ML052500560

Unit 1 - Technical Specifications (TS) Change 05-07 - Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity

ML052500560
Person / Time
Site:	Sequoyah
Issue date:	08/31/2005
From:	Pace P Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TVA-SQN-05-07
Download: ML052500560 (43)

Text

i Tennessee Valley Authority, Post Office Box 2000, Soddy-Daisy, Tennessee 37384-2000 August 31, 2005 TVA-SQN-TS-05-07 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN:

Document Control Desk Washington, D. C. 20555-0001 Gentlemen:

In the Matter of

)

Docket No. 50-327 Tennessee Valley Authority SEQUOYAH NUCLEAR PLANT (SQN) -

UNIT 1 -

TECHNICAL SPECIFICATIONS (TS)

CHANGE 05 "APPLICATION FOR TECHNICAL SPECIFICATION IMPROVEMENT REGARDING STEAM GENERATOR TUBE INTEGRITY" Pursuant to 10 CFR 50.90, Tennessee Valley Authority (TVA) is submitting a request for a TS change (TS-05-07) to License DPR-77 for SQN Unit 1.

The proposed TS change revises the TS requirements related to steam generator tube integrity.

The change is consistent with NRC-approved Revision 4 to Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-449, "Steam Generator Tube Integrity."

The availability of this TS improvement was announced in the Federal Register on May 6, 2005 (70 FR 24126), as part of the consolidated line item improvement process (CLIIP). provides a description of the proposed change and confirmation of applicability. provides the existing TS pages marked-up to show the proposed change. provides the applicable TS Bases pages associated with the TS change.

Paced on recycW DWe

U.S. Nuclear Regulatory Commission Page 2 August 31, 2005 TVA requests approval of the proposed license amendment by February 2006, to support site preparations for the Unit 1 Cycle 14 refueling outage that is scheduled to start in March 2006.

The proposed TS change provides significant outage savings for TVA due to the replacement steam generators that were installed on Unit 1 during the Unit 1 Cycle 12 refueling outage.

TVA anticipates submitting a similar change request for SQN Unit 2 soon, although there is no current operational need.

TVA requests that implementation of the revised TS be within 45 days of NRC approval.

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).

Additionally, in accordance with 10 CFR 50.91(b)(1), TVA is sending a copy of this letter and enclosures to the Tennessee State Department of Public Health.

There are no commitments contained in this submittal.

If you have any questions about this change, please contact me at 843-7170 or Jim Smith at 843-6672.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on this 3 1stday of August, 2005.

Sincerel P. L.

Pace Manager, Site Licensing and Industry Affairs

Enclosures:

1.

TVA Evaluation of the Proposed Changes

2.

Proposed Technical Specifications Changes (mark-up)

3.

Changes to Technical Specifications Bases Pages cc:

See page 3

U.S. Nuclear Regulatory Commission Page 3 August 31, 2005 Enclosures cc (Enclosures):

Framatome ANP, Inc.

P. 0. Box 10935 Lynchburg, Virginia 24506-0935 ATTN:

Mr. Frank Masseth Mr. Lawrence E. Nanney, Director Division of Radiological Health Third Floor L&C Annex 401 Church Street Nashville, Tennessee 37243-1532 Mr. Douglas V. Pickett, Senior Project Manager U.S. Nuclear Regulatory Commission Mail Stop 08G9 One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2739

ENCLOSURE 1 TENNESSEE VALLEY AUTHORITY (TVA)

SEQUOYAH NUCLEAR PLANT (SQN)

UNIT 1

1.0 DESCRIPTION

This letter is a request to amend Operating License DPR-77 for SQN Unit 1. The proposed changes would revise the requirements in technical specifications (TSs) related to steam generator tube integrity.

The changes are consistent with NRC approved Technical Specification Task Force (TSTF)

Standard Technical Specification Change Traveler, TSTF-449, "Steam Generator Tube Integrity," Revision 4. The availability of this TS improvement was announced in the Federal Register on May 2, 2005, as part of the consolidated line item improvement process (CLIIP).

2.0 PROPOSED CHANGE

Consistent with NRC-approved Revision 4 of TSTF-449, the proposed TS changes include:

• Revised TS definition of "LEAKAGE"

• Revised TS 3.4.6.2, "Reactor Coolant System, Operational Leakage"

• Revised TS 3.4.5, "Steam Generator (SG) Tube Integrity"

• New TS Administrative Controls Section 6.8.k, "Steam Generator (SG) Program"

• New TS Administrative Controls Section 6.9.1.16, "Steam Generator Tube Inspection Report" Proposed revisions to the TS Bases are also included in this application. 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 improvement.

The changes to the affected TS Bases pages will be incorporated in accordance with the TS Bases Control Program.

3.0 BACKGROUND

The background for this application is 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.

The applicable regulatory requirements and guidance associated with this application are adequately addressed by El-l

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.

4.0 TECHNICAL ANALYSIS

TVA has reviewed the safety evaluation (SE) published on March 2, 2005 (70 FR 10298), as part of the CLIIP Notice for Comment.

This included the NRC staff's SE, the supporting information provided to support 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 staff are applicable to SQN Unit 1 and justify this amendment for the incorporation of the changes to the SQN TSs.

5.0 REGULATORY SAFETY ANALYSIS 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), the NRC Notice for Comment published on March 2, 2005 (70 FR 10298), and Technical Specification Task Force (TSTF) 449, Revision 4.

5.1 Verification and Commitments The following information is provided to support the NRC staff's review of this amendment application:

Plant Name, Unit No.

Sequoyah, Unit 1 Steam Generator (SG)

Model(s):

Effective Full Power Years (EFPY) of service for currently installed SGs Tubing Material (e.g.,

600M, 600TT, 660TT)

Number of tubes per SG Number and percentage of tubes plugged in each SG Westinghouse Model 57 AG 3.33 EFPY 690TT 4983 SG 1 SG 2 SG 3 SG 4 14 6

6 5

(0.28%) (0.12%) (0.12%) (0.10%)

E1-2

Plant Name, Unit No.

Sequoyah, Unit 1 Plant Name, Unit No.

Sequoyah, Unit 1 Number of tubes repaired in each SG N/A Degradation mechanism(s) identified Current primary-to-secondary leakage limits Approved Alternate Tube Repair Criteria (ARC)

Approved SG Tube Repair Methods Performance criteria for accident leakage U-Bend Support Wear 150 gallons per day per SG and (600 gallons per day total leakage from four SGs) leak rates are evaluated at 70 degrees Fahrenheit.

Not Applicable Not Applicable Primary-to-secondary leak rate values assumed in SQN's licensing basis accident analysis is 0.1 gpm for the non-faulted SGs and 3.7 gpm for the faulted SG (assumed at 70 degrees Fahrenheit temperature condition).

The 3.7 gpm leakage limit is the approved plant analyses and includes alternate repair criteria (ARC).

SQN Unit 1 SGs were replaced and no longer apply ARC (Reference 3). Accordingly, the accident induced leakage is conservatively limited to 1 gpm for the faulted SG.

5.2 No Significant Hazards Consideration TVA has reviewed the proposed no significant hazards consideration determination published on March 2, 2005 (70 FR 10298), as part of the consolidated line item improvement process (CLIIP).

TVA has concluded that the proposed determination presented in the notice is applicable to SQN and the determination is hereby incorporated by reference to satisfy the requirements of 10 CFR 50.91(a).

E1-3

6.0 ENVIRONMENTAL EVALUATION TVA has reviewed the environmental evaluation included in the model SE published on March 2, 2005 (70 FR 10298), as part of the CLIIP.

TVA has concluded that the staff's findings presented in that evaluation are applicable to SQN and the evaluation is hereby incorporated by reference for this application.

7.0 PRECEDENT This application is being made in accordance with the CLIIP.

TVA is not proposing variations or deviations from the TS changes described in TSTF-449, Revision 4, or the NRC staff's model SE published on March 2, 2005 (70 FR 10298).

8.0 REFERENCES

1. Federal Register Notice -

Notice for Comment published on March 2, 2005 (70 CFR 10298)

2. Notice of Availability published on May 6, 2005 (70 FR 24126)

3. NRC letter to TVA dated March 4, 2003, "Sequoyah Nuclear Plant, Unit 1 -

Issuance of Amendment Regarding Deletion of Alternate Repair Criteria Requirements from the Technical Specifications" E1-4

ENCLOSURE 2 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT (SQN)

UNIT 1 Proposed Technical Specification Changes (mark-up)

E2-1

b.

Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or

c.

Reactor coolant system leakage through a steam generator to the secondary system MEMBER(S) OF THE PUBLIC

/

I(primary to secondary leakage)

I 1.17 DELETED OFFSITE DOSE CALCULATION MANUAL (ODCM) 1.18 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm/trip setpoints, and in the conduct of the Radiological Environmental Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Specifications 6.9.1.6 and 6.9.1.8.

OPERABLE - OPERABILITY 1.19 A system, subsystem, train, or component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s), and when all necessary attendant instrumentation, controls, a normal and an emergency electrical power source, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function(s) are also capable of performing their related support function(s).

OPERATIONAL MODE - MODE 1.20 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.1.

PHYSICS TESTS 1.21 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14.0 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

SEQUOYAH - UNIT 1 1-4 February 11, 2003 Amendment No. 12, 71, 148, 155, 169, 174, 178, 281 E2-2

PRESSURE BOUNDARY LEAKAGE 1.22 PRESSURE BOUNDARY LEAKAGE shall be leakage (except stoam'gonerator tube leakage) through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall.

PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) 1.23 The PTLR is the unit specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates and the LTOP arming temperature, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 6.9.1.15.

PROCESS CONTROL PROGRAM (PCP) 1.24 DELETED PURGE - PURGING 1.25 PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

QUADRANT POWER TILT RATIO 1.26 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.

RATED THERMAL POWER (RTP) 1.27 RATED THERMAL POWER (RTP) shall be a total reactor core heat transfer rate to the reactor coolant of 3455 MWt.

REACTOR TRIP SYSTEM (RTS) RESPONSE TIME 1.28 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its (RTS) trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by NRC.

REPORTABLE EVENT 1.29 DELETED November 9, 2004 SEQUOYAH - UNIT 1 1-5 Amendment No. 12, 71, 141, 148, 155, 201, 233, 251, 275, 276, 294, 297 E2-3

Remove Pages 3/4 4-6 thru -12 and replace with INSERT A EACTOR COOLANT SYSTEM 3/4 5 STEAM GENERATORS LIMITING ONDITION FOR OPERATION 3.4.5 Each st generator shall be OPERABLE.

APPLICABILITY: MADES 1, 2, 3 and 4.

ACTION:

With one or more steam genators inoperable, restore the inoperable generatos) to OPERABLE status prior to increasing Tavg above "0F.

SURVEILLANCE REQUIREMENTS 4.4.5.0 Each steam generator shall be e nstrated OPERABL by performance of the following augmented inservice inspection program and e requirements f Specification 4.0.5.

4.4.5.1 Steam Generator Sample Selection and ectio

-Each steam generator shall be determined OPERABLE during shutdown by selecting and insp tin t least the minimum number of steam generators specified in Table 4.4-1.

4.4.5.2 Steam Generator Tube Sample Selection d Insection - The steam generator tube minimum sample size, inspection result classification, and e corresp nding action required shall be as specified in Table 4.4-2. The inservice inspection of stea generator tub shall be performed at the frequencies specified in Specification 4.4.5.3 and the inspected tubes shall verified acceptable per the acceptance criteria of Specification 4.4.5.4. The tubes/ elected for each inse rce inspection shall include at least 3%

of the total number of tubes in all steamX enerators; the tubes selec d for these inspections shall be selected on a random basis except:

a.

Where experienc n similar plants with similar water che stry indicates critical areas to be inspected, th at least 50% of the tubes inspected shalle from these critical areas.

b.

The first sa le of tubes selected for each inservice inspection ubsequent to the preservic rnspection) of each steam generator shall include:

1.

All nonplugged tubes that previously had detectable wall pene ations (greater than 20%).

Tubes in those areas where experience has indicated potential problems.

SEQUOYAH - UNIT 1 3/4 4-6 E2-4

TEACTOR COOLANT SYSTEM S

VEILLANCE REQUIREMENTS (Continued)

/

3.

A tube inspection (pursuant to Specification 4.4.5.4.a.8) shall be perform on each selected tube. If any selected tube does not permit the passage the eddy current probe for a tube Inspection, this shall be recorded and an adj nt tube shall be selected and subjected to a tube inspection.

c.

The tubes selected as the second and third samples (if requ d by Table 4.4-2) durn each inservice inspection may be subjected to a partial tub nspection provided:

1.

The tubes selected for these sample) clude the tubes from tho areas of the tube sheet array where tubes wit imperfections were previ sly found.

2.

The inspections include thos portions of the tubes where imperfectio were previously found.

NOTE: ube degradation ide fied in the portion of the tube that is not a reactor coolant pr sure boundary e end up to the start of the tube-to-tubesheet weld) is euded fro Result and Action Required in Table 4.4-2.

The results of each sample inspection shall be c ed into one of the following three categories:

Category Insp tion Resu C-I Less than 5% of the total tubes inspected are degrade tubes and none of the inspected tubes are defective.

C-2 One or more tubes, but not more than 1% of the total tubes insp cted are defective, or between 5%

and 10% of the total tubes inspe ed are degraded tubes.

C-3 Mo than 10% of the total tubes inspected are degraded tubes or more an 1% of the inspected tubes are defective.

Note:

In all inspections, previously degraded tube ust exhibit significant (greater than 10%) further wall penetrations to be in uded in the above percentage calculations.

March 4, 2003 SEQUOYAH - UNIT 1 3/4 4-7 Amendment No. 189, 214, 222, 252, 282 E2-5

'REACTOR COOLANT SYSTEM SU\\'VEILLANCE REQUIREMENTS (Continued)/

4.4.5.3 Iection Frequencies - The above required inservice inspections of steam generator tes shall be performe t the following frequencies:

a.

first inservice inspection shall be performed after 6 Effective Full Po er Months but wit 24 calendar months of initial criticality. Subsequent inservice in ections shall be perfo ed at intervals of not less than 12 nor more than 24 calendar onths after the previou nspection. If two consecutive inspections following servi under AVT conditions not including the preservice inspection, result in all in ection results falling into the C-i tegory or if two consecutive inspections demon iate that previously observed deg dation has not continued and no additional d radation has occurred, the inspection interv may be extended to a maximum of on per 40 months.

b.

If the results of the i ervice inspection of a steam ge rator conducted in accordance with Table 4.4-2 at 40 onth intervals fall in Categ C-3, the inspection frequency shall be increased to at least o ce per 20 months. Th ncrease in inspection frequency shall apply until the subsequent spections satisfy t criteria of Specification 4.4.5.3.a; the interval may then be extende to a maximu f once per 40 months.

c.

Additional, unscheduled inservice~ spec, ns shall be performed on each steam generator in accordance with the firs mple inspection specified in Table 4.4-2 during the shutdown subsequent to any of tollowing conditions.

1.

Primary-to-secondary tu es leaks ot including leaks originating from tube-to-tube sheet welds) in e :ess of the Iits of Specification 3.4.6.2.

2.

A seismic occurree greater than the 0 rating Basis Earthquake.

3.

A loss-of-coo nt accident requiring actuation the engineered safeguards.

4.

A main s m line or feedwater line break.

SEQUOYAH - UNIT 1 3/4 4-8 E2-6

"ACTOR COOLANT SYSTEM SUR EILLANCE REQUIREMENTS (Continued) 4.4.5.4 cceptance Criteria

a.

As used in this Specification:

1.

Imperfe6tion means an exception to the dimensio, finish or contour of a tube from that required by fabrication drawings or s ecifications.

Eddy-current testing indications below 20% of the nominal tu wall thickness, if detectable, may be considered as imperfections.

2.

Degradation means a service-induced cking, wastage, wear or gene 1corrosion occurring on either inside or outside f a tube.

Degraded Tube means a tube co aining imperfections greater than or eq Ito 20% of the nominal wall thicknes caused by degradation.

4.

% De radation means the rcentage of the tube wall thickness affected or remold by degradation.

5.

Defet means an imp1 fection of such severity that it exceeds the plugging limit. A tube ntaining a fect is defective.

6.

PIucqinq Lit eans the imperfection depth at or beyond which the tube shall be removed fr service and is equal to 40% of the nominal tube wall thickness. Plugging Ii es not apply to that portion of the tube that is not within the pressure bound ry of the reactor coolant system (tube end up to the start of the tube-to-tubeeet weld.

7.

Un rviceable describ the condition of a tube if it leaks or contains a defect rge enough to affect i structural integrity in the event of an Operating Basi arthquake, a loss-of-coo nt accident, or a steam line or feedwater lin reak as specified in 4.4.5.3.c, bove.

8 Tube Inspection means an inspe ion of the steam generator tube from th point of entry (hot leg side) completely ar nd the U-bend to the top supp of the cold leg.

9.

Preservice Inspection means a tube insp ction of the full length f each tube in each steam generator performed by eddy c rent techniques prior to service establish a baseline condition of the tubing. This i pection shall be performed prior to initial POWER OPERATION using the equi ent and techniques expected to be used during subsequent inservice ins ctions.

March 4, 2003 SEQUOYAH - UNIT 1 3/4 4-9 Amendment No. 189. 214. 222. 252. 282 E2-7

'REACTOR COOLANT SYSTEM i

SU VEILLANCE REQUIREMENTS (Continued)/

\\

The steam generator shall be determined OPERABLE after completing the corres ongg

\\ ctions (plug all tubes exceeding the plugging limit and all tubes containing throug p 1al Wrcks) required by Table 4.4-2./

4.4.5.5 Reports

a.

Follow each inservice inspection of steam generator tubes, the nu er of tubes plugged in ach steam generator shall be reported to the Commissi within 15 days.

b.

The complete sults of the steam generator tube inservice insction shall be submitted to the Commissio in a Special Report pursuant to Specificat'n 6.9.2 within 12 months following completio the inspection. This Special Repo hall include:

1.

Number and ext t of tubes inspected.

2.

Location and percen f wall-thickness peetration for each indication of an imperfection.\\/

3.

Identification of tubes plugg

c.

Results of steam generator tube insp i

s which fall into Category C-3 shall be reported as a degraded condition pursuant to 10 R 50.3 prior to resumption of plant operation. The written followup of this report sh provide a scription of investigations conducted to determine cause of the tube d d

ctive measures taken to prevent recurrence.

March 4, 2003 SEQUOYAH - UNIT 1 314 4-10 Amendment No. 36, 214, 222, 276, 282 E2-8

TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION PRESE CE INSPECTION NO YES

\\

TwolThree l our Two Three Ffur No. of Steam nerators per Unit lIr Fr Three

/_

All One Two /

Two First Inservice Insp tion One' one' n;

One0 Second & Subsequent )sservice Inspections

/

Table Notation:

1.

The inservice inspection ay be limited to one steam enerator on a rotating schedule encompassing 3 N % of th tubes (where N is the n ber of steam generators in the plant) if the results of the first or pre us inspections indi te that all steam generators are performing in a like manner.

e that under so e circumstances, the operating conditions in one or more steam generators ay be foun to be more severe than those in other steam generators. Under such circumsta es the mple sequence shall be modified to inspect the most severe conditions.

2.

The other steam generator not inspec uring the first inservice inspection shall be inspected. The third and subsequen inspe ons should follow the instructions described in 1 above.

3.

Each of the other two steam nerators not insp ted during the first inservice inspections shall be inspected during th second and third insp tions. The fourth and subsequent inspections shall follow th instructions described in boe.

SEQUOYAH - UNIT 1 314 4-11 E2-9

TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION ST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3 RD SA LE

\\__

INSPECTION Sample Size Result Action Required Result Action Required Result Action

\\

/

Required A minimum of C-None NIA N/A/A NIA S Tubes K_/

per S.G.

C-2 P1 defective tubes C-1 None N/A N/A and spect additional Plug defective C-1 None 2S tub in this S.G.

C-2 tubes and in ect additional tubes C-2 Plug in this S

.defective tubes Perform C-3 action for C-3 result of first sample Perform action for C-C-3 result of first N/A N/A

\\/_

sample C-3 Inspect all tubes in All h this S.G. plug are -I None N/A N/A defective tubes and inspect 2S

/

tubes in each other Some S/Gs S.G.

C-2 but no Pe orm action for N/A NIA additional C-2 sult of S.G. are secon ample C-3

\\

Additional Inspect all s in S.G. is C-3 each S.G. an lug N/A N/A

/

defective tubes.

I NI S = 3-% Where s he number of steam generators in the unit, and n is the er of steam n

generators insp ted during an inspection.

May 24, 2002 SEQUOYAH - UNIT 1 3/4 4-12 Amendment No. 36, 276 E2-10

INSERT A REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY LIMITING CONDITION FOR OPERATION 3.4.5 SG tube integrity shall be maintained.

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

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS*:

a. With one or more SG tubes satisfying the tube repair criteria and not plugged in accordance with the Steam Generator Program, within 7 days verify tube integrity of the affected tube(s) is maintained until the next refueling outage or SG tube inspection, or be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in 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

b. Plug the affected tube(s) in accordance with the Steam Generator Program prior to startup following the next refueling outage or SG tube inspection.

SURVEILLANCE REQUIREMENTS 4.4.5.0 Verify steam generator tube integrity in accordance with the Steam Generator Program.

4.4.5.1 Verify that each inspected SG tube that satisfies the tube repair criteria is plugged in accordance with the Steam Generator Program prior to startup following a SG tube inspection.

• Separate Action entry is allowed for each SG tube.

SEQUOYAH - UNIT 1 3/4 4-6 E2-11

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:

a.

No PRESSURE BOUNDARY LEAKAGE,

b.

1 GPM UNIDENTIFIED LEAKAGE,

c.

150 gallons per day of primary-to-secondary leakage through any one steam generator, and

d.

10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System.

APPLICABILITY: MODES 1, 2, 3 and 4 1or with primary-to-secondary leakage not within limits,I ACTION:

a.

With any PRESSURE BOUNDARY LEAKAGE ein at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b.

With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY ithin the next 6_hours and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

E~fy S

EILLANCE REQUIREMENTS I or primary-to-secondary leaKage is within l

4.4.6.2\\ Reactor Coolant System leakage,&

verifiod to be within Rach of the above limits by performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.*

The provision of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.

4.4.6.2.2,Veify 6toeam genoRatOr tubo itogit'; i in aRaor.~dan6B with tho roqUiroMnRt6 ef ToehRical The above surveillance requirement is not applicable l to Primary-to-secondary leakage.

I

• Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.

August 4, 2000 Amendment No. 12, 214, 222, 259 SEQUOYAH - UNIT 1 3/4 4-14 E2-12

ADMINISTRATIVE CONTROLS j.

Technical Specification (TS) Bases Control Proqram This program provides a means for processing changes to the Bases of TSs.

a.

Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b.

Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

1.

A change in the TS incorporated in the license or

2.

A change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

c.

The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.

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).

I R

6.9 REPORTING REQUIREMENTS INSERT B ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted in accordance with 10 CFR 50.4.

STARTUP REPORT 6.9.1.1 DELETED 6.9.1.2 DELETED 6.9.1.3 DELETED ANNUAL REPORTS 6.9.1.4 Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year. The initial report shall be submitted prior to March I of the year following initial criticality.

6.9.1.5 DELETED I

11 A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station.

I SEQUOYAH - UNIT 1 6-11 April 5, 2005 Amendment No. 12, 32, 58, 72, 74, 148, 174, 233, 280, 300 E2-13

INSERT B

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 was 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 and/or plugged, to confirm that the performance criteria are being met.

b.

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

1.

Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, cooldown, and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance wih 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 design basis accident, other than a SG tube rupture, shall not exceed the maximum leakage rate established 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 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.

E2-14

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 SG 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.

E2-15

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (continued)

6.

WCAP-10054-P-A, Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code, August 1985, M Proprietary)

(Methodology for Specification 3/4.2.2 - Heat Flux Hot Channel Factor)

7.

WCAP-10266-P-A, Rev. 2, "THE 1981 REVISION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE", March 1987, & Proprietary).

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor).

8.

BAW-10227P-A, 'Evaluation of Advance Cladding and Structural Material (M5) in PWR Reactor Fuel," February 2000, (FCF Proprietary)

(Methodology for Specification 3/4.2.2 - Heat Flux Hot Channel Factor) 6.9.1.14.b The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

6.9.1.14.c THE CORE OPERATING LIMITS REPORT shall be provided within 30 days after cycle start-up (Mode 2) for each reload cycle or within 30 days of issuance of any midcycle revision of the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS (PTLR) REPORT 6.9.1.15 RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing, LTOP arming, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

Specification 3.4.9.1, 'RCS Pressure and Temperature (PIT) Limits' Specification 3.4.12, 'Low Temperature Over Pressure Protection (LTOP) System' 6.9.1.15.a The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1.

Westinghouse Topical Report WCAP-14040-NP-A, 'Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves."

2.

Westinghouse Topical Report WCAP-15293, "Sequoyah Unit I Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation."

3.

Westinghouse Topical Report WCAP-1 5984, 'Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Sequoyah Units I and 2."

6.9.1.15.b The PTLR shall be provided to the NRC within 30 days of issuance of any revision or supplement thereto.

SPOOSL REPORTS 6.9.2.1 Sp l reports shall be submitted within the time period specified for each report, in accordance wit CFR 50.4.

6.9.2.2 This specifica has been deleted.

November 9, 2004 SEQUOYAH - UNIT I 6-13a Amendment No. 52, 58, 72, 74, 117, 155, 223, 241, 258, 294, 297 E2-16

INSERT C STEAM GENERATOR TUBE INSPECTION REPORT 6.9.1.16 A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.9, Steam Generator (SG) Program. The report shall include:

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.

E2-17

ENCLOSURE 3 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT (SQN)

UNIT 1 Changes to Technical Specifications Bases Pages E3-1

REACTOR COOLANT SYSTEM BASES to cycle the valves when they have been closed to comply with the ACTION requirements. This precludes the need to cycle the valves with full system differential pressure or when maintenance is being performed to restore an inoperable PORV to operable status.

Testing of PORVs with a steam bubble in the pressurizer is considered to be a representive test for assessing PORV performance under normal operating conditions.

3/4.4.4 PRESSURIZER The limit on the maximum water volume in the pressurizer assures that the parameter is maintained within the normal steady state envelope of operation assumed in the SAR. The limit is consistent with the initial SAR assumptions. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance is sufficient to ensure that the parameter is restored to within its limit following expected transient operation. The maximum water volume also ensures that a steam bubble is formed and thus the RCS is not a hydraulically solid system. The requirement that 150 kw of pressurizer heaters and their associated controls be capable of being supplied electrical power from an emergency bus provides assurance that the plant will be able to control reactor coolant pressure and establish natural circulation conditions.

3/4.4.5 STEAM GENERATORS Surveeillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of thisthe RCS will be maintained. The program for inservice inspection of steam generator tubes is basedo fication of Regulatory Guide 1.83, Revision 1. Inservice inspection of steam generator tubing is essential i maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical dam rogressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inseri ection of steam generator tubing also provides a means of characterizing the nature and cause of any tubei so that corrective measures can be taken.

lINSER SEQUOYAH - UNIT 1 B 3/4 4-2a November 19, 1998 Amendment No. 12,157, 239 E3-2

\\EACTOR COOLANT SYSTEM ce plant is expected to beoperated in amanner such that the secondary coolant will bem mI ained within ths chemistry limits found to result in negligible corrosion of the steam generator tubes. If^h secondary lant chemistry is not maintained within these limits, localized corrosion may likely sult in stress corrosi cracking. The extent of cracking during plant operation would be limited by th limitation of steam generato ube leakage between the primary coolant system and the secondary coola system (primary-to-secon ry leakage = 150 gallons per day per steam generator). Cracks havin primary-to-secondary leakage ss than this limit during operation will have an adequate margin of fety to withstand the loads imposed du 'g normal operation and by postulated accidents. Sequoyah h demonstrated that primary-to-secondary leaage of 150 gallons per day steam generator can readily be etected by radiation monitors of steam genera r blowdown or condenser off-gas. Leakage in excess o this limit will require plant shutdown and an unsc eduled inspection, during which the leaking tubes w be located and plugged.

March rQUOYAH - UNIT 1 B 314 4-3 Amendment No. 36, 189, 214, 222, 282 E3-3

\\REACTOR COOLANT SYSTEM

/

BA ES Wastage-type defects are likely with proper chemistry treat nt of the secondary coolant.

However, even if a defect should de lop in service, it will be found ring scheduled inservice steam generator tube examinations. Pluggin will be required for all tube with imperfections exceeding the repair limit defined in Surveillance Requiremen 4.4.5.4.a. The portion the tube that the plugging limit does not apply to is the portion of the tube that is no within the RCS presure boundary (tube end up to the start of the tube-to-tubesheet weld). The tube end to tu to-tubesheet Ild portion of the tube does not affect structurar integrity of the steam generator tubes and ther ore indicaf ns found in this portion of the tube will be excluded from the Result and Action Required fo tube in ections. It is expected that any indications that extend from this region will be detected during the ch uled tube inspections. Steam generator tube inspections of operating plants have demonstrated t capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness.

March 4 03 S

UOA-UNIT I B33/4 4-4 Amendment No. 36, 189, 214, 222, 252,2 E3-4

page left blank ii 1

B 3/4 4-4a Amendment No. 36, 189, E3-5

SG Tube Integrity B 3/4.4.5 INSERT D B 3.4 REACTOR COOLANT SYSTEM B 3/4.4.5 Steam Generator (SG) Tube Integrity BASES BACKGROUND Steam generator (SG) tubes are small diameter, thin walled tubes that carry primary coolant through the primary to secondary heat exchangers.

The SG tubes have a number of important safety functions. Steam generator tubes are an integral part of the reactor coolant pressure boundary (RCPB) and, as such, are relied on to maintain the primary system's pressure and inventory. The SG tubes isolate the radioactive fission products in the primary coolant from the secondary system. In addition, as part of the RCPB, the SG tubes are unique in that they act as the heat transfer surface between the primary and secondary systems to remove heat from the primary system. This Specification addresses only the RCPB integrity function of the SG. The SG heat removal function is addressed by LCO 3.4.1.1, "Startup and Power Operation," LCO 3.4.1.2, "Hot Standby," LCO 3.4.1.3, 'Shutdown," and LCO 3.4.1.4, 'Cold Shutdown."

SG tube integrity means that the tubes are capable of performing their intended RCPB safety function consistent with the licensing basis, including applicable regulatory requirements.

Steam generator tubing is subject to a variety of degradation mechanisms.

Steam generator tubes may experience tube degradation related to corrosion phenomena, such as wastage, pitting, intergranular attack, and stress corrosion cracking, along with other mechanically induced phenomena such as denting and wear. These degradation mechanisms can impair tube integrity if they are not managed effectively. The SG performance criteria are used to manage SG tube degradation.

Specification 6.8.k, "Steam Generator (SG) Program," requires that a program be established and implemented to ensure that SG tube integrity is maintained. Pursuant to Specification 6.8.k, tube integrity is maintained when the SG performance criteria are met. There are three SG performance criteria: structural integrity, accident induced leakage, and operational leakage. The SG performance criteria are described in Specification 6.8.k. Meeting the SG performance criteria provides reasonable assurance of maintaining tube integrity at normal and accident conditions.

SEQUOYAH - UNIT 1 B 3/4 4-3 E3-6

SG Tube Integrity B 3/4.4.5 INSERT D (Continued)

BASES BACKGROUND (continued)

The processes used to meet the SG performance criteria are defined by the Steam Generator Program Guidelines (Ref. 1).

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 a 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 only briefly released to the atmosphere via safety valves and the majority is discharged to the main condenser.

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 is based on a primary to secondary leakage of 0.1 gallons per minute (gpm) for the non-faulted SGs and 3.7 gpm for the faulted SG. For purpose of this TS, the assumed accident induced leakage in the faulted SG is limited to 1.0 gpm which is bounded by the maximum leakage established by the plant safety analysis. 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.

SEQUOYAH - UNIT 1 B 3/4 4-3a E3-7

SG Tube Integrity B 3/4.4.5 INSERT D (Continued)

BASES LCO (continued)

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.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 (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.

SEQUOYAH - UNIT 1 B 3/4 4-3b E3-8

SG Tube Integrity B 3/4.4.5 INSERT D (Continued)

BASES LCO (continued)

Structural integrity requires that the primary membrane stress intensity in a tube not exceed the yield strength for all 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 III, 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. In the limiting accident analysis (MSLB), SG leakage is assumed to be 3.7 gpm for the faulted SG and 0.1 gpm for the non-faulted SGs. Limiting the allowable leakage in the faulted SG to 1.0 gpm ensures that the MSLB analysis remains conservative and bounding. 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 LOCA or a main steam line break. If this amount of leakage is due to more than one crack, the cracks are very small, and the above assumption is conservative.

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

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

SEQUOYAH - UNIT 1 B 3/4 4-3c E3-9

SG Tube Integrity B 3/4.4.5 INSERT D (Continued)

BASES 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.

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 (continued) 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 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. 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 following the next refueling outage or SG inspection.

SEQUOYAH - UNIT 1 B 3/4 4-3d E3-10

SG Tube Integrity B 3/4.4.5 INSERT D (Continued)

BASES ACTIONS (continued)

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 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.k contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections.

SEQUOYAH - UNIT 1 B 3/4 4-4 E3-11

SG Tube Integrity B 3/4.4.5 INSERT D (Continued)

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.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. 10 CFR 100.

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 1 B 3/4 4-4a E3-12

REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE BACKGROUND Components that contain or transport the coolant to or from the reactor core make up the reactor coolant system (RCS). Component joints are made by welding, bolting, rolling, or pressure loading, and valves isolate connecting systems from the RCS.

During plant life, the joint and valve interfaces can produce varying amounts of reactor coolant leakage, through either normal operational wear or mechanical deterioration. The purpose of the RCS Operational leakage LCO is to limit system operation in the presence of leakage from these sources to amounts that do not compromise safety. This LCO specifies the types and amounts of leakage.

10 CFR 50, Appendix A, GDC 30 (Ref. 1), requires means for detecting and, to the extent practical, identifying the source of reactor coolant leakage. Regulatory Guide 1.45 (Ref. 2) describes acceptable methods for selecting leakage detection systems.

The safety significance of leakage varies widely depending on its source, rate, and duration. Therefore, detecting and monitoring reactor coolant leakage into the containment area is necessary. Quickly separating the identified leakage from the unidentified leakage is necessary to provide quantitative information to the operators, allowing them to take corrective action should a leak occur that is detrimental to the safety of the facility and the public.

A limited amount of leakage inside containment is expected from auxiliary systems that cannot be made 100% leaktight. Leakage from these systems should be detected, located, and isolated from the containment atmosphere, if possible, to not interfere with RCS leakage detection.

This LCO deals with protection of the reactor coolant pressure boundary (RCPB) from degradation and the core from inadequate cooling, in addition to preventing the accident analyses radiation release assumptions from being exceeded. The consequences of violating this LCO include the possibility of a loss of coolant accident (LOCA).

APPLICABLE SAFETY ANALYSES Except for primary-to-secondary leakage, the safety analyses do not address operational leakage. However, other operational leakage is related to the safety analyses for LOCA; the amount of leakage can affect the probability of such an event. The safety analysis for an event resulting in steam discharge to the atmosphere assumea 1 gpm primary to Gocondary leakage as the initial condition.

hat primary-to-secondary leakage from all steam generators (SGs) is 0.4 gpm or increases to 4.0 gpm as a result of accident induced conditions.

I SEQUOYAH - UNIT 1 B 3/4 4-4g August 4, 2000 Amendment No. 36,189, 214, 222, 237, 259 E3-13

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

The FSAR (Ref. 3) analysis for 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 primary-to-secondary leakageis relatively inconsequential.

I from all four SGs I

A the entire l

e:Fbes more limiting for site radiation releases. The safety analysis for the SLB Iaccident a-s-s 3.7 gpm primary-to-secondary leakage in o01e generator as an initial condition. The dose consequences resulting from the B accident a well within the limits defined in 10 CFR 100.

/

accident induced 1 through the affected o 1.0 gpm in the I

Ir -

Limiting E leakage t Jol. tfrA 0 10URVtU 10 LIU la ul auvOLIVU with respect to the 3.7 gpm used in the SLB analyses.

I and 0.3 gpm through the non-affected generators l

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 sump level monitoring equipment can collectively SEQUOYAH - UNIT 1 B 3/4 4-4h May 22, 2003 Amendment No. 36, 189, 214, 222, 237, 259 E3-14

a' REACTOR COOLANT SYSTEM BASES detect within a reasonable time period. Violation of this LCO could result in continued degradation of the RCPB, if the leakage is from the pressure boundary.

C.

(SG)

Primarv-to-Secondarv Leakage through Any One Steam Generator I INSERT E A

e 150 gallons per day limit on one SG is based on the assumption that a sing ack leaking this amount would not propagate to a SGTR und e

stress con ons of a LOCA or a main steam line rupture. If le through many cracks, t acks are very small, and the above as ption is conservative.

The 150-gallons per day limit i orate oSurveillance 4.4.6.2.1 is more restrictive than the standard o g leakage limit and is intended to provide an additional margin commo a crack which might grow at a greater than expected r r unexpectedly exte utside the thickness of the tube support Hence, the reduced leakage I when combined with an effec eak rate monitoring program, provides ad al assura hat, should a significant leak be experienced, it will be

ected, a ie plant shut down in a timely manner.

d.

IDENTIFIED LEAKAGE Up to 10 gpm of IDENTIFIED LEAKAGE is considered allowable because leakage is from known sources that do not interfere with detection of UNIDENTIFIED LEAKAGE and is well within the capability of the RCS Makeup System. IDENTIFIED LEAKAGE includes leakage to the containment from specifically known and located sources, but does not include PRESSURE BOUNDARY LEAKAGE or controlled reactor coolant pump (RCP) seal leakoff (a normal function not considered leakage).

Violation of this LCO could result in continued degradation of a component or system.

APPLICABILITY In MODES 1, 2, 3, and 4, the potential for reactor coolant PRESSURE BOUNDARY LEAKAGE is greatest when the RCS is pressurized.

In MODES 5 and 6, leakage limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses and reduced potentials for leakage.

SEQUOYAH - UNIT I B 314 4-4i May 17, 2002 Amendment No. 36, 189, 214, 222, 237, 259 E3-15

REACTOR COOLANT SYSTEM BASES LCO 3/4.4.6.3, "RCS Pressure Isolation Valve (PIV) Leakage," measures leakage through each individual PIV and can impact this LCO. Of the two PIVs in series in each isolated line, leakage measured through one PIV does not result in RCS leakage when the other is leak tight. If both valves leak and result in a loss of mass from the RCS, the loss must be included in the allowable IDENTIFIED LEAKAGE.

ACTIONS Action a:

jor with primary to secondary leakage not within limits, If any PRESSURE BOUNDARY LEAKAGE exists,"(he reactor must be brought to lower pressure conditions to reduce the severity of the leakage and its potential consequences. It should be noted that leakage past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE. The reactor must be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 5 within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. This action reduces the leakage and also reduces the factors that tend to degrade the pressure boundary.

The allowed completion times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. In MODE 5, the pressure stresses acting on the RCPB are much lower, and further deterioration is much less likely.

Action b:

o UNIDENTIFIED LEAKAG

, ID NTIFIED LEAKAGE, or primary to socondary leakagein excess of the LCO I mits must be reduced to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

This completion time allows ti e to verify leakage rates and either identify UNIDENTIFIED LEAKAGE or educe leakage to within limits before the reactor must be shut down. This action is cessary to prevent further deterioration of the RCPB.

If UNIDENTIFIED LEAKAGE, DENTIFIED LEAKAGE, oF primary to eocondary leakage-cannot be reduced to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the reactor must be brought to lower pressure conditions to reduce the severity of the leakage and its potential consequences. The reactor must be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 5 within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. This action reduces the leakage and also reduces the factors that tend to degrade the pressure boundary.

The allowed completion times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. In MODE 5, the pressure stresses acting on the RCPB are much lower, and further deterioration is much less likely.

August 4, 2000 SEQUOYAH - UNIT 1 B 3/4 4-4j Amendment No. 36, 189, 214, 222, 237, 259 E3-16

REACTOR COOLANT SYSTEM BASES SURVEILLANCE REQUIREMENTS Surveillance 4.4.6.2.1 Verifying RCS leakage to be within the LCO limits ensures the integrity of the RCPB is maintained. PRESSURE BOUNDARY LEAKAGE would at first appear as UNIDENTIFIED LEAKAGE and can only be positively identified by inspection. It should be noted that leakage past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE. UNIDENTIFIED LEAKAGE and IDENTIFIED LEAKAGE are determined The by performance of an RCS water inventory balance. Primary to secondary leakage ic also moacurod by porformanco of an RCS water inventory balance in conjunction with effluent monitoring Within the secondary 6team and feedwater syctems.

T RCS water inventory balance must be met with the reactor at steady state The surveillance isf oper ng conditions (stable pressure, temperature, power level, pressurizer and dby a footnote. \\makeupnk levels, makeup, letdown, and RCP seal injection and return flows).

I~hoe footnote allowig that this SR is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after esta *shing steady state operation. The 12-hour allowance provides sufficient time tolect and process all necessary data after stable plant conditions are established. P ormance of this surveillance within the 12-hour allowance is required to maintain ompliance with the provisions of Specification 4.0.4.

Steady state operation is required to perform a proper inventory balance since calculations during maneuvering are not useful. For RCS operational leakage determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.

An early warning of PRESSURE BOUNDARY LEAKAGE or UNIDENTIFIED LEAKAGE is provided by the automatic systems that monitor the containment atmosphere radioactivity and the containment pocket sump level. It should be noted that leakage past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE.

These leakage detection systems are specified in LCO 3/4.4.6.1, "Leakage Detection Instrumentation."

INSERT F The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> frequency is a reasonable interval to trend leakage and recognizes the importance of early leakage detection in the prevention of accidents.

INETG1~

Surveillance 4.4.6.2.2 Th su rovides the means necessary to determine SG in an operational MOD uirement to demonstrat integrity in accordance with the Steam Genea ace Program emphasizes the importance of SG tube inter ough tnot be performed at normal o s.

SEQUOYAH - UNIT 1 B 314 4-4k August 4, 2000 Amendment No. 36,189, 214, 222, 237, 259 E3-17

i,

REACTOR COOLANT SYSTEM BASES REFERENCES

1.

10 CFR 50, Appendix A, GDC 30.

2.

Regulatory Guide 1.45, May 1973.

3.

FSAR, Section 15.4.3.

4.

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

5.

EPRI, Pressurized Water Reactor Primary-to-Secondary Leak Guidelines."

SEQUOYAH -UNIT 1 B 3/4 4-41 August 4, 2000 Amendment No. 36, 189, 214, 222, 237, 259 E3-18

INSERT E The limit of 150 gallons per day per SG is based on the operational leakage performance criterion in NEI 97-06, Steam Generator Program Guidelines (Ref. 4). The Steam Generator Program operational leakage performance criterion in NEI 97-06 states, "The RCS operational primary to secondary leakage through any one SG shall be limited to 150 gallons per day." The limit is based on operating experience with SG tube degradation mechanisms that result in tube leakage. The operational leakage rate criterion in conjunction with the implementation of the Steam Generator Program is an effective measure for minimizing the frequency of steam generator tube ruptures.

INSERT F The second footnote states that this SR is not applicable to primary to secondary leakage because leakage of 150 gallons per day cannot be measured accurately by an RCS water inventory balance.

INSERT G This SR verifies that primary to secondary leakage is less or equal to 150 gallons per day through any one SG. Satisfying the primary to secondary leakage limit ensures that the operational leakage performance criterion in the Steam Generator Program is met. If this SR is not met, compliance with LCO 3.4.5, "Steam Generator Tube Integrity," should be evaluated.

The 150 gallons per day limit is measured at 70 degrees Fahrenheit (Reference 5).

The operational leakage rate limit applies to leakage through any one SG. If it is not practical to assign the leakage to an individual SG, all the primary to secondary leakage should be conservatively assumed to be from one SG.

The surveillance is modified by a note which states that the surveillance is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation. For RCS primary to secondary leakage determination, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.

The surveillance frequency of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is a reasonable interval to trend primary to secondary leakage and recognizes the importance of early leakage detection in the prevention of accidents. The primary to secondary leakage is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with the EPRI guidelines (Ref.

5).

E3-19