ML11298A081

Response to NRC Request for Additional Information Regarding Application to Modify Technical Specifications for Replacement Steam Generators (TS-SQN-2011-01)

ML11298A081
Person / Time
Site:	Sequoyah
Issue date:	10/20/2011
From:	Krich R Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TS-SQN-2011-01
Download: ML11298A081 (19)

Text

Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402 R. M. Krich Vice President, Nuclear Licensing October 20, 2011 10 CFR 50.4 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:

Response to NRC Request for Additional Information Regarding Application to Modify Technical Specifications for Replacement Steam Generators (TS-SQN-201 1-01)

References:

1. Letter from TVA to NRC, "Application to Modify Technical Specifications for Replacement Steam Generators (TS-SQN-201 1-01)," dated July 15, 2011

2. Letter from NRC to TVA, "Sequoyah Nuclear Plant, Unit 2 - Request for Additional Information Regarding the Propopsed [sic] Technical Specification Changes for Replacement Steam Generator Tube Inspection Requirements (TAC No. ME6707)," dated October 14, 2011 By letter dated July 15, 2011 (Reference 1), the Tennessee Valley Authority (TVA) submitted a request for amendment to the Technical Specifications (TS) for Sequoyah Nuclear Plant (SQN),

Unit 2. The amendment request proposed to revise the TS inspection requirements for the steam generators to be installed during the SQN, Unit 2, refueling outage 18. In the Reference 2 letter, the NRC requested additional information regarding the proposed changes to the TS, and that a response be submitted within 30 days from the date of the letter. Enclosure 1 to this letter provides the response to the NRC request for additional information. Based on this response, Enclosure 2 includes modified proposed TS and Bases pages that were previously provided in the Reference 1 letter. No other changes have been made to the enclosure and attachments provided in the Reference 1 letter.

printed on recycled paper

U.S. Nuclear Regulatory Commission Page 2 October 20, 2011 TVA has determined that the information provided by this letter does not affect the no significant hazards considerations associated with the proposed TS changes provided in the reference letter. The proposed TS changes still qualify for a categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22(c)(9). Additionally, in accordance with 10 CFR 50.91(b)(1), TVA is sending a copy of this letter and the enclosures to the Tennessee State Department of Environment and Conservation.

There are no regulatory commitments associated with this submittal. Please address any questions regarding this request to Clyde Mackaman at 423-751-2834.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 20th day of October 2011.

Respectfully, R. M. Krich

Enclosures:

1. Response to Request for Additional Information Regarding Application to Modify Technical Specifications for Replacement Steam Generators (TS-SQN-201 1-01)

2. Modified Proposed Technical Specifications and Bases Pages cc (Enclosures):

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

ENCLOSURE 1 Response to Request for Additional Information Regarding Application to Modify Technical Specifications for Replacement Steam Generators (TS-SQN-201 1-01)

Response to Request for Additional Information Regarding Application to Modify Technical Specifications for Replacement Steam Generators (TS-SQN-201 1-01)

NRC Request for Additional Information (RAI)

By letter dated July 15, 2011 (Agencywide Document Access and Management System Accession No. MLII 199A212), Tennessee Valley Authority (the licensee),

submitted a license amendment request regardingproposedchanges to the technical specifications (TSs) to reflect steam generator(SG) replacement for Sequoyah Nuclear Plant, Unit 2. In orderto complete its review, the Nuclear Regulatory Commission (NRC) staff needs the following additionalinformation:

1. The proposedmodification to TS 6.8.4.k.b. 2 does not match the wording in TSTF-449 since the proposal indicatesthat accident induced leakage is not to exceed 1.0 gpm for the faulted steam generator(emphasis added), whereas TSTF-449 simply indicates that the accident induced leakage is not to exceed 1.0 gpm per steam generator. Please either modify your proposalto remove this discrepancyorjustify this discrepancy. The reason for this limit is discussed in RIS 2007-20, "Implementation of Primary-to-SecondaryLeakage Performance Criteria.

2. Changes to the Technical Specification Bases were provided for information only, however, the NRC staff noticed that several of the changes appearto go beyond what would be needed to reflect the replacementof the steam generators. In fact, some of the changes appearto remove changes made during the NRC staff's review of your submittal adopting TSTF-449. Please clarify these changes, which include the following:

a. Bases page B 3/4 4-3a: deletion of "orthe NRC approved licensing basis"

b. Bases page B 3/4 4-3i. deletion of "(Mode 4)"

c. Bases page B 3/4 4-3j: deletion of "(i.e., priorto HOT SHUTDOWN following a SG tube inspection)"

d. Bases page B 3/4 4-4f. deletion of "the expected leak rate following a steam line rupture is limited to below 3.7 gpm at atmospheric conditionsand 70 degrees F in the faulted loop, which will limit the calculated offsite doses to within 10 percent of the 10 CFR 100 guidelines" El-1

Response

The changes proposed in the TS and Bases to reflect the SG replacement for Sequoyah Nuclear Plant (SQN), Unit 2, were modeled after the changes that were approved by the NRC for SQN, Unit 1, in Amendment No. 306, dated February 23, 2006. At that time, the SQN, Unit 1, SGs had already been replaced with the same SGs that are currently scheduled to be installed in SQN, Unit 2, in the fall of 2012. As such, the SQN, Unit 2, license amendment request was developed to minimize differences between the TS and Bases of the two units, where possible. Based on the NRC RAI, the proposed changes to the SQN, Unit 2, TS and Bases were reviewed and the resulting response to the RAI is provided below. The changes (added text) in response to the NRC RAI are underlined (for emphasis only).

• In response to RAI item 1, the proposed change to TS 6.8.4.k.b.2 is revised to reflect the SQN accident analysis assumptions and is consistent with current SQN, Unit 2, TS 6.8.4.k.b.2 with respect to accident-induced leakage through intact SGs. Specifically, steam generator leakage through the intact SGs during a Steam Line Break accident is assumed to be 150 gallons per day per intact SG (i.e., 0.1 gallons per minute per intact SG). Therefore, TS 6.8.4.k.b.2 is revised to maintain the maximum allowable intact (or non-faulted) SG accident-induced leakage, consistent with the current SQN, Unit 2, TS.

2. Accident induced leakage performance criterion: The accident-induced leakage is not to exceed 1.0 gpm for the faulted SG and 0.1 .qmfor each of the non-faulted SGs. 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.

" With respect to RAI item 2.a (Bases page B 3/4 4-3a: deletion of "or the NRC approved licensing basis"), the proposed change to Bases B 3/4.4.5, "Steam Generator (SG) Tube Integrity," was made to reflect the SQN licensing basis for the dose consequence analyses associated with those design basis accidents and transients which include assumptions related to SG tube integrity.

Specifically, the phrase "or the NRC approved licensing basis" on page B 3/4 4-3a of the current SQN, Unit 2, Bases B 3/4.4.5 was proposed to be deleted. For SQN, the licensing basis dose consequence acceptance criteria for these events is provided in 10 CFR 50, Appendix A, General Design Criterion 19, "Control Room," and 10 CFR 100, "Reactor Site Criteria." There is no other "NRC approved licensing basis" for these events for SQN.

E1-2

• In response to RAI item 2.b (Bases page B 3/4 4-3i: deletion of "(Mode 4)"), the proposed change to Bases B 3/4.4.5, on current Bases page B 3/4 4-3i, is revised to maintain the phrase "(Mode 4)," consistent with the current SQN, Unit 2, Bases.

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

" In response to RAI item 2.c (Bases page B 3/4 4-3j: deletion of "(i.e., prior to HOT SHUTDOWN following a SG tube inspection)"), the proposed change to Bases B 3/4.4.5, on current Bases page B 3/4 4-3j, is revised to maintain the phrase

"(i.e., prior to HOT SHUTDOWN following a SG tube inspection)," consistent with the current SQN, Unit 2, Bases.

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.e., prior to HOT SHUTDOWN following a SG tube inspection).

" In response to RAI item 2.d (Bases page B 3/4 4-4f: deletion of "the expected leak rate following a steam line rupture is limited to below 3.7 gpm at atmospheric conditions and 70 degrees F in the faulted loop, which will limit the calculated offsite doses to within 10 percent of the 10 CFR 100 guidelines"), the proposed change to Bases B 3/4.4.6.2, "Operational Leakage," on current Bases page B 3/4 4-4f, is revised to maintain the phrase "the expected leak rate following a steam line rupture is limited to below 3.7 gpm at atmospheric conditions and 70 degrees F in the faulted loop, which will limit the calculated offsite doses to within 10 percent of the 10 CFR 100 guidelines," consistent with the current SQN, Unit 2, Bases.

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 expected leak rate following a steam line rupture is limited to below 3.7 qpm at atmospheric conditions and 70 deaqrees F in the faulted loop, which will limit the calculated offsite doses to within 10 percent of the 10 CFR 100 guidelines.

E1-3

ENCLOSURE 2 Modified Proposed Technical Specifications and Bases Pages The modified proposed Technical Specifications and Bases mark-ups and final typed pages provided in the attachments to this enclosure replace the corresponding pages provided in the attachments to the letter from TVA to NRC, "Application to Modify Technical Specifications for Replacement Steam Generators (TS-SQN-201 1-01)," dated July 15, 2011.

ATTACHMENT 1 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT UNIT 2 PROPOSED TS CHANGES (Mark-Ups)

Modified Page

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 3.0Pressure secondary against differential burst underandl normal steady

.... ep,-- ,,,, ,,state full

...NA .,, power, operation Iepd hM

.... primary-to--_ 11-, 1k ai terfate Fepar *r.Fter! disGc'-ed on- T-6 ' 6.8.4.k. G1 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.

For pmoreRto...etdotsd.iaee.trs orla

2. Accident induced leakage performance criterion: The accident-induced leakagen;a4--

is not to exceed 1.0 gpm for the faulted SG and 0.1 gpm for each of the non-faulted SGs.

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 6-10a Amendment No. 28, 50, 64, 66, 134, 165, 202, 207, 223, 231,265, 271,272, 276, 298, 305,

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

Modified Pages

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.

IfW .... to.m. ev

... lai,

,..+;.. r.m.8.+*,.,,

SG tube integ'rity is not being be brought to HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> and the affected tube(s) plugged prior to restart (Mode 4).

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,,

"27 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.e., prior to HOT SHUTDOWN following a SG tube inspection).

Mey 22, 2.*G SEQUOYAH - UNIT 2 B 3/4 4-3j 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 SLBwAth 1RCal; 1 A -- [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). Baseed em te peyiw6 mspetOO expected leak rate following a steam line Ei~EIIJ- rupture is limited to below 3.7 gpm at atmospheric conditions and 70°F in the faulted loop, which will limit the calculated offsite doses to within 10 percent of the 10 CFR 100 guidelines. IIf the r-" "jocted c','cYl dictd butino,'f.*..- ,.k 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 Mayt 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)

Modified Page

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 and 0.1 gpm for each of the non-faulted SGs. 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,

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

Modified Pages

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 refueling outage or 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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> and the affected tube(s) plugged prior to restart (Mode 4).

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 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 (i.e., prior to HOT SHUTDOWN following a SG tube inspection).

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 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 expected leak rate following a steam line rupture is limited to below 3.7 gpm at atmospheric conditions and 70 degrees F in the faulted loop, which will limit the calculated offsite doses to within 10 percent of thethe 10 CFR 100 guidelines.

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,