that the tube material in the RSGs is Alloy 690 Thermally Treated (TT) versus Alloy 600 in the existing SGs. The inspection requirements for Alloy 690 TT material permit longer periods between 100 percent tube population inspections and between individual SG inspections. Additionally, this LAR removes inspection requirements that are designated for specific damage conditions in the existing SGs, removes tube repair techniques approved by previous license amendments for the existing SGs, and removes inspection and reporting requirements specific to those repair techniques.

The elements of TSTF-449 regarding the Technical Specifications (TS) definition of Identified Leakage, TS definition of Pressure Boundary Leakage, and TS 3.4.6.2, Reactor Coolant System Operational Leakage, were incorporated into the SQN, Unit 2, TS by License Amendment No. 305. Those elements are consistent with this proposed change. Since those elements are integral to the overall maintenance of RSG tube integrity, reference to them in the Federal Register Notice (Reference 2) discussions applicable to Sections 3.0 through 5.0 below continues to remain appropriate.

2.0 DETAILED DESCRIPTION The proposed TS changes are as follows.

Revised TS Section 6.8.4.k, "Steam Generator (SG) Program" Revised TS Sections 6.9.1.16.2, 6.9.1.16.3, 6.9.1.16.4, and 6.9.1.16.5, "Steam Generator (SG) Tube Inspection Report" The revisions are necessary because of two factors. The inspection frequency for Alloy 690 TT tube material, as defined in TSTF-449, differs from the inspection frequency for Alloy 600, and the tube repair processes and products in the existing TS are not applicable to the RSGs.

TS Section 6.8.4.k is revised to change the tube inspection frequency, as specified in TSTF-449, from that applicable to Alloy 600 to that applicable to Alloy 690TT tube material.

TS Section 6.8.4.k is revised to delete information on repair techniques and inspection requirements specific to tube repairs.

TS Section 6.8.4.k is revised to delete inspection requirements that are designated for specific damage conditions in the existing SGs.

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TS Sections 6.9.1.16.2, 6.9.1.16.3, 6.9.1.16.4, and 6.9.1.16.5 are revised to remove reporting requirements associated with the deleted repair techniques.

Proposed revisions to the TS Bases are also included in this application for information only. As discussed in the NRC's model safety evaluation, adoption of the revised TS Bases associated with TSTF-449, Revision 4, is an integral part of implementing this TS change. The changes to the affected TS Bases pages will be incorporated in accordance with the SQN, Unit 2, TS Bases Control Program.

3.0 TECHNICAL EVALUATION

The Tennessee Valley Authority (TVA) has reviewed the safety evaluation published on March 2, 2005, (70 FR 10298) (Reference 2), as part of the Consolidated Line Item Improvement Program (CLIIP) Notice for Comment. This included the NRC's Safety Evaluation (SE), the supporting information provided in TSTF-449, and the changes associated with Revision 4 to TSTF-449. TVA has concluded that the justifications presented in the TSTF proposal and the SE prepared by the NRC are applicable to SQN, Unit 2, and justify this amendment for the incorporation of the changes to the SQN, Unit 2, TS.

Calculations performed by the manufacturer of the RSGs have confirmed the acceptability of the 40 percent tube plugging limit included in the existing TS. Those calculations were performed in accordance with the guidance and recommendations of Regulatory Guide 1.121 (Reference 4).

The proposed 100 percent inspection frequency and maximum interval for inspecting a RSG are in accordance with Revision 4 of TSTF-449 for steam generators with Alloy 690 TT tube material.

There are no repair processes currently approved for the RSGs tube material; therefore, the reference to repair, and the application of repair processes, in TS Section 6.8.4.k must be removed. Additionally, inspection requirements that are designated for specific damage conditions in the existing SGs must be removed from TS Section 6.8.4.k.

4.0 REGULATORY EVALUATION

A description of this proposed change and its relationship to applicable regulatory requirements and guidance was provided in the NRC.Notice of Availability published on May 6, 2005 (70 FR 24126) (Reference 3), the NRC Notice for Comment published on March 2, 2005 (70 FR 10298), and TSTF-449, Revision 4.

4.1 Applicable Regulatory Requirements /Criteria The applicable regulatory requirements/criteria associated with this application are adequately addressed by the NRC Notice of Availability published on May 6, 2005 (70 FR 24126), the NRC Notice for Comment published on March 2, 2005 (70 FR 10298), and TSTF-449, Revision 4.

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The following information concerning the RSGs is provided to support the NRC review of this LAR.

Plant Name, Unit Number:

Sequoyah Nuclear Plant (SQN), Unit 2 Steam Generator (G) Model(s):

Westinghouse Electric Company Model 57AG÷ Effective Full Power Years (EFPY) of New at startup from SQN, Unit 2, Refueling service for currently installed SGs:

Outage in Fall 2012.

Tubing Material (e.g., 600M. 600TT, 690TT):

690TT (Thermally Treated)

Number of tubes per SG:

4,983 Number and percentage of tubes plugged in SG 1 SG 2 SG 3 SG 4 each SG*:

0 0

Number of tubes repaired in each SG:

0 Degradation mechanism(s):

None Current primary-to-secondary leakage limits:

Per SG: 150 gallons per day through any one steam generator per TS 3.4.6.2 Approved Alternate Repair Criteria (ARC):

None.

Approved SG Tube Repair Methods:

None.

Performance criteria for accident leakage:

Primary-to-secondary leak rate values assumed in the limiting SQN, Unit 2, licensing basis accident analysis are 0.1 gallons per minute (gpm) for each of the non-faulted SGs and 1.0 gpm for the faulted SG.

There were no tubes plugged during fabrication. The tubes will be pre-service eddy current tested beginning in January 2012. Based on this testing, if tube plugging is required, TVA will provide the number and percentage of tubes plugged in each steam generator within 90 days following completion of eddy current testing.

4.2 Precedent This application is made in accordance with TSTF-449. TVA is not proposing any variations or deviations from the TS changes described in TSTF-449, Revision 4, or the NRC's model Safety Evaluation published on March 2, 2005, (70 FR 10298).

The proposed changes for the SQN, Unit 2, RSGs are similar to proposed changes included in the following license amendment requests.

Pacific Gas and Electric Company for the Diablo Canyon, Units 1 and 2, "License Amendment Request 07-01 Revision to Technical Specifications to Support Steam Generator Replacement," dated January 11, 2007 (Reference 5); Approved by NRC letter to Pacific Gas and Electric Company, dated January 8, 2008 (Reference 6)

Florida Power Corporation for the Crystal River, Unit 3, "Crystal River Unit 3 -

License Amendment Request #301, Revision 1: Application to Modify Improved Technical Specifications for Replacement Steam Generators and Response to Request for Additional Information (TAC NO. MD9547)," dated E-4

January 19. 2009 (Reference 7); Approved by NRC letter to Progress Energy, dated May 29, 2009 (Reference 8) 4.3 Significant Hazards Consideration The proposed License Amendment Request (LAR) revises the Sequoyah Nuclear Plant (SQN), Unit 2, Technical Specifications (TS) Section 6.8.4.k, "Steam Generator (SG) Program," and TS Sections 6.9.1.16.2, 6.9.1.16.3, 6.9.1.16.4, and 6.9.1.16.5, "Steam Generator (SG) Tube Inspection Report." The proposed changes are necessary to revise the current SQN, Unit 2, TS for Replacement Steam Generators (RSGs) to be installed in the fall of 2012 refueling outage.

TVA has evaluated the proposed LAR against the criteria of 10 CFR 50.92(c) to determine if any significant hazards consideration is involved. TVA's conclusion is based on its evaluation, in accordance with 10 CFR 50.91 (a)(1), of the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

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

Response: No The proposed change for RSGs continues to implement the current SG Program that includes performance criteria which provide reasonable assurance that the RSG tubing will retain integrity over the full range of operating conditions (including startup, operation in the power range, hot standby, cooldown, and all anticipated transients included in the design specifications). This change removes repair criteria from the SG Program that were approved by previous License Amendments for the existing SGs which are not applicable to the RSGs. It removes references to use of repairs and reporting of repair results in other TS sections. This change removes inspection requirements that are designated for specific damage conditions in the existing SGs. The change also revises the inspection interval for 100 percent inspections of SG tubes and the maximum interval for inspection of a single SG consistent with Technical Specification Task Force (TSTF)

Standard Technical Specification Traveler, TSTF-449, "Steam Generator Tube Integrity," Revision 4 for the Alloy 690 tube material in the RSGs. The revised inspection requirements are based on properties and experience with the improved Alloy 690 tube material. The revised inspection requirements will result in the same outcome that SG tube integrity will continue to be maintained.

This change continues to implement SG performance criteria for tube structural integrity, accident induced leakage, and operational leakage for the RSGs. Meeting the performance criteria provides reasonable assurance that the RSG tubing will remain capable of fulfilling its specific safety function of maintaining reactor coolant pressure boundary integrity throughout each operating cycle and in the unlikely event of a design basis accident (DBA).

The performance criteria are only a part of the SG Program required by the E-5

existing TS. The program, defined by NEI 97-06, "Steam Generator Program Guidelines," includes a framework that incorporates a balance of prevention, inspection, evaluation, repair, and leakage monitoring. These features will continue to be implemented as they are currently approved. The proposed changes do not, therefore, significantly increase the probability of an accident previously evaluated.

The consequences of DBAs are, in part, functions of the Dose Equivalent 1-131 in the primary coolant and the primary to secondary leakage rates resulting from an accident. Therefore, limits are included in the TS for Operational Leakage and for Dose Equivalent 1-131 in the primary coolant to ensure the plant is operated within its analyzed condition. The analysis of the limiting DBA assumes that the primary to secondary leak rate, after the accident, is 1 gallon per minute with no more than 150 gallons per day in any one SG, and that the reactor coolant activity levels of Dose Equivalent 1-131 are at the TS values before the accident. The proposed change to the SG inspection program does not affect the design of the SGs, their method of operation, operational leakage limits, or primary coolant chemistry controls.

The proposed change does not adversely impact any other previously evaluated DBA. In addition, the proposed changes do not affect the consequences of a main steam line break, rod ejection, a reactor coolant pump locked rotor event, or other previously evaluated accident. Therefore, the proposed change does not affect the consequences of a SG tube rupture accident and the probability of such an accident is unchanged.

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

Response: No The proposed license amendment does not affect the method of operation of the SGs, or the primary or secondary coolant chemistry controls. In addition, the proposed amendment does not impact any other plant system or component. The change modifies existing SG inspection requirements based on the RSG design and the properties and experience associated with their improved materials. The revised inspection requirements will result in the same outcome that SG tube integrity will continue to be maintained.

Therefore, the proposed change does not create the possibility of a new or different type of accident from any accident previously evaluated.

3. Does the proposed amendment 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 E-6

SG tubes isolate the radioactive fission products in the primary coolant from the secondary system. In summary, the safety function of a SG is maintained by ensuring the integrity of its tubes. SG tube integrity is a function of the design, environment, and the physical condition of the tube. The proposed change to the SG inspection program does not affect tube design or operating environment. The existing SG Program is maintained in this change. The repair criteria that are being removed are specific to the existing SGs and are not applicable to the RSGs. If tube defects are detected that exceed limits in the RSGs, then the tube will be removed from service. The effective tube plugging percentage will continue to be tracked for all plugging in each SG in accordance with TS Section 6.9.1.16.1 to ensure the heat transfer function of the SGs is not adversely affected. For the above reasons, the margin of safety is not changed and overall plant safety will be enhanced by the proposed change to the TS.

Based on the above, TVA concludes that the proposed amendment does not involve a 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.

4.4 Conclusion In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR Part 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment 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 assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

1.

Letter from NRC to TVA, "Sequoyah Nuclear Plant, Unit 2 - Issuance of Amendment Regarding Steam Generator Tube Integrity (TAC No. MD0145),"

dated May 22, 2007

2.

Federal Register Notice for Comment published on March 2, 2005 (70 FR 10298)

3.

Federal Register Notice of Availability published on May 6, 2005 (70 FR 24126)

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4.

U.S. Nuclear Regulatory Commission Regulatory Guide 1.121, "Bases for Plugging Degraded PWR Steam Generator Tubes," August 1976

5.

Letter from Pacific Gas and Electric Company to NRC, "License Amendment Request 07-01 Revision to Technical Specifications to Support Steam Generator Replacement," dated January 11, 2007

6.

Letter from NRC to Pacific Gas and Electric Company, "Diablo Canyon Power Plant, Unit Nos. 1 and 2 - Issuance of Amendments Re: Revise Technical Specifications to Support Steam Generator Replacement (TAC Nos. MD3992 and MD3993)," dated January 8, 2008

7.

Letter from Progress Energy to NRC, "Crystal River Unit 3 - License Amendment Request #301, Revision 1: Application to Modify Improved Technical Specifications for Replacement Steam Generators and Response to Request for Additional Information (TAC NO. MD9547)," dated January 19. 2009

8.

Letter from NRC to Progress Energy, "Crystal River Unit 3 - Issuance of Amendment Regarding the Revision of the Steam Generator Portion of the Technical Specifications to Reflect the Replacement of the Steam Generators (TAC No. MD9547)," dated May 29, 2009 E-8

ATTACHMENT I TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT UNIT 2 PROPOSED TS CHANGES (Mark-Ups)

ADMINISTRATIVE CONTROLS

d.

Proposed changes that meet the criteria of Specification 6.8.4.j.b above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

k.

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 provisions:

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.

Provisions for 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 SG tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, cooldown, and all anticipated transients included in the design specification) and design basis accidents (DBAs). This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and eXcept f"- f1... 4

,s 4.....

essee, hr.....h a,.,,,orpt,,,.,-, thp i *!ten~te r Far rtprag dagc'- ssed on TS 6. R 4.kýc.!fa-safety factor of 1.4 against burst applied to the DBA primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the DBAs, 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.

Ior more

a.

s eanln ra halb esta

2.

Accident induced leakage performance criterion: The accident-induced leakagefreGm all I so-toes,.

!,dn the 'A_;k;;ae ;;ttrh,--ted-tn the degrada~tton desr ibd.

kc!a, is not to exceed 1.0 gpm for the faulted SG "*~ 0.!.Xam fnr Lqnh nf th ' nnn-fa f'

tod"S*-'*

The primary-to-secondary accident induced leakage rate for any DBA, 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.

SEQUOYAH - UNIT 2 6-1Oa Amendment No. 28, 50, 64, 66, 134, 165, 202, 207, 223, 231,265, 271,272, 276, 298, 305,

ADMINISTRATIVE CONTROLS

3.

The operational leakage performance criterion is specified in Limiting Condition for Operation (LCO) 3.4.6.2, "Reactor Coolant System, Operational Leakage."

c.

Provisions for SG Tube Repair 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.

he following alternate tube repair criteria (ARC) may be applied as an alternative to the 40%

d thbased criteria:

1.

IC Generic Letter (GL) 95-05 Voltagqe-Based ARC (Tube Support Plate [TSP])

A vo e-based TSP repair criteria is used for the disposition of an alloy 600 SG ube for continu service that is experiencing predominately axially oriented ODSCC nfined within the ickness of the tube support plates (TSPs). At TSP intersections he repair criteria is de ribed below:

a)

SG tubes, w se degradation is attributed to ODSCC within the unds of the TSP with bobbin vol es less than or equal to 2.0 volts, will be allo ed to remain in service.

b)

SG tubes, whose d radation is attributed to ODSCC wit!* the bounds of the TSP with a bobbin voltage eater than 2.0 volts will be plug d, except as noted in Item c)

SG tubes, with indications of tential degradati attributed to ODSCC within the bounds of the TSP with a bobbi voltage grea than 2.0 volts, but less than or equal to the upper voltage repair limit (c culated cording to the methodology in GL 95-05 as supplemented), may remain in se iceii a rotating pancake coil inspection or comparable technology does not dete egradation.

d)

SG tubes with indications of ODSC degra tion with a bobbin coil voltage greater than the upper voltage repair lim'(calculated cording to the methodology in GL 95-05 as supplemented) w be plugged.

e)

If an unscheduled mid-cy e inspection is performed, e following mid-cycle repair limits apply instead oft limits identified in Items 6.8.. cla),.b),.c) and.d).

The mid-cycle rep limits are determined from the followin quations:

VMUR

ý VSL(CL - At)

CL (CL-At)

VMLRL =

MURL (VURL VLRL)

CL VURL

=

upper voltage repair limit Apme n No.

205,

~QUOYAH - UNIT.2 6-1lOb Amendment No. 305, SE

ADMINISTRATIVE CONTROLS I

II V LRL I~JVV~I V~JItO~~ ~~JOU 1~1 HIL

~~~v L R L I

W -V V

1 V Wr l

u u I * }. F QI I I I I I I I p e VMURL mid-cycle uppe voltage repair limit based on time into cycle VMLRL mid-c e lower voltage repair limit based on URL and time into cycle At l length of time since last scheduled i spection during which VURL and V L were implemented CL cycle length (the time betwe two scheduled SG inspectio

)

VSL stru ral limit voltage Gr verage growth rate per cycle length NDE

-95 percent cumulative probability allowance for nondestructive examination uncertainty (i.e., a value of 20 percent has been approved by NRC)

Implementation of these mid-cycle repair *mits should follow e same approach as in TS items 6.8.4.k.c.1.a),.b),.c) and.d).

2.

W* Methodology The following terms/defin* ons apply to the W*.

a)

Bottom of WEX X Transition (BWT) is the highest point of contact between the tube and tubesheet at, below the top of tubesheet (TTS), as determined by ed current testing.

b)

W* Dista e for the hot-leg tubesheet is the larger of the following two distan es as measur d from the TTS: (a) 8 inches below the TTS or (b) 7 inches below the ottom of the WE WIEX transition plus the uncertainty associated with determining the distance elow the tbo m of the WEXTEX transition as defined by WCAP-14797, Revision 2.

c)

W* distance for the cold-leg tubesheet is 10.5 inches below TTS.

Service induced flaws identified in the W* distance shall be plugged on detection. Flaws located below the W* distance may remain in service regardless of size.

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 OctNobr 19, 2009 QUOYAH - UNIT 2 6-1lOc Amendment No. 305, 315, 318, SE

ADMINISTRATIVE CONTROLS at the tube outlet, and that may satisfy the applicable tube repair 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 ;&at 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 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 replacement.

144, 108, 72 and thereafter,

2.

Inspect 100% of the tubes at sequential periods of 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs.

Insert A No SGs shall operate for more than ffectivefull power months or ý refueling outage (whichever is less) without being inspedeu7 2

three

3.

If crack indications are found in any SG tube, en e next inspec ion for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever 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.

4.

L 95-05 Voltage-Based ARC for TSP Indication ft in service as a result of application of the TSP voltage-based re criteria shall be inspec by bobbin coil probe every 24 effective full power mont r every refueling outage, w ver is less.

Implementation of the SG tube repair criteria require 00 percent bobbin coil inspection for hot-leg and cold-leg ntersections wn to the lowest cold-leg TSP with known ODSCC indications. The determin e lowest cold-leg TSP intersections having ODSCC indications shall be base th erformance of at least a 20 percent random sampling of tubes inspecte er their full le

5.

W* Inspection When the W*

hodology has been implemented, inspect 100 percent of nservice tubes in hot-leg tubesheet and 20 percent of the inservice tubes in the cold-e besheet re ges with the objective of detecting flaws that may satisfy the applicable tube repair eria f TS 6.8.4.k.c.2.

e. Provisions for Monitoring Operational Primary-to-Secondary Leakage.

1. Component Cyclic and Transient Limit This program provides controls to track the FSAR, Section 5.2.1, cyclic and transient occurrences to ensure that components are maintained within the design limits.

SEQUOYAH - UNIT 2 October 19, 2009 6-1Od Amendment No. 28, 34, 50, 64, 66, 107,134, 165, 207, 223, 231,271,272, 289,293, 305, 315, 318,

ADMINISTRATIVE CONTROLS STEAM GENERATOR (SG) TUBE INSPECTION REPORT (continued)

a. The scope of inspections performed on each SG,

b.

Active degradation 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 active degradation mechanism,

f.

Total number and percentage of tubes plugged to date,

g.

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

h.

The effective plugging percentage for all plugging in each SG.

I.-

6.

.16.2 A report shall be submitted within 90 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the steam generator progra (6.8.4.k) when voltage based alternate repair criteria have been applied. The repo all clude information described in Section 6.b of Attachment 1 to NRC Generic Le r 95-05, age-Based Repair Criteria for Westinghouse Steam Generator Tubes cted by Outsi Diameter Stress Corrosion Cracking."

6.9.1.16.3 For impleme tion of the voltage-based repair criteria for tube s ort plate (TSP) intersections, no~ the staff prior to initial entry into MODE 4 owing completion of an inspection performe accordance with Specification 6.8 "Steam Generator (SG)

Program," should any o e following conditions arise:

1) If circumferential crack-likei ications are ected at the TSP intersections.

2)

If indications are identified that ex beyond the confines of the TSP.

3) If indications are identified a e TSP ele ions that are attributable to primary water stress corrosion crackin 6.9.1.16.4 For implementation o ie'the calculated steam line brea eakage from the application of TSP alternate rep criteria and W* inspection methodology all be submitted within 90 days after initial entry into MODE 4 following completion an inspection performed in accorda e with Specification 6.8.4.k, "Steam Generator (SG) Pr ram." The report will include e number of indications within the tubesheet region, the locati of the indications (rel ye to the bottom of the WEXTEX transition [BWT] and UTS), the orien tion (axial,

.cumnferential, skewed, volumetric), the severity of each indication (e.g., near ough-wall or not through-wall), the side of the tube from which the indication initiated (inside Sequoyah - Unit 2 6-14a May 22, 0*-

Amendment No. 305,

ADMINISTRATIVE CONTROLS

6.

1.16.5 For implementation of the probability of prior cycle detection (POPCD) method, for the voltage-based repair criteria at tube support plate intersections, if the en of-cycle conditional tube rupture probability for a postulated main steam lin b

k, the projected primary to secondary leak rate during a postulated ain stea *e break, or the number of indications are under predicted he previous le operational assessment, the following shall be r, rted to the Commission t in 90 days after initial entry into MODE 4 f owing completion of inspection perfor in accordance with specification

.4.k, "Steam Generator Program."

1.

The assessment of the bable cau for the under prediction, proposed corrective actions, and any o

ended changes to probability of detection or growth methodol dicated by potential methods assessments.

2.

An assessment oe potential need to revi the alternate repair criteria analysis met s if: the burst probability is un redicted by more than 0.001 (i., 0 percent of the performance criteria) o n order of magnitude; or th ak rate is under predicted by more than 0.5 gall er minute (gpm) n order of magnitude.

An assessment of the potential need to increase the number of pre ed low voltage indications at the beginning of cycle if the total number of as-found indications in any SG are underestimated by greater than 15 percent or by greater than 150 indications.

SPECIAL REPORTS 6.9.2.1 Special reports shall be submitted within the time period specified for each report, in accordance with 10 CFR 50.4.

6.9.2.2 This specification has been deleted.

6.10 RECORD RETENTION (DELETED)

SEQUOYAH - UNIT 2 6-15 March 24, 200-8 Amendment No. 28, 44, 50, 64, 66, 107, 134,146,153,165,169, 206, 214, 223, 231, 249, 284, 309,

TS-SQN-201 1-01 TS INSERT Insert A:

In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period.

ATTACHMENT 2 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT UNIT 2 PROPOSED TS BASES CHANGES (Mark-Ups)

Reactor Coolant System B 3/4.4.5 REACTOR COOLANT SYSTEM BASES APPLICABLE SAFETY ANALYSES The steam generator tube rupture (SGTR) accident is the limiting design basis event for SG tubes and avoiding an SGTR is the basis for this specification. The analysis of an SGTR event assumes a bounding primary to secondary leakage rate equal to the operational leakage rate limits in LCO 3.4.6.2 "Operational Leakage," plus the leakage rate associated with a double-ended rupture of a single tube. The accident analysis for a SGTR assumes the contaminated secondary fluid is released to the atmosphere via safety valves. The main condenser isolates based on an assumed concurrent loss of off-site power.

The analysis for design basis accidents and transients other than a SGTR assume the SG tubes retain their structural integrity (i.e., they are assumed not to rupture). In these analyses, the steam discharge to the atmosphere depends on the accident and whether there are faulted SGs associated with the accident. For a steamline break (SLB), the maximum primary to secondary leakage under accident conditions is limited to 3.7 apm from the faulted SG and 0.1 gpm from each of the non-faulted SGs.

primary-to-secondary leak rate assumed dunn no more than l.

ome from sources th o been specifically exempted from the

=

y the NRC. The leakage attributed to the flaw ervice as a result otingg TS 6.8.4.

.2 havebe*en ýexemted from the 1.0 gpm limit slaff. For other accidents that assume a faulted SG (e.g., feedwater line break), the maximum primary to secondary leakage under accident conditions is limited to 1.0 gpm from the faulted SG and 0.1 gpm from each of the non-faulted SGs. For accidents in which there are no faulted SGs, the primary to secondary leakage is limited to 0.1 gpm from each SG. 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 3.4.8, "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 119 (Ref. 2). and 10 CFR 100 (Ref. 3).

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

_VýD-ased Alternate Repair Criteria (ARC) and W* Methodology a)

Voltage-Based A The voltage-based repair plement nce in Generic Letter (GL) 95-05 a applicable n o Westinghouse-e steam ors (SGs) with outside diameter stress corrosion crackina SEQUOYAH - UNIT 2 B 3/4 4-3a May22, 2v-Amendment No. 181, 211, 213, 243, 267, 291,305,

Reactor Coolant System B 3/4.4.5 REACTOR COOLANT SYSTEM BASES APPLICABLE SAFETY ANALYSIS (continued)

(ODSCC) located at the tube-to-tube support plate intersections. The oltage-based repair limits are not applicable to other forms of SG tube d

radation nor are they applicable to ODSCC that occurs at other loca 'ons within the SG. Additionally, the repair criteria apply only to indica'ons where the degradation mechanism is dominantly axial SCC with no *gnificant cracks extending outside the thickness of the pport plate. Re r to GL 95-05 for additional description of the degr ation morphology.

Implementation f voltage-based repair limits require a d rivation of the voltage structural iit from the burst versus voltage e irical correlation and then the subse ent derivation of the voltage re ir limit from the structural limit (whic then implemented by this rveillance).

The voltage structural limi is the voltage from e burst pressure/bobbin voltage correlation, at the 9 percent predicti n interval curve reduced to account for the lower 95/95 p cent tolera e bound for tubing material properties at 650OF (i.e., the 95 ercent I er tolerance limit curve). The voltage structural limit must be a ste downward to account for potential flaw growth during an ope ing interval and to account for NDE uncertainty. The upper voltage re i 'mit; VURL, is determined from the structural voltage limit by applin the f owing equation:

VURL :-- VSL - V R - VNDE where VGR represents e allowance for flaw gro between inspections and VNDE represents e allowance for potential so ces of error in the measurement of th bobbin coil voltage. Further disc ssion of the assumptions ne ssary to determine the voltage repair imit are discussed in G 95-05.

Themid-c e equation of TS 6.8.4.k.c.1.e should only be us during unplann inspection in which eddy current data is acquired for indicati ns at the tube support plates.

Sp cification 6.9.1.16.3 implements several reporting requirements r commended by GL 95-05 for situations which NRC wants to be notifi nior to returning the SGs to service. For 6.9.1.16.3 Items 2 and 3, May 22, 2097 SEQUOYAH - UNIT 2 B 3/4 4-3b Amendment No. 181, 211, 213, 243, 267, 291,305,

Reactor Coolant System B 3/4.4.5 REACTOR COOLANT SYSTEM BASES APPLICABLE SAFETY ANALYSIS (continued) are applicable only where alternate plugging criteria is being applied. For t e purposes of this reporting requirement, leakage and conditional burs pr ability can be calculated based on the as-found voltage distributiot F

rath than the projected end-of-cycle (EOC) voltage distribution (refer to GL 95-5 for more information) when it is not practical to complete ese calculati s using the projected EOC voltage distributions prior t b"

returning t SGs to service. Note that if leakage and conditio al burst probability w re calculated using the measured EOC voltage istribution for the purpos of addressing GL Sections 6.a. 1 and 6.a. reporting criteria, then the esults of the projected EOC voltage di

ibution should be provided per G Section 6.b(c) criteria.

For the operational as ssment, the Probability of rior Cycle Detection (POPCD) voltage based robability of detection OD) method, as approved by NRC letter da d March 24, 2008 s used to determine the beginning of cycle voltage di ributions. The OPCD method is an exception to the GL 95-05 gui nce that r uires the application of a POD of 0.6 to all previous bobbi indicat ns.

Tubes experiencing ODSCC within t thickness of the tube support plate are plugged by the criteria of 6.8.4. 0c.

b)

W* Methodology The W* criteria incorporate the guidance prov ed in WCAP-14797, Revision 2, 'Generic W*

ube Plugging Criteria r 51 Series Steam Generator Tubesheet gion WEXTEX Expansion." W* length is the length of tubing into t e tubesheet below the bottom f the WEXTEX transition (BWT) th precludes tube pullout in the eve of a complete circumferential s aration of the tube below the W* leng

. W* distance is the distance omn the top-of-tube sheet (TTS) to the bott of the W*

length includi g the distance from the TTS to the BWT and asurement uncertainti Indicati ns detected within the W* distance below the TTS, will be plug dupon detection. Tubes to which WCAP-1 4797 is applied ca epy rience through-wall degradation up to the limits defined in Revisio 2 w out increasing the probability of a tube rupture or large leakage even ube degradation of any type or extent below W* distance, March 24,20-082 SEQUOYAH - UNIT 2 B 3/4 4-3c Amendment No. 181, 211, 213, 243, 267, 291, 305, 309,

Reactor Coolant System B 3/4.4.5 REACTOR COOLANT SYSTEM BASES APPLICABLE SAFETY ANALYSIS (continued) including a complete circumferential separation of the tube, is acceptable.

applied at Sequoyah Nuclear Plant Unit 2, the W* methodology is us to define the required tube inspection depth into the hot-leg and ld-leg tubesheet, and is not used to permit degradations03n in the

distance to remain in service. Thus while primary to secon/ry leakage the W* distance need not be postulated, primary to seco dary leakage fr potential degradation below the W* distance will begc1 assumed for very inservice tube in the bounding SG.

c)

Calculatio of Operational Assessment (OA) Accident duced Leakage

/

The postulated leakag during a Steam Line Break (S B) shall be equal to the following equation Postulated SLB OA Leakage ARC GL 95-05 + A sumed Leakage 0"-8"<TTS +

Assumed Leakage 8"-12-<TTS +

sumed Leak ge >ý12" <Trs + All other sources of accident induced prim ry to sec ndary leakage.

Where: ARC GL 95-05 is the SLB OA I ka e for predominantly axially oriented outside diameter stress corr on cracking indications as determined from the methodology d c ed in GL 95-05 as revised by Technical Specification Change 0 6.

Assumed Leakage 0.-8. cT-s is t postulated leakage for undetected indications in SG tubes left i service between and 8 inches below the TTS for both the hot-leg a cold-leg tubesheet.

Assumed Leakage 8"- <-s is the conservatively ass ed OA leakage from the total of iden ied and postulated unidentified in ications in SG tubes left in servic between 8 and 12 inches below the S for both the hot-leg and cold-g tubesheet. This is 0.0045 gpm multipli d by the number of mdi tions. Postulated unidentified indications wil e

conservative assumed to be in one SG. The highest number f identified i ications left in service between 8 and 12 inches belo TTS in any one G will be included in this term.

Ass med Leakage >12' <TTS is the conservatively assumed OA leakage fo

°thbohund*IngSGatnubeslelfte inserviebtelow 12 i nche

0below t hmeuTTiS for \\

oth the hot-leg and cold-leg tubesheet. This is 0.00009 gpm multiplied by the number of tubes left in service in the least plugged SG. When no October 19, 20-09 SEQUOYAH - UNIT 2 B 3/4 4-3d Amendment No. 181, 211, 213, 243, 267, 291,305, 309, 318,

Reactor Coolant System B 3/4.4.5 REACTOR COOLANT SYSTEM BASES PWSCC tube indications are identified in the cold-leg tubesheet region B Scold-leg O A leakage is 0.0 gpm.

All ot r sources of accident induced primary to secondary leakage i the prima o secondary accident induced OA leakage from all other degradat n mechanisms other than the voltage based axial ODS C at tube supp plates repair criteria and W* leakage calculations determined the Operational Assessment.

d)

Calculati of Condition Monitoring (CM) Accident Ind ced Leakage The postulated lea ge during a SLB shall be equal to e following equation and is perfo ed for each steam generator:

Postulated SLB CM Leak ge = ARC GL 95-05 + Ass med Leakage 0.-8

<TTS

+

Assumed Leakage 8-rs + Assumed Leakag

>12- <Trs + All other sources of accident induced rimary to secon ry leakage.

Where: ARC GL95-05 is the SLB M leakage or predominantly axially oriented outside diameter stress orrosio cracking indications as determined from the methodology escr' ed in GL 95-05 as revised by Technical Specification Change

- 6.

Assumed Leakage 0--8".cT~s is the p tul ed CM leakage for indications detected in SG tubes between 0 d 8 in es below the TTS for both the hot-leg and cold-le ubesheet.

Assumed Leakage 8'-12" <TT~S the conservative assumed CM leakage from the total of identified d postulated uniden ffied indications in SG tubes left in service betw en 8 and 12 inches belo the UTS for both the hot-leg and cold-leg tu sheet. This is 0.0045 gpm ultiplied by the number of indications Assumed Leakag

>12"<TT is the conservatively assume CM leakage for the bounding S ubes in service 12 inches below the TTfr both the hot-leg and col -leg tubesheet. This is 0.00009 gpm multipl d by the number of tu s left in service in the SG. When no PWSCC t e

indications e identified in the cold-leg tubesheet region the co -leg CM leakage i

.0 gpm.

All ot r sources of accident induced primary to secondary leakage is he pri ry to secondary accident induced CM leakage from all other de radation mechanisms other than the voltage based axial ODSCC at t

e support plates repair criteria and W* leakage calculations as determined by Condition Monitoring.

October 19, 2009 SEQUOYAH - UNIT 2 B 3/4 4-3e Amendment No. 181, 211, 213, 243, 267, 291,305, 309, 318,

Reactor Coolant System B 3/4.4.5 REACTOR COOLANT SYSTEM BASES APPLICABLE SAFETY ANALYSIS (continued)

T ate calculated accident induced primary to seconda leakage from lication of all approved ARC (W*

age-based axial ODSCC at TSP) s reported to th in accordance with Technical Specification 6.9.1.16.

ombined calculated leak rate from all ARC and all oth ces of acci uced leakage must be less than the induced primary to secondary e rate ass the SLB accident analyses.

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

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

In the context of this specification, a 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.

A SG tube has tube integrity when it satisfies the SG performance criteria.

The SG performance criteria are defined in Specification 6.8.4.k "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.

There are three SG performance criteria: structural integrity, accident induced leakage, and operational leakage. Failure to meet any one of these criteria is considered failure to meet the LCO.

The structural integrity performance criterion provides a margin of safety against tube burst or collapse under normal and accident conditions, and ensures structural integrity of the SG tubes under all anticipated transients included in the design specification. Tube burst is defined as, "The gross structural failure of the tube wall. The condition typically corresponds to an unstable opening displacement (e.g., opening area increased in response to constant pressure) accompanied by ductile October 19, 2009 SEQUOYAH - UNIT 2 B 3/4 4-3f Amendment No. 181, 211, 213, 243, 267, 291,305,

Reactor Coolant System B 3/4.4.5 REACTOR COOLANT SYSTEM BASES LCO (continued)

(plastic) tearing of the tube material at the ends of the degradation." Tube collapse is defined as, "For the load displacement curve for a given structure, collapse occurs at the top of the load versus displacement curve where the slope of the curve becomes zero." The structural integrity performance criterion provides guidance on assessing loads that have a significant effect on burst or collapse. In that context, the term "significant" is defined as "An accident loading condition other than differential pressure is considered significant when the addition of such loads in the assessment of the structural integrity performance criterion could cause a lower structural limit or limiting burst/collapse condition to be established." For tube integrity evaluations, except for circumferential degradation, axial thermal loads are classified as secondary loads. For circumferential degradation, the classification of axial thermal loads as primary or secondary loads will be evaluated on a case-by-case basis.

The division between primary and secondary classifications will be based on detailed analysis and/or testing.

Structural integrity requires that the primary membrane stress intensity in a tube not exceed the yield strength for all American Society of Mechanical Engineers (ASME) Code,Section III, Service Level A (normal operating conditions), and Service Level B (upset or abnormal conditions) transients included in the design specification. This includes safety factors and applicable design basis loads based on ASME Code, Section IIl, Subsection NB (Ref. 4) and Draft Regulatory Guide 1.121 (Ref. 5).

The accident induced leakage performance criterion ensures that the primary to secondary leakage caused by a design basis accident, other than a SGTR, is within the accident analysis assumptions. The accident Applicable analyses assumptions are discussed in thel*

t Safety Analyses section. The accident induced leakage rate includes any primary to secondary leakage existing prior to the accident in addition to primary to secondary leakage induced during the accident.

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 3.4.6.2, "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 a SGTR under the stress conditions of a loss-of-coolant accident (LOCA) or a MSLB. If this amount of leakage is due to more than one crack, the cracks are very small, and the above assumption is conservative.

Octobor 19, 2009 SEQUOYAH - UNIT 2 B 3/4 4-3g Amendment No. 181, 211, 213, 243, 267, 291,305,

Reactor Coolant System B 3/4.4.5 REACTOR COOLANT SYSTEM BASES ACTIONS (continued)

If the evaluation determines that the affected tube(s) have tube integrity, Action (a) allows plant operation to continue until the next refueling outage or SG inspection provided the inspection interval continues to be supported by an operational assessment that reflects the affected tubes.

However, the affected tube(s) must be plugged prior to startup following the next refueling outage or SG inspection. This allowed time is acceptable since operation until the next inspection is supported by the operational assessment.

m" intane

y SG tu e

P",ev the

R dVt

-q0

-q At'/l' nh w

d c'

1A#7 be brought to HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months and the affected tube(s) plugged prior to restart The action times are reasonable, based on operating experience, to reach the desired plant condition from full power in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 4.4.5.0 REQUIREMENTS During shutdown periods the SGs are inspected as required by this SR and the Steam Generator Program. NEI 97-06, Steam Generator Program Guidelines (Ref. 1), and its referenced EPRI Guidelines, establish the content of the Steam Generator Program. Use of the Steam Generator Program ensures that the inspection is appropriate and consistent with accepted industry practices.

During SG inspections a condition monitoring assessment of the SG tubes is performed. The condition monitoring assessment determines the "as found" condition of the SG tubes: The purpose of the condition monitoring assessment is to ensure that the SG performance criteria have been met for the previous operating period.

The Steam Generator Program determines the scope of the inspection and the methods used to determine whether the tubes contain flaws satisfying the tube repair 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 May 22, 20(9 SEQUOYAH - UNIT 2 B 3/4 4-3i Amendment No. 181, 211, 213, 243, 267, 291,305,

Reactor Coolant System B 3/4.4.5 REACTOR COOLANT SYSTEM BASES SURVEILLANCE REQUIREMENTS (continued) specifies the inspection methods to be used to find potential degradation.

Inspection methods are a function of degradation morphology, nondestructive examination (NDE) technique capabilities, and inspection locations.

The Steam Generator Program defines the frequency of SR 4.4.5.0. 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 6.8.4.k contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections.

SR 4.4.5.1 During an SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by plugging.

The tube repair criteria delineated in Specification 6.8.4.k 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 repair 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 this surveillance 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 I.,

r-+.,RWT 1=

,TN

Nng a SG W May 22, 29OO-SEQUOYAH - UNIT 2 B 3/4 4-3j Amendment No. 181, 211, 213, 243, 267, 291,305,

Reactor Coolant System B 3/4.4.5 REACTOR COOLANT SYSTEM BASES REFERENCES

1.

NEI 97-06, "Steam Generator Program Guidelines."

2.

10 CFR 50 Appendix A, GDC 19.

3.

10CFR100.

4.

ASME Boiler and Pressure Vessel Code,Section III, Subsection NB.

5.

Draft Regulatory Guide 1.121, "Basis for Plugging Degraded Steam Generator Tubes," August 1976.

6.

EPRI, "Pressurized Water Reactor Steam Generator Examination

(' iidrlpinpq " --NRC Generic Letter 95-05, "Voltage Based Repair Criteria for c

Ba s

de e a g

Weikhigbouse Steam Generator Tubes Affected by Outsidd Is Diameter Streas-Corrosion Cracking."

by si echn a

8.

NRC letter to TVA-dated April 9, 1997, "Issu of Technical ai r

urit 99 0

1 Specification Amendme fý tph e S ah Nuclear Plant, Units 1 ri 6ý e Se ah Nuclea Plant' U n it:

'in and 2 (TAC Nos. M9 a-

=

9 (TS 96-05)."

for

9.

NRC letter to TVA dat "ay 3,,2005, 0 ah Nuclear Plant,

'g1 997 su of T ic Unit 2 - lssýý FAPmendmEt. Regarding Ch s to the

ýnspec it' cope for the Steam Generator Tubes (TA MC5212) ct'3-06)."

SEQUOYAH - UNIT 2 B 3/4 4-3k May-22, 2097 Amendment No. 181, 211, 213, 243, 267, 291,305,

REACTOR COOLANT SYSTEM BASES APPLICABLE SAFETY ANALYSES (continued)

Primary to secondary leakage is a factor in the dose releases outside containment resulting from a steam generator tube rupture or a steam line break (SLB) accident. To a lesser extent, other accidents or transients also involve secondary steam release to the atmosphere. The leakage contaminates the secondary fluid.

The FSAR (Ref. 3) analysis for steam generator tube rupture (SGTR) assumes the contaminated secondary fluid is released via safety valves for up to 30 minutes. Operator action is taken to isolate the affected steam generator within this time period. The 0.4 gpm operational primary to secondary leakage safety analysis assumption is relatively inconsequential.

The SLBlw~th

,Ri ap,,lglIs more limiting for site radiation releases.

The safety analysis for the SLB accident assumes a 3.7 gpm primary to secondary leakage through the affected generator and 0.3 gpm through the non-affected generators as an initial condition. The dose consequences resulting from the SLB accident are well within the limits defined in 10 CFR 100 or the staff approved licensing basis (i.e., a small fraction of these limits). Ba DE uncertainties, bobbin coil voltage distribution and crack rate Trom the pr inspection, the expected leak rate followi eam line rupture is limited to

.7 gpm at atmospheric ions and 70OF in the Faulted loop, which will limit t lated doses to within 10 percent of he 10 CFR 100 guidelines. If the d cycle distribution of crack ndications results in pri secondary leaiZag ter than 3.7 gpm in the aulted loop dun ostulated steam line break event, a Is tubes must be emove service in order to reduce the postulated primary-to-e r

m line break leakage to below 3.7 gm.

The RCS operational leakage satisfies Criterion 2 of the NRC Policy Statement.

LCO RCS operational leakage shall be limited to:

a.

PRESSURE BOUNDARY LEAKAGE No PRESSURE BOUNDARY LEAKAGE is allowed, being indicative of material deterioration. Leakage of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher leakage.

Violation of this LCO could result in continued degradation of the RCPB.

Leakage past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE.

b.

UNIDENTIFIED LEAKAGE One gpm of UNIDENTIFIED LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment pocket MayS 22, 2500 SEQUOYAH - UNIT 2 B 3/4 4-4f Amendment No. 211, 213, 227, 250, 305,

ATTACHMENT 3 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT UNIT 2 PROPOSED TS CHANGES (Final Typed)

ADMINISTRATIVE CONTROLS

d.

Proposed changes that meet the criteria of Specification 6.8.4.j.b above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

k.

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 provisions:

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.

Provisions for 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 SG tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, cooldown, and all anticipated transients included in the design specification) and design basis accidents (DBAs). 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 DBA primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the DBAs, 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 accident-induced leakage is not to exceed 1.0 gpm for the faulted SG. The primary-to-secondary accident induced leakage rate for any DBA, 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.

3.

The operational leakage performance criterion is specified in Limiting Condition for Operation (LCO) 3.4.6.2, "Reactor Coolant System, Operational Leakage."

SEQUOYAH - UNIT 2 6-1Oa Amendment No. 28, 50, 64, 66, 134, 165, 202, 207, 223,231,265, 271,272, 276, 298, 305,

ADMINISTRATIVE CONTROLS

c.

Provisions for SG Tube Repair 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 repair 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 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 replacement.

2.

Inspect 100% of the tubes at sequential periods of 144, 108, 72 and thereafter, 60 effective 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 remaining 50% by the refueling outage nearest the end of the period. No SGs shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected.

3.

If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever 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.

1.

Component Cyclic and Transient Limit This program provides controls to track the FSAR, Section 5.2.1, cyclic and transient occurrences to ensure that components are maintained within the design limits.

SEQUOYAH - UNIT 2 6-10Ob Amendment No. 305, 315, 318,

ADMINISTRATIVE CONTROLS STEAM GENERATOR (SG) TUBE INSPECTION REPORT (continued)

a.

The scope of inspections performed on each SG,

b.

Active degradation 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 active degradation mechanism,

f.

Total number and percentage of tubes plugged to date,

g.

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

h. The effective plugging percentage for all plugging in each SG.

SEQUOYAH - UNIT 2 6-14a Amendment No. 305,

ADMINISTRATIVE CONTROLS SPECIAL REPORTS 6.9.2.1 Special reports shall be submitted within the time period specified for each report, in accordance with 10 CFR 50.4.

6.9.2.2 This specification has been deleted.

6.10 RECORD RETENTION (DELETED)

SEQUOYAH - UNIT 2 6-15 Amendment No. 28, 44, 50, 64, 66, 107, 134,146,153,165,169,206,214,223,231, 249,284,309,

ATTACHMENT 4 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT UNIT 2 PROPOSED BASES CHANGES (Final Typed)

Reactor Coolant System B 3/4.4.5 REACTOR COOLANT SYSTEM 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 specification.

ANALYSES The analysis of an SGTR event assumes a bounding primary to secondary leakage rate equal to the operational leakage rate limits in LCO 3.4.6.2 "Operational Leakage," plus the leakage rate associated with a double-ended rupture of a single tube. The accident analysis for a SGTR assumes the contaminated secondary fluid is released to the atmosphere via safety valves.

The main condenser isolates based on an assumed concurrent loss of off-site power.

The analysis for design basis accidents and transients other than a SGTR assume the SG tubes retain their structural integrity (i.e., they are assumed not to rupture). In these analyses, the steam discharge to the atmosphere depends on the accident and whether there are faulted SGs associated with the accident.

For a steamline break (SLB), the maximum primary to secondary leakage under accident conditions is limited to 3.7 gpm from the faulted SG and 0.1 gpm from each of the non-faulted SGs. For other accidents that assume a faulted SG (e.g.,

feedwater line break), the maximum primary to secondary leakage under accident conditions is limited to 1.0 gpm from the faulted SG and 0.1 gpm from each of the non-faulted SGs. For accidents in which there are no faulted SGs, the primary to secondary leakage is limited to 0.1 gpm from each SG. 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 3.4.8, "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) and 10 CFR 100 (Ref. 3).

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 repair criteria be plugged in accordance with the Steam Generator Program.

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

In the context of this specification, a 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.

A SG tube has tube integrity when it satisfies the SG performance criteria. The SG performance criteria are defined in Specification 6.8.4.k "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.

SEQUOYAH - UNIT 2 B 3/4 4-3a Amendment No. 181, 211, 213, 243, 267, 291, 305, 309, 318,

Reactor Coolant System B 3/4.4.5 REACTOR COOLANT SYSTEM BASES LCO (continued)

There are three SG performance criteria: structural integrity, accident induced leakage, and operational leakage. Failure to meet any one of these criteria is considered failure to meet the LCO.

The structural integrity performance criterion provides a margin of safety against tube burst or collapse under normal and accident conditions, and ensures structural integrity of the SG tubes under all anticipated transients included in the design specification. Tube burst is defined as, "The gross structural failure of the tube wall. The condition typically corresponds to an unstable opening displacement (e.g., opening area increased in response to constant pressure) accompanied by ductile (plastic) tearing of the tube material at the ends of the degradation." Tube collapse is defined as, "For the load displacement curve for a given structure, collapse occurs at the top of the load versus displacement curve where the slope of the curve becomes zero." The structural integrity performance criterion provides guidance on assessing loads that have a significant effect on burst or collapse. In that context, the term "significant" is defined as "An accident loading condition other than differential pressure is considered significant when the addition of such loads in the assessment of the structural integrity performance criterion could cause a lower structural limit or limiting burst/collapse condition to be established." For tube integrity evaluations, except for circumferential degradation, axial thermal loads are classified as secondary loads. For circumferential degradation, the classification of axial thermal loads as primary or secondary loads will be evaluated on a case-by-case basis.

The division between primary and secondary classifications will be based on detailed analysis and/or testing.

Structural integrity requires that the primary membrane stress intensity in a tube not exceed the yield strength for all American Society of Mechanical Engineers (ASME) Code,Section III, Service Level A (normal operating conditions), and Service Level B (upset or abnormal conditions) transients included in the design specification. This includes safety factors and applicable design basis loads based on ASME Code, Section Ill, Subsection NB (Ref. 4) and Draft Regulatory Guide 1.121 (Ref. 5).

SEQUOYAH - UNIT 2 B 3/4 4-3b Amendment No. 181, 211, 213, 243, 267, 291,305,

Reactor Coolant System B 3/4.4.5 REACTOR COOLANT SYSTEM BASES LCO (continued)

The accident induced leakage performance criterion ensures that the primary to secondary leakage caused by a design basis accident, other than a SGTR, is within the accident analysis assumptions. The accident analyses assumptions are discussed in the Applicable Safety Analyses section. The accident induced leakage rate includes any primary to secondary leakage existing prior to the accident in addition to primary to secondary leakage induced during the accident.

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 3.4.6.2, "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 a SGTR under the stress conditions of a loss-of-coolant accident (LOCA) or a SLB. 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 MODES 1, 2, 3, or 4.

Reactor coolant system (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 clarifying footnote that Action (a) may be entered independently for each SG tube. This is acceptable because the actions provide appropriate compensatory measures for each affected SG tube. Complying with the actions may allow for continued operation, and subsequent affected SG tubes are governed by subsequent action entry, and application of associated actions.

SEQUOYAH - UNIT 2 B 3/4 4-3c Amendment No. 181, 211, 213, 243, 267, 291,305,

Reactor Coolant System B 3/4.4.5 REACTOR COOLANT SYSTEM BASES ACTIONS (continued)

Actions (a) and (b)

Action (a) applies if it is discovered that one or more SG tubes examined in an inservice inspection satisfy the tube repair criteria but were not plugged in accordance with the Steam Generator Program as required by SR 4.4.5.1. 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 repair 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 a 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 refueling outage or SG tube inspection. If it is determined that tube integrity is not being maintained until the next SG inspection, Action (a) requires unit shutdown and Action (b) requires the affected tube(s) be plugged.

An allowed time of 7 days is sufficient to complete the evaluation while minimizing the risk of plant operation with a SG tube that may not have tube integrity.

If the evaluation determines that the affected tube(s) have tube integrity, Action (a) allows plant operation to continue until the next refueling outage or SG inspection provided the inspection interval continues to be supported by an operational assessment that reflects the affected tubes.

However, the affected tube(s) must be plugged prior to startup following the next refueling outage or SG inspection. This allowed time is acceptable since operation until the next inspection is supported by the operational assessment.

If SG tube integrity is not being maintained, the reactor must be brought to HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months and the affected tube(s) plugged prior to restart.

The action times are reasonable, based on operating experience, to reach the desired plant condition from full power in an orderly manner and without challenging plant systems.

SEQUOYAH - UNIT 2 B 3/4 4-3d Amendment No. 181, 211, 213, 243, 267, 291,305,

Reactor Coolant System B 3/4.4.5 REACTOR COOLANT SYSTEM BASES SURVEILLANCE SR 4.4.5.0 REQUIREMENTS During shutdown periods the SGs are inspected as required by this SR and the Steam Generator Program. NEI 97-06, Steam Generator Program Guidelines (Ref. 1), and its referenced EPRI Guidelines, establish the content of the Steam Generator Program. Use of the Steam Generator Program ensures that the inspection is appropriate and consistent with accepted industry practices.

During SG inspections a condition monitoring assessment of the SG tubes is performed. The condition monitoring assessment determines the "as found" condition of the SG tubes. The purpose of the condition monitoring assessment is to ensure that the SG performance criteria have been met for the previous operating period.

The Steam Generator Program determines the scope of the inspection and the methods used to determine whether the tubes contain flaws satisfying the tube repair 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, nondestructive examination (NDE) technique capabilities, and inspection locations.

The Steam Generator Program defines the frequency of SR 4.4.5.0. 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 6.8.4.k contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections.

SEQUOYAH - UNIT 2 B 3/4 4-3e Amendment No. 181, 211, 213, 243, 267, 291,305,

Reactor Coolant System B 3/4.4.5 REACTOR COOLANT SYSTEM BASES SURVEILLANCE REQUIREMENTS (continued)

SR 4.4.5.1 During an SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by plugging.

The tube repair criteria delineated in Specification 6.8.4.k 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 repair 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 this surveillance 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.

REFERENCES

1.

NEI 97-06, "Steam Generator Program Guidelines."

2.

10 CFR 50 Appendix A, GDC 19.

3.

10CFR100.

4.

ASME Boiler and Pressure Vessel Code,Section III, Subsection NB.

5.

Draft Regulatory Guide 1.121, "Basis for Plugging Degraded Steam Generator Tubes," August 1976.

6.

EPRI, "Pressurized Water Reactor Steam Generator Examination Guidelines."

SEQUOYAH - UNIT 2 B 3/4 4-3f Amendment No. 181, 211, 213, 243, 267, 291,305,

Reactor Coolant System B 3/4.4.5 REACTOR COOLANT SYSTEM BASES APPLICABLE SAFETY ANALYSES (continued)

Primary to secondary leakage is a factor in the dose releases outside containment resulting from a steam generator tube rupture or a steam line break (SLB) accident. To a lesser extent, other accidents or transients also involve secondary steam release to the atmosphere. The leakage contaminates the secondary fluid.

The FSAR (Ref. 3) analysis for steam generator tube rupture (SGTR) assumes the contaminated secondary fluid is released via safety valves for up to 30 minutes. Operator action is taken to isolate the affected steam generator within this time period. The 0.4 gpm operational primary to secondary leakage safety analysis assumption is relatively inconsequential.

The SLB is more limiting for site radiation releases. The safety analysis for the SLB accident assumes a 3.7 gpm primary to secondary leakage through the affected generator and 0.3 gpm through the non-affected generators as an initial condition. The dose consequences resulting from the SLB accident are well within the limits defined in 10 CFR 100 or the staff approved licensing basis (i.e.,

a small fraction of these limits).

The RCS operational leakage satisfies Criterion 2 of the NRC Policy Statement.

LCO RCS operational leakage shall be limited to:

a.

PRESSURE BOUNDARY LEAKAGE No PRESSURE BOUNDARY LEAKAGE is allowed, being indicative of material deterioration. Leakage of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher leakage.

Violation of this LCO could result in continued degradation of the RCPB.

Leakage past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE.

b.

UNIDENTIFIED LEAKAGE One gpm of UNIDENTIFIED LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment pocket SEQUOYAH - UNIT 2 B 3/4 4-4f Amendment No. 211, 213, 227, 250, 305,

v t e Letter Sequence Request
TAC:MD0145, Steam Generator Tube Integrity (Approved, Closed)
Initiation Request, Request Acceptance... Supplement Administration Questions 6 June 2006 7 November 2006 28 February 2007 Answers 7 August 2006 Results Approval, Approval Other:
MONTHYEAR2006-08-30 [Table View]

Person / Time
ML11199A212
Site:	Sequoyah (DPR-079)
Issue date:	07/15/2011
From:	Krich R Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TAC MD0145
Download: ML11199A212 (43)
v • d • e

ML11199A212

Text

Subject:

Reference:

Enclosure:

Subject:

SUMMARY

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

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