NL-06-0990, Southern Nuclear Operating Company Response to NRC Questions Regarding the Vogtle Electric Generating Plant Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity

Southern Nuclear Operating Company Response to NRC Questions Regarding the Vogtle Electric Generating Plant Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity

ML061570032
Person / Time
Site:	Vogtle
Issue date:	06/05/2006
From:	Grissette D Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-06-0990
Download: ML061570032 (27)

Text

Don E. Grissette Vice President Southern Nuclear Operating Company, Inc.

40 lnverness Center Parkway Post Office Box 1295 Birmingham, Alabama 35201 Tel 205.992.6474 Fax 205.992.0341 June 5, 2006 Docket Nos.:

50-424 50-425 Energy to Serve Your World" NL-06-0990 U. S. Nuclear Regulatory Commission A'ITN: Document Control Desk Washington, D. C. 20555-0001 Southern Nuclear Operating Company Response to NRC Questions Regarding the Vogtle Electric Generating Plant Application for Technical S~ecification Imvrovement Regarding Steam Generator Tube Integrity Ladies and Gentlemen:

On May 4,2006, Southern Nuclear Operating Company (SNC) received five questions by facsimile from the NRC staff concerning the March 29,2006, Vogtle Electric Generating Plant (VEGP) application for technical specification (TS) improvement regarding steam generator tube integrity, consistent with NRC-approved Revision 4 to TSTF-449. The SNC response to these questions is enclosed. contains the SNC response to the NRC questions, and Enclosures 2 and 3 contain the revised markup and clean typed TS and TS Bases pages, respectively, that were affected by the SNC response. Changes from the previous submittal are identified by "clouded" markings in the Enclosure 2 markup pages. The SNC response to the NRC questions does not affect the 10 CFR 50.92 evaluation or any other documentation supporting the SNC TSTF-449 submittal dated March 29,2006.

Mr. D. E. Grissette states he is a Vice President of Southern Nuclear Operating Company, is authorized to execute this oath on behalf of Southern Nuclear Operating Company and to the best of his knowledge and belief, the facts set forth in this letter are true.

This letter contains no NRC commitments. If you have any questions, please advise.

Respectfully submitted, SOUTHERN NUCLEAR OPERATING COMPANY

'..1!"P,%n E. Grissette

U. S. Nuclear Regulatory Commission NL-06-0990 Page 2

Enclosures:

1.

SNC Response to NRC Questions Regarding the Vogtle Electric Generating Plant Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity

2.

Revised Markup Technical Specification and Technical Specification Bases Pages Resulting From the SNC Response to NRC Questions

3.

Revised Clean Technical Specification and Technical Specification Bases Pages Resulting From the SNC Response to NRC Questions cc:

Southern Nuclear Operating Comvany Mr. J. T. Gasser, Executive Vice President Mr. T. E. Tynan, General Manager - Plant Vogtle RType: CVC7000 U. S. Nuclear Regulatory Commission Dr. W. D. Travers, Regional Administrator Mr. C. Gratton, NRR Project Manager - Vogtle Mr. G. J. McCoy, Senior Resident Inspector - Vogtle State of Georgia Mr. L. C. Barrett, Commissioner - Department of Natural Resources SNC Response to NRC Questions Regarding the Vogtle Electric Generating Plant Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity SNC Response to NRC Questions Regarding the Vogtle Electric Generating Plant Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity

1.

NRC Ouestion Page 1 of Insert 5.5.9 in your proposed technical specification (TS) states that leakage would not exceed 1 gallon-per-minute (gpm) per steam generator (SG) except for specific types of degradation at specific locations as described in paragraph c of the Steam Generator Program. However, the actual alternate tube repair criteria that may be applied is described in TS Section 5.5.9.c.l.

Please discuss your plans to modify your TSs to more precisely reflect the location of the applicable alternate tube repair criteria. Since proposed TS Section 5.5.9.c.1 does not specify what sources of accident induced leakage shall be limited to 1 gpm, discuss your plans to clarify the accident induced leakage performance criterion. For example, "Leakage from all sources, excluding the leakage attributed to the degradation described in TS Section 5.5.9.c.1, is not to exceed 1 gpm per SG."

The staff also notes that the 1" paragraph in page B3.4.17-4 in your proposed Bases may also need to be revised to clarify this issue.

SNC Response Southern Nuclear Operating Company (SNC) includes leakage attributed to the degradation described in TS Section 5.5.9.c.1 within the 1 gprn per SG limit.

Therefore, the statement ", except for specij?~

types of degradation at specljic locations as described in paragraph c of the Steam Generator Program " will be deleted from the last sentence in TS Section 5.5.9.b.2. In addition, the statement

", except for specijic types of degradation at spec@ locations where the NRC has approved greater accident induced leakage " will be deleted from the first sentence on page B3.4.17-4 of the TS Bases.

2.

NRC Ouestion On page 5.5-1 1 of your current TS, the last sentence in your definition of plugging limit states that: "For Unit 2 during refueling outage 11 and the subsequent operating cycle, degradation identified in the portion of the tube from the top of the hot leg tubesheet to 17 inches below the top of the tubesheet shall be plugged upon detection." Upon reviewing your proposed new TS, the staff found that in your Insert 5.5.9, the requirement for plugging on detection tubes with flaws in the top 17 inches of the tubesheet is missing. Please provide the basis for why this requirement is no longer needed or modify your proposed TS to be consistent with your current TS with respect to this issue.

The staff also notes that your Bases may also need to be revised to clarify this issue (page B3.4.17-2).

SNC Response to NRC Questions Regarding the Vogtle Electric Generating Plant Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity SNC Response SNC will add the sentence "For Unit 2 during Refieling Outage I 1 and the subsequent operating cycle, degradation identged in the portion ofthe tube fiom the top of the hot leg tubesheet to 17 inches below the top of the tubesheet shall be plugged upon detection. " to TS Section 5.5.9.c.1 and to TS Bases B3.4.17, LC0 Section, 3rd paragraph. '

3.

NRC Ouestion In your proposed TS Bases, page B3.4-13-2, it appears that three paragraphs are missing from the Applicable Safety Analyses section. The paragraphs are the following:

a) "Primary to secondary LEAKAGE is a factor on 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 steam generator tube rupture (SGTR). The leakage contaminates the secondary fluid."

b) "The FSAR (Ref. 3) analysis for SGTR assumes the contaminated secondary fluid is only briefly released via safety valves and the majority is steamed to the condenser. The (1 gpm) primary to secondary LEAKAGE safety analysis assumption is relatively inconsequential."

c) "The SLB is more limiting for site radiation releases. The safety analysis for the SLB accident assumes the entire (1 gpm) primary to secondary LEAKAGE is through the affected generator as an initial condition. The dose consequences resulting from the SLB accident are well within the limits defined in 10 CFR 100 or the staff approved licensing basis (i.e., a small fraction of these limits)."

Please discuss why these three paragraphs (or similar paragraphs) are not in your Bases (or alternatively discuss your plans to modify your Bases to incorporate these paragraphs in order to be fully consistent with the TSTF-449).

SNC Response to NRC Questions Regarding the Vogtle Electric Generating Plant Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity SNC Response The three paragraphs are not part of TSTF-449. TSTF-449 was incorporated into a version of a Bases that is different from the version adopted for use by Vogtle Electric Generating Plant (VEGP). The current VEGP Bases does not contain the three paragraphs; therefore, these three paragraphs were not included in the SNC submittal.

4.

NRC Ouestion On page 3.4.13-2 of your proposed TS, there are three Notes associated with Surveillance Requirement (SR) 3.4.13.1 that are later described in detail on pages B3.4.13-4 and B3.4.13-5 of your proposed Bases. Upon reviewing the SR description, the staff found a discrepancy between the number of Notes and their description. For example, the second paragraph of Section SR 3.4.13.1 contains a sentence that reads: "The Surveillance is modified by two Notes."

The "two" should be a "three." In addition, the 5" paragraph of the same section states: "Note 2 states that this SR is not applicable to primary to secondary LEAKAGE because of 150 gallons per day cannot be measured accurately by an RCS (reactor coolant system) water inventory balance." This description corresponds to Note 3 of the S R Please discuss your plans to modify the Bases so that it accurately reflects your TSs Notes identification number and its respective description.

All references to "two Notes" and "Note 2" will be changed to "three Notes" and "Note 3" on pages B3.4.13-4 and B3.4.13-5. In addition, reference to "a Note" on page B3.4.13-5 will be changed to "Note 2."

5.

NRC Ouestion In your marked up version of TS Section 3.4.17, there are several references to

"[or repaired]." In your re-typed TS pages, the "[or repaired]" is removed.

Please confirm that the "[or repaired]" should not have been included in the marked up version of TS Section 3.4.17.

SNC Response The references to "[or repaired]" will be deleted from the mark-up version of TS Section 3.4.17.

Revised Markup Technical Specification and Technical Specification Bases Pages Resulting From the SNC Response to NRC Questions

SG Tube Integrity 3.4.17 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.17 Steam Generator (SO) Tube Integrity LC0 3.4.17 SG tube integrity shall be maintained.

AND satisfying ttm tube repair cdtaia shall be accordance with the Steam Generator Progrem.

APPLICABILITY:

MODES 1,2,3, and 4, ACTIONS

---NOTE-Separate Condition entry is allowed for each SG tube.

CONDITION 1

REQUIRED ACTION 1 COMPMONTlME A. One or more SG tubes satisfying the tube repair plugged h the Steam Generator Program.

A1 V e f i tube integrity of the affected tube(s) is maintained until the next refueling outage or SG tube inspection.

ANR A2 PI tu the Steam Generator Program.

7 days Prior to entering MODE 4 following the I next refueling outege or SG tube inspection

8. Required Action and associated Completion Tlme of Candltion A not met.

SG tube Integrity not mafntahd.

t 8.1 Be In MODE 3.

AW B.2 Be In MODE 5.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 36 hours Vogtle Units 1 and 2 3.4.17-1 Amendment No. xx (Unlt I)

Amendment No. xx (Unit 2)

SG Tube Integrity 3.4.1 7 SURVEILLANCE REQUIREMENTS SURVEILLANCE I

FREQUENCY SR 3.4.17.1 Verify SG tube integrity in accordance with the Steam Generator Program.

In accordance with the Steam Generator Program Vogtle Units 1 and 2 SR 3.4.17.2 Verify that each inspected S isfies the tube repair criteria is plugge accordance with the Steam

m.

Amendment No. xx (Unit 1)

Amendment No. xx (Unit 2)

Prior to entering MODE 4 following a SG tube inspection

A Steam Generator Program shall be established and lmplemiented to ensure that SO tube Integrity is malntalned. In addltlon, the Steam Generatw Program shall Include the following provisions:

a.

Proddons for condition monitoring assessments. Condition monitoring asse86ment means an evaluation of the *as found conditkn of the tubing with respect to the performance criteria for sbuctural lntegdty and acddent Induced leakage. The 'as found condition refers to the condition of the tubing during an SG inspection outage, as determined from the InseNtce inspection results or by other means, prior to the plugging of tubes. Condition monbdng assessments shall be conducted dMng each outage during whkh the SO tubes are inspected or plugged, to confirm that the performance aiteda 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, addent induced leakage, and operational LEAKAGE.

Structural integrity performance crltetion: All in-senrice steam generator tubes shell retaln structural integrity over the full range of normal operating conditions (Including startup, operation In the power range, hot standby, and oool down and all antldpeted transients Included In the design specffication) and U g n bask accidents. This indudes retaining a safety factor of 3.0 agalnst burst under normal steady state full p o w operatbn pdmary-to-secondary pressure dlfferenbial and a safety factor of 1.4 agalnst burst applied to the design basis accklent pdmary-to-secondery pressure dtfferentlals. Apart from the ebove requirements, additional bading conditions assodated with the deslgn bask accidents, or combination of accidents In accordance with the design and licensing basis, shaH also be evaluated to determine If the associated bads contribute significantly to burst or collapse. In the assessment of tube integrity, those bads that do slgntficantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

2.

Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakape rate assumed in the acddent anahrsis in terms of total

3.

The operational LEAKAGE performance criterion b spedfied in LC0 3.4.1 3, 'RCS Operational LEAKAGE."

c.

Provisions for SQ tube repair criteria. Tubes found by inse~ce inspection to contain flaws wlth a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.

Page 1

I depth based-criteria:

For Unit 2 during Refueling Outage 11 and the subsequent operating cycle, degradation found in the portion of the tube below 17 inches from the top of the hot leg tubesheet does not require plugging.

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., volumetrlc 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. For Unit 2 during Refueling Outage 11 and the subsequent operating cycle, the portion of the tube below 17 inches from the top of the hot leg tubesheet is excluded. 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 whim 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 120,90, 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 48 effective full power months or two 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 nondestructive 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.

Page 2

RCS Operational LEAKAGE B 3.4.13 from the RCS, the loss must be included in the allowable identified LEAKAGE.

1 I

ACTIONS

&I Unidentified LEAKAGES; identified LEAKAGE-;-

in excess of the LC0 limits 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 time to verify leakage rates and either identify unidentified LEAKAGE or reduce LEAKAGE to within limits before the reactor must be shut down. This action is necessary to prevent further deterioration of the RCPB.

B.l and 8.2 If any pressure boundary LEAKAGE exists, or ~rimarv to secondary LEAKAGE is nat *in lirnit,or if unidentified Q&AW&&

identified 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. 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 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-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 canditlans 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.

SURVEILLANCE SR 3.4.13.1 REQUIREMENTS Verifyhg RCS LEAKAGE to be within the LC0 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 by performance of an RCS water inventory balance. &hwyh The RCS water inventory balance must be performed with the reactor at steady state operating conditions. The Surveillance is modified bv otes. Note 1 states-that

? -

this SR is not required to be 1' I

Vogtle Units 1 and 2 B 3.4.13-4 Revision No. 0

RCS Operational LEAKAGE B 3.4.13 Vogtle Units 1 and 2 performed in MODES 3 and 4 until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of steady state operation I have been established. In all cases, this SR is required to be performed prior to entering MODE 2 to ensure the assessment of RCS leakage prior to critical operation.

Steady state operation is required to perform a proper invent0 balance: calculations during maneuvering are not useful an-requires the Surveillance to be performed when steady state IS established. 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 sump level. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. These leakage detection systems are specified in LC0 3.4.15, "RCS Leakage Detection Instrumentation."

measured awuraietv bv an RCS wter inventow balance.

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. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency after steady state operation has been achieved provides for those situations where a transient occurs, and the duration of the transient is such that the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Frequency plus the 25% extension allowed by SR 3.0.2 would be exceeded. In this event, the SR would be due within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after steady state operation has been reestablished.

SR 3.4.13.2 This SR verifies that ~rimarv to secondaw LEAKAGE is less than or eaual to 150 aallons per dav throuah anr one SG. Satisfvina the prirnaw to secundaw LEAKAGE limit ensures that the a~erational LE3KAGE ~erformancs criterion in the Steam Generator Prwram is met. If this SR is not met. corn~liance with LC0 3.4.1 7. 'Steam Generator Tube Intsnritv," should be evaluated. The 150 aallons per dav limit is measured at room kmoerature as described in Reference anv one SG. If it is not ~racticai to assian the LEAKAGE to an individual SG. all the ~rirnarv to secondary LEAKAGE should be conservativelv assumed to be from one SG.

The Surveillance is modified bv a Note which states that the Surveillance is not reauired 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 ooeration. For RCS wirnarv to RCS Dressure, temmrature, oowef level. ~ressusizer and rnakeu~

B 3.4.13-5 Revision No. 0

SO Tube Integrity B 3.4.17 BASES APPUCABLE The eteam generator tube ~pture (SGTR) acddent Is the limltlng ddgn SAFETY bae'i event for SG tubes and avofdhg an SGTR is the baeis for thls ANALYSES Spedflcatbn. The analysls of a SGTR event assumes a bounding prlmary to secondary LEAKAGE rate equal to the operational LEAKAQE rate limlts In LC0 3.4.1 3, 'RCS Operational LEAKAGE,' plus the leakage rate assodated with a double-ended rupture of a dngle tube. The accldent analysis for a SGTR assumes the contaminated secondary fluid Is only brlefly released to the atmosphere vla safety valves and the majority is d'lscharged to the maln condenser.

The analysis for design basis accidents and transients other than a SOTR assume the SG tubes retain thelr structural lntegrlty (Le., they are assumed not to rupture.) In these enalyses, the steam discharge to the atmosphere Is based on the total prlmary to secondary L E A W E from all SGs of 1 gallon per mlnute or Is assumed to Increase to 1 gallon per minute as a result of accident induced condltbns. For accidents that do not lnvohre fuel damage, the primary coolant activity level of DOSE EQUIVALENT 1-131 la assumed to be equal to the LC0 3.4.16, "RCS Specific Activity," Ilmlts. For accidents that assume fuel damage, the primary coolant adivity is a function of the amount of activity released f tom the damaged fuel. The dose consequences of these events are within the llmb of GDC 19 (Ref. 2), 10 CFR 100 (Ref, 3) or the NRC approved licensing basla (e.g., a smell fraction of these limits).

Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(~)(2)(1).

The LC0 requires that SQ tube Integrity be maintained. 'The LC0 also requlres that all SG tubes that satisfy the repalr criteria be plugged In accordance with the Steam Generator Program.

During an SO inspection, any inspected tube that satisfies the Steam Generator Program repalr crltetla Is removed from sendce by plugglng. tf a tube was determined to satisfy the repair crlterla but was not plugged, the tube may still have tube Integrity.

In the context of this Specification, a SG tube Is defined as the entlre length of the tube, including the tube wall, between the tube-to-tubesheet weld at the tube Inlet and the tubeto-tubesheet weld et the tube outlet.

For Unit 2 during Refueling Outage 11 and the subsequent operating cycle, degradaffon found In the portion of the tube below 1 the top of the hot leg tubesheet does not require plugglng.

during Refueling Outage 11 and the subsequent operatlng cycle, the portion of the tube below 17 Inches from the top of the hot leg tubesheet is excluded from tube Inspections (Ref, 7). The tube-to-tubesheet weM is not considered part of the tube.

SG Tube Integrity B 3.4.17 BASES LC0 (continued) anal 3 '

LEAKAGE existing prior to the accident in additlon to primary to secondary LEAKAGE induced during the accident.

'The operatlonal LEAKAGE performance criterion provides an observable indication of SG tube conditions during plant operation. The limit on operational LEAKAGE is contained in LC0 3.4.13, "RCS Operational LEAKAGE," and llmits 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 differentlal across the tubes is large. Large differential pressures across SG tubes can only be experienced in MODE 1,2,3, or 4.

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

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

P.1 and A.2 Condition A applies if it is discovered that one or more SG tubes examined in an inservice Inspection satlsfy the tube repair criteria but were not plugged in accordance with the Steam Generator Program as required by SR 3.4.172.

An evaluation of SG tube integrity of the affected tube(s) must be made. Steam generator tube integrity is based on meeting the SO performance criteria described in the Steam Generator Program. The SG repair criteria deflne limits on SO tube degradation that allow for flaw growth between inspections while still providing assurance that the SO performance criteria will continue to be met. In order to determine if a SO tube that should have been plugged Vogtle Units 1 and 2 B 3.4.1 7-4 Rev. X.X Revised Clean Technical Specification and Technical Specification Bases Pages Resulting From the SNC Response to NRC Questions

INSERT 5.5.9 A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions:

a.

Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged, to confirm that the performance criteria are being met.

b.

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.

Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down 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 with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

2.

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

3.

The operational LEAKAGE performance criterion is specified in LC0 3.4.13, "RCS 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.

Page 1

The following alternate tube repair criteria may be applied as an alternative to the 40%

depth based criteria:

1.

For Unit 2 during Refueling Outage 11 and the subsequent operating cycle, degradation found in the portion of the tube below 17 inches from the top of the hot leg tubesheet does not require plugging.

For Unit 2 during Refueling Outage 11 and the subsequent operating cycle, degradation identified in the portion of the tube from the top of the hot leg tubesheet to 17 inches below the top of the tubesheet shall be plugged upon detection.

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. For Unit 2 during Refueling Outage 11 and the subsequent operating cycle, the portion of the tube below 17 inches from the top of the hot leg tubesheet is excluded. 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.

lnspect 100% of the tubes in each SG during the first refueling outage following SG replacement.

2.

lnspect 100% of the tubes at sequential periods of 120, 90, 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 48 effective full power months or two 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 nondestructive 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.

Page 2

Programs and Manuals 5.5 5.5 Programs and Manuals Steam Generator (SG) Pronram (continued)

2.

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

3. The operational LEAKAGE performance criterion is specified in LC0 3.4.13, "RCS 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.

The following alternate tube repair criteria may be applied as an alternative to the 40% depth based criteria:

1. For Unit 2 during Refueling Outage 11 and the subsequent operating cycle, degradation found in the portion of the tube below 17 inches from the top of the hot leg tubesheet does not require plugging.

For Unit 2 during Refueling Outage 11 and the subsequent operating cycle, degradation identified in the portion of the tube from the top of the hot leg tubesheet to 17 inches below the top of the tubesheet shall be plugged upon detection.

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. For Unit 2 during Refueling Outage 11 and the subsequent operating cycle, the portion of the tube below 17 inches from the top of the hot leg tubesheet is excluded. 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 (continued)

Vogtle Units 1 and 2 Amendment No.

(Unit 1)

Amendment No.

(Unit 2)

RCS Operational LEAKAGE B 3.4.13 from the RCS, the loss must be included in the allowable identified LEAKAGE.

I I

ACTIONS A. 1 Unidentified LEAKAGE or7 identified LEAKAGE--

in excess of the LC0 limits 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 time to verify leakage rates and either identify unidentified LEAKAGE or reduce LEAKAGE to within limits before the reactor must be shut down. This action is necessary to prevent further deterioration of the RCPB.

6.1 and 8.2 If any pressure boundary LEAKAGE exists, or ~rimarv to secondarv LEAKAGE is not within limit, or if unidentified &fMAG&

identified L

E A

K A

G E

E 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. 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 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-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.

SURVEILLANCE SR 3.4.13.1 REQUIREMENTS Verifying RCS LEAKAGE to be within the LC0 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 by performance of an RCS water inventory balance. Pmwyte The RCS water inventory balance must be performed with the reactor at steady state operating conditions. The Surveillance is modified bv three Notes. Note 1 state-that this SR is not required to be 1 '

performed in MODES 3 and 4 until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of steady state operation 1

(continued)

Vogtle Units 1 and 2 B 3.4.1 3-4 Revision No. 0

RCS Operational LEAKAGE B 3.4.13 have been established. In all cases, this SR is required to be performed prior to entering MODE 2 to ensure the assessment of RCS leakage prior to critical operation.

Steady state operation is required to perform a proper inventory balance; calculations during maneuvering are not useful and Note 2 1

requires the Surveillance to be performed when steady state is established. 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 sump level. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. These leakage detection systems are specified in LC0 3.4.15, "RCS Leakage Detection Instrumentation."

Note 3 states that this SR is not applicable to primaw to secondary LEAKAGE because LEAKAGE of 150 qallons per dav cannot be measured accuratelv by an RCS water inventow balance.

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. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency after steady state operation has been achieved provides for those situations where a transient occurs, and the duration of the transient is such that the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Frequency plus the 25% extension allowed by SR 3.0.2 would be exceeded. In this event, the SR would be due within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after steady state operation has been reestablished.

This SR verifies that primarv to secondarv LEAKAGE is less than or equal to 150 nallons per dav throuah anv one SG. Satisfvina the primarv to secondarv LEAKAGE limit ensures that the operational LEAKAGE performance criterion in the Steam Generator Proqram is met. If this SR is not met, compliance with LC0 3.4.17, "Steam Generator Tube Intearitv," should be evaluated. The 150 qallons per dav limit is measured at room temperature as described in Reference

5. The operational LEAKAGE rate limit applies to LEAKAGE throuah anv one SG. If it is not practical to assiqn the LEAKAGE to an individual SG. all the primarv to sewndaw LEAKAGE should be wnservativelv assumed to be from one SG.

The Surveillance is modified bv 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 steadv state operation. For RCS primarv to secondarv LEAKAGE determination, steadv state is defined as stable RCS pressure, temperature, power level. pressurizer and makeup (continued)

Vogtle Units 1 and 2 B 3.4.13-5 Revision No. 0

SG Tube Integrity B 3.4.17 BASES APPLICABLE The steam generator tube rupture (SGTR) accident is the limiting design SAFETY basis event for SG tubes and avoiding an SGTR is the basis for this ANALYSES Specification. The analysis of a SGTR event assumes a bounding primary to secondary LEAKAGE rate equal to the operational LEAKAGE rate limits in LC0 3.4.1 3, "RCS 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 the total primary to secondary LEAKAGE from all SGs of 1 gallon per minute or is assumed to increase to 1 gallon per minute as a result of accident induced conditions. 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 LC0 3.4.16, "RCS Specific Activity," limits. For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaged fuel. The dose consequences of these events are within the limits of GDC 19 (Ref. 2), 10 CFR 100 (Ref. 3) or the NRC approved licensing basis (e.g., a small fraction of these limits).

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

'The LC0 requires that SG tube integrity be maintained. The LC0 also requires that all SG tubes that satisfy the repair criteria be plugged in accordance with the Steam Generator Program.

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

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

For Unit 2 during Refueling Outage 11 and the subsequent operating cycle, degradation found in the portion of the tube below 17 inches from the top of the hot leg tubesheet does not require plugging. For Unit 2 during Refueling Outage 11 and the subsequent operating cycle, degradation identified in the portion of the tube from the top of the hot leg tubesheet to 17 inches below the top of the tubesheet shall be plugged upon detection. For Unit 2 during Refueling Outage 11 and the subsequent operatirrg cycle, the portion of the tube below 17 inches from Vogtle Units 1 and 2 B 3.4.1 7-2 Rev. X.X

SG Tube Integrity B 3.4.1 7 BASES LC0 (continued) the top of the hot leg tubesheet is excluded from tube inspections (Ref. 7). 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 5.5.9, "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.

Structural integrity requires that the primary membrane stress intensity in a tube not exceed the yield strength for all ASME Code, Section Ill, Service Level A (normal operating conditions) and Service Level B (upset or abnormal conditions) transients included in the design specification.

This includes safety factors and applicable design basis loads based on ASME Code, Section Ill, Subsection NB (Ref. 4) and Draft Regulatory Guide 1.121 (Ref. 5).

Vogtle Units 1 and 2 B 3.4.17-3 Rev. X.X

SG Tube Integrity B 3.4.17 BASES LC0 (continued)

The accident induced leakage performance criterion ensures that the primary to secondary LEAKAGE caused by a design basis accident, other than a SGTR, is within the accident analysis assumptions. The accident analysis assumes that accident induced leakage does not exceed 1 gpm per SG. 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 LC0 3.4.1 3, "RCS Operational LEAKAGE," and limits primary to secondary LEAKAGE through any one SG to 150 gallons per day. This limit is based on the assumption that a single crack leaking this amount would not propagate to 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

~otential for LEAKAGE.

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

A.l and A.2 Condition 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 3.4.1 7.2. An evaluation of SG tube integrity of the affected tube(s) must be made. Steam generator tube integrity is based on meeting the SG performance criteria described in the Steam Generator Program. The SG repair criteria define limits on SG tube Vogtle Units 1 and 2 Rev. X.X

SG Tube Integrity B 3.4.17 BASES ACTIONS (continued) 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, Condition B applies.

A Completion 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, Required Action A.2 allows plant operation to continue until the next refueling outage or SG inspection provided the inspection interval continues to be supported by an operational assessment that reflects the affected tubes. However, the affected tube(s) must be plugged prior to entering MODE 4 following the next refueling outage or SG inspection.

'This Completion Time is acceptable since operation until the next inspection is supported by the operational assessment.

B.l and 8.2 If the Required Actions and associated Completion Times of Condition A are not met or if SG tube integrity is not being maintained, 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 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

The allowed Completion Times are reasonable, based on operating experience, to reach the desired plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.17.1 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. I),

and its referenced EPRl 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.

Vogtle Units 1 and 2 B 3.4.17-5 Rev. X.X

SG Tube Integrity B 3.4.1 7 BASES SURVEILLANCE REQUIREMENTS (continued)

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. lnspection 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.

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

The Steam Generator Program defines the Frequency of SR 3.4.17.1.

The Frequency is determined by the operational assessment and other limits in the SG examination guidelines (Ref. 6). The Steam Generator Program uses information on existing degradations and growth rates to determine an inspection Frequency that provides reasonable assurance that the tubing will meet the SG performance criteria at the next scheduled inspection. In addition, Specification 5.5.9 contains prescriptive requirements concernirlg inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections.

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 Speci.fication 5.5.9 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.

Vogtle Units 1 and 2 B 3.4.17-6 Rev. X.X

SG Tube Integrity B 3.4.1 7 BASES SURVEILLANCE REQUIREMENTS (continued)

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

REFERENCES

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

2.

10 CFR 50 Appendix A, GDC 19.

4.

ASME Boiler and Pressure Vessel Code, Section Ill, 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."

7. License Amendment Nos. 138 and 1 17, "Vogtle Electric Generating Plant Units 1 and 2, Re: Issuance of Amendments Regarding the Steam Generator Tube Surveillance Program (TAC Nos. MC8078 and MC8079)," September 21,2005.

Vogtle Units 1 and 2 B 3.4.17-7 Rev. X.X