ML022960514

Unit 1 - Technical Specification (TS) Change No. 02-02, Steam Generator (SG) Alternate Repair Criteria (ARC) Deletion and SG Inspection Interval Revision - Partial Withdrawal

ML022960514
Person / Time
Site:	Sequoyah
Issue date:	10/10/2002
From:	Salas P Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TVA-SQN-TS-02-02
Download: ML022960514 (15)

Text

Tennessee Valley Authority, Post Office Box 2000, Soddy-Daisy, Tennessee 37384-2000 October 10, 2002 TVA-SQN-TS-02-02 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 Gentlemen:

In the Matter of ) Docket No. 50-321 Tennessee Valley Authority SEQUOYAH NUCLEAR PLANT (SQN) - UNIT 1 - TECHNICAL SPECIFICATION (TS) CHANGE NO. 02-02, STEAM GENERATOR (SG)

ALTERNATE REPAIR CRITERIA (ARC) DELETION AND SG INSPECTION INTERVAL REVISION - PARTIAL WITHDRAWAL

Reference:

TVA letter to NRC dated March 29, 2002, "Sequoyah Nuclear Plant (SQN) - Unit 1 - Technical Specification (TS) Change No. 02-02, Steam Generator (SG) Alternate Repair Criteria (ARC)

Deletion And SG Inspection Interval Revision" By the reference letter, TVA submitted a license amendment for SQN Unit 1 TSs. The proposed license amendment revised SG TS 3/4.4.5 to eliminate surveillance requirements (SRs) associated with two alternate repair criteria. The proposed change also included a revision to SR 3/4.4.5.3.a that provided a 40-month SG inspection interval.

Based on discussion between TVA and NRC, it is our understanding that the NRC is not ready to approve a 40-month SG inspection interval without collection of one operating cycle of SG inspection data to assess tube integrity for newly replaced SGs. Accordingly, TVA is withdrawing the requested change for a 40-month SG inspection interval at this time. All other aspects of the proposed TS change remain unchanged.

Nrn~don mcW p0

U.S. Nuclear Regulatory Commission Page 2 October 10, 2002 The no significant hazards consideration determination (NSHCD) previously provided by TVA' s reference letter remains bounding and does not alter any conclusions of the NSHCD.

Accordingly, Enclosure 1 provides those TS pages and associated Bases pages relevant to removal of the ARC.

Enclosure 2 provides revised NSHCD pages that remove the evaluation language associated with the 40-month SG inspection interval.

This letter is being sent in accordance with NRC RIS 2001-05.

No commitments are contained in this letter. If you have any questions about this change, please telephone me at (423) 843-7170 or J. D. Smith at (423) 843-6672.

" -and Industry Affairs Manager

=~*oc r iled and sworn tp before me n 1. of

• ar* Public t

My Commission Expires August 12, 2006 Enclosures JDS:DVG:PMB Enclosures cc (Enclosures):

Mr. Lawrence E. Nanney, Director (w/o Enclosures)

Division of Radiological Health Third Floor L&C Annex 401 Church Street Nashville, Tennessee 37243-1532 Mr. Frank Masseth Framatome ANP, Inc.

3315 Old Forest Road P. 0. Box 10935 Lynchburg, VA 24506-0935

ENCLOSURE 1 TENNESSEE VALLEY AUTHORITY SEQUOYAH PLANT (SQN)

UNIT 1 TECHNICAL SPECIFICATION (TS) CHANGE 02-02 PARTIAL WITHDRAWAL MARKED PAGES I. AFFECTED PAGE LIST Page 5 of SQN Operating License (License Condition 2.C.9.d) 3/4 4-7 3/4 4-8 3/4 4-9 3/4 4-9a 3/4 4-9b 3/4 4-10 3/4 4-10a B3/4 4-3 B3/4 4-4 B3/4 4-4a II. MARKED PAGES See attached.

El-I

(9) Steam Generator Inspection (Section 5.3 1)

(a) Prior to March 1, 1981, TVA shall provide to the NRC the results of its tests to determine the feasibility of using a steam generator camera device.

(b) Prior to start-up after the frist refueling, TVA must install inspection ports in each steam generator if the results of the camera device inspection are not satisfactory to the NRC; (c) Prior to start-up after the first refueling, TVA will plug Row 1 of the steam generator tubes, if required by NRC.

(d) By May 20, 1997, P44 shall establish a steamn generator inspection program that is in accordance w;ith the commitments listed in Enclosure 2 to the TVA,letter to the CoarmNmision O this subject dated March 12, 1997, as modified by ,,

letters dated March 17, 1097 anRd May 14, 2001 (10) Water Chemistry Control Program (Section 5 3 2)

This requirement has been deleted.

(11) Neqative Pressure in the Auxiliary Buildingq Secondary Containment Enclosure (ABSCE)

(Section 6 2 3)

After the final ABSCE configuration is determined, TVA must demonstate to the satisfaction of the NRC that a negative pressure of 0.25 inches of water gauge can be maintained in the spent fuel storage area and in the esf pump room.

(12) Environmental Qualification (Section 7 2 2)

(a) No late than November 1, 1980, TVA shall submit information to show compliance with the requirement of NUREG-0588, "Interim Staff Position on Environmental Qualification of Safety-Related Elecctrical Equipment," for safety related equipment exposed to a harsh environment. Implementation shall be in accordance with NUREG-0588 by June 30, 1982.

(b) By no later than December 1, 1980, complete and auditable records must be available and maintained at a central location which describe the environmental qualification method used for all safety-related electrical equipment in sufficient detail to document the degree of compliance with the DOR Guidelines or NUREG-0588 Thereafter, such records should be updated and maintained current as equipment is replaced, further tested, or otherwise further qualified to document complete compliance by June 30, 1982.

July 18, 2001 Amendment No. 75, 222, 270 E1-2

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) I I 3 A tube inspection (pursuant to Specification 4.4.5.4.a.8) shall be performed on each selected tube. If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube I DeletedI ,1, shall be selected and subiected to a tube insDection voltage-based repair cri en cted bybobbincoil probe during allI future refueling outages

4. n i rvice as aresult of application of the tube support plate

c. The tubes selected as the second and third samples (if required by Table 4 4-2) during each inservice inspection may be subjected to a partial tube inspection provided:

1. The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found.

2. The inspections include those portions of the tubes where imperfections were previously found Deleted NOTE: Tube degradation identified in the portion of the tube that is not a reactor coolant pressure boundary (tube end up to the start of the tube-to tubesheet weld) is excluded from the Result and Action Required in Table 4 4-2.

d. lementation of the steam generator tube/tube support plate repair criteria requires a 10 nt bobbin coil inspection for hot-leg and cold-leg tube support plae I ections down to the lowest cold-leg tube support plate with known outside diameter stre rosion cracking (ODSCC) indications The determination of the lowest cold-leg tube supp te intersections having ODSCC indications shall be based on the performance of at least ercent random sampling of tubes inspected over their full length.

e. Inspection of dented tube support plate intersections will be pe d in accordance with WCAP 15128, Revision 2, dated February 2000 as supplemen~ted by tter to NRC dated March 2, 2000. This alternate repair criteria is applicable to Cycle 11 and 12 operation.

The results of each sample inspection shall be classified into one of the following three categories Category Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.

C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.

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

Note In all inspections, previously degraded tubes must exhibit significant (greater than 10%) further wall penetrations to be included in the above percentage calculations March 8, 2000 SEQUOYAH - UNIT 1 3/4 4-7 Amendment No. 189, 214, 222, 252 E1-3

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMEýNTS (Continued) 4 4.5 4 Acceptance Criteria a As used in this Specification:

1. Imperfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawings or specifications.

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

2. Degradation means a service-induced cracking, wastage, wear or general corrosion occurring or either inside or outside of a tube.

3. Degraded Tube means a tube containing imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation.

4.  % Degradation means the percentage of the tube wall thickness affected or removed by degradation.

5 Defect means an imperfection of such severity that it exceeds the plugging limit.

A tube containing a defect is defective.

6 Plugqqing Limit means the imperfection depth at or beyond which the tube shall be removed from service and is equal to 40% of the nominal tube wall thickness. Plugging limit does not apply to that portion of the tube that is not within the pressure boundar of ther ratr coolant system ('tube end tin to the start of the tube-to-tubesheet weldl This definition does not apply to tube

-plate intersections if the vo tage-based repair criteria are being applied Refer to*-4..*,,for the repair limit applicable to these intersections For Cycl 1ad1 pe~gr , eiition does not apply for axial PWSCC iniaino otosteef whi~-r4Qtdwithin the thickness of dented tub upr ltswihehbtamxmum de-p~TM-gt"*grhn or equal to 40 percent of the initial tube wall thickness Refer to 4.4 5.4.a.1 arlmt I aDDlicable to these intersections.

7. Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3.c, above.

8. Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg

9. Preservice Inspection means a tube inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service establish a baseline condition of the tubing. This inspection shall be performed prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections March 8, 2000 SEQUOYAH - UNIT 1 3/4 4-9 Amendment No. 189, 214, 222, 252 E1-4

DeleeEntire H]~

(Continued)

COOLANT EACTOR COOLANT SYSTEM REQUIREMENTS SU__EILLANCE REQUIREMENTS_(Continued)

SU*EILLANCE

10. Tube Support Plate Plugginq Limit is used for the disposition of an alloy 600

• steam generator tube for continued service that is experiencing predominately axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates. At tube support plate intersections, the plugging (repair) limit is based on maintaining steam generator tube serviceability as described below.

Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with bobbin voltages less than or equal to the lower voltage repair "imit(Note 1), will be allowed to remain sIervice.

b. Stea generator tubes, whose degradation is attributed to outside diamet stress corrosion cracking within the bounds of the tube support plate with bobbin voltage greater than the lower voltage repair limit (Note 1), w be repaired or plugged, except as noted in 4 4.5 4.a 10.c below.

c Steam generator bes, with indications of potential degradation attributed to outside iameter stress corrosion-cracking within the bounds of the tub upport plate with a bobbin voltage greater than the lower voltage repair it (Note 1), but less than or equal to upper voltage repair limit (Note may remain in service ifa rotating pancake coil inspection does not det tdegradation Steam generator tubes, with indications of outside diamete tress corrosion-cracking degradation with a bobbin coil voltage greater than the upper oltage repair limit (Note 2) will be plugged or repaired.

d. Not applicable to SQN.

e. If an unscheduled mid-cycle inspection is pe rmed, the following mid cycle repair limits apply instead of the limits ide ified in 44 5 4.a.10.a, S4 .5 4.a 10."b, and 4.4.5.4.a 10".c.

The rnnid-cycle repair limits are determined from the following equations:

V .... = VSL l.O+NDE+Gr (CLLAIAt)

CL V = VL L- - vL (

J U/At9CL April 9, 7 SEQUJOYAH - UNIT 1 3/4 4-9a Amendment No. 189, 214, 2 E1-5

Delete ren age REACTOR COOLANT SYSTEM S VEILLANCE REQUIREMENTS (Continued) here

- upper voltage repair limit LRL - lower voltage repair limit VN- mid-cycle upper voltage repair limit based on time into cycle VMLRL - mid-cycle lower voltage repair limit based on VMuRL and time into cycle At length of time since last scheduled inspection during which VUn and VLRL were implemented CL - cycle length (the time between two scheduled steam generator inspections)

VSL tructural limit voltage Gr ave ge growth rate per cycle length NDE 95-per nt cumulative probability allowance for nondestructive examina n uncertainty (i e., a value of 20-percent has been approved NRC)

Implementation of these mid-cycle repair limits should f ow the same approach as in TS 4.4.5 4.a 10.a, 4.4.5 4.a.1O.b, and 4 4.5 4.a 10.c Note 1: The lower voltage repair limit is 1.0 volt for 3/4-inch d meter tubing or 2.0 volts for 7/8-inch diameter tubing.

Note 2: The upper voltage repair limit is calculated according to th ethodology in GL 90-05 as supplemented VURL may differ at the TSPs and flow distribu n baffle

11. Primary Water Stress Corrosion Crackinq (PWSCC) Tube oort Plate Plugging Limit is used for the disposition of an Alloy 600 steam generator tube Vr continued service that is experiencing predominantly axially oriented PWSCC at dented qbe support plate intersections as described in WCAP 15128, Revision 2, dated Fekuary 2000 as supplemented by TVA's letter to NRC dated March 2, 2000. This ernate repair criteria is applicable to Cycle 11 and 12 operation.

March 8, 200 SEQUOYAH - UNIT 1 3/4 4-9b Amendment No, 189, 214, 222, 252 E1-6

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) b The steam generator shall be determined OPERABLE after completing the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 4.4-2.

4.4.5.5 Reports a Following each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission within 15 days

b. The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a Special Report pursuant to Specification 6.9 2 within 12 months following completion of the inspection. This Special Report shall include

1. Number and extent of tubes inspected.

2. Location and percent of wall-thickness penetration for each indication of an imperfection.

3 Identification of tubes plugged.

c. Results of steam generator tube inspections which fall into Category C-3 shall be reported pursuant to Specification 6 6 1 prior to resumption of plant operation The written followup of this report shall provide a description of De¶etedinvestigations measures taken conducted to determine cause of the tube degradation and corrective to prevent recurrence.

d. r implementation of the voltage-based repair criteria to tube support plate intersections, noti e staff prior to returning the steam generators to service should any of the followin nitons arise.

1. If estim leakage based on the projected end-of-cycle (or if not practical using the actual m ured end-of-cycle) voltage distribution exceeds the leak limit (determined fro e licensing basis dose calculation for the postulated main steam line break) for next operating cycle.

2. If circumferential crack-like in-* ations are detected at the tube support plate intersections

3. If indications are identified that extend be d the confines of the tube support plate.

4. If indications are identified at the tube support plate el. tions that are attributable to primary water stress corrosion cracking.

5. If the calculated conditional burst probability based on the projecte end-of-cycle (or if not practical, using the actual measured end-of-cycle oltage distribution exceeds I1X 10 2 , notify the NRC and provide an assessment o e safety significance of the occurrence.

April 9, 1997 SEQUOYAH - UNIT 1 3/4 4-10 Amendment No. 36, 214, 222 E1-7

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

e. For i ntation of the depth-based repair criteria for axial PWSCC at

- dented TSPs, Its of the condition monitoring and operational assessments will be I Deleted J _

reported to the NRC wit i a s following completion of the inspection. The report will in bulations of indications found in the inspection, tabulations of tubes repaired and left in servic the ARC, and growth rate distributions for indications found in the inspection as w e growth distributions used to establish the tube repair limits. Any corrective actions found ne in the event that condition monitoring requirements are not met will be identified in the re March 8, 2000 SEQUOYAH - UNIT 1 3/4 4-10a Amendment No. 252 E1-8

REACTOR COOLANT SYSTEM BASES The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage = 150 gallons per day per steam generator). Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Sequoyah has demonstrated that primary-to-secondary leakage of 150 gallons per day steam generator can readily be detected by radiation monitors of steam generator blowdown or condenser off-gas Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged The voltage-based repair limits of SR 4.4.5 implement the guidance in GL-95-05 and are app le only to Westinghouse-designed steam generators (S/Gs) with outside diameter stress corrosion cking (ODSCC) located at the tube-to-tube support plate intersections The voltage-based repair limits a ot applicable to other forms of S/G tube degradation nor are they applicable to ODSCC that occurs at oth ocations within the SIG. Additionally, the repair criteria apply only to indications where the degradation mech sm is dominantly axial ODSCC with no significant cracks extending outside the thickness of the support p Refer to GL 95-05 for additional description of the degradation morphology.

Implementation of SR 4.4.5 rei es a derivation of the voltage structural limit from the burst versus voltage empirical correlation and the e subsequent derivation of the voltage repair limit from the structural limit (which is then implemented by th urveillance).

The voltage structural limit is the voltage from th urst pressurelbobbin voltage correlation, at the 95 percent prediction interval curve reduced to account for t wer 95/95 percent tolerance bound to tubing material properties at 6500F (i.e., the 95 percent LTL curv The voltage structural limit must be adjusted downward to account for potential flaw growth during an ope 'ng interval and to account for NDE uncertainty. The upper voltage repair limit; VURL, is determined from structural voltage limit by applying the following equation VURL = VSL - VGR - VNDE where VGR represents the allowance for flaw growth between inspections and VNDE represents allowance for potential sources of error in the measurement of the assumptions necessary to determine the voltage repairoflimit the are bobbin coil voltage.

discussed Further disc in GL 95-05. ions April 9, 1997 SEQUOYAH - UNIT I B 3/4 4-3 Amendment No. 36, 189, 214, 222 E1-9

REACTOR COOLANT SYSTEM BASES e mid-cycle equation of SR 4.4.5.4 a 10.e should only be used during unplanned inspection in which eddy t data is acquired for indications at the tube support plates.

SR 4 4.5.5 impleme veral reporting requirements recommended by GL 95-05 for situations which NRC wants to be notified prior urning the SfGs to service. For SR 4.4.5 5.d., Items 3 and 4, indications are applicable only where alterna ing criteria is being applied. For the purposes of this reporting requirement, leakage and conditional burs bility can be calculated based on the as-found voltage distribution rather than the projected end-of-cycle vo istribution (refer to GL 95-05 for more information) when it is not practical to complete these calculations usi projected EOC voltage distributions prior to returning the SIGs to service. Note that if leakage and co. nal burst probability were calculated using the measured EOC voltage distribution for the purposes of ad GIL Sections 6.a.1 and 6 a.3 reporting criteria, then the results of the projected EOC voltage distribution sh e provided per GL Section 6 b(c) criteria.

Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant.

However, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations. Plugging will be required for all tubes with imperfections exceeding the repair limit defined in Surveillance Requirement 4.4.5 4.a. The portion of the tube that the plugging limit does not apply to is the portion of the tube that is not within the RCS pressure boundary (tube end up to the start of the tube-to-tubesheet weld). The tube end to tube-to-tubesheet weld portion of the tube does not affect structural integrity of the steam generator tubes and therefore indications found in this portion of the tube will be excluded from the Result and Action Required for tube inspections It is expected that any indications that extend from this region will be detected during the scheduled tube inspections Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness Tubes experiencing outside diameter stress corrosion cracking within the thickness of the tube supp late are plugged or repaired by the criteria of 4.4.5.4.a 10 The ste enerator tube repair limits for primary water stress corrosion cracking (PWSCC) of SR 4.4.5 represents team generator tube alternate repair criteria for greater than or equal to 40 percent deep PWSCC indications ich are located within the thickness of tube support plates The repair bases for PWSCC are not applicable ther types of localized tube wall degradation located at the tube-to-tube support plate intersections The ARC includes completion of a c 'tion monitoring assessment to determine the end-of-cycle (EOC) condition of the tube bundle. An operation ssessment is completed to determine the need for tube repair on a forward-fit basis. The ARC is based o e use of crack depth profiles obtained from Plus Point analyses. Burst pressures and leak rates are calcula room depth profiles by searching the total crack length for the partial length that results in the lowest burst sure and the longest length that would tear through-wall at steam-line break conditions. The repair ba for PWSCC at dented TSP intersections is obtained by projecting the crack profile to the end of the ne erating cycle and determining if the projected profile meets the requirements of WCAP 15128, .on 2, dated February 2000 as supplemented by TVA's letter to NRC dated March 2, 2000 The following ides the limits and bases for repair established in the WCAP analyses.

March 8, 2000 SEQUOYAH - UNIT 1 B 3/4 4-4 Amendment No. 36, 189, 214, 222, 252 El-1O

Contents of this page deleted REACTOR COOLANT SYSTEM BASES Freespan Indication Repair Limits' The tube will be repaired if the crack length outside the dented TSP is >40% maximum depth Crack L gth Limit for >40% Maximum Depth The ack length limit for >40% maximum depth indications is defined as 0.375 inch from the centerline of t TSP. This limit defines the edges of the TSP thickness of 0.75 inch for Model 51 SIGs. It is acceptable for e crack to extend to both edges of the TSP as long as the maximum depth of the crack outside the TSP is 0% maximum depth and the requirements for EOC conditions are acceptable Operational Assessme Repair Bases If the indication sati es the above maximum depth and length requirements, the repair bases is then obtained by projecting th rack profile to the end of the next operating cycle and determining the burst pressure and leakage for t projected profile. The burst pressure and leakage is compared to the requirements in WCAP 15128, Re ion 2, dated February 2000 as supplemented by TVA's letter to NRC dated March 2, 2000. Separate anal es are required for the total crack length and the length outside the TSP due to differences in requirements If the projected EOC requirements are satisfied, the tube will be left in service.

The results of the condition monitorin nd operational assessments will be reported to the NRC within 120 days following completion of the inspe ion Whenever the results of any steam generato ubing inservice inspection fall into Category C-3, these results will be promptly reported to the Commissi pursuant to Specification 6 6.1 prior to resumption of plant operation. Such cases will be consid d by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory exa nations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if nec sary March 8, 2000 SEQUOYAH - UNIT 1 B 3/4 4-4a Amendment No 36, 189, 214, 222, 252 El-1I

ENCLOSURE 2 TENNESSEE VALLEY AUTHORITY SSEQUOYAH PLANT (SQN)

UNIT 1 TS CHANGE 02-02 PARTIAL WITHDRAWAL NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION TVA has concluded that operation of SQN Unit 1, in accordance with the proposed change to the technical specifications (TS) and License Condition, does not involve a significant hazards consideration. TVA's conclusion is based on its evaluation, in accordance with 10 CFR 50.91(a) (1), of the three standards set forth in 10 CFR 50.92(c).

TVA is proposing to modify SQN Unit 1 TS 3/4.4.5, "Steam Generators" to delete surveillance requirements (SRs) that describe steam generator (SG) tube plugging limits for two alternate repair criteria (ARC). The first ARC is for axial outside diameter stress corrosion cracking (ODSCC) at non-dented tube support plates and the second ARC is for axial primary water stress corrosion cracking (PWSCC) at dented tube support plates. TVA' s proposed amendment removes both ARCs through the deletion of the following SRs: SR 4.4.5.2.b.4, 4.4.5.2.d, 4.4.5.2.e, a portion of 4.4.5.4.a.6, 4.4.5.4.a.10, 4.4.5.4.a.11, 4.4.5.5.d, and 4.4.5.5.e. TVA' s proposed removal of these SRs for ARC reestablishes standard tube plugging criteria within the TS for SQN Unit 1. Returning to the standard TS 40 percent through-wall tube plugging limit is inherently more conservative.

Included with the above change is deletion of License Condition 2.C.9.d that references prior TVA commitment letters for SG inspection. The TVA letters and their commitments will no longer apply following replacement of the Unit 1 SGs.

A. The proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

TVA' s proposed TS amendment does not compromise limits associated with SG tube integrity. TVA' s proposed change removes existing SG tube plugging criteria (i.e., ARC) from the TS and reestablishes the standard TS criteria (40 percent through-wall criteria). This change is inherently more conservative.

E2-1

The proposed revision does not alter plant equipment, test methods or operating practices. The proposed change continues to provide controls for safe operation of SQN SGs within the required limits. The proposed change does not contribute to events or assumptions associated with postulated design basis accidents (i.e., SG tube rupture). The proposed change does not affect operator indicators or actions required to diagnose or mitigate a SG tube rupture accident. The proposed revisions continue to maintain the required safety functions. Accordingly, the probability of an accident or the consequences of an accident previously evaluated is not increased.

B. The proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

TVA' s proposed amendment removes existing repair criteria and incorporates the more conservative TS limit for SG tube plugging (i.e., plug tubes with degradation depths equal to or greater than 40 percent through-wall). This change will not give rise to new failure modes. The failure of a SG tube to maintain leakage integrity during operation is an analyzed event in the SQN Updated Final Safety Analysis Report.

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

C. The proposed amendment does not involve a significant reduction in a margin of safety.

TVA' s proposed TS amendment is conservative with respect to the margin of safety. The margin of safety is preserved through ensuring structural integrity and leakage integrity of the SG tubes.

TVA' s proposed change to remove ARC from the TS does not compromise structural integrity or leakage integrity of SG tubes. The proposed change invokes the standard TS tube plugging criteria limit (40 percent through-wall criteria) which is inherently more conservative.

The proposed change does not affect the plant conditions, setpoints, or safety limits that could result in precursors to accidents or degrade accident mitigation systems. Plant system safety functions are not altered by the proposed change. Consequently, the proposed TS revisions does not reduce the margin of safety.

E2-2