RC-13-0047, License Amendment Request - LAR 12-03567, Application to Revise Tech Specs to Adopt TSTF-510, Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection, Using Consolidated Line Item Improvement..

License Amendment Request - LAR 12-03567, Application to Revise Tech Specs to Adopt TSTF-510, Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection, Using Consolidated Line Item Improvement..

ML13095A106
Person / Time
Site:	Summer
Issue date:	04/02/2013
From:	Gatlin T South Carolina Electric & Gas Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RC-13-0047
Download: ML13095A106 (34)

Text

Thomas D. Gatlin Vice President,Nuclear Operations 803.345.4342 A SCANA COMPANY April 2, 2013 RC-13-0047 U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555-0001

Subject:

VIRGIL C. SUMMER NUCLEAR STATION UNIT 1 (VCSNS)

DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 LICENSE AMENDMENT REQUEST - LAR 12-03567 Application to Revise Technical Specifications to Adopt TSTF-510, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," Using The Consolidated Line Item Improvement Process Pursuant to 10 CFR 50.90, South Carolina Electric & Gas Company (SCE&G), acting for itself and as an agent for South Carolina Public Service Authority, is herby submitting the following request for an amendment to the Technical Specifications (TS) for VCSNS.

The proposed amendment would modify TS requirements regarding steam generator tube inspections and reporting as described in TSTF-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection." provides a description and assessment of the proposed changes, the requested confirmation of applicability, and plant-specific verifications. Attachment 2 provides the existing TS pages marked up to show the proposed changes. Attachment 3 provides revised TS pages.

Approval of the proposed amendment is requested by April 4, 2014 in support of the next scheduled refueling outage scheduled for April 2014. Once approved, the amendment shall be implemented within 60 days. There are no new or revised commitments associated with this proposed change.

Virgil C.Summer Station - Post Office Box 88 - Jenkinsville, SC - 29065

• F(803) 345-5209

Document Control Desk CR-12-03567 RC-13-0047 Page 2 of 2 In accordance with 10CFR50.91 a copy of this application is being provided to the designated South Carolina Official.

The proposed change has been reviewed and approved by both the VCSNS Plant Safety Review Committee and the VCSNS Nuclear Safety Review Committee.

If you should have any questions, please contact Mr. Bruce L. Thompson at (803) 931-5042.

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

Executed On ThomatY.Gfi JMG/TDG/bq Attachments:

1. Description and Assessment

2. Proposed Technical Specification Changes (Mark-Up)

3. Revised Technical Specification Pages

4. List of Regulatory Commitments cc: K. B. Marsh NRC Resident Inspector S. A. Byrne S. E. Jenkins J. B. Archie Paulette Ledbetter N. S. Cams K. M. Sutton J. H. Hamilton NSRC J. W. Williams RTS (CR-12-03567)

W. M. Cherry File (813.20)

V. M. McCree PRSF (RC-13-0047)

E. A. Brown

Document Control Desk CR-12-03567 RC-13-0047 Page 1 of 4 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) Unit 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 ATTACHMENT I DESCRIPTION AND ASSESSMENT

1.0 DESCRIPTION

The proposed change revises Technical Specification (TS) 3.4.5, "Steam Generator Tube Integrity," 6.8.4.k, "Steam Generator (SG) Program," and 6.9.1.12, "Steam Generator Tube Inspection Report." The proposed changes are needed to address implementation issues associated with the inspection periods, and address other administrative changes and clarifications.

The proposed amendment is consistent with TSTF-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection."

2.0 ASSESSMENT 2.1 Applicability of Published Safety Evaluation South Carolina Electric & Gas Company (SCE&G) has reviewed TSTF-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," (ADAMS Accession No. ML110610350) and the Model Safety Evaluation dated October 19, 2011 (ADAMS Accession No. ML112101513) as identified in the Federal Register Notice of Availability, dated October 27, 2011 (76 FR 66763). As described in the subsequent paragraphs, SCE&G has concluded that the justifications presented in TSTF-510 and the Model Safety Evaluation prepared by the Nuclear Regulatory Commission (NRC) staff is applicable to VCSNS Unit 1 and justify this amendment for incorporation of the changes to the VCSNS TS.

2.2 Optional Changes and Variations SCE&G is not proposing any technical variations or deviations from the TS changes described in TSTF-510, Revision 2, or the applicable parts of the NRC staff's Model Safety Evaluation. However, SCE&G is proposing the following

Document Control Desk CR-12-03567 RC-1 3-0047 Page 2 of 4 administrative variations from the TS changes described in TSTF-51 0, Revision 2.

The VCSNS Unit 1 TS numbering system is different than the Standard Technical Specifications (STS) NUREG-1431 Revision 4.0 on which TSTF-510 was based. Specifically, the "Steam Generator (SG) Program" in the VCSNS Unit 1 TS is numbered 6.8.4.k rather than 5.5.9, the "Steam Generator Tube Integrity" TS is numbered 3.4.5 rather than 3.4.20, and the "Steam Generator Tube Inspection Report" is numbered 6.9.1.12 rather than 5.5.9. These differences are administrative and do not affect the applicability of TSTF-510 to the VCSNS and the TS numbering scheme.

In addition, within Section 4.0 of TSTF-510 Revision 2, there are three versions for adjusting the inspection sample based on tube material (one each for 600MA tubing, 600TT tubing, and 690TT tubing) contain the following statement:

"If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube repair criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated."

This administrative error was identified in a February NRC-TSTF meeting and documented in a letter from the TSTF to the NRC dated March 28, 2012 (TSTF letter No. 12-09). The underlined phrase above should state "tube plugging [or repair] criteria," consistent with the other changes made in TSTF-510-A revision

2. SCE&G is changing the phrase to "tube plugging criteria" as reflected within this amendment request, Attachment 2 titled "Insert - A." This change is administrative and should not result in this application being removed from the Consolidated Line Item Improvement Process (CLIIP).

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Determination SCE&G requests adoption of an approved change to the plant specific Technical Specifications for VCSNS Unit 1, to revise TS 6.8.4.k, "Steam Generator (SG)

Program," TS 6.9.1.12, "Steam Generator Tube Inspection Report," and TS 3.4.5, "Steam Generator Tube Integrity," to address inspection periods and other administrative changes and clarifications.

As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below:

Document Control Desk CR-12-03567 RC-1 3-0047 Page 3 of 4

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

Response: No.

The proposed change revises the Steam Generator (SG) Program to modify the frequency of verification of SG tube integrity and SG tube sample selection. A steam generator tube rupture (SGTR) event is one of the design basis accidents that are analyzed as part of a plant's licensing basis. The proposed SG tube inspection frequency and sample selection criteria will continue to ensure that the SG tubes are inspected such that the probability of a SGTR is not increased. The consequences of a SGTR are bounded by the conservative assumptions in the design basis accident analysis. The proposed change will not cause the consequences of a SGTR to exceed those assumptions. The proposed change to reporting requirements and clarifications of the existing requirements have no affect on the probability or consequences of SGTR.

Therefore, it is concluded that this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed changes to the SG Program will not introduce any adverse changes to the plant design basis or postulated accidents resulting from potential tube degradation. The proposed change does not affect the design of the SGs or their method of operation. In addition, the proposed change does not impact any other plant system or component.

Therefore, it is concluded that this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Document Control Desk CR-12-03567 RC-13-0047 Page 4 of 4

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The SG tubes in pressurized water reactors are an integral part of the reactor coolant pressure boundary and, as such, are relied upon to maintain the primary system's pressure and inventory. As part of the reactor coolant pressure boundary, the SG tubes are unique in that they are also relied upon as a heat transfer surface between the primary and secondary systems such that residual heat can be removed from the primary system. In addition, the SG tubes also isolate the radioactive fission products in the primary coolant from the secondary system. In summary, the safety function of a SG is maintained by ensuring the integrity of its tubes.

Steam generator tube integrity is a function of the design, environment, and the physical condition or the tube. The proposed change does not affect tube design of operating environment. The proposed change will continue to require monitoring of the physical condition of the SG tubes such that there will not be a reduction in the margin of safety compared to the current requirements.

Therefore, it is concluded that the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, SCE&G concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

4.0 ENVIRONMENTAL EVALUATION The proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement.

However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.

Document Control Desk CR-12-03567 RC-1 3-0047 Page 1 of 15 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) Unit 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 ATTACHMENT 2 PROPOSED TECHNICAL SPECIFICATION CHANGES (MARK-UP)

SCE&G - EXPLANATION OF CHANGES Page Affected Description of Change Reason for Change Section.

3/4 4-11 3/4.4.5 Replace repair with plugging TSTF-510 Revision 2 6-12d 6.8.4.k Replace repair with plugging TSTF-510 Revision 2 6-12e 6.8.4.k Replace repair with plugging and TSTF-510 Revision 2 clarification 6-12f 6.8.4.k Updated inspection frequency TSTF-510 Revision 2 6-12g 6.8.4.k Repagination TSTF-510 Revision 2 6-12h 6.8.4.k Repagination TSTF-510 Revision 2 6-16b 6.9.1.12 Reporting requirements change TSTF-510 Revision 2 B 3/4 4-3a Bases Replace repair with plugging TSTF-510 Revision 2 B 3/4 4-3c Bases Replace repair with plugging TSTF-510 Revision 2 B 3/4 4-3d Bases Replace repair with plugging TSTF-510 Revision 2 B 3/4 4-3e Bases Replace repair with plugging TSTF-510 Revision 2

Document Control Desk CR-12-03567 RC-1 3-0047 Page 2 of 15 314.4.5 STEAM GENERATOR TUBOE INTEGRITY UMING CONDITION FOR OPERATION 3.4.5 Steam generator tube Integ^t shall be maintained.

AND 10 AN steam generator k*A*sab Vg Owal the tuu . hctia be lugoged in accordance with toe Steam Generator Program.

ABJQWlJ: MOME 1,2.3 and 4.

The ACTIONS may be entered separately for each ste~ generatorUw

a. W~i one or more steamn generator tUbes satisfyingth tUbe apeb criteria and not .0 Plugged Inaccordance with the Sream Generator Program,

1. W"h 7 days veiFy tube k of the affected tube) ismantained n legftt nebt refunerln tage or steam generator tube pection, or be In HOT STANDBY vAhn th ned in 6 howe and COLD SHUTDOWN wwUWn the folloing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, and

2. PVig the affected tube(s) Inaccordance w4h the Steam Ge*nrt Program prior to entering HOT SHUTDOWN followingtuewnet refueling oLh tge or steam ge nerao i"ub pecton. 10 0

b. WiNh steam generator ube Integft not maintained, be In HOT STANDBY 6 WIN hos and In COLD SHUTDOWN Win fth next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQIUIREMAENTS 4.4.5.1 Ver~fy steam generator tube kntegrity Inaccordance wit the Steam Generator pugn Program.; lugn 4.4.5.2 Verif tOa each Inspected stam generator tub that saftisis the tbew pairrilteria is plugged in accordance with the Steam Generator Program prio to entering NOT SHUTDOWN following a steam generator tube Inspection.

SUMMER - UNIT"I 31444-11 Amendment No.-tP

Document Control Desk CR-12-03567 RC-13-0047 Page 3 of 15 AMIrSTRATWE CONTROLS

1. TO This proarm provides; mrun for processing dwanges to ie Basm of tiese Tecvrkai Spedccad"n.

1. Changes to e Bae shall be made under apprqo admnitrailve Control and re~vies

2. Lcenree masy makt chnos to Beam vwthout prior NRO approal provided th cwges do not rmeure ete of ft flowing:

a) A change intie T1 Incorporated In the license or b) Achne to the pdated SAR or basn that require. NRC approval pursuant to 10 CFR 60.50.

3. The BSea Contol Programn saH contain provisions to insureW that ie Bas ae maintained consistent Wit the FSAR.

4. Proposed canges th meet ftcter e of Speofcation 6..4.Lb above sta be revitwed and approved pio to hniplemWnlan Chan"e to ie Bases kismented without prior NRC approval shall be provided to the NRC on e frequecy consistent with 10 CFA 50.71(e).

a"R~Ift Cociant Pumo FM* WgOMct Program T program sl prWode for O inspection of eoc reactor coolant pump flyWe per "e -31i. of Regulatory Posio CA.b of Regulaory Guide 1.14, Revision 1, August 1M7.

In leu of Positions CA.b(1) and CA.b(2), a qualed i-pla UT examination over the volume from the Inner bore of the flywheel to Oh circle one-hal of the outer radius or a surface examinaon (MT andlor PT) of exposed surfaces of the removed flywheels may be oonducted at 20 year intervals.

k. Sturn G BmgMM 0 A Steam Generator Program shall be established and Implemented to ensure tat steam generator (SG) tu integrity Ismailatned. In addition, the Steam Generator Program shal Include the Wooing p.lwlu.

1. Provisions for condition monitoring assessments. condition monitoring assssent means an evakation of the as found" condition ol the luting with respect to the perormanc cdteri for structr'l Integrityand ccitdent induced lakage. The "as OXNd condition refers to fte condition of Ie tubing during a SG Inspection oulag, as determined from toe kwvioe Inspection results or by otte means, prior to Oe plugging of tubes. CoWdlion monitoring assessments s"e be conducted during each outag during wi t SG tusm ae Inspected or plugged to confirm tha the performance criteria are being met.

SUMR- UNIT I 6-12d . Amendment No. in" * .-

Document Control Desk CR-12-03567 RC-13-0047 Page 4 of 15 ADMNISTRATIVE CONTROL

2. Perfornmanecriteri fr SG beMo intleity. Steam gWWor be int*egriy OWui be m~raintid by meeting the perfounwice cri~tea for tUbe atoUra)8 kbtekY, 40 accident Induced Wekage and opeao eaMag.

a) Strucium ntegty peulormaie critedon. All inS SG tubessha retain structure integrity ove Me U rae of normal operat"n condom (including start.u, Operast*in te power rag, HOT STANDY. ad addesign basis aciet.This includes retaining a sawt atr o13.0 (3dslt;&P) agairst burst undler normal steady sota ful povw operation

-. e-ý primary4o-aeondary pressure dWfftl &id a sdfely factor d 1.4 agains burst apple to the deig basi accdent plsyt-ecdx pressure dffewrti. Apl t frorn the above requirements, addtonal Cc oain condions assocted wit th desip basi aockiet, or a-ooml~lnitin of accidents in accordance whth the desig and loensin basis shall also be evaluated to determnine 0 the associated loads contribute slgncatly to burst or colapse. In ,th assessment of tibe 0 interity, tthe loads that do sIgnfcantly aff burst or collapse " be CO Cl determined and assessed In combination with fte boads due to pressure wOt a safely hactof 0112 on toe combined priwy load and 1.0 on matal seodr losý o:~ b) Accident Induced leakage performance cteion. The primary-to-

.4 0~CI,.

0c- secondary acddentduced leakage rate for any deirn basis acckent, U other than a SG lube rupture, shal not exed ft lkage at assumod o (U in the accident analysis interms of total leakage rats foral SOs and le g rate for an hdlivdi,,ual SG. Acckient Inducld leakaoo is not to excee 1 Opm per So.

c) The operational leakaep performance criterion Isspecifid InLOO 3&4.6.2

• Reactor Coolant System Operational Lsaalksp- -Eng

3. ProvIsin for SO tube =- %fts md by inservice inaetion to cmoin fk.as with a depth equal to or exceeding 40% of the nominul ubew wl tiknB esMsbe p*ugged.

4. Provislons f SO ube ks tn. Per c SQ tutubeinspection s bhael perdfoed. The number and portions of the tubes inspected and me*-ods of plugging lrpectionshalbe perlored with*t* oblec of detetg flaws of any "

W axi mawtretia aw) tatmay be preen volmotItes.

Vw lan Otho ibe, from weld at the tube inet to t weld at the tube outlet, and that may aisl the applicable m critera. The tubaeto-tubeshetwd Is not part of the tube. In addition to rmee th requirements of 4.a, 4.b, and 4.0 below, the ipection scope, Inspection metxos and inspecion Intervals shil" be soch as to ensure ta SO_-*-_-_

tube kntgy':*--* is maimtined hohl be unti the nexttoS"eremin peftormd An-.and inpecti.eftyWe

" A degra~dation assessment II II SUMMER - UINIT I 6-120 SUMMER UNIT6-12eAmendmenWN..VlJ

Document Control Desk CR-12-03567 RC-13-0047 Page 5 of 15 ADMINSTRATIVE COWTROLi klWuon *Ro o which he tube miy be suscepifle wd, boWd on ths ssssmmnt So dfnlem* whIchI lnmisd methods need to he e d aW at We locatns.

0 a) Inspec 100% of t tubw In each 90 duark the first l age insta llaio C

CU

-o -o flowing installation U) .2 4- 0 0 b)

U) W~rhr 6 p~wmuiti.

e~~efii hefirt z.eiii al du CU CL n~i he ii...-e I, bon u er hn7 h.elsrit Inser-D c; dicallion wee I"W Inany SG ube, ahe fthnext lwpecton for she not emeed 24 afsec* fuN power monhs ortis refusing outge p b ungnosic con-dekt*adclve c, eik, hor t ko eIeerin e hakloW n VWdic #e a aec le icaton Isnot asmsatd wt a ca $s).VM ft indicamonled not be brtedeasa crack.

L. 5~. Provisions for monhoring operational plewy-t~oxxeendy lmieakg.

._0

.4 t AproWm shell be estaishd to IrpMesunt the %foloVlgrpeq d teeing of Engineered Safety Feature (FSF)Ikter vesrao n sstems at Ohe frequencies specifle In acordenem wih Reg my GuLde 1.52, Revisio 2. and ASME NSIO-

._ 1069.

UC

> U) 1. Dsmonm"Wf for each of Oi ESF ysy n hat ankipset Wtoftheihgh Offiiecy pUt*cue I* (HIEPA) ltMeS shows a p VerNon md systeM bypas !

' 0.06% when tistd In ccordance with Regui Guide 1.52. Revision 2, and

_ ASME N510-194N a the ftsysm flowete specifid below 1 10%.

ESF VenWation Sysiem Floukn Con*o Room Reeclor SBdkn

-nweW F"Vion System oing Units 21,270 $CFM 60,270 ACFM SUMR - UNrT I 6-l2( Anmedment No. 409,

Document Control Desk CR-12-03567 RC-1 3-0047 Page 6 of 15 Insert - A b) After the first refueling outage following SG installation, inspect each SG at least every 72 effective full power months or at least every third refueling outage (whichever results in more frequent inspections). In addition, the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in 1), 2), 3), and 4) below. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube plugging criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period.

Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage.

1) After the first refueling outage following SG installation, inspect 100% of the tubes during the next 144 effective full power months.

This constitutes the first inspection period;

2) During the next 120 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period;

3) During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the third inspection period; and

4) During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the fourth and subsequent inspection periods.

Document Control Desk CR- 2-03567 RC-1 3-0047 Page 7 of 15 For Repagination Purposes Only AUNISTRATIVE CONTROLS

2. Demonstial for each of te ESF system s an Ww

• est of wedwsel adsorber shows a penetatio end system bpass 4 0.06% when 1t1,d in accmelsnc wh Reguoy Gte 1.52, Revision 2, and ASME N510-96 at the system llowra specified below *10%.

ESF Vaetfion Syslem Fl-rat.

conrol Room Emerprncy Fitratlon System 21.270 SCFM

3. Demonstrate beach of fthe .syst*ems* O a sboralory les of a sample of st the chaca adsow , wten obWke as desc*bed In PReuaoy Guide 1.52, Revin 2, shows the methyl Wdid penes tione th # emthuene specfied below when tesled in accordance wih ASTM D3803-1909 st te mleiture of 30C (WF)and Ohe relatIve humdty specIfied below.

ESF Veniltalion Syslern Pentation RH Face Velodty (fps)

Contol Room 2.2S% 70% 0,667

4. DOsmonaere for e*ch of le ESF systems ta the pnesue drop asmon the conmbined HEPA SUM, 0 pefller, and liehart d oa dmobers Islees l*in lie value spedled below when tested In accordance with Reguiilouy Guide 1.52. RevIsion 2, en ASME NS-I 0 at lw sysem lowrae specHed below

• 10%.

ESF Vonilatio System Delft P Flowae CoWlll Rom < In. W.G. 21,270 SCFM Reactor B&ldn Coolig Units <3 n W.G. 60,2o ACFM The provslons of SR 4.0.2 and SR 4.0.3 are applice to Oi VFTP test frequencies.

m. Cvnt Pm En-A Control Room Envope (cRE) Habitbity Prmgrm sha be eslablishmend Implemented la ensur tha CRE habitability Ismantinulae such that, wMt an OPERABLE Conro Room Ermrgency Fillration System (CREFS), CRE oeoanft can conrol 11e reacto sael* undew norma contins and maintain I In a ska condiltion following a radiological event, hazadouschemical release ore smoke challenge. The pwrogram siw ensure that adequate radiation protection IsproVIded to

-wr acess and occupancy of the CRE under design basis accident (DON) condilions wihout pesonnel receiving radiation eaxsures in excess od 5 mm whole bodyoreet loanypwtoft Nobody for e duation of te accdent The program shall kwlue the follow"n elements:

SUMMER - UNIT I 6-12g SAmendment No. *-

Document Control Desk CR-12-03567 RC-1 3-0047 Page 8 of 15 P For Repagination Purposes Only ADMINSTRATIVE CONTROLS

1. The definition of the CRE and the CRE boundary.

2. Requirements for maItralning the CRE boundary In Its design condition kxcluing configuration control and preventive maintenance.

3. Requirements for (I)determining the unfiltered air Inleakage past the CRE boundary Into the CRE inaocordance with the tei methods and at the

• Frequencies specifned Sections CA and C2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reatos,'

Revision 0, May 2003. and (0)assessing CRE habilabilly at the Frequencies specified In Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.

4. Measurement, at designated locations, of the CRE pressure relative to all external arm adjacent to te CRE boundary during the presrIzation mode of operation by one train of te CREFS, operating at the flow rat required by th Ventalion Filter Testing Program (VIFTP). at a Frequency of 36 months on a STAGGERED TEST BASIS such that one train is tested every 18 months. The results sihal be trended and used as part of the 1 month assessment of the ,

GRE boundary. 0

5. The quantitative kmits on unfiltered air Inlealtage ,Int the CRE. Thee lnits shall be stated In a manner to allow direct comparison to te unfterd air Inleakage measured by the testing described in paragraph 6.8.4.m.3. The unlftered ar inleakage "il for radiological challenges Is the inleakage flow rate assumed in the kmesing basis analyses of DBA consequences. Uniltored air inleakage Oirn for hazardous chemicals must ensure lhst exposure of CRE occupants to hens hazards wi* be within fte assumplios in the licensing basis.

6. The provtslons of SR 4.0.2 are applicable to the Frequencies for assessing CRE habltablity, determining CRE unftered inleakage, and measuring CRE pressure' and asseing the CRE boundary as required by paragraphs 6.8.4.m.3 and 6.8.4,m.4, respectively.

SUMMER - U~iT 1 6-12h Amendment No. "&

Document Control Desk CR-12-03567 RC-13-0047 Page 9 of 15 ADMINISTRATIVE CONTROLS MF...ENATOR I=B INPETAIN BEPORT 69.1.12 A reipot shadbe subreitted* wi" 180 days after the kinwtalent klo MODE 4 foloiong competion d an inspection perlormed in accordanoe YAM Specification 6.SAJ The rpoW shal incfde:

a. The soeof inspections performed on each SG,

b. ftU badation mec*w*ms found, 00, 4.-

c. Nondesructive examlnation teU niques utzed for each degradon mechanism,

d. Location. orientation (i Neer), and measured sizes (if availle) of serv*ce kxdued

0. Number of tubes pLugged during twe ispeclon outaep for each a-&,.dg0radetion mechanism, g.1' 7Te resuts Of condItion rnoMIGtoun, Inctdin fte resuts of tue pub aNd w-stu too".n The number and percentage of tubes plugged to date, and the effective plugging percentage in each steam generator, and 40 SUJMMM*l~ - UNIT 1 6-16b) SAndment No. "g.

Document Control Desk CR-12-03567 RC-1 3-0047 Page 10 of 15 I For Information Only REACTOR C SYSTEM No Changes Required BASES 3/4.45 STEAM GENERATOR TUBE INTEGRITY S*ean generar (SG) tubes are small diameter, thn waled tubes that carry to ý74-seor hea exchaqe The SG ubes

• ugh~m f*r a =coav ofth.,brCo UIC,&IMM tubes are an ikt l,part of th*e reactor coolant pressum boundary (RCPB) and, as suck are reW on to maintain toe primary system's pressure and inventory. The SG tubes isatle to radoacive fission products in the primary coolant from the secondary system. Inaddition. as part of the RCPB, the SG tubes ae unique in that they act as the heat Wansfer surface between the prmary and secondary systems to remove heat from th primay 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. IReactor Coolant St Reactor Coolant Loops and Coolant Circulation, Start and Power ieradon LCO 3A.12.

Reador Coolant System, Hot Standby,- LCO 34.1.3. Reactor Coolant System. Hot Shutdown: and LCO 3.4.1.4.1, Reactor Coolant System, Cold Shutdown-Loops Filled.

SG tube iti means that the tubes are capable of performing their intended RCPB safety fiction consistent with the icensmg bas including applicable regulatory reciements.

SG tubing is subjec to a variety of degradaton mechanisms. SG tubes may expenence tube degada related tocorrosion phenomena, such as wastage, piting, i tergranularattack arid stress corrosion craln gWong with othered phenomena such as denting and war. These degradation nudianisms can impair degradatin.

used to maage SG tube Specification 6.8.k, 'Steam Generator Program, requires that a program be established diplemented to ensure that SG tube integrity is mairied. Pursuant to Specification 6.8.4.k, tube integrity is mantaied when the SG performance criteria are met There are tlhree SG performance criteria: srucWual integrity, accident induced leakage, and operational leakage. The SG performance criteria are described in Speciication 6.8.4A. Meeting the SG performance criteria provides reasonable assurance of maintaining tube integrity at normal and accident condcions.

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

SUMNLER - UNIT I B 3/4 4-3 Amendment No. 3664, 69, W6MS, 166, RN07-001

Document Control Desk CR-12-03567 RC-13-0047 Page 11 of 15 REACTOR COOLANT SYSTEM BASES STEAM GENERATOR TUBE INTEGRITY (Conltied)

The steam generator tube rump (SGTR) accident is the dsgn basis evert for SG tubes and avoiding a SGIR is the basis for this SpecificaO. The accident analysis for a SGTR event accounts for a boundin primatyto-secondary leakage rate equal to I gpm and the leakage rate associated with a double-ended niture of a single tube. Contanmited flud ina nupured steam generaior is only briefly released to the atmosphere as steam via the man steam safety valves. To maxinie its contuibution to the dose releases, the entire 1 gpm pnma (>-tsecondauy leakage is assumed to occur in the intact steam generators where it can be released during the subsequent cooldown of the plant The analyses for design basis accidents and transients other than a SGTR assume the SG tubes retain their structural integrity (ie., they are assumed not to rupture). In these analyses the steam discharge to the atmosphere is based on the total pnri to-secondary leakage from all SGs of I gpm, or is assumed to increase to 1gpm as a result of accident induced condition For accidents that do rot involve fuel damage, te primary coolant activiy level of DOSE EQUIVALENT 1-131 is assumed to be Treler than or equal to the knits in LCO 3.4.8, *ReactorCoolant System, Specific Activity. For accidents that assume fel damage, the primary coolant actvity is a function of the amount of activity released from the damaged fuel. The dose consequences of these events am within the knits of GDC 19 (Reference 2), 10 CFR 50.67 (Reference 3) or the NRC approved lcensing basis (e.g.. a small fraction of these knits).

Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(cX2)().

lbe LCO requires that SG tube in ybe maintained. The LCO also requir that all SG tubes that satisfy therýcitnri" - a -- withtheSleam CI~a~ Program- lggn During a SG inspection. any . 0350cW W fies t Steam Generator criteria is removed from service by pkiugging. Ifa tube was deternmed to criteria b was not pluged. the tube may still have tube integrity. Refer ngIntocotdofth Specificato a SG tube is defined as the entire length of the tb~e, induding the tube wall between thewedathtueiltndg tube-

• weld at the tube outlet The tueo e4weld is not considevi part ofthe tube.

A SG tube has tube integrity when i satisfies the SG pformanc criteria. The SG performance criteria are defuned in Specification 6.8A.k and describe acceptable SG tube performance. The Stam Generator Program also provides the evaluation process for deternining confomance with the SG pfmance criteria.

SUMMER - UNIT 1 B 3/4 4-3a Amendment No. BRN 07 091,

Document Control Desk CR-1 2-03567 RC-13-0047 Page 12 of 15 REACTOR COLANT For Information Only No Changes Required BASES STEAM GENERATOR TUBE INTEGRITY XContinuecl)

There are three SG peftman atea. structural integrty, accident induced leakage, and operational leakage. Failure to meet any one of these arieria is considered abihe to meet the LCO.

The structural irtegry pertrmance criterion provides a margin of safety against tube burst or collapse under normal and accident comdIlions, and ensures structurl interity of the SG tubes under al anticpated transiernt included inthe design specificalin. Tube burst is defined as, The gross structural lfure of the tube wall. The condition typically coresponds to an unstable opernng displacement (e.g. opening area increased inresponse to constant pressure) accomarved by ductile (plastic) teaing of lie ube nmatal at the ends of I*e degradation. Tube collapse is defined as, "For the load displacement curve for a given structure, collapse occus at the top of the load verses displacement curve where the slope of the curve becomes zero. The structural integiy peftmance critei provides gui e on assessing loads t have a signlicant e~et on burst or collapse. Inthat conext, the term 'significant is deltied as

'An accident loading condition olter than differential pressure Is considered signlicant when the addition of such loads inthe assessrneet of the structural integrity perlormanm criterion could cause a lower sumckial limit or Witling burstcollapse condition to be established. For tubeiutegrity evaluations, except for cacumerential degradation, axial thermal loads are classified as secondary loads. For acrcunerential degradation, the classification of axial thermal loads as primr or secondary loads will be evaluated on a case-by-case basis. The division betwen primary and secondary classificatio will be based on detailed analysis andfor testing.

Sbuctural ntegrity requires that the rimnary mentmae ss 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) transmits included Inthe design specilication- This Includes safetyfctors arnd applicable design basis loads based on ASME Code.Section III, Subsection NB (Reference 4) and Dralt Regulatory Guide 1.121 (Refenence 5).

The accident Induced leakage perlomiance criterion ensures hat the pdray-secondary leakage caused by a design basis accident, ofher than a SGTR is within the accident analysis assuwt .ions The accident analy assumes that accident induced lakage does not exceed I gpm total from all SGs. The accident induced leakage rate includes any priniaryto-secondlary leakage exising pwor to the accident in addition to primato-sed ary leakage induced during the accident.

The opeAioal leakage performance criterion provides an observab indication of SG tube conditions during plant operation. The limit on operational leakage is contained inLCO 3A.62 and limits pmaryo-se leakage through any one SG to 150 gallons per day. This lINit is based on the assumption that a single crack leaking this amount would not propagate to a SGTR under the stress condlions of a LOCA or a main stem line break. If this aroniunt of leakage is due to more than one crack, the crack are very small, and the above assumtlion is conservatve.

SUMR - UNIT I B 34 431 Amendment No. BRN-07-001

Document Control Desk CR-12-03567 RC-1 3-0047 Page 13 of 15 REACTOR COL* SYSTEMA BASES STEAM GEERTOR TUBE OffTEG;RIT Cn" u Steam generator tube tegrty ischallenged when the pressure derena aaross the tubes Is large. Large differential pressures acrsstSo tubes can orny be exedre inMODE 1,2,3, or4.

RCS conditions are far kess challengno InMODES 5 and 6 than durng MODES 1, 2, 3.and 4. In Modes 5 and 6, Mffeential pmssurm is low, resuting inlower stresses and reduced potental for learone.

Actons The ACTIONS am modifed by a Note clarifying that the Cao ons may be entered in*dendenw for each SG tube. This is acceptable because the required ACTIONS provide appropriate compensatory actions for each affected SG tube.

Compi with the req*red ACTIONS may alow for confirmed operation,and sbseqent acted SG tubes are governed by subsequent Condiion entry and application of associated required ACTIONS.

aL The Concition applies if it is disovered that one or more SG tubes examined in an Inservce Inspecton bIutv~

weeot PR in " -InGener~ Programn as rMikW by 4.4.52. An evaluation of SG tube integty of the kegrity is based on plugging I=tu-e(s) must be made.

SG perforrmance Steam critria generator describd Intube th Sla Ge orw Program. . ateria delne limits on SG tube degradation that allow for flaw growth between inspections wtile still providing assurance that the SG performance crita will continue to be meL In order to delemine If a S(3 tube that should have been plugged has tube integrity, an evaluation utbe completed that dermosates at the 03 performan critria will contiue bo be met until the next rlofueing outage or SG tube inspection.

The tube Irdeglty debrminaton is based on the estimated condition of the tube at the time the situation isdscMe and the estimated growth of the degradation prior to the next SG tube inspection. If it is diemined tha tube integrity is not being maintained, LCO 3.4.5 Action b. applies.

A completion time of seven days is sufficient to complete the evaluatlon while mininizing the risk of plant operation with a SG tube that may not have tube integrity.

Ifthe evaluation determes that the affected tube(s) have tube inlegrily, the ACTION statement allows plait opeation to continue until the nxt refueling outage or S( inspection provided the Inspection interval continues to be supported by an operational assessment that reflects the atlcted tubes Howeve, the aftcted tube(s) must be plugged prior to entering MODE 4 olow*ig the next reftyeng outage or SG inspection. This completion time is acceptable since operation until the next Inspection is supported by the operational assessment.

SUMER- UNIT I 8 Y/4 4-3c: UAmendment No. IRMOBTO1

Document Control Desk CR-12-03567 RC-1 3-0047 Page 14 of 15 REACTOR COOLANT SYSTEM BASES STE=.AM GENERATOR TUBE INTEGRITY fodze ACTIONS lcrkwd

b. 9f the required acions and associated coopetion times of LCO 3.4.5 Action

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 te next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> .

The allowed conmletion times are reasonable, based on operating expenence, to reach the desired plant conditions from u power conditions in an orderly manwn and without challenging plant systems.

Surveilanoe Recuremenits (SR 4.4.5.1 During shutdown periods the SGs are inspected as required by tit SR and the Steam Generator Program. NEI 97-06, Steam Generator Program Guidelines (Reference 1), and iis relerenced EPRI Guidelines, establish the corntt d the Steam Generator Progranm Use of the Steam Generator Program enures that the irspection is appropriate and consistent with accepted industry practices. 00 A condition oxnton assessment of the SG btbes Is pertwmed duing SG inspecions. The condition monitorg assessment detemines the as fowd Condition of the SG tubes. The puqxse of the condition monoring assessment Is to enue that the SG pertinmance criteria have been met for the previous operating period.

The Steam Generator Program determnes the sCOpe of te inspection and the plugging ushdted to delenaie whether the ft contain sting the te r glaws criteriL Inspection scope (Le., wich tubes or areas of tubing within the SO are to be inspected) isa function f exising and potential degradation locations. The Steam Generator Program also specifies the inspecion Mhods to be used to end potential degradation. Inspection methods are a Aiaion of degradation morphology.

nondestructive examnation (NOE) techunique caablities, and inspection locations.

The Steam Generaor Program delines the frequency of SR 4.4.5.1. The frequency is determined by the operational assessment and other liits in the SG examination guidelines (Reforence 6). The Steam Generator Program uses infrmation on existi degradations and growth rates to determine an inspection frequency that provides reasonable assurance that theubing will meet the SG perormance criteria at the next scheduJed inspection. In addition, Specifcation 6.8.4-k contains prescriptive reqrements concerning irpection Intervals to provide added assurance that the SG perlfrmance cntena will be met between scheduled inspectins.

If crack indications are found in any SG tube, the maximum inspection interval for all affected and potentially affected SGs is restricted by Specification 6.8.4.k until subsequent inspections support extending the inspection interval.

SUMER- UNIT 1 0 W 4-3d Affwxkrwd No. 6FV+ff-M4

Document Control Desk CR- 2-03567 RC-1 3-0047 Page 15 of 15 REACTOR COOLANT SYSTEM BASES STEAM GENERATOR TUBE INTEGRITY (c ,ontned SurQellance RwdurmenIs Yu)

Durn a SG i nspecton any nspted tube th satisfies t Steam criteria is refovd from service by pugng*. The tube critea delineated inSpecification 6.8A.k ae Intended to ensure Ott tubes accepted for continued service satisfy tfe SG prormance criteria wth alowace for error in the flaw size mesrement and for fitur flaw grwth In addition the tube crtia, in c a-t ton with otr efremnts of the Steam Genwat rograna ensure t1a the SG performance critera wl continue to be met until the next ipbtion of the subect tube(s). Reference I povides guVdance for pero opeational assessments to verify that the tubes remainng in service 01 continue to meet the SG perormance criteria.

The ftecuency of "Prior to enterig MODE 4 foloinga a SO plugging that the Survelilance has been completed and al tubes meeting the#,*" =e -

plumed piwr to subjectg the SG tubes to swinlcat pimay-to-scndary pressure driferential.

1. NEI 97-06, 'Stean Generator Program Guideines' 2 10 CFR 50, Appendix A,GDC 19, 'Control Room-

3. 10 CFR 50.67, "Accdent Source Term"

4. ASME Boiler and Pressure Vessel Code, Sectio III, Subsection NB

5. RegulatoW Guide 1.121, -Basis for Plugging Degraded Steam Generator Tubes,- August 1976

6. EPRI TR-107569, 'Pressurized Water Reactor Steam Generaor Exaination Guideines SUMMR- UNIT I B 3(4 4-3e Amendment No. imN-Q7OM.

ORNONO

Document Control Desk CR-12-03567 RC-13-0047 Page 1 of 15 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) Unit I DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 ATTACHMENT 3 REVISED TECHNICAL SPECIFICATION PAGES Proposed Technical Specification Changes Summary Replace the following pages of the Appendix A to Operating License Number NPF-12, Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages Insert Pages 3/44-11 3/44-11 6-12d 6-12d 6-12e 6-12e 6-12f 6-12f 6-12g 6-12g 6-12h 6-12h 6-16b 6-16b B 3/4 4-3a B 3/4 4-3a B 3/4 4-3c B 3/4 4-3c B 3/4 4-3d B 3/4 4-3d B 3/4 4-3e B 3/4 4-3e

REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATOR TUBE INTEGRITY LIMITING CONDITION FOR OPERATION 3.4.5 Steam generator tube integrity shall be maintained.

AND All steam generator tubes satisfying the tube plugging criteria shall be plugged in accordance with the Steam Generator Program.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

The ACTIONS may be entered separately for each steam generator tube.

a. With one or more steam generator tubes satisfying the tube plugging criteria and not plugged in accordance with the Steam Generator Program,

1. Within 7 days verify tube integrity of the affected tube(s) is maintained until the next refueling outage or steam generator tube inspection, or be in HOT STANDBY within the next 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 />, and

2. Plug the affected tube(s) in accordance with the Steam Generator Program prior to entering HOT SHUTDOWN following the next refueling outage or steam generator tube inspection.

b. With steam generator tube integrity not maintained, 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 />.

SURVEILLANCE REQUIREMENTS 4.4.5.1 Verify steam generator tube integrity in accordance with the Steam Generator Program.

4.4.5.2 Verify that each inspected steam generator tube that satisfies the tube plugging criteria is plugged in accordance with the Steam Generator Program prior to entering HOT SHUTDOWN following a steam generator tube inspection.

SUMMER - UNIT 1 3/4 4-11 Amendment No. 47 ADMINISTRATIVE CONTROLS

i. Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

1. Changes to the Bases shall be made under appropriate administrative control and reviews.

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

a) A change in the TS incorporated in the license or b) A change to the updated FSAR or bases that requires NRC approval pursuant to 10 CFR 50.59.

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

4. Proposed changes that meet the criteria of Specification 6.8.4.i.2.b above shall be reviewed and approved 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).

j. Reactor Coolant Pump Flywheel Inspection Program This program shall provide for the inspection of each reactor coolant pump flywheel per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1, August 1975.

In lieu of Positions C.4.b(1) and C.4.b(2), a qualified in-place UT examination over the volume from the inner bore of the flywheel to the circle one-half of the outer radius or a surface examination (MT and/or PT) of exposed surfaces of the removed flywheels may be conducted at 20 year intervals.

k. Steam Generator Program A Steam Generator Program shall be established and implemented to ensure that steam generator (SG) tube integrity is maintained. In addition, the Steam Generator Program shall include the following:

1. 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 a 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.

SUMMER - UNIT 1 6-12d Amendment No. 47-47,

ADMINISTRATIVE CONTROLS

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

a) Structural integrity performance criterion. All inservice SG tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, HOT STANDBY, and cooldown), all anticipated transients included in the design specification, and design basis accidents. This includes retaining a safety factor of 3.0 (3deltaP) 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.

b) 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. Accident induced leakage is not to exceed 1 gpm per SG.

c) The operational leakage performance criterion is specified in LCO 3.4.6.2, "Reactor Coolant System Operational Leakage."

3. Provisions for SG tube plugging 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.

4. 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 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 plugging criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of 4.a, 4.b, and 4.c 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. A degradation assessment 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.

SUMMER - UNIT 1 6-12e Amendment No. 4-97-,

ADMINISTRATIVE CONTROLS a) Inspect 100% of the tubes in each SG during the first refueling outage following SG installation.

b) After the first refueling outage following SG installation, inspect each SG at least every 72 effective full power months or at least every third refueling outage (whichever results in more frequent inspections). In addition, the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in 1), 2), 3), and 4) below. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube plugging criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period. Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage.

1) After the first refueling outage following SG installation, inspect 100%

of the tubes during the next 144 effective full power months. This constitutes the first inspection period;

2) During the next 120 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period;

3) During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the third inspection period; and

4) During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the fourth and subsequent inspection periods.

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

5. Provisions for monitoring operational primary-to-secondary leakage.

SUMMER - UNIT 1 6-12f Amendment No. 483-

ADMINISTRATIVE CONTROLS Ventilation Filter Testing Program (VFTP)

A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies specified in accordance with Regulatory Guide 1.52, Revision 2, and ASME N51 0-1989.

1. Demonstrate for each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass

< 0.05% when tested in accordance with Regulatory Guide 1.52, Revision 2, and ASME N510-1989 at the system flowrate specified below +/- 10%.

ESF Ventilation System Flowrate Control Room Emergency Filtration System 21,270 SCFM Reactor Building Cooling Units 60,270 ACFM

2. Demonstrate for each of the ESF systems that an inplace test of the charcoal adsorber shows a penetration and system bypass < 0.05% when tested in accordance with Regulatory Guide 1.52, Revision 2, and ASME N510-1989 at the system flowrate specified below +/- 10%.

ESF Ventilation System Flowrate Control Room Emergency Filtration System 21,270 SCFM

3. Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of 30 0 C (86 0F) and the relative humidity specified below.

ESF Ventilation System Penetration RH Face Velocity (fps)

Control Room <2.5% 70% 0.667

4. Demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters, the prefilters, and the charcoal adsorbers is less than the value specified below when tested in accordance with Regulatory Guide 1.52, Revision 2, and ASME N510-1989 at the system flowrate specified below

+ 10%.

ESF Ventilation System Delta P Flowrate Control Room <6 in. W.G. 21,270 SCFM Reactor Building Cooling Units <3 in. W.G. 60,270 ACFM The provisions of SR 4.0.2 and SR 4.0.3 are applicable to the VFTP test frequencies.

SUMMER - UNIT 1 6-12g Amendment No. 483

ADMINISTRATIVE CONTROLS

m. Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Filtration System (CREFS), CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem whole body or its equivalent to any part of the body for the duration of the accident. The program shall include the following elements:

1. The definition of the CRE and the CRE boundary.

2. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.

3. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.A and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.A and C.2 of Regulatory Guide 1.197, Revision 0.

4. Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CREFS, operating at the flow rate required by the Ventilation Filter Testing Program (VFTP), at a Frequency of 36 months on a STAGGERED TEST BASIS such that one train is tested every 18 months. The results shall be trended and used as part of the 18 month assessment of the CRE boundary.

5. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph 6.8.4.m.3. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences.

Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

6. The provisions of SR 4.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs 6.8.4.m.3 and 6.8.4.m.4, respectively.

SUMMER - UNIT 1 6-12h Amendment No. 4 6 0T

ADMINISTRATIVE CONTROLS STEAM GENERATOR TUBE INSPECTION REPORT 6.9.1.12 A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 6.8.4.k. The report shall include:

a. The scope of inspections performed on each SG,

b. 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 degradation mechanism,

f. The number and percentage of tubes plugged to date, and the effective plugging percentage in each steam generator, and

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

SUMMER - UNIT 1 6-16b Amendment No. 4797

REACTOR COOLANT SYSTEM BASES STEAM GENERATOR TUBE INTEGRITY (Continued)

Applicable Safety Analyses The steam generator tube rupture (SGTR) accident is the limiting design basis event for SG tubes and avoiding a SGTR is the basis for this Specification. The accident analysis for a SGTR event accounts for a bounding primary-to-secondary leakage rate equal to 1 gpm and the leakage rate associated with a double-ended rupture of a single tube. Contaminated fluid in a ruptured steam generator is only briefly released to the atmosphere as steam via the main steam safety valves. To maximize its contribution to the dose releases, the entire 1 gpm primary-to-secondary leakage is assumed to occur in the intact steam generators where it can be released during the subsequent cooldown of the plant.

The analyses 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 I gpm, or is assumed to increase to 1 gpm 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 greater than or equal to the limits in LCO 3.4.8, "Reactor Coolant System, Specific Activity." 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 (Reference 2), 10 CFR 50.67 (Reference 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(c)(2)(ii).

Limitinq Condition for Operation (LCO)

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

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

Refer to Action a. below.

In the context of this Specification, a SG tube is defined as the entire length of the tube, including the tube wall between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet. The tube-to-tubesheet weld is not considered part of the tube.

A SG tube has tube integrity when it satisfies the SG performance criteria. The SG performance criteria are defined in Specification 6.8.4.k and describe acceptable SG tube performance. The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria.

SUMMER - UNIT 1 B 3/4 4-3a Amendment No. BRN 07 00!,

BRN*-41*001,

REACTOR COOLANT SYSTEM BASES STEAM GENERATOR TUBE INTEGRITY (Continued)

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.

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. The Condition applies if it is discovered that one or more SG tubes examined in an Inservice Inspection satisfy the tube plugging criteria but were not plugged in accordance with the Steam Generator Program as required by Surveillance Requirement 4.4.5.2. 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 plugging 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, LCO 3.4.5 Action b. applies.

A completion time of seven 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, the ACTION statement 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.

SUMMER - UNIT 1 B 3/4 4-3c Amendment No. ORN 07 001,

REACTOR COOLANT SYSTEM BASES STEAM GENERATOR TUBE INTEGRITY (Continued)

ACTIONS (Continued)

b. If the required actions and associated completion times of LCO 3.4.5 Action

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 the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-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 Requirements (SR) 4.4.5.1 During shutdown periods the SGs are inspected as required by this SR and the Steam Generator Program. NEI 97-06, Steam Generator Program Guidelines (Reference 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.

A condition monitoring assessment of the SG tubes is performed during SG inspections. 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 method used to determine whether the tubes contain flaws satisfying the tube plugging 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 4.4.5.1. The frequency is determined by the operational assessment and other limits in the SG examination guidelines (Reference 6). The Steam Generator Program uses information on existing degradations and growth rates to determine an inspection frequency that provides reasonable assurance that the tubing will meet the SG performance criteria at the next scheduled inspection. In addition, Specification 6.8.4.k contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections. If crack indications are found in any SG tube, the maximum inspection interval for all affected and potentially affected SGs is restricted by Specification 6.8.4.k until subsequent inspections support extending the inspection interval.

SUMMER - UNIT 1 B 3/4 4-3d Amendment No. BRN 07- 001,

REACTOR COOLANT SYSTEM BASES STEAM GENERATOR TUBE INTEGRITY (Continued)

Surveillance Requirements (Continued) 4.4.5.2 During a SG inspection, any inspected tube that satisfies the Steam Generator Program plugging criteria is removed from service by plugging. The tube plugging criteria delineated in Specification 6.8.4.k are intended to ensure that tubes accepted for continued service satisfy the SG performance criteria with allowance for error in the flaw size measurement and for future flaw growth. In addition, the tube plugging 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 "Prior to entering MODE 4 following a SG inspection" ensures that the Surveillance has been completed and all tubes meeting the plugging 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, "Control Room"

3. 10 CFR 50.67, "Accident Source Term"

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

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

6. EPRI TR-107569, "Pressurized Water Reactor Steam Generator Examination Guidelines" SUMMER - UNIT 1 B 3/4 4-3e Amendment No. 8RN 07-00, 4 1 00*,

lRN

Document Control Desk CR-12-03567 RC-13-0047 Page 1 of 1 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS)

DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 ATTACHMENT 4 LIST OF REGULATORY COMMITMENTS The following table identifies those actions committed to by Virgil C. Summer Nuclear Station (VCSNS) in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments. Please direct questions regarding these commitments to Mr. Bruce L. Thompson at (803) 931-5042.

Commitment Due Date None