Text

Proposed License Amendment Request Consolidated Line Item Improvement Process Technical Specification Improvement Regarding Steam Generator Tube Integrity
	ML061430062
Person / Time
Site:	North Anna
Issue date:	05/22/2006
From:	Grecheck E; Virginia Electric & Power Co (VEPCO)
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	06-403
	Download: ML061430062 (89)
	v • d • e

VIRGINIA ELECTRIC AND POWERCOMPANY RIC~MOND,VIRGINIA 23261 May 22, 2006 U.S. Nuclear Regulatory Commission Serial No.06-403 Attention: Document Control Desk NL&OS/ETS RO Washington, D.C. 20555 Docket Nos. 50-3381339 License Nos. NPF-417 NORTH ANNA POWER STATION UNITS 1 AND 2 PROPOSED LICENSE AMENDMENT REQUEST CONSOLIDATED LINE ITEM IMPROVEMENT PROCESS TECHNICAL SPECIFICATION IMPROVEMENT REGARDING STEAM GENERATOR TUBE INTEGRITY Pursuant to 10 CFR 50.90, Virginia Electric and Power Company (Dominion) requests amendments, in the form of changes to the Technical Specifications (TS) to Facility Operating License Numbers NPF-4 and NPF-7 for North Anna Power Station Units 1 and 2, respectively. The proposed amendment would revise the TS requirements related to steam generator tube integrity. The changes are consistent with NRC-approved Revision 4 to Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-449, "Steam Generator Tube Integrity."

The availability of this TS improvement was announced in the Federal Register on May 6, 2005 (70 FR 24126) as part of the consolidated line item improvement process (CLIIP). A discussion of the proposed TS changes is provided in Attachment 1. The marked-up and proposed TS pages are provided in Attachments 2 and 3, respectively.

The associated Bases changes are provided in Attachments 4 and 5 for information only and will be implemented in accordance with the TS Bases Control Program and 10 CFR 50.59.

The proposed changes have been reviewed and approved by the Station Nuclear Safety and Operating Committee.

Dominion requests approval of the license amendments by March 31, 2007 with a 180-day implementation period.

If you have any questions or require additional information, please contact Mr. Thomas Shaub at (804) 273-2763.

Very truly yours, Eugene S. Grecheck Vice President - Nuclear Support Services

Serial No.06-403 Docket Nos. 50-3381339 Page 2 of 3 Attachments

1. Description and Assessment

2. Mark-up of Technical Specifications Changes

3. Proposed Technical Specifications Changes

4. Mark-up of Technical Specifications Bases Changes

5. Proposed Technical Specifications Bases Changes Commitments made in this letter: None cc: U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW Suite 237185 Atlanta, Georgia 30303 Mr. J. E. Reasor, Jr.

Old Dominion Electric Cooperative lnr~sbrookCorporate Center 4201 Dominion Blvd.

Suite 300 Glen Allen, Virginia 23060 Commissioner Bureau of Radiological Health 15100 East Main Street Suite 240 Richmond, Virginia 23218 Mr. J. T. Reece NFlC Senior Resident Inspector North Anna Power Station Mr. S. R. Monarque NHC Project Manager U. S. Nuclear Regulatory Commission On~eWhite Flint North 11555 Rockville Pike Mail Stop 8-HI 2 Rockville, MD 20852

Serial No.06-403 Docket Nos. 50-3381339 Page 3 of 3 COMMONWEALTH OF VIRGINIA )

1 COUNTY OF HENRICO 1 The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Eugene S. Grecheck, who is Vice President -

Nuclear Support Services, of Virginia Electric and Power Company. He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that Company, and that the statements in the document are true to the best of his knowledge and belief. n Acknowledged before me t h i s ~ ~ ? % ' d a ~of 2006.

My Commission Expires:

(SEAL)

Serial No.06-403 Docket Nos. 50-3381339 Attachment 1 Description and Assessment North Anna Power Station Units 1 and 2 Virginia Electric and Power Company (Dominion)

Serial No.06-403 Docket Nos. 50-3381339 Description and Assessment

1.0 INTRODUCTION

The proposed license amendment revises the requirements in Technical Specifications (TS) related to steam generator tube integrity. The changes are consistent with NRC approved Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-449, "Steam Generator Tube Integrity," Revision 4. The availability of this technical specification improvement was announced in the Federal Register on May 6, 2005 as part of the consolidated line item improvement process (CLII P).

2.0 DESCRIPTION

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

Revised TS definition of LEAKAGE Revised TS 3.4.1 3, "RCS [Reactor Coolant System] Operational Leakage" New TS 3.4.20, "Steam Generator (SG) Tube Integrity" Revised TS 5.5.8, "Steam Generator (SG) Program" Revised TS 5.6.7, "Steam Generator Tube Inspection Report" Proposed revisions to the TS Bases are also included in this application. As discussed in the NRC's model safety evaluation, adoption of the revised TS Bases associated with TSTF-449, Revision 4 is an integral part of implementing this TS improvement.

The changes to the affected TS Bases pages will be incorporated in accordance with the TS Bases Control Program after approval of the license amendment.

3.0 BACKGROUND

The back.ground for this application is adequately addressed by the NRC Notice of Availability published on May 6, 2005 (70 FR 24126), the NRC Notice for Comment published1 on March 2, 2005 (70 FR 10298)) and TSTF-449, Revision 4.

4.0 REGULATORY REQUIREMENTS AND GUIDANCE The applicable regulatory requirements and guidance associated with this application are adequately addressed by the NRC Notice of Availability published on May 6, 2005 (70 FR 24126), the NRC Notice for Comment published on March 2, 2005 (70 FR 10298), and TSTF-449, Revision 4.

5.0 TECHNICAL ANALYSIS

Virginia Electric and Power Company (Dominion) has reviewed the safety evaluation (SE) published on March 2, 2005 (70 FR 10298) as part of the CLllP Notice for Comment. This included the NRC staff's SE, the supporting information provided to support TSTF-449, and the changes associated with Revision 4 to TSTF-449.

Page 1 of 3

Serial No.06-403 Docket Nos. 50-3381339 Dominion has concluded that the justifications presented in the TSTF proposal and the SE prepared by the NRC staff are applicable to North Anna Power Station Units 1 and 2 and justify this amendment for the incorporation of the changes to the North Anna Units 1 and 2 TS.

6.0 REGULATORY ANALYSIS

A description of this proposed change and its relationship to applicable regulatory requirements and guidance was provided in the NRC Notice of Availability published on May 6, 2005 (70 FR 24126), the NRC Notice for Comment published on March 2, 2005 (70 FR 10298), and TSTF-449, Revision 4.

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

Plant Name, Unit No. North Anna Power Station (NAPS) Units 1 and 2 Steam Generator Model(s): Westinghouse Model 54F; 3-Loop Effective Full Power Years (EFPY) of NAPS 1 service for currently installed SGs 11.9 EFPY at last inspection in spring 2006 NAPS 2 9.1 EFPY at last inspection in fall 2005 Tubing Material Number of tubes per SG Number and percentage of tubes NAPS 1 NAPS 2 plugged in each SG SG A - 0 (0.00%) SG A - 1 (0.03%)

Number of tubes repaired in each SG Current primary to secondary leakage TS Admin. Control Limit limits: Per SG: 50rgpd 50 gpd Leakage is evaluated at what Total: 1 gPm 150 gpd temperature condition?

At room temperature (~70°F)and normal atmosphere pressure 11 4.7/in2 \

Approved Alternate Tube Repair Criteria None (ARC):

Approved SG Tube Repair Methods None Primary to secondary leak rate values assumed in licensing Performance criteria for accident basis accident analysis, including assumed temperature leakage conditions.

1 gpm total SG leakage at room temgerature (~70°F)and normal atmosphere pressure (l4.7lin )

Page 2 of 3

Serial No.06-403 Docket Nos. 50-3381339 7.0 NO SIGNIFICANT HAZARDS CONSIDERATION Dominion has reviewed the proposed no significant hazards consideration determination published on March 2, 2005 (70 FR 10298) as part of the CLIIP.

Dominior~has concluded that the proposed determination presented in the notice is applicable to North Anna Power Station Units 1 and 2 and the determination is hereby incorporated by reference to satisfy the requirements of 10 CFR 50.91 (a).

8.0 ENVIRONMENTAL EVALUATION Dominior~ has reviewed the environmental evaluation included in the model SE published on March 2, 2005 (70 FR 10298) as part of the CLIIP. Dominion has concluded that the staff's findings presented in that evaluation are applicable to North Anna Power Station Units 1 and 2, and the evaluation is hereby incorporated by reference for this application.

9.0 PRECEDENT This application is being made in accordance with the CLIIP. Dominion is not proposing variations or deviations from the TS changes described in TSTF-449, Revision 4, or the NRC staff's model SE published on March 2, 2005 (70 FR 10298).

10.0 REFERENCES

Federal Register Notices:

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

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

Page 3 of 3

Serial No.06-403 Docket Nos. 50-3381339 Attachment 2 Mark-up of Technical Specifications Changes North Anna Power Station Units 1 and 2 Virginia Electric and Power Company (Dominion)

TECHNICAL SPECIFICATIONS TABLE OF CONTENTS 3.4 REACTOR COOLANT SYSTEM (RCS) ( c o n t i n u e d )

3.4.10 P r e s s u r i z e r S a f e t y Valves ............. .3.4.1 0 - 1 3.4.11 P r e s s u r i z e r Power Operated Re1 ie f Valves (PORVS) .................... .3.4.1 1-1 3.4.12 Low Temperature Overpressure P r o t e c t i o n (LTOP)

System . . . . . . . . . . . . . . . . . . . . .3.4.1 2.1 3.4.13 RCS O p e r a t i o n a l LEAKAGE . . . . . . . . . . . . . . .3.4.1 3 - 1 3.4.14 RCS Pressure I s o l a t i o n Valve (PIv) Leakage . . . . .3.4.1 4 - 1 3.4.15 RCS Leakage D e t e c t i o n I n s t r u m e n t a t i o n . . . . . . . .3.4.1 5 - 1 3.4.16 RCS S p e c i f i c A c t i v i t y . . . . . . . . . . . . . . . .3.4.1 6 - 1 3.4.17 RCS Loop I s o l a t i o n Valves . . . . . . . . . . . . . .3.4.1 7 - 1 3.4.18 RCS I s o l a t e d Loop S t a r t u p . . . . . . . . . . . . . .3.4.1 8 - 1

. V 3.4.19 RCS Loops-Test Exceptions . . . . . . . . . . . . . .3.4.1 9 - 1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) . . . . . . . . . . 3.5.1-1 3.5.1 Accumulators . . . . . . . . . . . . . . . . . . . . 3.5.1-1 3.5.2 ECCS-Operating . . . . . . . . . . . . . . . . . . . 3.5.2.1 3.5.3 ECCS-Shutdown . . . . . . . . . . . . . . . . . . . . 3.5.3.1 3.5.4 Refuel i n g Water Storage Tank (RWST) . . . . . . . . . 3.5.4-1 3.5.5 Seal I n j e c t i o n Flow . . . . . . . . . . . . . . . . . 3.5.5-1 3.5.6 Boron I n j e c t i o n Tank (BIT) . . . . . . . . . . . . . 3.5.6.1 CONTAINMENT SYSTEMS . . . . . . . . . . . . . . . . . . . 3.6.1.1 Containment . . . . . . . . . . . . . . . . . . . . . 3.6.1.1 Containment A i r Locks . . . . . . . . . . . . . . . . 3.6.2.1 Containment I s o l a t i o n Valves . . . . . . . . . . . . 3.6.3.1 Containment Pressure . . . . . . . . . . . . . . . . 3.6.4.1 Containment A i r Temperature . . . . . . . . . . . . . 3.6.5.1 Quench Spray (QS) System . . . . . . . . . . . . . . 3.6.6. 1 R e c i r c u l a t i o n Spray (RS) System . . . . . . . . . . . 3.6.7.1 Chemical A d d i t i o n System . . . . . . . . . . . . . . 3.6.8.1 25 3.7 PLANT SYSTEMS . . . . . . . . . . . . . . . . . . . . . . 3.7.1.1 3.7.1 Main Steam S a f e t y Valves (MSSVs) . . . . . . . . . . 3.7.1.1

. . . . . . . . . . . 3.7.2.1 3.7.2 3.7.3 Main Steam T r i p Valves (MSTVs)

Main Feedwater I s 0 1 a t i o n Valves (MFIVs) Main . .

Feedwater Pump D i scharge Val ves (MFPDVs)

Main Feedwater R e g u l a t i n g Valves (MFRVs) and Main Feedwater Regul a t i n g Bypass Valves (MFRBVS) . . . . . . . . . . . . . . . . . . . . 3.7.3.1 3.7.4 Steam Generator Power Operated Re1 ie f Valves (SG PORVS) . . . . . . . . . . . . . . . . . . . 3.7 1 3.7.5 Auxi 1 i a r y Feedwater (AFW) System . . . . . . . . . . 3.7.5.1 3.7.6 Emergency Condensate Storage Tank (ECST) . . . . . . 3.7.6.1 3.7.7 Secondary Speci f i c A c t i v i t y . . . . . . . . . . . . . 3.7.7.1 3.7.8 S e r v i c e Water (SW) System . . . . . . . . . . . . . . 3.7.8.1 3.7.9 U l t i m a t e Heat S i n k (UHS) . . . . . . . . . . . . . . 3.7.9.1 N o r t h Anna U n i t s 1 and 2 ii

Definitions 1.1 1.1 D e f i n i t i o n s E-AVERAGE DISINTEGRATION E s h a l l be t h e average (weighted i n p r o p o r t i o n t o ENERGY t h e concentration o f each r a d i onucl ide i n t h e r e a c t o r coolant a t t h e time o f sampling) o f t h e sum o f t h e average beta and gamma energies p e r d i s i n t e g r a t i o n ( i n MeV) f o r isotopes, o t h e r than iodines, w i t h h a l f l i v e s > 15 minutes, making up a t l e a s t 95% o f t h e t o t a l noniodine a c t i v i t y i n t h e coolant.

ENGINEERED SAFETY The ESF RESPONSE TIME s h a l l be t h a t t i m e i n t e r v a l FEATURE (ESF) RESPONSE from when t h e monitored parameter exceeds i t s ESF TIME: a c t u a t i o n s e t p o i n t a t t h e channel sensor u n t i 1 t h e ESF equipment i s capable o f performing i t s s a f e t y f u n c t i o n (i.e., t h e valves t r a v e l t o t h e i r r e q u i r e d p o s i t i o n s , pump discharge pressures reach t h e i r r e q u i r e d values, etc.) Times s h a l l i n c l u d e d i e s e l generator s t a r t i n g and sequence 1oadi ng de1ays , where appl i c a b l e. The response t i m e may be measured by means o f any s e r i e s o f sequential, overlapping, o r t o t a l steps so t h a t t h e e n t i r e response t i m e i s measured. I n l i e u o f measurement, response time may be v e r i f i e d f o r s e l e c t e d components provided t h a t t h e components and method01ogy f o r v e r i f i c a t i o n have been p r e v i o u s l y reviewed and approved by t h e NRC.

LEAKAGE LEAKAGE s h a l l be:

a. I d e n t i f i e d LEAKAGE

1. LEAKAGE, such as t h a t from pump seals o r valve packing (except r e a c t o r cool a n t pump (RCP) seal water i n j e c t i o n o r l e a k o f f j , t h a t i s captured and conducted t o c o l l e c t i o n systems o r a sump o r c o l l e c t i n g tank;

2. LEAKAGE i n t o t h e containment atmosphere from sources t h a t a r e both s p e c i f i c a l l y l o c a t e d and known e i t h e r n o t t o i n t e r f e r e w i t h t h e o p e r a t i on o f 1eakage d e t e c t i o n systems o r n o t t o be pressure boundary LEAKAGE; o r N o r t h Anna U n i t s 1 and 2 1.1-3 Amendments mtmb

Definitions 1.1 1.1 D e f i n i t i o n s LEAKAGE b. U n i d e n t i f i e d LEAKAGE (continued)

A l l LEAKAGE (except RCP seal water i n j e c t i o n o r l e a k o f f ) t h a t i s n o t i d e n t i f i e d LEAKAGE;

c. Pressure Boundary LEAKAGE e e o z a LEAKAGE (except& EAKA noni sol able f aul t i n an RCS component body, p i p e w a l l , o r vessel w a l l .

MASTER RELAY TEST A MASTER RELAY TEST s h a l l c o n s i s t o f e n e r g i z i n g a l l master r e l a y s i n t h e channel r e q u i r e d f o r channel OPERABILITY and v e r i f y i ng t h e OPERABILITY o f each r e q u i r e d master r e l a y . The MASTER RELAY TEST s h a l l i n c l u d e a c o n t i n u i t y check of each associated r e q u i r e d slave re1ay. The MASTER RELAY TEST may be performed by means o f any s e r i e s of sequenti a1 , overlapping, o r t o t a l steps.

MODE A MODE s h a l l correspond t o any one i n c l u s i v e combination o f core r e a c t i v i t y c o n d i t i o n , power 1eve1 , average r e a c t o r cool a n t temperature, and r e a c t o r vessel head c l osure b o l t t e n s i o n i n g s p e c i f i e d i n Table 1.1-1 w i t h f u e l i n t h e r e a c t o r vessel.

A system, subsystem, t r a i n , component, o r device s h a l l be OPERABLE o r have OPERABILITY when i t i s capabl e o f performing i t s speci f i ed safety f u n c t i o n ( s ) and when a1 1 necessary attendant instrumentation, c o n t r o l s , normal o r emergency e l e r t r i c a l power, cool i n g and seal water, l u b r i c a t i o n , and o t h e r auxi 1i a r y equipment t h a t a r e r e q u i r e d f o r t h e system, subsystem, t r a i n ,

component, o r device t o perform i t s s p e c i f i e d s a f e t y f u n c t i o n (s) a r e a1 so capabl e o f p e r f ormi ng t h e i r r e l a t e d support f u n c t i o n ( s ) .

PHYSICS TESTS PHYSICS TESTS s h a l l be those t e s t s performed t o measure t h e fundamental n u c l e a r c h a r a c t e r i s t i c s o f t h e r e a c t o r core and r e l a t e d instrumentation.

These t e s t s are:

a. Described i n Chapter 14, I n i t i a l Tests and Operation, o f t h e UFSAR; (continued)

North Anna U n i t s 1 and 2 1.1-4 Amendments 9STfTTT),

RCS Operati onal LEAKAGE 3.4.13 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.13 RCS Operati onal LEAKAGE LC0 3.4.13 RCS o p e r a t i o n a l LEAKAGE s h a l l be 1 i m i t e d t o :

a. No pressure boundary LEAKAGE;

b. 1 gpm u n i d e n t i f i e d LEAKAGE;

c. 10 gpm i d e n t i f i e d LEAKAGE; d

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

ACT l ONS COMPLETION TIME A. d RCS EAKAGEnotwithin 1i m i t s f o r reasons A.l Reduce LEAKAGE t o within limits.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months o t h e r than pressure % .-

boundary LEAKAG

/

Grr LPAE~GCT

-b Y C I W ~ ~ V ~

I

< 6 B. Required A c t i o n and B.l BeinMODE3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Compl e t i on Time o f C o n d i t i o n A -AND n o t met.

8.2 Be i n MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Pressure boundary LEAKAGE e x i s t s .

North Anna U n i t s 1 and 2 3.4.13-1 Amendments

RCS O p e r a t i onal LEAKAGE

\

.paMq A

G GLL-SURVEILLANCE REQUIREMENTS

/ SURVEILLANCE -

s FREQUENCY C, SR 3.4.13.1 s ------------------- NOT offdi.,------------------

(, o t r e q u i r e d t o be performed u n t i l 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months f t e r e s t a b l ishment o...........................................

peration.

state V e r i f y RCS o p e r a t i o n a l LEAKAGE i s w i t h i n 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months l i m i t s by performance o f RCS water i n v e n t o r y b a l ance.

N o r t h Anna U n i t s 1 and 2 3.4.13-2

Ew TS 3 , 4 . z O 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.20 Steam Generator (SG) Tube Integrity LC0 3.4.20 SG tube integrity shall be maintained.

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

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

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more SG tubes A.l Verify tube integrity of the 7 days satisfying the tube repair affected tube(s) is criteria and not plugged maintained until the next in accordance with the refueling outage or SG Steam Generator tube inspection.

Program.

A.2 Plug the affected tube(s) in accordance with the Steam Prior to entering Generator Program. MODE 4 following the next refueling outage or SG tube inspection B. Req~~ired Action and B.l Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A not AND met.

B.2 Be in MODE 5 .

OR 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SG tube integrity not maintained.

3.4.20-1 Amendments

SURVEILLANCE REQUIREMENTS SURVEILLANCE ( FREQUENCY SR 3.4.2!0.1 Verify SG tube integrity in accordance with the In accordance Steam Generator Program. with the Steam Generator Program SR 3.4.20.2 Verify that each inspected SG tube that satisfies the Prior to entering tube repair criteria is plugged in accordance with the MODE 4 Steam Generator Program. following a SG tube inspection Amendments

Programs and Manuals 5.5 5.5 Proqrams and Manual s 5.5.7 Inservi ce Testing Program This program provides controls f o r inservice t e s t i n g of ASME Code Class 1, 2, and 3 components. The program shall include the f 01 1owi ng :

a. Testing frequencies specified in the ASME Code f o r Operation and Mai ntenance of Nuclear Power Pl ants and appl i cabl e Addenda as follows:

ASME Code f o r Operation and Maintenance of Nuclear Power Plants and appl i cabl e Addenda Requi red Frequencies f o r termi no1 ogy f o r inservice performing i nservi ce testing activities testing activities Weekly A t l e a s t once per 7 days Monthly A t l e a s t once per 31 days Quarterly or every 3 months A t l e a s t once per 92 days Semiannually or every 6 months A t l e a s t once per 184 days Every 9 months A t l e a s t once per 276 days Yearly or annual ly A t l e a s t once per 366 days Biennially or every 2 years A t l e a s t once per 731 days

b. The provisions of SR 3.0.2 a r e applicable t o the above required Frequencies f o r performing i nservice t e s t i n g a c t i v i t i e s ;

c. The provisions of SR 3.0.3 a r e applicable t o inservice t e s t i n g a c t i v i t i e s ; and

d. Nothing in the ASME Code f o r Operation and Maintenance of Nuclear Power Plants shall be construed t o supersede the requirements of any TS.

5.5.8 Steam Generator (SG) Program C P North Anna Units 1 and 2 5.5-5 Amendments

Programs and Manual s 5.5 5.5 Proarams and Manuals North Anna Uni t s 1 and 2 5.5-6 Amendments

Programs and Manual s 5.5 5.5 Proarams and Manuals North Anna Units 1 and 2 5.5-7 Amendments

Programs and Manuals 5.5 5.5 Programs and Manuals North Anna Units 1 and 2 5.5-8 Amendments +%-

Programs and Manual s 5.5 Norlth Anna Units 1 and 2 5.5-9 Amendments v

Programs and Manuals 5.5 North Anna Units 1 and 2 5.5-10 Amendments v

Programs and Manual s 5.5 N o r t h Anna U n i t s 1 and 2 Amendments -Tk

INSERT 5.5.8 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:

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 whicli the SG tubes are inspected or plugged to confirm that the performance criteria are being met.

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 for all SGs.

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

Page 1

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 phxled-Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

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

2. Inspect 100% of the tubes at sequential periods of 144, 108, 72, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected.

3. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack@),then the indication need not be treated as a crack.

e. Provisions for monitoring operational primary to secondary LEAKAGE.

Page 2

Reporting Requi rements 5.6 5.6 Reporting Requi rements 5.6.5 CORE OPERATING LIMITS REPORT (COLR)

b. (continued)

14. BAW-10199P-A, "The BWU C r i t i c a l Heat F l ux C o r r e l a t i o n s . "

15. BAW-10170P-A, " S t a t i s t i c a l Core Design f o r M i x i n g Vane Cores."

16. EMF-2103 (P) (A), "Real is t i c Large Break LOCA Method01ogy f o r Pressurized Water Reactors. "

17. EMF-96-029 (P) (A), "Reactor Analysis System f o r PWRs."

18. BAW-10168P-A, "RSG LOCA BWNT Loss-of-Coolant Accident Evaluation Model f o r R e c i r c u l a t i n g Steam Generator Plants,"

Vol ume II on1y (SBLOCA model s) .

c. The core o p e r a t i n g 1 i m i t s s h a l l be determined such t h a t a l l appl i c a b l e 1 i m i t s (e.g., f u e l thermal mechanical 1 i m i t s , c o r e thermal hydraul ic 1i m i t s , Emergency Core Cool i n g Systems (ECCS) l i m i t s , n u c l e a r l i m i t s such as SDM, t r a n s i e n t a n a l y s i s l i m i t s ,

and accident a n a l y s i s 1i m i t s ) o f t h e s a f e t y a n a l y s i s a r e met.

d. The COLR, i n c l u d i n g any midcycle r e v i s i o n s o r supplements, s h a l l be provided upon issuance f o r each r e l o a d c y c l e t o t h e NRC.

5.6.6 PAM Report When a r e p o r t i s r e q u i r e d by Condition B o f LC0 3.3.3, "Post Accident M o n i t o r i n g (PAM) Instrumentation," a r e p o r t s h a l l be submitted w i t h i n t h e f o l l o w i n g 14 days. The r e p o r t s h a l l o u t l i n e t h e cause o f t h e i n o p e r a b i l i t y , and t h e plans and schedule f o r r e s t o r i n g t h e i n s t r u m e n t a t i o n channels o f t h e Function t o OPERABLE s t a t u s .

5.6.7 Steam Generator Tube I n s p e c t i o n ~ e p b r t North" Anna U n i t s 1 and 2 5.6-4 Amendments 239/220

Reporting Requi rements 5.6 5.6 Reporting Requirements North Anna Units 1 and 2 Amendments -EB$Vl+

INSERT 5.6.7 A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.8, "Steam Generator (SG) Program." The report shall include:

The scope of inspections performed on each SG, Active degradation mechanisms found, Nondestructive examination techniques utilized for each degradation mechanism, Location, orientation (if linear), and measured sizes (if available) of service induced indications, Number of tubes plugged during the inspection outage for each active degradation mechanism, Tota.1number and percentage of tubes plugged to date, The results of condition monitoring, including the results of tube pulls and in-situ testing, and The effective plugging percentage for all plugging in each SG.

Serial No.06-403 Docket Nos. 50-3381339 Attachment 3 Proposed Technical Specifications Changes North Anna Power Station Units 1 and 2 Virginia Electric and Power Company (Dominion)

TECHNICAL SPECIFICATIONS TABLE OF CONTENTS REACTOR COOLANT SYSTEM (RCS) ( c o n t i n u e d )

P r e s s u r i z e r S a f e t y Valves . . . . . . . . . . . . . .3.4.1 0.1 P r e s s u r i z e r Power Operated R e l i e f Valves (PORVS) . . . . . . . . . . . . . . . . . . . . .3.4.1 1.1 Low Temperature Overpressure P r o t e c t i o n (LTOP)

System . . . . . . . . . . . . . . . . . . . . .3.4.1 2-1 RCS O p e r a t i o n a l LEAKAGE . . . . . . . . . . . . . . .3.4.1 3-1 RCS Pressure I s o l a t i o n Valve (PIV) Leakage . . . . .3.4.1 4.1 RCS Leakage D e t e c t i o n I n s t r u m e n t a t i o n . . . . . . . .3.4.1 5-1 RCS S p e c i f i c A c t i v i t y . . . . . . . . . . . . . . . .3.4.1 6.1 RCS Loop I s o l a t i o n Valves . . . . . . . . . . . . . .3.4.1 7.1 RCS I s o l a t e d Loop S t a r t u p . . . . . . . . . . . . . .3.4.1 8.1 RCS Loops-Test Exceptions . . . . . . . . . . . . . .3.4.1 9 - 1 Steam Generator (SG) Tube I n t e g r i t y . . . . . . . . .3.4.2 0.1 1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) . . . . . . . . . . 3.5.1. 1 3.5.1 Accumulators . . . . . . . . . . . . . . . . . . . . 3.5.1-1 3.5.2 ECCS-Operating .......... . . . . . . . . . 3.5.2.1 3.5.3 ECCS-Shutdown . . . . . . . . . . . . . . . . . . . . 3.5.3-1 3.5.4 Refuel ing Water Storage Tank (RWST) . . . . . . . . . 3.5.4-1 3.5.5 Seal I n j e c t i o n Flow . . . . . . . . . . . . . . . . . 3.5.5-1 3.5.6 Boron I n j e c t i o n Tank (BIT) .. . . . . . . . . . . . 3.5.6-1 CONTAINMENT SYSTEMS . . . . . . . . . . . . . . . . . . . 3.6.1.1 Containment . . . . . . . . . . . . . . . . . . . . . 3.6.1.1 Containment A i r Locks . . . . . . . . . . . . . . . . 3.6.2.1 Containment I s o l a t i o n Valves . . . . . . . . . . . . 3.6.3.1 Containment Pressure . . . . . . . . . . . . . . . . 3.6.4.1 Containment A i r Temperature . . . . . . . . . . . . . 3.6.5.1 Quench Spray (QS) System . . . . . . . . . . . . . . 3.6.6.1 R e c i r c u l a t i o n Spray (RS) System . . . . . . . . . . . 3.6.7.1 Chemical A d d i t i o n System . . . . . . . . . . . . . . 3.6.8.1 PLANT SYSTEMS . . . . . . . . . . . . . . . . . . . . . . 3.7.1.1 Main Steam S a f e t y Valves (MSSVs) . . . . . . . . . . 3.7.1.1 Main Steam T r i p Valves (MSTVs) . . . . . . . . . . . 3.7.2.1 Main Feedwater I s o l a t i o n Valves (MFIVs) , Main Feedwater Pump Discharge Valves (MFPDVs)

Main Feedwater R e g u l a t i n g Valves (MFRVs) and Main Feedwater R e g u l a t i n g Bypass Valves (MFRBVs) . . . . . . . . . . . . . . . . . . . . 3.7.3.1 Steam Generator Power Operated R e l i e f Valves (SG PORVS) . . . . . . . . . . . . . . . . . . . 3.7.4.1 Auxi 1i a r y Feedwater (AFW) System . . . . . . . . . . 3.7.5.1 Emergency Condensate Storage Tank (ECST) . . . . . . 3.7.6.1 Secondary S p e c i f i c A c t i v i t y . . . . . . . . . . . . . 3.7.7.1 S e r v i ce Water (SW) System . . . . . . . . . . . . . . 3.7.8.1 U l t i m a t e Heat S i n k (UHS) . . . . . . . . . . . . . . 3.7.9.1 N o r t h Anna U n i t s 1 and 2 ii

Definitions 1.1 1.1 D e f i n i t i o n s E-AVERAGE DISINTEGRATION E s h a l l be t h e average (weighted i n p r o p o r t i o n t o ENERGY t h e c o n c e n t r a t i o n o f each r a d i o n u c l i d e i n t h e r e a c t o r c o o l a n t a t t h e t i m e o f sampling) of t h e sum o f t h e average b e t a and gamma e n e r g i e s p e r d i s i n t e g r a t i o n ( i n MeV) f o r i s o t o p e s , o t h e r t h a n i o d i n e s , w i t h h a l f l i v e s > 15 minutes, making up a t l e a s t 95% o f t h e t o t a l n o n i o d i n e a c t i v i t y i n the coolant.

ENGINEERED SAFETY The ESF RESPONSE TIME s h a l l be t h a t t i m e i n t e r v a l FEATURE (ESF) RESPONSE f r o m when t h e m o n i t o r e d parameter exceeds i t s ESF TIME a c t u a t i o n s e t p o i n t a t t h e channel sensor u n t i l t h e ESF equipment i s capable o f p e r f o r m i n g i t s s a f e t y f u n c t i o n ( i .e., t h e v a l v e s t r a v e l t o t h e i r r e q u i r e d p o s i t i o n s , pump d i scharge pressures r e a c h t h e i r r e q u i r e d values, e t c . ) . Times s h a l l i n c l u d e d i e s e l g e n e r a t o r s t a r t i n g and sequence l o a d i n g delays, where appl ic a b l e . The response t i m e may be measured by means o f any s e r i e s o f s e q u e n t i a l ,

o v e r l a p p i n g , o r t o t a l s t e p s so t h a t t h e e n t i r e response t i m e i s measured. I n li e u o f measurement, response t i m e may be v e r i f i e d f o r s e l e c t e d components p r o v i d e d t h a t t h e components and method01 ogy f o r v e r i f i c a t i o n have been p r e v i o u s l y reviewed and approved b y t h e NRC.

LEAKAGE LEAKAGE s h a l l be:

a. I d e n t i f i e d LEAKAGE

1. LEAKAGE, such as t h a t f r o m pump s e a l s o r v a l ve p a c k i n g (except r e a c t o r c o o l a n t pump (RCP) seal w a t e r i n j e c t i o n o r l e a k o f f ) , t h a t i s c a p t u r e d and conducted t o c o l l e c t i o n systems o r a sump o r c o l l e c t i n g tank;

2. LEAKAGE i n t o t h e containment atmosphere f r o m sources t h a t a r e b o t h s p e c i f i c a l l y l o c a t e d and known e i t h e r n o t t o i n t e r f e r e w i t h t h e o p e r a t i o n of leakage d e t e c t i o n systems o r n o t t o be p r e s s u r e boundary LEAKAGE; o r

3. Reactor Coolant System (RCS) LEAKAGE t h r o u g h a steam g e n e r a t o r t o t h e Secondary System

( p r i m a r y t o secondary LEAKAGE); I (continued)

N o r t h Anna U n i t s 1 and 2 1.1-3

Definitions 1.1 1.1 D e f i n i t i o n s LEAKAGE b. U n i d e n t i f i e d LEAKAGE

( c o n t i nued)

A1 1 LEAKAGE (except RCP seal w a t e r i n j e c t i o n o r l e a k o f f ) t h a t i s n o t i d e n t i f i e d LEAKAGE; c . Pressure Boundary LEAKAGE LEAKAGE (except p r i m a r y t o secondary LEAKAGE) t h r o u g h a n o n i s o l a b l e f a u l t i n an RCS component I

body, p i p e w a l l , o r vessel w a l l .

MASTER RELAY TEST A MASTER RELAY TEST s h a l l c o n s i s t o f e n e r g i z i n g a1 1 master r e l a y s i n t h e channel r e q u i r e d f o r channel OPERABILITY and v e r i f y i ng t h e OPERABILITY o f each r e q u i r e d master r e l a y . The MASTER RELAY TEST s h a l l i n c l u d e a c o n t i n u i t y check o f each associ a t e d r e q u i r e d s l ave r e 1 ay The MASTER RELAY TEST may be performed by means o f any s e r i e s o f s e q u e n t i a l , o v e r l a p p i n g , o r t o t a l steps.

MODE A MODE s h a l l correspond t o any one i n c l u s i v e combination o f c o r e r e a c t i v i t y c o n d i t i o n , power 1eve1 , average r e a c t o r c o o l a n t temperature, and r e a c t o r vessel head c l o s u r e b o l t t e n s i o n i n g s p e c i f i e d i n Table 1.1-1 w i t h f u e l i n t h e r e a c t o r vessel.

OPERABLE-OPERABILITY A system, subsystem, t r a i n , component, o r d e v i c e s h a l l be OPERABLE o r have OPERABILITY when i t i s capable o f p e r f o r m i n g i t s s p e c i f i e d s a f e t y f u n c t i o n (s) and when a1 1 necessary a t t e n d a n t i n s t r u m e n t a t i o n , c o n t r o l s, normal o r emergency e l e c t r i c a l power, c o o l i n g and seal water, 1u b r i c a t i o n , and o t h e r a u x i 1 i a r y equipment t h a t a r e r e q u i r e d f o r t h e system, subsystem, t r a i n ,

component, o r d e v i c e t o p e r f o r m i t s s p e c i f i e d s a f e t y f u n c t i o n ( s ) a r e a1 so capable o f p e r f o r m i n g t h e i r r e 1a t e d s u p p o r t f u n c t i o n ( s ) .

PHYSICS TES'TS PHYSICS TESTS s h a l l be t h o s e t e s t s performed t o measure t h e fundamental n u c l e a r c h a r a c t e r i s t i c s o f t h e r e a c t o r c o r e and r e l a t e d i n s t r u m e n t a t i o n .

These t e s t s are:

a. Described i n Chapter 14, I n i t i a l Tests and Operation, o f t h e UFSAR; (continued)

N o r t h Anna U n i t s 1 and 2 1.1-4

RCS O p e r a t i o n a l LEAKAGE 3.4.13 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1.3 RCS O p e r a t i o n a l LEAKAGE LC0 3.4.13 RCS o p e r a t i o n a l LEAKAGE s h a l l be l i m i t e d t o :

a. No p r e s s u r e boundary LEAKAGE;

b. 1 gpm u n i d e n t i f i e d LEAKAGE;

c. 10 gpm i d e n t i f i e d LEAKAGE;

d. 150 g a l l o n s p e r day p r i m a r y t o secondary LEAKAGE t h r o u g h any one steam g e n e r a t o r (SG) . I APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. RCS o p e r a t i o n a l A.l Reduce LEAKAGE t o LEAKAGE n o t w i t h i n l i m i t s f o r reasons w i t h i n 1i m i t s .

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months I

o t h e r than pressure boundary LEAKAGE o r p r i m a r y t o secondary LEAKAGE.

B. Required A c t i o n and B.l Be i n MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months a s s o c i a t e d Completion Time o f C o n d i t i o n A AND n o t met.

B.2 BeinMODE5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Pressure boundary LEAKAGE e x i s t s .

P r i m a r y t o secondary LEAKAGE n o t w i t h i n limit.

N o r t h Anna U n i t s 1 and 2

RCS O p e r a t i o n a l LEAKAGE SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.13.1 ------------------- NOTES-------------------

1. Not r e q u i r e d t o be performed u n t i 1 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months a f t e r e s t a b l i s h m e n t o f steady state operation.

2. Not a p p l i c a b l e t o p r i m a r y t o secondary LEAKAGE.

V e r i f y RCS o p e r a t i o n a l LEAKAGE i s w i t h i n 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months l i m i t s by performance o f RCS w a t e r i n v e n t o r y balance.

SR 3.4.1:1.2 ------------------- NOTE--------------------

Not r e q u i r e d t o be performed u n t i l 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months a f t e r establishment o f steady s t a t e operation.

V e r i f y p r i m a r y t o secondary LEAKAGE i s 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months I150 g a l l o n s p e r day t h r o u g h any one SG. I N o r t h Anna U n i t s 1 and 2

SG Tube I n t e g r i t y 3.4.20 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.20 Steam Generator (SG) Tube l ntegr i t y LC0 3.4.20 SG tube i ntegr i t y sha 1 l be ma i n t a i ned .

A1 1 SG tubes s a t i s f y i ng t h e tube repa i r c r i t e r i a sha l 1 be p l ugged i n accordance w i t h t h e Steam Generator Program.

APPLICABILITY: MODES1, 2, 3, a n d 4 .

ACT l ONS COND l T l ON REQU l RED ACT l ON COMPLET l ON T l ME A. One o r more SG tubes A. 1 V e r i f y tube i n t e g r i t y 7 days s a t i s f y i n g t h e tube o f the affected repa i r c r i t e r i a and tube (s) i s ma i nta i ned n o t p l ugged i n u n t i l the next accordance w i t h the refuel ing outage o r SG Steam Generator tube inspection.

Program.

AND A. 2 PI ug the affected Pr i o r t o tube (s) i n accordance entering MODE 4 w i t h t h e Steam f o l lowing the Generator Program. next r e f u e l i ng outage or SG tube i nspect i on North Anna U n i t s 1 and 2 3.4.20-1

SG Tube l ntegr i t y 3.4.20 ACT l ONS I

COND l T l ON I REQU IRED ACT l ON COMPLETION TIME B. Required Action and B.l Be i n MODE 3 . 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months assolc i ated Comp I e t i on Time o f Condition A AND n o t met.

B.2 Be inMODE5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SG tube i n t e g r i t y not ma i nta i ned .

SURVE l LLANCE REQU l REMENTS SURVE l LLANCE I FREQUENCY SR 3.4.20.1 V e r i f y SG tube i ntegr i t y i n accordance w i t h I n accordance t h e Steam Generator Program. w i t h t h e Steam Generator Program SR 3 . 4 .;!0.2 V e r i f y t h a t each inspected SG tube t h a t Pr i o r t o s a t i s f i e s the tube r e p a i r c r i t e r i a i s entering MODE 4 p l ugged i n accordance w i t h t h e Steam f o l lowing a SG Generator Program. tube i nspect i on North Anna U n i t s 1 and 2 3.4.20-2

Programs and Manual s 5.5 5.5 Proqrams and Manual s 5.5.:7 Inservice Testing Program This program provides controls f o r inservice testing of ASME Code Class 1, 2 , and 3 components. The program shall include the fol 1owing :

a. Testing frequencies specified in the ASME Code f o r Operation and Maintenance of Nuclear Power Plants and applicable Addenda as fol 1 ows :

ASME Code f o r Operation and Maintenance of Nuclear Power Plants and applicable Addenda Required Frequencies for termi no1 ogy f o r i nservi ce performing inservice testina activities testina activities Weekly A t l e a s t once per 7 days Monthly A t l e a s t once per 31 days Quarterly or every 3 months A t l e a s t once per 92 days Semiannual ly or every 6 months A t l e a s t once per 184 days Every 9 months A t l e a s t once per 276 days Yearly or annually A t l e a s t once per 366 days Biennially or every 2 years A t l e a s t once per 731 days

b. The provisions of SR 3.0.2 are applicable t o the above required Frequencies f o r performing i nservi ce testing a c t i v i t i e s ;

c. The provisions of SR 3.0.3 are applicable t o inservice testing a c t i v i t i e s ; and

d. Nothing i n the ASME Code f o r Operation and Maintenance of Nuclear Power Plants shall be construed t o supersede the requirements of any TS.

5.5.8 Steam Generator (SG) Program A Steam Generator Program shall be established and implemented t o ensure t h a t SG tube integrity i s maintained. In addition, the Steam Generator Program shall incl ude the fol lowing provisions :

a. Provisions f o r condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect t o the performance c r i t e r i a for structural i n t e g r i t y and accident induced leakage. The "as found" condition refers t o the condition of the t u b i n g during a SG inspection outage, as determined from the inservice (cont i nued)

North Anna Units 1 and 2 5.5-5

Programs and Manual s 5.5 5.5 Programs and Manuals 5.5.8 Steam

- Generator (SG) Program

a. (conti nued) inspection r e s u l t s o r by o t h e r means, p r i o r t o t h e plugging of tubes. Condition moni t o r i ng assessments s h a l l be conducted during each outage during which t h e SG tubes a r e inspected o r plugged t o confirm t h a t t h e performance c r i t e r i a a r e being met.

b. Performance c r i t e r i a f o r SG tube i n t e g r i t y . SG tube i n t e g r i t y s h a l l be maintained by meeting t h e performance c r i t e r i a f o r tube s t r u c t u r a l i n t e g r i t y , accident induced leakage, and operational LEAKAGE.

S t r u c t u r a l i n t e g r i t y performance c r i t e r i o n : All in-service steam generator tubes s h a l l r e t a i n s t r u c t u r a l i n t e g r i t y over t h e f u l l range of normal operati ng condi t i ons ( i ncl udi ng s t a r t u p , operation in t h e power range, hot standby, and cool down and a1 1 a n t i c i p a t e d t r a n s i e n t s included in t h e design s p e c i f i c a t i o n ) and design b a s i s accidents. This includes r e t a i n i n g a s a f e t y f a c t o r of 3.0 a g a i n s t burst under normal steady s t a t e f u l l power operation primary t o secondary pressure d i f f e r e n t i a l and a s a f e t y f a c t o r of 1 . 4 a g a i n s t burst applied t o t h e design b a s i s accident primary t o secondary pressure d i f f e r e n t i a l s . Apart from t h e above requirements, additional loading conditions associated with t h e design b a s i s accidents, o r combination of accidents in accordance with t h e design and 1i censi ng b a s i s , s h a l l a1 so be eval uated t o determine i f t h e associated 1 oads c o n t r i b u t e s i g n i f i c a n t l y t o burst o r collapse. In t h e assessment of tube i n t e g r i t y , those loads t h a t do s i g n i f i c a n t l y a f f e c t burst o r c o l l a p s e s h a l l be determined and assessed in combination with t h e loads due t o pressure with a s a f e t y f a c t o r of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

2. Accident induced leakage performance c r i t e r i o n : The primary t o secondary accident induced leakage r a t e f o r any design b a s i s accident, o t h e r than a SG tube rupture, s h a l l not exceed t h e leakage r a t e assumed in t h e accident a n a l y s i s in terms of t o t a l leakage r a t e f o r a l l SGs and leakage r a t e f o r an individual SG. Leakage i s not t o exceed 1 gpm f o r a1 1 SGs.

3. The operational LEAKAGE performance c r i t e r i o n i s s p e c i f i e d in LC0 3.4.13, "RCS Operational LEAKAGE."

North Anna Units 1 and 2 5.5-6

Programs and Manual s 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) Program ( c o n t i n u e d )

c. P r o v i s i o n s f o r SG t u b e r e p a i r c r i t e r i a . Tubes found by i n s e r v i c e i n s p e c t i o n t o c o n t a i n f l a w s w i t h a d e p t h equal t o o r exceeding 40% o f t h e nominal t u b e w a l l t h i c k n e s s s h a l l be plugged.

d. P r o v i s i o n s f o r SG t u b e i n s p e c t i o n s . P e r i o d i c SG t u b e i n s p e c t i o n s s h a l l be performed. The number and p o r t i o n s o f t h e t u b e s i n s p e c t e d and methods o f i n s p e c t i o n s h a l l be performed w i t h t h e o b j e c t i v e o f d e t e c t i n g f l a w s o f any t y p e (e.g., v o l u m e t r i c f l a w s , a x i a l and c i r c u m f e r e n t i a l cracks) t h a t may be p r e s e n t a l o n g t h e l e n g t h o f t h e tube, f r o m t h e t u b e - t o - t u b e s h e e t weld a t t h e t u b e i n l e t t o t h e t u b e - t o - t u b e s h e e t weld a t t h e t u b e o u t l e t ,

and t h a t may s a t i s f y t h e a p p l i c a b l e t u b e r e p a i r c r i t e r i a . The t u b e - t o - t u b e s h e e t weld i s n o t p a r t o f t h e tube. I n a d d i t i o n t o meeting t h e requirements o f d.1, d.2, and d.3 below, t h e i n s p e c t i o n scope, i n s p e c t i o n methods, and i n s p e c t i o n i n t e r v a l s s h a l l be such as t o ensure t h a t SG t u b e i n t e g r i t y i s m a i n t a i n e d u n t i l t h e n e x t SG i n s p e c t i o n . An assessment o f d e g r a d a t i o n s h a l l be performed t o determine t h e t y p e and l o c a t i o n o f f l a w s t o which t h e tubes may be s u s c e p t i b l e and, based on t h i s assessment, t o d e t e r m i n e which i n s p e c t i o n methods need t o be employed and a t what l o c a t i o n s .

1. I n s p e c t 100% o f t h e tubes i n each SG d u r i n g t h e f i r s t r e f u e l ing outage f o l 1owing SG rep1 acement .

2. I n s p e c t 100% o f t h e tubes a t s e q u e n t i a l p e r i o d s o f 144, 108, 72, and, t h e r e a f t e r , 60 e f f e c t i v e f u l l power months. The f i r s t s e q u e n t i a l p e r i o d s h a l l be considered t o b e g i n a f t e r t h e f i r s t i n s e r v i c e i n s p e c t i o n o f t h e SGs. I n a d d i t i o n ,

i n s p e c t 50% o f t h e tubes by t h e r e f u e l i n g outage n e a r e s t t h e m i d p o i n t o f t h e p e r i o d and t h e r e m a i n i n g 50% by t h e r e f u e l i n g outage n e a r e s t t h e end o f t h e p e r i o d . No SG s h a l l o p e r a t e f o r more t h a n 72 e f f e c t i v e f u l l power months o r t h r e e r e f u e l i n g outages (whichever is 1ess) w i t h o u t b e i n g i n s p e c t e d .

3. I f c r a c k i n d i c a t i o n s a r e found i n any SG tube, t h e n t h e n e x t i n s p e c t i o n f o r each SG f o r t h e d e g r a d a t i o n mechanism t h a t caused t h e c r a c k i n d i c a t i o n s h a l l n o t exceed 24 e f f e c t i v e f u l l power months o r one r e f u e l i n g outage (whichever i s l e s s ) . I f d e f i n i t i v e i n f o r m a t i o n , such as f r o m e x a m i n a t i o n o f a p u l l e d tube, d i a g n o s t i c n o n - d e s t r u c t i v e t e s t i n g , o r e n g i n e e r i n g e v a l u a t i o n i n d i c a t e s t h a t a crack-1 ike i n d i c a t i o n i s not associated w i t h a crack(s), then the i n d i c a t i o n need n o t be t r e a t e d as a crack.

N o r t h Anna U n i t s 1 and 2 5.5-7

Programs and Manual s 5.5 5.5 Programs and Manuals 5.5.8 Steam

- Generator (SG) Program (continued)

e. Provisions f o r monitoring operational primary t o secondary LEAKAGE.

5.5.9 Secondary

- Water Chemistry Program This program provides c o n t r o l s f o r monitoring secondary water chemistry t o i n h i b i t SG tube degradation and low pressure t u r b i n e d i s c s t r e s s corrosion cracking. The program s h a l l include:

a. I d e n t i f i c a t i o n of a sampl ing schedule f o r t h e c r i t i c a l v a r i a b l e s and control p o i n t s f o r t h e s e v a r i a b l e s ;

. I d e n t i f i c a t i o n of t h e procedures used t o measure t h e values of the critical variables;

c. I d e n t i f i c a t i o n of process sampling p o i n t s , which s h a l l include monitoring t h e discharge of t h e condensate pumps f o r evidence of condenser i n 1 eakage;

d. Procedures f o r t h e recording and management of d a t a ;

e. Procedures d e f i n i n g c o r r e c t i v e a c t i o n s f o r a1 1 o f f control point chemistry conditions; and

f. A procedure i d e n t i f y i n g t h e a u t h o r i t y responsible f o r t h e i n t e r p r e t a t i o n of t h e d a t a and t h e sequence and timing of a d m i n i s t r a t i v e e v e n t s , which i s required t o i n i t i a t e c o r r e c t i v e action.

5.5.10 V e n t i l a t i o n F i l t e r Testing Program (VFTP)

A program s h a l l be e s t a b l i s h e d t o implement t h e following required t e s t i n g of Engineered Safety Feature (ESF) f i l t e r venti 1 a t i o n systems i n general conformance with t h e frequencies and requirements of Regulatory P o s i t i o n s C.5.a, C.5.c, C.5.d, and C.6.b of Regulatory Guide 1.52, Revision 2 , March 1978, and ANSI N510-1975.

a. Demonstrate f o r each of t h e ESF systems t h a t an i n p l a c e t e s t of t h e high e f f i c i e n c y p a r t i c u l a t e a i r (HEPA) f i l t e r s shows a p e n e t r a t i o n and system bypass < 1.0% when t e s t e d i n accordance (continued)

North Anna Units 1 and 2 5.5-8

Programs and Manuals 5.5 5.5 Prosrams and Manuals 5.5.10 V e n t i l a t i o n F i l t e r T e s t i n g Program (VFTP)

a. ( c o n t inued) w i t h R e g u l a t o r y P o s i t i o n s C.5.a and C.5.c o f R e g u l a t o r y Guide 1.52, R e v i s i o n 2, March 1978, and ANSI N510-1975 a t t h e system f l owrate s p e c i f i e d be1 ow.

ESF V e n t i 1 a t i on System F l owrate Main C o n t r o l Room/Emergency Switchgear 1000 + 10% cfm Room (MCRIESGR) Emergency Vent i1a t ion System (EVS)

Emergency Core Cool ing System (ECCS) Nominal Pump Room Exhaust A i r Cleanup System accident flow (PREACS) f o r a single t r a i n actuation Nominal a c c i d e n t f l o w f o r a s i n g l e t r a i n a c t u a t i o n i s g r e a t e r t h a n t h e minimum r e q u i r e d cool i n g f l o w f o r ECCS equipment o p e r a t i o n , and I 39,200 cfrn, which i s t h e maximum f l o w r a t e p r o v i d i n g an adequate r e s i d e n c e t i m e w i t h i n t h e charcoal adsorber.

b. Demonstrate f o r each o f t h e ESF systems t h a t an i n p l a c e t e s t o f t h e charcoal adsorber shows a p e n e t r a t i o n and system bypass

< 1.0% when t e s t e d i n accordance w i t h R e g u l a t o r y P o s i t i o n s C.5.a and C.5.d o f R e g u l a t o r y Guide 1.52, R e v i s i o n 2, March 1978, and ANSI N510-1975 a t t h e system f l o w r a t e s p e c i f i e d below.

ESF V e n t i 1a t i o n System F l owrate MCRIESGR EVS 1000 + 10% cfm ECCS PREACS Nominal a c c i d e n t f l o w f o r a single t r a i n actuation Nominal a c c i d e n t f l o w f o r a s i n g l e t r a i n a c t u a t i o n i s g r e a t e r t h a n t h e minimum r e q u i r e d c o o l i n g f l o w f o r ECCS equipment o p e r a t i o n , and 5 39,200 cfm, which i s t h e maximum f l o w r a t e p r o v i d i n g an adequate r e s i d e n c e t i m e w i t h i n t h e charcoal adsorber.

c. Demonstrate f o r each o f t h e ESF systems t h a t a l a b o r a t o r y t e s t o f a sample o f t h e charcoal adsorber, when o b t a i n e d as d e s c r i b e d i n R e g u l a t o r y P o s i t i o n C.6. b o f R e g u l a t o r y Guide 1.52, R e v i s i o n 2, March 1978, shows t h e methyl i o d i d e p e n e t r a t i o n l e s s t h a n t h e

( c o n t inued)

N o r t h Anna U n i t s 1 and 2 5.5-9

Programs and Manual s 5.5 5.5 Programs and Manuals 5.5.10 V e n t i l a t i o n F i l t e r T e s t i n g Program (VFTP)

c. (conti wed) v a l ue speci f ied be1 ow when t e s t e d i n accordance w i t h ASTM D3803-1989 a t a temperature o f 30°C (86°F) and r e 1 a t i v e humi d i t y speci f ied be1 ow.

ESF V e n t i l a t i o n System Penetration RH MCR/ESGR EVS 2.5% 7 0%

ECCS PREACS 5% 7 0%

d. Demonstrate f o r each of t h e ESF systems t h a t t h e p r e s s u r e d r o p across t h e combined HEPA f i l t e r s , t h e p r e f i l t e r s , and t h e charcoal adsorbers i s l e s s t h a n t h e v a l u e s p e c i f i e d below when t e s t e d i n accordance w i t h ANSI N510-1975 a t t h e system f l o w r a t e s p e c i f i e d below.

ESF V e n t i l a t i o n System Delta P F l owrate MCRIESGR EVS 4 i n c h e s W.G. 1000 + 10% cfm ECCS PREACS 5inchesW.G. 539,200cfm

e. Demonstrate t h a t t h e h e a t e r s f o r each o f t h e ESF systems d i s s i p a t e > t h e v a l u e s p e c i f i e d below when t e s t e d i n accordance w i t h ASME N510-1975.

ESF V e n t i l a t i o n System Wattaqe MCR/ESGR EVS 3.5 kW The p r o v i s i o n s o f SR 3.0.2 and SR 3.0.3 a r e a p p l i c a b l e t o t h e VFTP t e s t frequencies.

5.5.11 E x p l o s i v e Gas and Storage Tank R a d i o a c t i v i t y M o n i t o r i n g Program T h i s program p r o v i d e s c o n t r o l s f o r p o t e n t i a l l y e x p l o s i v e gas m i x t u r e s c o n t a i n e d i n t h e Gaseous Waste System, t h e q u a n t i t y o f r a d i o a c t i v i t y c o n t a i n e d i n gas s t o r a g e tanks, and t h e q u a n t i t y o f r a d i o a c t i v i t y c o n t a i n e d i n u n p r o t e c t e d o u t d o o r 1i q u i d s t o r a g e t a n k s .

The gaseous r a d i o a c t i v i t y q u a n t i t i e s s h a l l be determined f o l 1owing t h e methodology i n Branch Technical P o s i t i o n (BTP) ETSB 11-5,

" P o s t u l a t e d R a d i o a c t i v e Release due t o Waste Gas System Leak o r (continued)

N o r t h Anna U n i t s 1 and 2 5.5-10

Programs and Manuals 5.5 5.5 Proqrams and Manuals 5.5.11 E x p l o s i v e Gas and Storage Tank R a d i o a c t i v i t y M o n i t o r i n g Program

( c o n t i nued)

F a i l u r e " . The l i q u i d radwaste q u a n t i t i e s s h a l l be determined i n accordance w i t h Standard Review Plan, S e c t i o n 15.7.3, " P o s t u l a t e d R a d i o a c t i v e Re1 ease due t o Tank F a i 1u r e s " .

The program s h a l l i n c l u d e :

a. The 1 i m i t s f o r c o n c e n t r a t i o n s o f hydrogen and oxygen i n t h e Gaseous Waste System and a s u r v e i l l a n c e program t o ensure t h e l i m i t s a r e m a i n t a i n e d . Such l i m i t s s h a l l be a p p r o p r i a t e t o t h e system's d e s i g n c r i t e r i a ( i .e., whether o r n o t t h e system i s designed t o w i t h s t a n d a hydrogen e x p l o s i o n ) ;

b. A s u r v e i l l a n c e program t o ensure t h a t t h e q u a n t i t y o f r a d i o a c t i v i t y c o n t a i n e d i n each gas s t o r a g e t a n k i s l e s s t h a n t h e amount t h a t would r e s u l t i n a whole body exposure o f L 0.5 rem t o any i n d i v i d u a l i n an u n r e s t r i c t e d area, i n t h e event o f an u n c o n t r o l 1ed r e 1 ease o f t h e t a n k s ' c o n t e n t s ; and

c. A s u r v e i l l a n c e program t o ensure t h a t t h e q u a n t i t y o f r a d i o a c t i v i t y c o n t a i n e d i n each o f t h e f o l l o w i n g o u t d o o r t a n k s t h a t a r e n o t surrounded by 1 i n e r s , d i k e s , o r w a l l s, capable o f h o l d i n g t h e t a n k s ' c o n t e n t s and t h a t do n o t have t a n k o v e r f l o w s and s u r r o u n d i ng area d r a i n s 1 iq u i d radwaste i o n exchanger system i s l e s s t h a n t h e amount t h a t would r e s u l t i n c o n c e n t r a t i o n s g r e a t e r t h a n t h e l i m i t s o f 10 CFR 20, Appendix B, Table 2, Column 2, e x c l u d i n g t r i t i u m , a t t h e n e a r e s t p o t a b l e w a t e r s u p p l y and t h e n e a r e s t s u r f a c e w a t e r s u p p l y i n an u n r e s t r i c t e d area, i n t h e event o f an u n c o n t r o l l e d r e l e a s e o f t h e t a n k s ' c o n t e n t s :

1. R e f u e l i n g Water Storage Tank;

2. Casing C o o l i n g Storage Tank;

3. PG Water Storage Tank;

4. Boron Recovery Test Tank; and

5. Any O u t s i d e Temporary Tank.

The p r o v i s i o n s o f SR 3.0.2 and SR 3.0.3 a r e a p p l i c a b l e t o t h e E x p l o s i v e Gas and Storage Tank R a d i o a c t i v i t y M o n i t o r i n g Program s u r v e i l l a n c e frequencies.

N o r t h Anna U n i t s 1 and 2 5.5-11

Programs and Manuals 5.5 5.5 Proqrams and Manuals 5.5.12 Cliesel Fuel O i l T e s t i n g Program A d i e s e l f u e l o i l t e s t i n g program t o implement r e q u i r e d t e s t i n g o f b o t h new f u e l o i l and s t o r e d f u e l o i l s h a l l be e s t a b l i s h e d . The program s h a l l in c l ude sampl ing and t e s t i n g r e q u i rements, and acceptance c r i t e r i a , a l l i n accordance w i t h a p p l i c a b l e ASTM Standards. The purpose o f t h e program i s t o e s t a b l i s h t h e f o l l o w i n g :

a. A c c e p t a b i l i t y o f new f u e l o i l f o r use p r i o r t o a d d i t i o n t o s t o r a g e t a n k s by d e t e r m i n i n g t h a t t h e f u e l o i l has:

1. an API g r a v i t y o r an a b s o l u t e s p e c i f i c g r a v i t y w i t h i n limits,

2. a f l a s h p o i n t and k i n e m a t i c v i s c o s i t y w i t h i n l i m i t s f o r ASTM 2D f u e l o i 1, and

3. w a t e r and sediment 5 0.05%.

b. W i t h i n 3 1 days f o l l o w i n g a d d i t i o n o f t h e new f u e l o i l t o s t o r a g e t a n k s v e r i f y t h a t t h e p r o p e r t i e s o f t h e new f u e l o i l , o t h e r t h a n t h o s e addressed i n a. above, a r e w i t h i n l i m i t s f o r ASTM 2D f u e l oil;

c. Total p a r t i c u l a t e concentration o f the stored f u e l o i l i s I 10 mg/l when t e s t e d every 92 days i n accordance w i t h ASTM D-2276, Method A-2 o r A-3; and cl. The p r o v i s i o n s o f SR 3.0.2 and SR 3.0.3 a r e a p p l i c a b l e t o t h e D i e s e l Fuel O i 1 T e s t i n g Program t e s t i n g Frequencies.

5.5.13 Techni c a l Speci f ic a t i o n s (TS) Bases C o n t r o l Program T h i s program p r o v i d e s a means f o r p r o c e s s i n g changes t o t h e Bases o f t h e s e Technical S p e c i f i c a t i o n s .

a,. Changes t o t h e Bases o f t h e TS s h a l l be made under a p p r o p r i a t e a d m i n i s t r a t i v e c o n t r o l s and reviews.

b. Licensees may make changes t o Bases w i t h o u t p r i o r NRC approval p r o v i d e d t h e changes do n o t r e q u i r e e i t h e r o f t h e f o l l o w i n g :

1. a change i n t h e TS i n c o r p o r a t e d i n t h e l i c e n s e ; o r

( c o n t inued)

N o r t h Anna U n i t s 1 and 2 5.5-12

Programs and Manual s 5.5 5.5 Programs and Manuals 5.5.1.3 Technical S p e c i f i c a t i o n s (TS) Bases C o n t r o l Program ( c o n t i n u e d )

b. (continued)

2. a change t o t h e UFSAR o r Bases t h a t r e q u i r e s NRC approval p u r s u a n t t o 10 CFR 50.59.

c. The Bases C o n t r o l Program s h a l l c o n t a i n p r o v i s i o n s t o ensure t h a t t h e Bases a r e m a i n t a i n e d c o n s i s t e n t w i t h t h e UFSAR.

d. Proposed changes t h a t meet t h e c r i t e r i a o f S p e c i f i c a t i o n 5.5.13b above s h a l l be reviewed and approved by t h e NRC p r i o r t o implementation. Changes t o t h e Bases implemented w i t h o u t p r i o r NRC approval s h a l l be p r o v i d e d t o t h e NRC on a frequency c o n s i s t e n t w i t h 10 CFR 5O.7l(e).

5.5.1.4 S a f e t y F u n c t i o n D e t e r m i n a t i o n Program (SFDP)

T h i s program ensures l o s s o f s a f e t y f u n c t i o n i s d e t e c t e d and a p p r o p r i a t e a c t i o n s taken. Upon e n t r y i n t o LC0 3.0.6, an e v a l u a t i o n s h a l l be made t o determine i f 1oss o f s a f e t y f u n c t i o n e x i s t s .

A d d i t i o n a l l y , o t h e r a p p r o p r i a t e a c t i o n s may be t a k e n as a r e s u l t o f t h e s u p p o r t system i n o p e r a b i l i t y and corresponding e x c e p t i o n t o e n t e r i n g supported system C o n d i t i o n and Required A c t i o n s . T h i s program implements t h e r e q u i r e m e n t s o f LC0 3.0.6. The SFDP s h a l l contain the following:

a. P r o v i s i o n s f o r c r o s s t r a i n checks t o ensure a l o s s o f t h e c a p a b i l i t y t o p e r f o r m t h e s a f e t y f u n c t i o n assumed i n t h e a c c i d e n t a n a l y s i s does n o t go undetected;

b. Provisions f o r ensuring t h e p l a n t i s maintained i n a safe condition i f a loss o f function condition exists;

c. P r o v i s i o n s t o ensure t h a t an i n o p e r a b l e supported s y s t e m ' s Completion Time i s n o t i n a p p r o p r i a t e l y extended as a r e s u l t o f m u l t i p l e s u p p o r t system i n o p e r a b i l i t i e s ; and

d. O t h e r a p p r o p r i a t e 1 i m i t a t i o n s and remedi a1 o r compensatory actions.

A l o s s o f s a f e t y f u n c t i o n e x i s t s when, assuming no c o n c u r r e n t s i n g l e f a i l u r e , no c o n c u r r e n t l o s s o f o f f s i t e power o r l o s s o f o n s i t e d i e s e l g e n e r a t o r ( s ) , a s a f e t y f u n c t i o n assumed i n t h e a c c i d e n t (continued)

N o r t h Anna U n i t s 1 and 2 5.5-13

Programs and Manual s 5.5 5.5 Progra.ms and Manuals 5.5.14 S a f e t y F u n c t i o n Determi n a t i o n Program (SFDP) ( c o n t i n u e d )

a n a l y s i s cannot be performed. F o r t h e purpose o f t h i s program, a l o s s o f s a f e t y f u n c t i o n may e x i s t when a s u p p o r t system i s inoperabl e, and:

a. A r e q u i r e d system redundant t o t h e system(s) supported by t h e i n o p e r a b l e s u p p o r t system i s a l s o i n o p e r a b l e ; o r

b. A r e q u i r e d system redundant t o t h e system(s) i n t u r n supported by t h e i n o p e r a b l e supported system i s a l s o i n o p e r a b l e ; o r

c. A r e q u i r e d system redundant t o t h e s u p p o r t system(s) f o r t h e supported systems (a) and (b) above i s a l s o i n o p e r a b l e .

The SFDP i d e n t i f i e s where a l o s s o f s a f e t y f u n c t i o n e x i s t s . I f a l o s s o f s a f e t y f u n c t i o n i s determined t o e x i s t by t h i s program, t h e a . p p r o p r i a t e C o n d i t i o n s and Required A c t i o n s o f t h e LC0 i n which t h e l o s s o f s a f e t y f u n c t i o n e x i s t s a r e r e q u i r e d t o be entered. When a l o s s o f s a f e t y f u n c t i o n i s caused by t h e i n o p e r a b i l i t y o f a s i n g l e T e c h n i c a l S p e c i f i c a t i o n s u p p o r t system, t h e a p p r o p r i a t e C o n d i t i o n s a.nd Required A c t i o n s t o e n t e r a r e t h o s e o f t h e support system.

5.5.15 <:ontainment Leakage Rate T e s t i n g Program

a. A program s h a l l e s t a b l i s h t h e leakage r a t e t e s t i n g o f t h e containment as r e q u i r e d by 10 CFR 50.54(0) and 10 CFR 50, Appendix J, O p t i o n B, as m o d i f i e d by approved exemptions. T h i s program s h a l l be i n accordance w i t h t h e g u i d e l i n e s c o n t a i n e d i n Regul a t o r y Guide 1.163, "Performance-Based Contai nment Leak-Test Program," dated September 1995 as m o d i f i e d by t h e f o l l o w i n g e x c e p t ion :

N E I 94-01-1995, S e c t i o n 9.2.3: The f i r s t U n i t 1 Type A t e s t -

performed a f t e r t h e A p r i l 3, 1993 Type A t e s t s h a l l be performed no l a t e r t h a n A p r i l 2, 2008.

bf. The c a l c u l a t e d peak containment i n t e r n a l p r e s s u r e f o r t h e d e s i g n b a s i s l o s s o f c o o l a n t a c c i d e n t , Pa, i s 44.1 p s i g . The containment d e s i g n p r e s s u r e i s 45 p s i g .

c. The maximum a l l o w a b l e containment leakage r a t e , La, a t Pa, s h a l l be 0.1% o f containment a i r w e i g h t p e r day.

(continued)

N o r t h Anna U n i t s 1 and 2 5.5-14

Programs and Manual s 5.5 5.5 Programs and Manuals 5.5.15 Containment Leakage Rate T e s t i n g Program ( c o n t i n u e d )

d. Leakage Rate acceptance c r i t e r i a a r e :

1. P r i o r t o e n t e r i n g a MODE where containment OPERABILITY i s r e q u i r e d , t h e containment leakage r a t e acceptance c r i t e r i a are:

< 0.60 La f o r t h e Type B and Type C t e s t s on a Maximum Path B a s i s and I 0.75 La f o r Type A t e s t s .

D u r i n g o p e r a t i o n where containment OPERABILITY i s r e q u i r e d ,

t h e containment leakage r a t e acceptance c r i t e r i a a r e :

I 1.0 La f o r o v e r a l l containment leakage r a t e and < 0.60 La f o r t h e Type B and Type C t e s t s on a Minimum Path B a s i s .

2. O v e r a l l a i r l o c k l e a k a g e r a t e t e s t i n g acceptance c r i t e r i o n i s I 0.05 La when t e s t e d a t > Pa.

e. The p r o v i s i o n s o f SR 3.0.3 a r e a p p l i c a b l e t o t h e Containment Leakage Rate T e s t i n g Program.

f. N o t h i n g i n t h e s e T e c h n i c a l S p e c i f i c a t i o n s s h a l l be c o n s t r u e d t o m o d i f y t h e t e s t i n g Frequencies r e q u i r e d by 10 CFR 50, Appendix J.

N o r t h Anna U n i t s 1 and 2

Report i ng Requ i rements 5.6 5 . 6 Reporting Requirements CORE OPERAT l NG L l M l TS REPORT (COLR)

b. (cont i nued)

14. BAW-10199P-A, "The BWU C r i t i ca l Heat F l ux Correl a t i ons. "

15. BAW-10170P-A, "Stat i s t i ca l Core Des i gn f o r M i x i ng Vane Cores. "

16. EMF-2103 (P) (A), "Rea l i s t i c Large Break LOCA Methodo l ogy f o r Pressur i zed Water Reactors. "

17. EMF-96-029 (P) (A) , "Reactor Ana l ys i s System f o r PWRs . "

18. BAW-10168P-A, "RSG LOCA - BWNT Loss-of-Coolant Accident Eva l uat i on Model f o r Rec i rcu l a t i ng Steam Generator Pl ants, "

Vol ume I I on l y (SBLOCA model s) .

c. The core operating l i m i t s s h a l l be determined such t h a t a l l applicable l i m i t s (e.g., f u e l thermal mechanical l i m i t s , core therma l hydrau l i c l i m i t s , Emergency Core Cool i ng Systems (ECCS) l i m i t s , nuclear l i m i t s such as SDM, t r a n s i e n t analysis l i m i t s ,

and acc i dent ana l y s i s l i m i t s ) o f t h e safety ana l ys i s are met.

d. The COLR, including any midcycle revisions o r supplements, s h a l l be provided upon issuance f o r each re1oad cycle t o t h e NRC.

PAM Report When a r e p o r t i s required by Condition B o f LC0 3.3.3, "Post Acc i dent Mon it o r i ng (PAM) I nstrumentat i on, " a r e p o r t sha l l be submitted w i t h i n t h e f o l l o w i n g 14 days. The r e p o r t s h a l l o u t l i n e t h e cause o f t h e i n o p e r a b i l i t y , and t h e plans and schedule f o r r e s t o r i n g t h e instrumentation channels o f t h e Function t o OPERABLE s t a t u s .

Steam Generator Tube l nspect i on Report A r e p o r t s h a l l be submitted w i t h i n 180 days a f t e r t h e i n i t i a l e n t r y i n t o MODE 4 f o l l o w i n completion o f an inspection performed i n 8

accordance w i t h t h e pec i f i c a t i on 5.5.8, "Steam Generator (SG)

Program. " The r e p o r t sha l l i nc l ude :

a. The scope o f i nspect i ons performed on each SG,

b. Act i ve degradat i on mechan i sms found, North Anna U n i t s 1 and 2 5.6-4

R e p o r t i n g Requirements 5.6 5.6 R e p o r t i n g Requirements 5.6.;7 Steam Generator Tube I n s p e c t i o n Report ( c o n t i n u e d )

c. N o n d e s t r u c t i v e examination t e c h n i q u e s u t i l i z e d f o r each d e g r a d a t i o n mechanism,

d. L o c a t i o n , o r i e n t a t i o n ( i f 1 i n e a r ) , and measured s i z e s ( i f a v a i 1a b l e ) o f s e r v i c e induced i n d i c a t i o n s ,

e. Number o f tubes plugged d u r i n g t h e i n s p e c t i o n outage f o r each a c t i v e d e g r a d a t i o n mechanism,

f. T o t a l number and percentage o f tubes plugged t o date,

g. The r e s u l t s o f c o n d i t i o n m o n i t o r i n g , i n c l u d i n g t h e r e s u l t s o f t u b e p u l l s and i n - s i t u t e s t i n g , and

h. The e f f e c t i v e p l u g g i n g percentage f o r a l l p l u g g i n g i n each SG N o r t h Anna U n i t s 1 and 2

Serial No.06-403 Docket Nos. 50-3381339 Attachment 4 Mark-up of Technical Specifications Bases Changes (For Information Only)

North Anna Power Station Units 1 and 2 Virginia Electric and Power Company (Dominion)

REACTOR COOLANT SYSTEM (RCS) ( c o n t i nued)

Low Temperature Overpressure P r o t e c t i o n (LTOP) System . . . . . . . . . . . . . . . . .B 3.4.12.1 RCS Operational LEAKAGE . . . . . . . . . . . . . .B 3.4.13.1 RCS Pressure I s o l a t i o n Valve (PIV) Leakage . . . .B 3.4.14.1 RCSLeakageDetection Instrumentation . . . . . . .B3.4.1 5.1 RCS S p e c i f i c A c t i v i t y . . . . . . . . . . . . . . .B 3.4.16.1 RCS Loop I s o l a t i o n Valves . . . . . . . . . . . . .B 3.4.17.1 RCS I s o l a t e d Loop S t a r t u p . . . . . . . . . . . . .B 3.4.18.1 RCS Loops-Test Exceptions . . . . . . . . . . . . .B 3.4.19.1 EMERGENCY CORE COOLING SYSTEMS (ECCS) . . . . . . . . . B 3.5.1.1 Accumulators . . . . . . . . . . . . . . . . . . . B 3.5.1.1 ECCS-Operati ng . . . . . . . . . . . . . . . . . . B 3.5.2.1 ECCS-Shutdown . . . . . . . . . . . . . . . . . . . B 3.5.3.1 Refuel ing Water Storage Tank (RWST) . . . . . . . . B 3.5.4.1 Seal I n j e c t i o n Flow . . . . . . . . . . . . . . . . B 3.5.5.1 Boron I n j e c t i o n Tank (BIT) . . . . . . . . . . . . B 3.5.6.1 CONTAINMENT SYSTEMS . . . . . . . . . . . . . . . . . . B 3.6.1.1 Containment . . . . . . . . . . . . . . . . . . . . B 3.6.1.1 Containment A i r Locks . . . . . . . . . . . . . . . B 3.6.2.1 Containment I s o l a t i o n Valves . . . . . . . . . . . B 3.6.3.1 Containment Pressure . . . . . . . . . . . . . . . B 3.6.4.1 Containment A i r Temperature . . . . . . . . . . . . B 3.6.5.1 Quench Spray (QS) System . . . . . . . . . . . . . B 3.6.6.1 R e c i r c u l a t i o n Spray (RS) System . . . . . . . . . . B 3.6.7.1 Chemical A d d i t i o n System . . . . . . . . . . . . . B 3.6.8.1 Y

PLANT SYSTEMS ..................

Main Steam Safety Valves (MSSVs) ......

Main Steam T r i p Valves (MSTVs)

Main Feedwater Is01 a t i on Valves (MFIVs) Main . .

Feedwater Pump Discharge Valves (MFPDVs)

Main Feedwater Regul a t i n g Valves (MFRVs) and Main Feedwater Regulating Bypass Val ves (MFRBVs) .............

Steam Generator Power Operated R e l i e f Valves (SG PORVS) ...............

A u x i l i a r y Feedwater (AFW) System ......

Emergency Condensate Storage Tank (ECST) ..

Secondary Speci f i c A c t i v i t y .........

Service Water (SW) System ..........

U l t i m a t e Heat Sink (UHS) ..........

Main Control Room/Emergency S w i tchgear Room I

(MCR ESGR) Emergency Vent i1a t ion System (EVS -MODES 1. 2. 3. and 4 Mai n Control Room/Emergency Swi tchgear Room (MCRIESGR) A i r Conditioning System (ACS)

North Anna U n i t s 1 and 2 ii Revision-

RCS Loops-MODES 1 and 2 B 3.4.4 BASES APPLICABLE Both transient and steady s t a t e analyses have been performed SAFETY ANALYSES t o establish the e f f e c t of flow on the departure from (continued) nucleate boi 1ing (DNB) . The transient and accident analyses f o r the unit have been performed assuming three RCS loops a r e in operation. The majority of the unit safety analyses a r e based on i n i t i a l conditions a t high core power o r zero power.

The accident analyses that are most important t o RCP operation are the complete loss of forced reactor flow, single reactor coolant pump locked r o t o r , p a r t i a l loss of forced reactor flow, and rod withdrawal events (Ref. 1).

The DNB analyses assume normal three loop operation.

Uncertainties in key unit operating parameters, nuclear and thermal parameters, and fuel fabrication parameters a r e considered s t a t i s t i c a l l y such t h a t there i s a t l e a s t a 95 percent probability t h a t DNB will not occur f o r the 1imi ting power rod. Key unit parameter uncertainties are used t o determine the unit departure from nucleate boi 1i ng r a t i o (DNBR) uncertainty. This DNBR uncertainty, combined with the DNBR limi t , establishes a design DNBR value which must be met in unit safety analyses and i s used t o determine the pressure and temperature Safety Limit (SL). Since the parameter uncertainties a r e considered in determining the design DNBR value, the unit safety analyses are performed using values of i n p u t parameters without uncertainties. Therefore, nominal operating values f o r reactor coolant flow a r e used in the accident analyses.

The unit i s designed t o operate with a l l RCS loops in operation t o maintain DNBR above the 1imit during a l l normal operations and anticipated transients. By ensuring heat t r a n s f e r in the nucleate boiling region, adequate heat t r a n s f e r i s provided between the fuel cladding and the reactor coolant.

RCS Loops-MODES 1 and 2 s a t i s f y Criterion 2 of 10 CFR 50.36(c) (2) (i i ) .

The purpose of t h i s LC0 i s t o require an adequate forced flow r a t e f o r core heat removal. Flow i s represented by the number of RCPs in operation f o r removal of heat by the SGs. To meet safety anal ysi s acceptance c r i t e r i a f o r DNBR, three pumps a r e required a t rated power.

An OPERABLE RCS loop consists of an OPERABLE RCP in operation providing forced flow f o r heat transport and an OPERABLE SG,

+ h. Generator Survpi 11-North Anna Units 1 and 2 B 3.4.4-2 Revision-

RCS LOOPS-MODE 3 B 3.4.5 e

BASES LC0 U t i l i z a t i o n o f t h e Note i s p e r m i t t e d p r o v i d e d t h e f o l l o w i n g (cont inued) c o n d i t i o n s a r e met, along w i t h any o t h e r c o n d i t i o n s imposed by i n i t i a l s t a r t u p t e s t procedures:

a. No o p e r a t i o n s a r e p e r m i t t e d t h a t would d i l u t e t h e RCS boron c o n c e n t r a t i o n w i t h c o o l a n t a t boron c o n c e n t r a t i o n s l e s s t h a n r e q u i r e d t o ensure t h e SDM o f LC0 3.1.1, t h e r e b y m a i n t a i n i n g t h e margin t o c r i t i c a l i t y . Boron r e d u c t i o n w i t h cool a n t a t boron c o n c e n t r a t i o n s 1ess t h a n r e q u i r e d t o assure t h e SDM i s m a i n t a i n e d i s p r o h i b i t e d because a u n i form c o n c e n t r a t i o n d i s t r i b u t i on throughout t h e RCS cannot be ensured when i n n a t u r a l c i r c u l a t i o n ;

and

b. Core o u t l e t temperature i s m a i n t a i n e d a t l e a s t 10°F below s a t u r a t i o n temperature, so t h a t no vapor bubble may form and p o s s i b l y cause a n a t u r a l c i r c u l a t i o n f l o w obstruction.

An OPERABLE RCS l o o p c o n s i s t s o f one OPERABLE RCP and one OPERABLE SGj 9 which has t h e minimum w a t e r l e v e l s p e c i f i e d i n SR 3.4.5.2. An RCP i s OPERABLE i f i t i s capable o f b e i n g powered and i s a b l e t o p r o v i d e f o r c e d f l o w i f required.

APPLICABILITY I n MODE 3, t h i s LC0 ensures f o r c e d c i r c u l a t i o n o f t h e r e a c t o r c o o l a n t t o remove decay heat f r o m t h e c o r e and t o p r o v i de p r o p e r boron mi x i ng .

Operation i n o t h e r MODES i s covered by:

LC0 3.4.4, "RCS Loops-MODES 1 and 2" ;

LC0 3.4.6, "RCS LOOPS-MODE 4";

LC0 3.4.7, "RCS Loops-MODE 5, Loops F i l l e d " ;

LC0 3.4.8, "RCS Loops-MODE 5, Loops Not F i l l e d " ;

LC0 3.9.5, "Residual Heat Removal (RHR) and Coolant C i r c u l at ion-High Water Level " (MODE 6) ; and LC0 3.9.6, "Residual Heat Removal (RHR) and Cool a n t C i r c u l a t i on-Low Water Level " (MODE 6) .

N o r t h Anna U n i t s 1 and 2 3 3.4.5-3

RCS Loops-MODE 4 B 3.4.6 BASES LC0 I 280°F. T h i s r e s t r a i n t i s t o prevent a low temperature

( c o n t inued) o v e r p r e s s u r e e v e n t due t o a thermal t r a n s i e n t when an RCP i s started.

An OPERABLE RCS l o o p i s comprised o f an OPERABLE RCP and an S i m i l a r l y f o r t h e RHR System, an OPERABLE RHR l o o p i s comprised o f an OPERABLE RHR pump capable o f p r o v i d i n g f o r c e d f l o w t o an OPERABLE RHR h e a t exchanger. RCPs and RHR pumps a r e OPERABLE i f t h e y a r e capable o f beirig powered and are able t o provide forced flow i f required.

APPLICABILITY I n MODE 4, t h i s LC0 ensures f o r c e d c i r c u l a t i o n o f t h e r e a c t o r c o o l a n t t o remove decay h e a t f r o m t h e c o r e and t o p r o v i d e p r o p e r boron m i x i n g . One l o o p o f e i t h e r RCS o r RHR p r o v i d e s s u f f i c i e n t c i r c u l a t i o n f o r t h e s e purposes. However, two l o o p s c o n s i s t i n g o f any combination o f RCS and RHR l o o p s a r e r e q u i r e d t o be OPERABLE t o p r o v i d e redundancy f o r h e a t removal .

O p e r a t i o n i n o t h e r MODES i s covered by:

LC0 3.4.4, "RCS Loops-MODES 1 and 2";

LC0 3.4.5, "RCS Loops-MODE 3";

LC0 3.4.7, "RCS Loops-MODE 5, Loops F i l l e d " ;

LC0 3.4.8, "RCS Loops-MODE 5, Loops Not F i l l e d " ;

LC0 3.9.5, "Residual Heat Removal (RHR) and Coolant C i r c u l ation-High Water Level " (MODE 6) ; and LC0 3.9.6, "Residual Heat Removal (RHR) and Coolant C i r c u l ation-Low Water Level " (MODE 6).

ACTIONS I f one r e q u i r e d l o o p i s i n o p e r a b l e , redundancy f o r h e a t removal i s l o s t . A c t i o n must be i n i t i a t e d t o r e s t o r e a second RCS o r RHR l o o p t o OPERABLE s t a t u s . The immediate Completion Time r e f l e c t s t h e importance o f m a i n t a i n i n g t h e a v a i l a b i 1 it y o f two p a t h s f o r h e a t removal.

N o r t h Anna U n i t s 1 and 2 B 3.4.6-3 R e v i s i o n +Q-

RCS Loops-MODE 5, Loops F i 1l e d B 3.4.7 BASES LC0 U t i l i z a t i o n o f Note 1 i s p e r m i t t e d provided t h e f o l l o w i n g (cont inued) c o n d i t i o n s a r e met, along w i t h any o t h e r c o n d i t i o n s imposed by i n i t i a l s t a r t u p t e s t procedures:

a. No operations a r e p e r m i t t e d t h a t would d i l u t e t h e RCS boron c o n c e n t r a t i o n w i t h coolant a t boron concentrat i ons l e s s than r e q u i r e d t o meet t h e SDM o f LC0 3.1.1, t h e r e f o r e m a i n t a i n i n g t h e margin t o c r i t i c a l i t y . Boron r e d u c t i o n w i t h cool ant a t boron concentrations 1ess than r e q u i r e d t o assure t h e SDM i s maintained i s p r o h i b i t e d because a u n i f o r m concentration d i s t r i b u t i o n throughout t h e RCS cannot be ensured when i n n a t u r a l c i r c u l a t i o n ;

and

b. Core o u t l e t temperature i s maintained a t l e a s t 10°F below s a t u r a t i o n temperature, so t h a t no vapor bubble may form and p o s s i b l y cause a n a t u r a l c i r c u l a t i o n f l o w obstruction.

Note 2 a l l o w s one RHR l o o p t o be i n o p e r a b l e f o r a p e r i o d o f up t o 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , provided t h a t t h e o t h e r RHR l o o p i s OPERABLE and i n operation. This permits p e r i o d i c survei 1lance t e s t s t o be performed on t h e inoperable l o o p d u r i n g t h e o n l y time when such t e s t i n g i s safe and possible.

Note 3 r e q u i r e s t h a t t h e secondary s i d e water temperature o f each SG be I 50°F above each o f t h e RCS c o l d l e g temperatures b e f o r e t h e s t a r t o f a r e a c t o r coolant pump (RCP) w i t h an RCS c o l d l e g temperature I 280°F. This r e s t r i c t i o n i s t o prevent I a low temperature overpressure event due t o a thermal t r a n s i e n t wnen an RCP i s s t a r t e d .

Note 4 provides f o r an o r d e r l y t r a n s i t i o n from MODE 5 t o MODE 4 d u r i n g a planned heatup by permi t t i n g removal o f RHR loops from o p e r a t i o n when a t l e a s t one RCS l o o p i s i n operation. This Note provides f o r t h e t r a n s i t i o n t o MODE 4 where an RCS l o o p i s p e r m i t t e d t o be i n o p e r a t i o n and replaces t h e RCS c i r c u l a t i o n f u n c t i o n provided by t h e RHR loops w i t h c i r c u l a t i o n provided by an RCP.

RHR pumps a r e OPERABLE if they a r e capable o f being powered and a r e a b l e t o p r o v i d e f l o w i f required. Al+ fPHMtE SG can perform as a heat s i n k v i a n a t u r a l c i r c u l a t i o n when i t has an adequate water l e v e l and i s 0PERABLE.i-North Anna U n i t s 1 and 2 B 3.4.7-3 Revision .Pe?Y

RCS Operational LEAKAGE B 3.4.13 BASES APPLICABLE Except f o r primary t o secondary LEAKAGE, t h e s a f e t y analyses SAFETY ANALYSES do n o t address o p e r a t i o n a l LEAKAGE. However, o t h e r operational LEAKAGE i s r e l a t e d t o t h e s a f e t y analyses f o r LOCA; t h e amount of leakage can a f f e c t t h e p r o b a b i l i t y o f such an event. The safety a n a l y s i s f o r an event r e s u l t i n g i n secondary steam release t o t h e atmosphere, such as a steam I

generator tube r u p t u r e (SGTR) The 1eakage contaminates t h e secondary f l u i d .

The UFSAR (Ref. 3) a n a l y s i s f o r SGTR assumes t h e contaminated secondary f l u i d is released v i a power operated re1 i e f valves o r s a f e t y valves. The source term i n t h e primary system coolant i s t r a n s p o r t e d t o t h e a f f e c t e d (ruptured) steam generator by t h e break flow. The a f f e c t e d steam generator discharges steam t o t h e environment f o r 30 minutes u n t i l t h e generator i s manually i s o l a t e d . The 1 gpm primary t o secondary LEAKAGE t r a n s p o r t s t h e source term t o t h e unaffected steam generators. Releases continue throush t h e unaffected steam senerators u n t i 1 t h e Residual Heat Removal System i s p l a c e d - i n service.

The MSLB i s l e s s l i m i t i n g f o r s i t e r a d i a t i o n releases t h a n t h e SGTR. The s a f e t y a n a l y s i s f o r t h e MSLB a c c i d e n t assumes 1 gpm primary t o secondary LEAKAGE as an i n i t i a l c o n d i t i o n .

The dose consequences r e s u l t i n g from t h e MSLB and SGTR accidents a r e w i t h i n t h e l i m i t s d e f i n e d i n t h e s t a f f approved 1icensi ng basi s .

The RCS operational LEAKAGE s a t i s f i e s C r i t e r i o n 2 o f 10 CFR 50.36(c) (2) (ii) .

LC0 RCS operational LEAKAGE s h a l l be 1i m i t e d t o :

a. Pressure Boundary LEAKAGE No pressure boundary LEAKAGE is a1 1owed, being. i n d i c a t i v e o f m a t e r i a l d e t e r i o r a t i o n . LEAKAGE o f t h i s t y p e i s unacceptable as t h e l e a k i t s e l f could cause f u r t h e r d e t e r i o r a t i o n , r e s u l t i n g i n h i g h e r LEAKAGE. V i o l a t i o n o f (continued)

N o r t h Anna U n i t s 1 and 2 B 3.4.13-2 revision^

RCS Operational LEAKAGE B 3.4.13 BASES LC0 a. Pressure Boundary LEAKAGE (continued)

(continued) t h i s LC0 could r e s u l t i n continued degradation o f t h e RCPB. LEAKAGE past seals and gaskets i s n o t pressure boundary LEAKAGE.

U n i d e n t i f i e d LEAKAGE One g a l l o n per minute (gpm) o f u n i d e n t i f i e d LEAKAGE i s a1 1owed as a reasonabl e mi nimum d e t e c t a b l e amount t h a t t h e containment a i r monitoring and containment sump 1eve1 m o n i t o r i n g equipment can d e t e c t w i t h i n a reasonable t i m e period. V i o l a t i o n o f t h i s LC0 could r e s u l t i n continued degradation o f t h e RCPB, i f t h e 'LEAKAGE i s from t h e pressure boundary.

I d e n t i f i e d LEAKAGE Up t o 10 gpm o f i d e n t i f i e d LEAKAGE i s considered a1 lowable because LEAKAGE i s from known sources t h a t do n o t i n t e r f e r e w i t h d e t e c t i o n o f u n i d e n t i f i e d LEAKAGE and i s w e l l w i t h i n t h e c a p a b i l i t y o f t h e RCS Makeup System.

I d e n t i f i e d LEAKAGE includes LEAKAGE t o t h e containment from s p e c i f i c a l l y known and 1ocated sources, b u t does n o t i n c l u d e pressure boundary LEAKAGE o r c o n t r o l 1ed r e a c t o r coolant pump (RCP) seal l e a k o f f (a normal f u n c t i o n n o t considered LEAKAGE). V i 01a t i on o f t h i s LC0 could r e s u l t i n continued degradation o f a component o r system.

@ Primary t o Secondary LEAKAGE through Any One SG -

U North Anna U n i t s 1 and 2 , B 3.4.13-3 Revision %

RCS Operati onal LEAKAGE B 3.4.13 BASES APPLICABILITY I n MODES 1, 2, 3, and 4, t h e p o t e n t i a l f o r RCPB LEAKAGE i s g r e a t e s t when t h e RCS i s pressurized.

I n MODES 5 and 6, LEAKAGE l i m i t s a r e n o t r e q u i r e d because t h e r e a c t o r coolant pressure i s f a r lower, r e s u l t i n g i n 1ower stresses and reduced p o t e n t i a1 s f o r LEAKAGE.

LC0 3.4.14, "RCS Pressure I s o l a t i o n Valve ( P I V ) Leakage,"

measures 1eakage through each i n d i v i d u a l PIV and can impact t h i s LCO. O f t h e two PIVs i n s e r i e s i n each i s o l a t e d l i n e ,

1eakage measured through one PIV does n o t r e s u l t i n RCS LEAKAGE when t h e o t h e r i s l e a k t i g h t . I f both valves l e a k and r e s u l t i n a l o s s of mass from t h e RCS, t h e l o s s must be in c l uded i n t h e a1 lowabl e i d e n t i f i e d LEAKAGE.

ACTIONS -

A. 1

-ni p

U n i d e n t i f i e d LEAKAGE i d e n t i f i e d LEAKAGEi w-p&my t c excess o f t h e LC0 l i m i t s must be reduced t o w i t h i n l i m i t s w i t h i n 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . T h i s Completion Time a l l o w s t i m e t o v e r i f y leakage r a t e s and e i t h e r i d e n t i f y u n i d e n t i f i e d LEAKAGE o r reduce LEAKAGE t o w i t h i n l i m i t s b e f o r e t h e r e a c t o r must be shut down. T h i s a c t i o n i s necessary t o prevent f u r t h e r d e t e r i o B . l and B.2 F

I f any p essure boundary LEAKAGE, i d e n t i f i e d LEAKAGE, r . r n r ; n ; r r \ r r y cannot be reduced t o w i t h i n l i m i t s w i t h i n 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , t h e r e a c t o r must be brought t o lower pressure c o n d i t i o n s t o LEAW&

reduce t h e s e v e r i t y o f t h e LEAKAGE and i t s p o t e n t i a l consequences. I t should be noted t h a t LEAKAGE p a s t s e a l s and gaskets i s n o t pressure boundary LEAKAGE. The r e a c t o r must be brought t o MODE 3 w i t h i n 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 w i t h i n 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . This a c t i o n reduces t h e LEAKAGE and a1 so reduces t h e f a c t o r s t h a t tend t o degrade t h e pressure boundary.

The a1 1owed Completion Times a r e reasonabl e, based on o p e r a t i n g experience, t o reach t h e r e q u i r e d u n i t c o n d i t i o n s from f u l l power c o n d i t i o n s i n an o r d e r l y manner and w i t h o u t c h a l l e n g i n g u n i t systems. I n MODE 5, t h e pressure stresses a c t i n g on t h e RCPB a r e much lower, and f u r t h e r d e t e r i o r a t i o n i s much l e s s l i k e l y .

North Anna U n i t s 1 and 2 B 3.4.13-4 Revision-

RCS Operati onal LEAKAGE B 3.4.13 BASES SURVEILLANCE SR 3.4.13.1 REOU IREMENTS Verifying RCS LEAKAGE t o be within the LC0 l i m i t s ensures the i n t e g r i t y of the RCPB i s maintained. Pressure boundary LEAKAGE would a t f i r s t appear as unidentified LEAKAGE and can only be positively identified by inspection. I t should be noted t h a t LEAKAGE past seals and gaskets i s not pressure boundary LEAKAGE. Unidentified LEAKAGE and identified LEAKAGE are determined by performance of an RCS water inventory balance. f4wmy t c V E ' s ?; s+

The RCS water inventory balance must be met with t h e reactor a t steady s t a t e operating conditions (stab1 e temperature, pressurizer and makeup tank 1evel s , makeup RCP seal injection and return flows) .

Note#- t h a t t h i s SR i s not k e performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months a f t e r establishing operation. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance provides t o collect and process a l l necessary data a f t e r stable plant conditions are established.

Steady s t a t e operation i s required t o perform a proper inventory balance since calculations during maneuvering are not useful. For RCS operational LEAKAGE determination by water inventory balance, steady s t a t e i s defined as s t a b l e RCS pressure, temperature, power l e v e l , pressurizer and makeup tank levels, makeup and 1etdown, and RCP seal injection and return flows.

An early warning of pressure boundary LEAKAGE or ur~identifiedLEAKAGE i s provided by the automatic systems t h a t monitor the containment atmosphere radioactivity and t h e containment sump level. I t should be noted t h a t LEAKAGE past seals and gaskets i s not pressure boundary LEAKAGE.

These leakage detection systems a r e specified in LC0 3.4.15, "RCS Leakage - Detection Instrumentation."

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Frequency i s a reasonable interval t o trend LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents.

'North Anna Units 1 and 2 B 3.4.13-5 Revision p

RCS O p e r a t i o n a l LEAKAGE B 3.4.13 BASES SURVEILLANCE SR 3.4.13.2 REFERENCES 1. UFSAR, S e c t i o n 3.1.26.

2. -Regulatory Guide 1.45, May 1973.

3. UFSAR, Chapter 15.

- / 's N o r t h Anna U n i t s 1 and 2 Revision 0

INSERT B 3.4.13 A that primary to secondary LEAKAGE from all steam generators (SGs) is one gallon per minute or increases to one gallon per minute as a result of accident induced conditions. The LC0 requirement to limit primary to secondary LEAKAGE through any one SG to less than or equal to 150 gallons per day is significantly less than the conditions assumed in the safety analysis.

INSERT B 3.4.13 B The limit of 150 gallons per day per SG is based on the operational LEAKAGE performance criterion in NEI 97-06, Steam Generator Program Guidelines (Ref. 4). The Steam Generator Program operational LEAKAGE performance criterion in NEI 97-06 states, The RCS operational primary to secondary leakage through any one SG shall be limited to 150 gallons per day." The limit is based on operating experience with SG tube degradation mechanisms that result in tube leakage. The operational leakage rate criterion in conjunction with the implementation of the Steam Generator Program is an effective measure for minimizing the frequency of steam generator tube ruptures.

INSERT B 3.4.13 C Note 2 states that this SR is not applicable to primary to secondary LEAKAGE because LEAKAGE of 150 gallons per day cannot be measured accurately by an RCS water inventory balance.

INSERT B 3.4.13 D This SR verifies that primary to secondary LEAKAGE is less than or equal to 150 gallons per day through any one SG. Satisfying the primary to secondary LEAKAGE limit ensures that the operational LEAKAGE performance criterion in the Steam Generator Program is met. If this SR is not met, compliance with LC0 3.4.20, "Steam Generator Tube Integrity," should be evaluated.

The 150 gallons per day limit is measured at room temperature as described in Reference 5.

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

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

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

5).

INSERT B 3.4.13 E

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

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

SG Tube lntegr B 3.4.

B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.20 Steam Generator (SG) Tube l ntegr i t y BASES BACKGROUND Steam generator (SG) tubes are sma l l d i ameter, t h i n wa l l ed tubes t h a t carry primary coolant through the primary t o secondary heat exchangers. The SG tubes have a number o f important safety functions. SG tubes are an integral p a r t o f the reactor coolant pressure boundary (RCPB) and, as such, are r e l i e d on t o maintain the primary system's pressure and i nventory . The SG tubes isol ate the rad ioact i ve f i ss i on products i n the primary cool ant from the secondary system.

I n addition, as p a r t of the RCPB, the SG tubes are unique i n t h a t they a c t as the heat transfer surface between t h e primary and secondary systems t o remove heat from the p r imary system. Th i s Spec i f i c a t ion addresses on l y the RCPB integr it y f u n c t i on of the SG. The SG heat remova l funct i on i s addressed by LC0 3.4.4, "RCS Loops-MODES 1 and 2 , "

LC0 3.4.5, "RCS Loops-MODE 3," LC0 3.4.6, "RCS Loops-MODE 4, " and LC0 3.4.7, "RCS Loops-MODE 5, Loops Fi I led."

SG tube i n t e g r i t y means t h a t the tubes are capable o f performi ng the i r intended RCPB safety funct ion cons is t e n t with the licensing basis, including applicable regulatory requ i rements .

SG tubing i s subject t o a variety o f degradation mechani sms.

SG tubes may experience tube degradation related t o corrosion phenomena, such as wastage, p i tt i ng , i ntergranu l a r attack, and stress corrosion cracking, along w i t h other mechanically induced phenomena such as denting and waar.

These degradat i on mechan i sms can i mpa i r tube i ntegr it y i f they are not managed e f f e c t i v e l y . The SG performance c r i t e r i a are used t o manage SG tube degradation.

Spec if i cat i on 5.5.8, "Steam Generator (SG) Program, "

requires t h a t a program be established and implemented t o ensure t h a t SG tube i n t e g r i t y i s maintained. Pursuant t o Spec if icat i on 5.5.8, tube i ntegr it y i s ma i nta i ned when the SG performance c r i t e r i a are met. There are three SG performance c r i t e r ia : structura l i ntegr it y , acc ident i nduced l eakage, and operat i ona l LEAKAGE. The SG performance c r i t e r i a are described i n Specification 5.5.8. Meeting the (cont i nued)

North Anna Units 1 and 2 B 3.4.20-1

SG Tube I n t e g r i t y B 314.20 BASES BACKGROUND SG performance c r i t e r i a p r o v i des reasonable assurance o f (continued) maintaining tube i n t e g r i t y a t normal and accident conditions.

The processes used t o meet t h e SG performance c r i t e r i a a r e defined by t h e Steam Generator Program Gui del ines (Ref. 1).

APPLICABLE The steam generator tube r u p t u r e (SGTR) accident i s t h e SAFETY ANALYSES l i m i t i n g basis event f o r SG tubes and avoiding a SGTR i s t h e basis f o r t h i s S p e c i f i c a t i o n . The a n a l y s i s o f a SGTR event assumes a bounding primary t o secondary LEAKAGE r a t e o f 1 gpm, which i s conservative w i t h respect t o t h e operational LEAKAGE r a t e l i m i t s i n LC0 3.4.13, "RCS Operational LEAKAGE," p l u s t h e leakage r a t e associated w i t h a double-ended r u p t u r e o f a s i n g l e tube. The UFSAR a n a l y s i s f o r SGTR assumes t h e contaminated secondary f l u i d i s re1eased v i a power operated re1 i e f valves o r s a f e t y valves.

The source term i n t h e primary system coolant i s transported t o t h e a f f e c t e d (ruptured) steam generator by t h e break flow. The a f f e c t e d steam generator discharges steam t o t h e environment f o r 30 minutes u n t i l t h e generator i s manually i s o l a t e d . The 1 gpm primary t o secondary LEAKAGE t r a n s p o r t s t h e source term t o t h e unaffected steam generators. Releases continue through t h e unaffected steam generators u n t i l t h e Residual Heat Removal System i s placed i n service.

The a n a l y s i s f o r design b a s i s accidents and t r a n s i e n t s o t h e r than a SGTR assume t h e SG tubes r e t a i n t h e i r s t r u c t u r a l i n t e g r i t y (i.e., they are assumed n o t t o rupture.) I n these analyses, t h e steam djscharge t o t h e atmosphere i s based on t h e t o t a l primary t o secondary LEAKAGE from a1 1 SGs o f 1 g a l l o n per minute o r i s assumed t o increase t o 1 g a l l o n p e r minute as a r e s u l t o f accident induced conditions. For accidents t h a t do n o t i n v o l v e f u e l damage, t h e primary coolant a c t i v i t y l e v e l o f DOSE EQUIVALENT 1-131 i s assumed t o be equal t o t h e LC0 3.4.16, "RCS S p e c i f i c A c t i v i t y , "

l i m i t s . For accidents t h a t assume fuel damage, t h e primary coolant a c t i v i t y i s a f u n c t i o n o f t h e amount o f a c t i v i t y released from t h e damaged f u e l . The dose consequences o f these events are w i t h i n t h e 1 i m i t s o f GDC 19 (Ref. 2),

10 CFR 50.67 (Ref. 3) o r RG 1.183 (Ref. 4), as appropriate.

SG tube i n t e g r i t y s a t i s f i e s C r i t e r i o n 2 o f 10 CFR 50.36 (c) (2) (i i) .

North Anna U n i t s 1 and 2 B 3.4.20-2

SG Tube l n t e a r i t v B 3y4.26 BASES LC0 The LC0 requires t h a t SG tube i n t e g r i t y be maintained. The LC0 also requ i r e s t h a t a I I SG tubes t h a t s a t i s f y t h e repa i r c r i t e r i a be plugged i n accordance w i t h t h e Steam Generator Program.

During an SG inspection, any inspected tube t h a t s a t i s f i e s t h e Steam Generator Program r e p a i r c r i t e r i a i s removed from service by plugging. I f a tube was determined t o s a t i s f y t h e r e p a i r c r i t e r i a b u t was n o t plugged t h e tube may s t i l l have tube i ntegr i t y .

In t h e context o f t h i s Spec if i c a t i on, a SG tube i s d e f i ned as t h e e n t i r e length o f t h e tube, including t h e tube wal l between t h e tube-to-tubesheet weld a t t h e tube i n l e t and t h e tube-to-tubesheet weld a t t h e tube o u t l e t . The tube-to-tubesheet weld i s n o t considered p a r t o f t h e tube.

A SG tube has tube i n t e g r i t y when i t s a t i s f i e s t h e SG performance c r i t e r i a . The SG performance c r i t e r i a are defined i n S p e c i f i c a t i o n 5 . 5 . 8 , "Steam Generator Program,"

and describe acceptable SG tube performance. The Steam Generator Program a l so provides the eva I u a t i on process f o r determ i n i ng conformance w i t h t h e SG performance c r i t e r i a .

There are three SG performance c r i t e r i a : s t r u c t u r a I i n t e g r i t y , accident induced leakage, and operationa LEAKAGE. Fa i l u r e t o meet any one o f these c r i t e r i a S considered f a i l u r e t o meet the LCO.

The s t r u c t u r a l i n t e g r i t y performance c r i t e r i o n provi des a marg i n o f safety against tube b u r s t o r cot lapse under. norma I and acc i dent cond i t i ons, and ensures s t r u c t u r a I i ntegr i t y o f t h e SG tubes under a i I a n t i c i p a t e d t r a n s i e n t s included i n t h e design s p e c i f i c a t i o n . Tube b u r s t i s defined as, "The gross s t r u c t u r a l f a i l u r e o f t h e tube w a l l . The condition t y p i c a l l y corresponds t o an unstable opening displacement (e. g . , o en i ng area i ncreased i n response t o constant 7

pressure accompan i ed by duct i I e (p l a s t i c) t e a r i ng o f t h e tube material a t t h e ends o f t h e degradation. " Tube cot lapse i s defined as, "For t h e load displacement curve f o r a given s t r u c t u r e , cot lapse occurs a t t h e top o f t h e load versus displacement curve where t h e slope o f t h e curve becomes zero." The s t r u c t u r a l i n t e g r i t y performance c r i t e r i o n provides guidance on assessing loads t h a t have a s i g n i f icant effect on b u r s t o r col lapse. I n t h a t context, t h e term

" s i g n i f i c a n t " i s defined as "An accident loading condition other than d i f f e r e n t i a l pressure i s considered s i g n i f i c a n t (cont i nued)

North Anna U n i t s 1 and 2 B 3.4.20-3

SG Tube I n t e g r i t y B 3.4.20 BASES LC0 when t h e a d d i t i o n o f such l o a d s i n t h e assessment o f t h e (continued) s t r u c t u r a l i n t e g r i t y performance c r i t e r i o n c o u l d cause a 1ower s t r u c t u r a l 1 i m i t o r 1 i m i t i n g b u r s t / c o l l a p s e c o n d i t i o n t o be e s t a b l i s h e d . " For tube i n t e g r i t y e v a l u a t i o n s , except f o r c i r c u m f e r e n t i a l degradation, a x i a l thermal l o a d s a r e c l a s s i f i e d as secondary loads. For c i r c u m f e r e n t i a1 degradation, t h e c l a s s i f i c a t i o n o f a x i a l thermal 1oads as p r i m a r y o r secondary loads w i l l be e v a l u a t e d on a case-by-case b a s i s . The d i v i s i o n between p r i m a r y and secondary c l a s s i f i c a t i o n s w i l l be based on d e t a i 1ed a n a l y s i s and/or t e s t i n g .

S t r u c t u r a l i n t e g r i t y r e q u i r e s t h a t t h e p r i m a r y membrane s t r e s s i n t e n s i t y i n a t u b e n o t exceed t h e y i e l d s t r e n g t h f o r a l l ASME Code, S e c t i o n 111, S e r v i c e Level A (normal o p e r a t i n g c o n d i t i o n s ) and S e r v i c e Level B (upset o r abnormal c o n d i t i o n s ) t r a n s i e n t s i n c l u d e d i n t h e design s p e c i f i c a t i o n .

Thi s in c l udes s a f e t y f a c t o r s and appl icab1 e design b a s i s l o a d s based on ASME Code, S e c t i o n 111, Subsection NB (Ref. 5) and D r a f t Regulatory Guide 1.121 (Ref. 6).

The a c c i d e n t induced 1eakage performance c r i t e r i o n ensures t h a t t h e p r i m a r y t o secondary LEAKAGE caused by a design b a s i s accident, o t h e r than a SGTR, i s w i t h i n t h e a c c i d e n t a n a l y s i s assumptions. The a c c i d e n t a n a l y s i s assumes t h a t a c c i d e n t induced leakage does n o t exceed 1 gpm. The a c c i d e n t induced leakage r a t e i n c l u d e s any p r i m a r y t o secondary LEAKAGE e x i s t i n g p r i o r t o t h e a c c i d e n t i n a d d i t i o n t o p r i m a r y t o secondary LEAKAGE induced d u r i n g t h e a c c i d e n t .

The o p e r a t i o n a l LEAKAGE performance c r i t e r i o n p r o v i d e s an observable i n d i c a t i o n o f SG tube c o n d i t i o n s d u r i ng p l a n t o p e r a t i o n . The l i m i t on o p e r a t i o n a l LEAKAGE i s c o n t a i n e d i n LC0 3.4.13, "RCS Operational LEAKAGE," and 1 i m i t s p r i m a r y t o secondary LEAKAGE through any one SG t o 150 g a l l o n s p e r day.

T h i s l i m i t i s based on t h e assumption t h a t a s i n g l e c r a c k l e a k i n g t h i s amount would n o t propagate t o a SGTR under t h e s t r e s s c o n d i t i o n s o f a LOCA o r a main steam l i n e break. I f t h i s amount o f LEAKAGE i s due t o more than one crack, t h e cracks a r e v e r y small, and t h e above assumption i s conservative.

APPLICABILITY SG tube i n t e g r i t y i s challenged when t h e p r e s s u r e d i f f e r e n t i a 1 across t h e tubes i s 1 arge. Large d i f f e r e n t i a1 pressures across SG tubes can o n l y be experienced i n MODE 1, 2, 3, o r 4.

(continued)

N o r t h Anna U n i t s 1 and 2 B 3.4.20-4

SG Tube I n t e g r i t y B 3.4.20 BASES APPLl CAB1L l TY SG i n t e g r i t y l i m i t s are not provided i n MODES 5 and 6 s i nce (cont i nued) RCS conditions are f a r less challenging than i n MODES 5 and 6 than during MODES 1, 2, 3, and 4 . I n MODES 5 and 6, primary t o secondary d i f f e r e n t i a l pressure i s low, r e s u l t i n g i n lower stresses and reduced potential f o r LEAKAGE.

ACT l ONS The ACTIONS are modified by a Note c l a r i f y i n g t h a t separate Conditions entry i s permitted f o r each SG tube. This i s acceptable because the Required Actions provide appropriate compensatory a c t i ons f o r each affected SG tube. Comply i ng w i t h t h e Requ i red Act i ons may a l l ow f o r cont i nued operat i on, and subsequent affected SG tubes are governed by subsequent Cond i t i on entry and app l ic a t ion o f assoc i ated Requ i red Act i ons .

A . l and A.2 Condition A applies i f i t i s discovered t h a t one or more SG tubes examined i n an inservice inspection s a t i s f y the tube r e p a i r c r i t e r i a but were not plugged i n accordance with the Steam Generator Program as required by SR 3.4.20.2. An eva I uat i on o f SG tube i ntegr it y o f the affected tube(s) must be made. Steam generator tube i n t e g r i t y i s based on meeting the SG performance c r i t e r i a described i n the Steam Generator Program. The SG repa i r c r i t e r i a def i ne I i m i t s on SG tube degradation t h a t allow f o r flaw growth between inspections while s t i l l providing assurance t h a t the SG performance c r i t e r i a w i l l continue t o be met. I n order t o determine i f a SG tube t h a t shou I d have been plugged has tube i n t e g r i t y , an eva l uat i on must be completed t h a t demonstrates t h a t the SG performance c r i t e r i a w i l l continue t o be met u n t i l the next refuel i ng outage o r SG tube inspect ion. The tube i n t e g r i t y determination i s based on the estimated condition o f the tube a t the time the s i t u a t i o n i s discovered and the estimated growth o f the degradation p r i o r t o the next SG tube inspection. I f i t i s determined t h a t tube i n t e g r i t y i s not being maintained, Condition 6 applies.

A Completion Time of 7 days i s s u f f i c i e n t t o complete the evaluation while minimizing the r i s k o f p l a n t operation with a SG tube t h a t may not have tube i n t e g r i t y .

If the eva l uat i on determ i nes t h a t the affected tube(s) have tube i ntegr i t y , Requ i red Action A. 2 a l lows p l ant operat i on t o continue u n t i l the next r e f u e l i n g outage or SG inspection provided the inspection interval continues t o be supported (cont i nued)

North Anna Units I and 2 B 3.4.20-5

SG Tube 1 ntegr i t y B 3.4.20 BASES ACT lONS A . 1 and A . 2 (continued) by an operational assessment t h a t r e f l e c t s the affected tubes. However, the affected tube(s) must be plugged p r i o r t o entering MODE 4 f o l lowing t h e next refuel i ng outage o r SG i nspect ion. Th i s Comp l e t i on Ti me i s acceptab 1 e s i nce operation u n t i l the next inspection i s supported by the operational assessment.

B . l and B.2 I f the Requ i red Actions and assoc i ated Completion Times o f Condition A are n o t met o r if SG tube i n t e g r i t y i s not being maintained, the reactor must be brought t o MODE 3 w i t h i n 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 w i t h i n 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The allowed Completion Times are reasonable, based on operating experience, t o reach the desired p l a n t cond i t i o n s from f u l l power conditions i n an o r d e r l y manner and w i t h o u t challenging p l a n t systems.

SURVEILLAN REQU l REMEN During shutdown periods the SGs are inspected as required by t h i s SR and the Steam Generator Program. NEI 97-06, Steam Generator Program Guide 1 i nes (Ref. 1) , and i t s referenced EPRl Guide l i nes, estab l i sh the content o f the Steam Generator Program. Use o f the Steam Generator Program ensures t h a t the inspection i s appropriate and consistent with accepted industry practices.

During SG inspectims a condition monitoring assessment o f the SG tubes i s performed. The condition monitoring assessment determines the "as found" condition o f the SG tubes. The purpose o f the condition monitoring assessment i s t o ensure t h a t the SG performance c r i t e r i a have been met f o r the previous operating period.

The Steam Generator Program determines the scope o f the inspection and the methods used t o determine whether the tubes contain flaws s a t i s f y i n g the tube r e p a i r c r i t e r i a .

l nspect ion scope ( i . e . , wh i ch tubes o r areas o f tub i ng w i t h i n the SG are t o be inspected) i s a function o f e x i s t i n g and potential degradation locations. The Steam Generator Program also specifies the inspection methods t o be used t o f i n d potential degradation. Inspection methods are a North Anna Units I and 2 B 3.4.20-6

SG Tube I n t e g r i t y B 3.4.20 BASES SURVE l LLANCE SR 3.4.20.1 (continued)

REQU l REMENTS f u n c t i o n o f degradation morphology, non-destructive exam i n a t i on (NDE) techn i que capab i l i t i es, and i nspect i on locations.

The Steam Generator Program defines the Frequency o f SR 3.4.20.1. The Frequency i s determined by t h e operational assessment and other l i m i t s i n the SG examination guide l i nes (Ref. 7 ) . The Steam Generator Program uses i nformat i on on e x i s t i n g degradations and growth rates t o determine an i nspect ion Frequency t h a t prov i des reasonable assurance t h a t t h e tubing w i l l meet the SG performance c r i t e r i a a t t h e next scheduled inspection. I n a d d i t i o n , S p e c i f i c a t i o n 5 . 5 . 8 .

contains p r e s c r i p t i v e requirements concerning inspection i n t e r v a l s t o provide added assurance t h a t the SG performance c r i t e r i a w i l l be met between scheduled inspections.

Dur i ng an SG i nspect i on, any i nspected tube t h a t s a t i s f i es the Steam Generator Program r e p a i r c r i t e r i a i s removed from service by plugging. The tube r e p a i r c r i t e r i a delineated i n S p e c i f i c a t i o n 5.5.8 are intended t o ensure t h a t tubes accepted f o r cont i nued serv i ce s a t i s f y t h e SG performance c r i t e r i a w i t h allowance f o r e r r o r i n t h e f l a w s i z e measurement and f o r f u t u r e f l a w growth. I n a d d i t i o n , the tube r e p a i r c r i t e r i a , i n conjunction w i t h other elements o f t h e Steam Generator Program, ensure t h a t t h e SG performance c r i t e r i a w i I l cont i nue t o be met u n t i l the next i nspect i on o f t h e subject tube(s) . Reference I provi des gu idance f o r performing operational assessments t o v e r i f y t h a t t h e tubes rema i n i ng i n service w i l l cont i nue t o meet the SG performance c r i t e r i a .

The Frequency o f p r i o r t o entering MODE 4 following a SG inspection ensures t h a t the Surveillance has been completed and a l l tubes meeting t h e r e p a i r c r i t e r i a are plugged p r i o r t o subjecting the SG tubes t o s i g n i f i c a n t primary t o secondary pressure d i f f e r e n t i a l .

REFERENCES 1 . NE I 97-06, "Steam Generator Program Gu i del i nes. "

2 . 10 CFR 50 Appendix A, GDC 19.

3 . 10 CFR 50.67.

North Anna U n i t s 1 and 2 B 3.4.20-7

SG Tube I n t e g r i t y B 3.4.20 BASES REFERENCES 4 . RG 1.183, July 2000.

(cont i nued) 5 . ASME Bo i I e r and Pressure Vessel Code, Sect i on I I I ,

Subsect i on NB .

6. D r a f t Regulatory Guide 1.121, "Basis f o r Plugging Degraded Steam Generator Tubes," August 1976.

7 . EPRI, "Pressurized Water Reactor Steam Generator Examination Guidelines."

North Anna Units 1 and 2 B 3.4.20-8

Serial No.06-403 Docket Nos. 50-3381339 Attachment 5 Proposed Technical Specifications Bases Changes (For Information Only)

North Anna Power Station Units 1 and 2 Virginia Electric and Power Company (Dominion)

TECHNICAL SPECIFICATIONS BASES TABLE OF CONTENTS REACTOR COOLANT SYSTEM (RCS) ( c o n t i n u e d )

Low Temperature Overpressure P r o t e c t i o n (LTOP) System . . . . . . . . . . . . . . . . .B 3.4.12.1 RCS O p e r a t i o n a l LEAKAGE . . . . . . . . . . . . . .B 3.4.13.1 RCS P r e s s u r e I s o l a t i o n Valve (PIV) Leakage . . . . B 3.4.14.1 RCS Leakage D e t e c t i o n I n s t r u m e n t a t i o n . . . . . . .B 3.4.15.1 RCS S p e c i f i c A c t i v i t y . . . . . . . . . . . . . . .B 3.4.16.1 RCS Loop I s o l a t i o n Valves . . . . . . . . . . . . . B 3.4.17.1 RCS I s o l a t e d Loop S t a r t u p . . . . . . . . . . . . . B 3.4.18.1 RCS Loops-Test E x c e p t i o n s . . . . . . . . . . . . . B 3.4.19.1 Steam G e n e r a t o r (SG) Tube I n t e g r i t y . . . . . . . .B 3.4.20.1 (

EMERGENCY CORE COOLING SYSTEMS (ECCS) . . . . . . . . . B 3.5.1.1 Accumulators . . . . . . . . . . . . . . . . . . . B 3.5.1.1 ECCS-Operating . . . . . . . . . . . . . . . . . . B 3.5.2.1 ECCS-Shutdown . . . . . . . . . . . . . . . . . . . B 3.5.3.1 R e f u e l i n g Water S t o r a g e Tank (RWST) . . . . . . . . B 3.5.4.1 Seal I n j e c t i o n Flow . . . . . . . . . . . . . . . . B 3.5.5.1 Boron I n j e c t i o n Tank (BIT) . . . . . . . . . . . . B 3.5.6.1 CONTAINMENT SYSTEMS . . . . . . . . . . . . . . . . . . B 3.6.1.1 Containment . . . . . . . . . . . . . . . . . . . . B 3.6.1.1 Containment A i r Locks . . . . . . . . . . . . . . . B 3.6.2.1 Containment I s o l a t i o n Valves . . . . . . . . . . . B 3.6.3.1 Containment P r e s s u r e . . . . . . . . . . . . . . . B 3.6.4.1 Containment A i r Temperature . . . . . . . . . . . . B 3.6.5.1 Quench Spray (QS) System . . . . . . . . . . . . . B 3.6.6.1 Reci r c u l a t i on Spray (RS) System . . . . . . . . . . B 3.6.7.1 Chemical A d d i t i o n System . . . . . . . . . . . . . B 3.6.8.1 PLANT SYSTEMS . . . . . . . . . . . . . . . . . . . . . B 3.7.1.1 Main Steam S a f e t y Valves (MSSVs) . . . . . . . . . B 3.7.1.1 Main Steam T r i p Valves (MSTVs) . . . . . . . . . . B 3.7.2.1 Main Feedwater I s o l a t i o n Valves (MFIVs) , Main Feedwater Pump D i scharge Val ves (MFPDVs)

Main Feedwater R e g u l a t i n g Valves (MFRVs) and Main Feedwater R e g u l a t i n g Bypass Valves (MFRBVs) . . . . . . . . . . . . . . . . B 3.7.3.1 Steam G e n e r a t o r Power Operated R e l i e f Valves (SG PORVs) . . . . . . . . . . . . . . . . . . B 3.7.4.1 Auxi 1 i a r y Feedwater (AFW) System . . . . . . . . . B 3.7.5.1 Emergency Condensate Storage Tank (ECST) . . . . . B 3.7.6.1 Secondary S p e c i f i c A c t i v i t y . . . . . . . . . . . . B 3.7.7.1 S e r v i c e Water (SW) System ............. B 3.7.8.1 U l t i m a t e Heat S i n k (UHS) ............. B 3.7.9.1 Main C o n t r o l Room/Emergency Swi t c h g e a r Room I

(MCR ESGR) Emergency V e n t i l a t i o n System (EVS -MODES 1, 2, 3, and 4 . . . . . . .

Main C o n t r o l Room/Emergency Swi t c h g e a r Room

. . .B 3.7.10.1 (MCR/ESGR) A i r C o n d i t i o n i n g System (ACS) . . .B 3.7.11.1 N o r t h Anna U n i t s 1 and 2 ii

RCS Loops-MODES 1 and 2 B 3.4.4 BASES APPLICABLE Both t r a n s i e n t and steady s t a t e analyses have been performed SAFETY ANALYSES t o e s t a b l i s h t h e e f f e c t o f f l o w on t h e d e p a r t u r e f r o m (continued) n u c l e a t e b o i 1 ing (DNB) . The t r a n s i e n t and a c c i d e n t analyses f o r t h e u n i t have been performed assuming t h r e e RCS l o o p s a r e i n o p e r a t i o n . The m a j o r i t y o f t h e u n i t s a f e t y analyses a r e based on i n i t i a l c o n d i t i o n s a t h i g h c o r e power o r z e r o power.

The a c c i d e n t analyses t h a t a r e most i m p o r t a n t t o RCP o p e r a t i o n a r e t h e complete l o s s o f f o r c e d r e a c t o r f l o w ,

s i n g l e r e a c t o r c o o l a n t pump l o c k e d r o t o r , p a r t i a l l o s s o f f o r c e d r e a c t o r f l o w , and r o d w i t h d r a w a l events (Ref. 1).

The DNB analyses assume normal t h r e e l o o p o p e r a t i o n .

U n c e r t a i n t i e s i n key u n i t o p e r a t i n g parameters, n u c l e a r and thermal parameters, and f u e l f a b r i c a t i o n parameters a r e c o n s i d e r e d s t a t i s t i c a l l y such t h a t t h e r e i s a t l e a s t a 95 p e r c e n t p r o b a b i l i t y t h a t DNB w i l l n o t o c c u r f o r t h e l i m i t i n g power rod. Key u n i t parameter u n c e r t a i n t i e s a r e used t o determine t h e u n i t departure from nucleate b o i l i n g r a t i o (DNBR) u n c e r t a i n t y . Thi s DNBR u n c e r t a i n t y , combined w i t h t h e DNBR 1 i m i t, e s t a b l i s h e s a d e s i g n DNBR v a l u e which must be met i n u n i t s a f e t y analyses and i s used t o determine t h e p r e s s u r e and temperature S a f e t y L i m i t (SL) . Since t h e parameter u n c e r t a i n t i e s a r e c o n s i d e r e d i n d e t e r m i n i n g t h e d e s i g n DNBR value, t h e u n i t s a f e t y analyses a r e performed u s i n g v a l u e s o f i n p u t parameters w i t h o u t u n c e r t a i n t i e s . Therefore, nominal o p e r a t i n g v a l u e s f o r r e a c t o r c o o l a n t f l o w a r e used i n t h e a c c i d e n t analyses.

The u n i t i s designed t o o p e r a t e w i t h a l l RCS l o o p s i n o p e r a t i o n t o m a i n t a i n DNBR above t h e 1 i m i t d u r i n g a1 1 normal o p e r a t i o n s and a n t i c i p a t e d t r a n s i e n t s . By e n s u r i n g h e a t t r a n s f e r i n t h e n u c l e a t e b o i l i n g r e g i o n , adequate h e a t t r a n s f e r i s p r o v i d e d between t h e f u e l c l a d d i n g and t h e r e a c t o r cool a n t .

RCS Loops-MODES 1 and 2 s a t i s f y C r i t e r i o n 2 o f 10 CFR 50.36 (c) (2) ( i i) .

The purpose o f t h i s LC0 i s t o r e q u i r e an adequate f o r c e d f l o w r a t e f o r c o r e h e a t removal. Flow i s r e p r e s e n t e d by t h e number o f RCPs i n o p e r a t i o n f o r removal o f h e a t by t h e SGs. To meet s a f e t y a n a l y s i s acceptance c r i t e r i a f o r DNBR, t h r e e pumps a r e r e q u i r e d a t r a t e d power.

An OPERABLE RCS l o o p c o n s i s t s o f an OPERABLE RCP i n o p e r a t i o n p r o v i d i n g f o r c e d f l o w f o r h e a t t r a n s p o r t and an OPERABLE SG. I N o r t h Anna U n i t s 1 and 2 B 3.4.4-2

RCS Loops-MODE 3 B 3.4.5 BASES LC0 U t i l i z a t i o n o f t h e Note i s p e r m i t t e d p r o v i d e d t h e f o l l o w i n g

( c o n t i nued) c o n d i t i o n s a r e met, a l o n g w i t h any o t h e r c o n d i t i o n s imposed by i n i t i a l s t a r t u p t e s t procedures:

a. No o p e r a t i o n s a r e p e r m i t t e d t h a t would d i l u t e t h e RCS boron c o n c e n t r a t i o n w i t h c o o l a n t a t boron c o n c e n t r a t i o n s l e s s t h a n r e q u i r e d t o ensure t h e SDM o f LC0 3.1.1, t h e r e b y m a i n t a i n i n g t h e m a r g i n t o c r i t i c a l i t y . Boron r e d u c t i o n w i t h c o o l a n t a t boron c o n c e n t r a t i o n s 1ess t h a n r e q u i r e d t o assure t h e SDM i s m a i n t a i n e d i s p r o h i b i t e d because a u n i f o r m c o n c e n t r a t i o n d i s t r i b u t i o n t h r o u g h o u t t h e RCS cannot be ensured when i n n a t u r a l c i r c u l a t i o n ;

and

b. Core o u t l e t temperature i s m a i n t a i n e d a t l e a s t 10°F below s a t u r a t i o n temperature, so t h a t no vapor bubble may f o r m and p o s s i b l y cause a n a t u r a l c i r c u l a t i o n f l o w obstruction.

An OPERABLE RCS l o o p c o n s i s t s o f one OPERABLE RCP and one OPERABLE SG, which has t h e minimum w a t e r l e v e l s p e c i f i e d i n I SR 3.4.5.2. An RCP i s OPERABLE i f i t i s capable o f b e i n g powered and i s able t o provide forced flow i f required.

APPLICABILITY I n MODE 3, t h i s LC0 ensures f o r c e d c i r c u l a t i o n o f t h e r e a c t o r c o o l a n t t o remove decay h e a t f r o m t h e c o r e and t o p r o v i d e p r o p e r boron m i x i n g .

O p e r a t i o n i n o t h e r MODES i s covered by:

LC0 3.4.4, "RCS Loops-MODES 1 and 2";

LC0 3.4.6, "RCS Loops-MODE 4";

LC0 3.4.7, "RCS Loops-MODE 5, Loops F i l l e d " ;

LC0 3.4.8, "RCS Loops-MODE 5, Loops Not F i l l e d " ;

LC0 3.9.5, "Residual Heat Removal (RHR) and Coolant C i r c u l a t i o n - H i g h Water L e v e l " (MODE 6); and LC0 3.9.6, "Residual Heat Removal (RHR) and Coolant C i r c u l ation-Low Water Level " (MODE 6).

ACTIONS I f one r e q u i r e d RCS l o o p i s i n o p e r a b l e , redundancy f o r h e a t removal i s l o s t . The Required A c t i o n i s r e s t o r a t i o n o f t h e r e q u i r e d RCS l o o p t o OPERABLE s t a t u s w i t h i n t h e Completion (continued)

N o r t h Anna U n i t s 1 and 2 B 3.4.5-3

RCS Loops-MODE 3 B 3.4.5 BASES ACTIONS A.1 ( c o n t inued)

Time o f 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . T h i s t i m e a1 lowance i s a j u s t i f i e d p e r i o d t o be w i t h o u t t h e redundant, nonoperating l o o p because a s i n g l e l o o p i n o p e r a t i o n has a heat t r a n s f e r c a p a b i l i t y g r e a t e r than t h a t needed t o remove t h e decay heat produced i n t h e r e a c t o r c o r e and because o f t h e low p r o b a b i l i t y o f a f a i l u r e i n t h e remaining l o o p o c c u r r i n g d u r i n g t h i s p e r i o d .

I f r e s t o r a t i o n i s n o t p o s s i b l e w i t h i n 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , t h e u n i t must be brought t o MODE 4. I n MODE 4, t h e u n i t may be p l a c e d on t h e Residual Heat Removal System. The a d d i t i o n a l Completion Time o f 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months i s compatible w i t h r e q u i r e d o p e r a t i o n s t o a c h i e v e cool down and d e p r e s s u r i z a t i o n from t h e e x i s t i n g u n i t c o n d i t i o n s i n an o r d e r l y manner and w i t h o u t c h a l l e n g i n g u n i t systems.

C.l, C.2, and C.3 Iftwo r e q u i r e d RCS loops a r e i n o p e r a b l e o r a r e q u i r e d RCS l o o p i s n o t i n o p e r a t i o n , except as d u r i n g c o n d i t i o n s p e r m i t t e d by t h e Note i n t h e LC0 s e c t i o n , p l a c e t h e Rod C o n t r o l System i n a c o n d i t i o n i n c a p a b l e o f r o d w i t h d r a w a l (e. g., a1 1 CRDMs must be de-energi zed by opening t h e RTBs o r de-energi z i ng t h e MG s e t s ) . A1 1 o p e r a t i o n s in v o l v i ng i n t r o d u c t i o n o f c o o l a n t i n t o t h e RCS w i t h boron c o n c e n t r a t i o n l e s s t h a n r e q u i r e d t o meet t h e minimum SDM o f LC0 3.1.1 must be suspended, and a c t i o n t o r e s t o r e one o f t h e RCS loops t o OPERABLE s t a t u s and o p e r a t i o n must be i n i t i a t e d . Boron d i 1u t i o n r e q u i r e s f o r c e d c i r c u l a t i o n f o r p r o p e r m i x i n g , and opening t h e RTBs o r d e - e n e r g i z i n g t h e MG s e t s removes t h e p o s s i b i l i t y o f an i n a d v e r t e n t r o d w i t h d r a w a l . Suspending t h e i n t r o d u c t i o n o f c o o l a n t i n t o t h e RCS o f cool a n t w i t h boron c o n c e n t r a t i o n 1ess t h a n r e q u i r e d t o meet t h e minimum SDM o f LC0 3.1.1 i s r e q u i r e d t o assure c o n t i n u e d s a f e o p e r a t i o n . With c o o l a n t added w i t h o u t f o r c e d c i r c u l a t i o n , unmixed cool a n t c o u l d be i n t r o d u c e d t o t h e core, however c o o l a n t added w i t h boron c o n c e n t r a t i o n meeting t h e minimum SDM m a i n t a i n s a c c e p t a b l e margin t o s u b c r i t i c a l o p e r a t i o n s . The immediate Completion Time r e f l e c t s t h e importance o f m a i n t a i n i n g o p e r a t i o n f o r heat removal. The a c t i o n t o r e s t o r e must be c o n t i n u e d u n t i l one l o o p i s r e s t o r e d t o OPERABLE s t a t u s and o p e r a t i o n .

N o r t h Anna U n i t s 1 and 2 B 3.4.5-4

RCS Loops-MODE 4 B 3.4.6 BASES LC0 5 2 8 0 ° F . T h i s r e s t r a i n t i s t o p r e v e n t a low temperature (continued) o v e r p r e s s u r e event due t o a thermal t r a n s i e n t when an RCP i s started.

An OPERABLE RCS l o o p i s comprised o f an OPERABLE RCP and an OPERABLE SG, which has t h e minimum w a t e r l e v e l s p e c i f i e d i n SR 3.4.6.2.

I S i m i l a r l y f o r t h e RHR System, an OPERABLE RHR l o o p i s comprised o f an OPERABLE RHR pump capable o f p r o v i d i n g f o r c e d f l o w t o an OPERABLE RHR h e a t exchanger. RCPs and RHR pumps a r e OPERABLE i f t h e y a r e capable o f b e i n g powered and are able t o provide forced flow i f required.

APPLICABILITY I n MODE 4, t h i s LC0 ensures f o r c e d c i r c u l a t i o n o f t h e r e a c t o r c o o l a n t t o remove decay h e a t f r o m t h e c o r e and t o p r o v i d e p r o p e r boron m i x i n g . One l o o p o f e i t h e r RCS o r RHR p r o v i d e s s u f f i c i e n t c i r c u l a t i o n f o r t h e s e purposes. However, two l o o p s c o n s i s t i n g o f any combination o f RCS and RHR l o o p s a r e r e q u i r e d t o be OPERABLE t o p r o v i d e redundancy f o r h e a t removal.

O p e r a t i o n i n o t h e r MODES i s covered by:

LC0 3.4.4, "RCS Loops-MODES 1 and 2";

LC0 3.4.5, "RCS Loops-MODE 3";

LC0 3.4.7, "RCS Loops-MODE 5, Loops F i l l e d " ;

LC0 3.4.8, "RCS Loops-MODE 5, Loops Not F i l l e d " ;

LC0 3.9.5, "Residual Heat Removal (RHR) and Cool a n t C i r c u l ation-High Water Level " (MODE 6) ; and LC0 3.9.6, "Residual Heat Removal (RHR) and Coolant C i r c u l ation-Low Water Level " (MODE 6 ) .

ACTIONS I f one r e q u i r e d l o o p i s i n o p e r a b l e , redundancy f o r h e a t removal i s l o s t . A c t i o n must be i n i t i a t e d t o r e s t o r e a second RCS o r RHR l o o p t o OPERABLE s t a t u s . The immediate Completion Time r e f l e c t s t h e importance o f m a i n t a i n i n g t h e a v a i l a b i l i t y o f two p a t h s f o r h e a t removal.

N o r t h Anna U n i t s 1 and 2 B 3.4.6-3

RCS Loops-MODE 5, Loops F i l l e d B 3.4.7 BASES LC0 U t i l i z a t i o n o f Note 1 i s p e r m i t t e d p r o v i d e d t h e f o l l o w i n g

( c o n t inued) c o n d i t i o n s a r e met, along w i t h any o t h e r c o n d i t i o n s imposed by i n i t i a l s t a r t u p t e s t procedures:

a. No o p e r a t i o n s a r e p e r m i t t e d t h a t would d i l u t e t h e RCS boron c o n c e n t r a t i o n w i t h c o o l a n t a t boron c o n c e n t r a t i o n s l e s s t h a n r e q u i r e d t o meet t h e SDM o f LC0 3.1.1, t h e r e f o r e m a i n t a i n i n g t h e m a r g i n t o c r i t i c a l i t y . Boron r e d u c t i o n w i t h c o o l a n t a t boron c o n c e n t r a t i o n s l e s s t h a n r e q u i r e d t o assure t h e SDM i s m a i n t a i n e d i s p r o h i b i t e d because a u n i f o r m c o n c e n t r a t i o n d i s t r i b u t i o n t h r o u g h o u t t h e RCS cannot be ensured when i n n a t u r a l c i r c u l a t i o n ;

and

b. Core o u t l e t temperature i s m a i n t a i n e d a t l e a s t 10°F below s a t u r a t i o n temperature, so t h a t no vapor bubble may f o r m and p o s s i b l y cause a n a t u r a l c i r c u l a t i o n f l o w obstruction.

Note 2 a l l o w s one RHR l o o p t o be i n o p e r a b l e f o r a p e r i o d o f up t o 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , p r o v i d e d t h a t t h e o t h e r RHR l o o p i s OPERABLE and i n o p e r a t i o n . T h i s p e r m i t s p e r i o d i c s u r v e i l 1ance t e s t s t o be performed on t h e i n o p e r a b l e l o o p d u r i n g t h e o n l y t i m e when such t e s t i n g i s s a f e and p o s s i b l e .

Note 3 r e q u i r e s t h a t t h e secondary s i d e w a t e r temperature o f each SG be 5 50°F above each o f t h e RCS c o l d l e g temperatures b e f o r e t h e s t a r t o f a r e a c t o r c o o l a n t pump (RCP) w i t h an RCS c o l d l e g temperature 5 280°F. T h i s r e s t r i c t i o n i s t o p r e v e n t a low temperature o v e r p r e s s u r e e v e n t due t o a thermal t r a n s i e n t when an RCP i s s t a r t e d .

Note 4 p r o v i d e s f o r an o r d e r l y t r a n s i t i o n from MODE 5 t o MODE 4 d u r i n g a planned heatup by p e r m i t t i n g removal o f RHR l o o p s f r o m o p e r a t i o n when a t l e a s t one RCS l o o p i s i n o p e r a t i o n . T h i s Note p r o v i d e s f o r t h e t r a n s i t i o n t o MODE 4 where an RCS l o o p i s p e r m i t t e d t o be i n o p e r a t i o n and r e p l a c e s t h e RCS c i r c u l a t i o n f u n c t i o n p r o v i d e d b y t h e RHR l o o p s w i t h c i r c u l a t i o n p r o v i d e d by an RCP.

RHR pumps a r e OPERABLE i f t h e y a r e capable o f b e i n g powered and a r e a b l e t o p r o v i d e f l o w i f r e q u i r e d . A SG can p e r f o r m as I a h e a t s i n k v i a n a t u r a l c i r c u l a t i o n when i t has an adequate w a t e r l e v e l and i s OPERABLE. I N o r t h Anna U n i t s 1 and 2 B 3.4.7-3

RCS O p e r a t i o n a l LEAKAGE B 3.4.13 BASES APPLICABLE Except f o r p r i m a r y t o secondary LEAKAGE, t h e s a f e t y analyses SAFETY ANALYSES do n o t address o p e r a t i o n a l LEAKAGE. However, o t h e r o p e r a t i o n a l LEAKAGE i s r e l a t e d t o t h e s a f e t y analyses f o r LOCA; t h e amount o f leakage can a f f e c t t h e p r o b a b i l i t y o f such an event. The s a f e t y a n a l y s i s f o r an e v e n t r e s u l t i n g i n steam d i s c h a r g e t o t h e atmosphere assumes t h a t p r i m a r y t o secondary LEAKAGE f r o m a1 1 steam g e n e r a t o r s (SGs) i s one g a l l o n p e r m i n u t e o r i n c r e a s e s t o one g a l l o n p e r m i n u t e as a r e s u l t o f a c c i d e n t induced c o n d i t i o n s . The LC0 r e q u i r e m e n t t o l i m i t p r i m a r y t o secondary LEAKAGE t h r o u g h any one SG t o l e s s t h a n o r equal t o 150 g a l l o n s p e r day i s s i g n i f i c a n t l y l e s s t h a n t h e c o n d i t i o n s assumed i n t h e s a f e t y a n a l y s i s .

P r i m a r y t o secondary LEAKAGE i s a f a c t o r i n t h e dose r e l e a s e s o u t s i d e containment r e s u l t i n g f r o m a main steam l i n e break (MSLB) a c c i d e n t . O t h e r a c c i d e n t s o r t r a n s i e n t s i n v o l v e secondary steam r e l e a s e t o t h e atmosphere, such as a steam g e n e r a t o r t u b e r u p t u r e (SGTR). The leakage contaminates t h e secondary f 1u i d.

The UFSAR (Ref. 3) a n a l y s i s f o r SGTR assumes t h e contaminated secondary f l u i d i s r e l e a s e d v i a power o p e r a t e d r e 1 i e f v a l v e s o r s a f e t y v a l v e s . The source t e r m i n t h e p r i m a r y system c o o l a n t i s t r a n s p o r t e d t o t h e a f f e c t e d

( r u p t u r e d ) steam g e n e r a t o r by t h e break f l o w . The a f f e c t e d steam g e n e r a t o r d i s c h a r g e s steam t o t h e environment f o r 30 m i n u t e s u n t i l t h e g e n e r a t o r i s manually i s o l a t e d . The 1 gpm p r i m a r y t o secondary LEAKAGE t r a n s p o r t s t h e source t e r m t o t h e u n a f f e c t e d steam g e n e r a t o r s . Re1eases c o n t i n u e t h r o u g h t h e u n a f f e c t e d steam g e n e r a t o r s u n t i l t h e Residual Heat Removal System i s p l a c e d i n s e r v i c e .

The MSLB i s l e s s l i m i t i n g f o r s i t e r a d i a t i o n r e l e a s e s t h a n t h e SGTR. The s a f e t y a n a l y s i s f o r t h e MSLB a c c i d e n t assumes 1 gpm p r i m a r y t o secondary LEAKAGE as an i n i t i a l c o n d i t i o n .

The dose consequences r e s u l t i n g f r o m t h e MSLB and SGTR accidents are w i t h i n the l i m i t s defined i n the s t a f f approved l i c e n s i n g b a s i s .

The RCS o p e r a t i o n a l LEAKAGE s a t i s f i e s C r i t e r i o n 2 o f 10 CFR 50.36 (c) (2) ( i i) .

N o r t h Anna U n i t s 1 and 2 B 3.4.13-2

RCS O p e r a t i o n a l LEAKAGE B 3.4.13 BASES RCS o p e r a t i o n a l LEAKAGE s h a l l be l i m i t e d t o :

a. Pressure Boundary LEAKAGE No p r e s s u r e boundary LEAKAGE is a1 1owed, b e i n g i n d i c a t i v e o f m a t e r i a l d e t e r i o r a t i o n . LEAKAGE o f t h i s t y p e i s unacceptable as t h e l e a k i t s e l f c o u l d cause f u r t h e r d e t e r i o r a t i o n , r e s u l t i n g i n h i g h e r LEAKAGE. V i o l a t i o n o f t h i s LC0 c o u l d r e s u l t i n c o n t i n u e d d e g r a d a t i o n o f t h e RCPB. LEAKAGE p a s t s e a l s and gaskets i s n o t p r e s s u r e boundary LEAKAGE.

b. U n i d e n t i f i e d LEAKAGE One g a l l o n p e r m i n u t e (gpm) o f u n i d e n t i f i e d LEAKAGE i s a l l o w e d as a reasonable minimum d e t e c t a b l e amount t h a t t h e containment a i r m o n i t o r i n g and containment sump 1eve1 m o n i t o r i n g equipment can d e t e c t w i t h i n a reasonable t i m e p e r i o d . V i o l a t i o n o f t h i s LC0 c o u l d r e s u l t i n c o n t i n u e d d e g r a d a t i o n o f t h e RCPB, i f t h e LEAKAGE i s f r o m t h e p r e s s u r e boundary.

c. I d e n t i f ied LEAKAGE Up t o 10 gpm o f i d e n t i f i e d LEAKAGE i s c o n s i d e r e d a l l o w a b l e because LEAKAGE i s f r o m known sources t h a t do n o t i n t e r f e r e w i t h d e t e c t i o n o f u n i d e n t i f i e d LEAKAGE and i s w e l l w i t h i n t h e c a p a b i l i t y o f t h e RCS Makeup System.

I d e n t i f i e d LEAKAGE i n c l u d e s LEAKAGE t o t h e containment f r o m speci f i c a l l y known and 1ocated sources, b u t does n o t i n c l u d e p r e s s u r e boundary LEAKAGE o r c o n t r o l l e d r e a c t o r c o o l a n t pump (RCP) seal l e a k o f f (a normal f u n c t i o n n o t considered LEAKAGE). V i o l a t i o n o f t h i s LC0 c o u l d r e s u l t i n c o n t i n u e d d e g r a d a t i o n o f a component o r system.

d. P r i m a r y t o Secondary LEAKAGE t h r o u g h Any One SG The l i m i t o f 150 g a l l o n s p e r day p e r SG i s based on t h e o p e r a t i o n a l LEAKAGE performance c r i t e r i o n i n NEI 97-06, Steam Generator Program Guide1 ines (Ref. 4 ) . The Steam Generator Program o p e r a t i o n a l LEAKAGE performance c r i t e r i o n i n NEI 97-06 s t a t e s , "The RCS o p e r a t i o n a l p r i m a r y t o secondary leakage t h r o u g h any one SG s h a l l be l i m i t e d t o 150 g a l l o n s p e r day." The l i m i t i s based on o p e r a t i n g e x p e r i e n c e w i t h SG t u b e d e g r a d a t i o n mechanisms t h a t r e s u l t i n t u b e leakage. The o p e r a t i o n a l leakage (continued)

N o r t h Anna U n i t s 1 and 2 B 3.4.13-3

RCS O p e r a t i o n a l LEAKAGE B 3.4.13 BASES

d. Primary t o Secondary LEAKAGE through Any One SG

( c o n t inued) r a t e c r i t e r i o n i n c o n j u n c t i o n w i t h t h e implementation of t h e Steam Generator Program i s an e f f e c t i v e measure f o r m i n i m i z i n g t h e frequency of steam g e n e r a t o r t u b e ruptures.

APPLICABILITY I n MODES 1, 2, 3, and 4, t h e p o t e n t i a l f o r RCPB LEAKAGE i s g r e a t e s t when t h e RCS i s p r e s s u r i z e d .

I n MODES 5 and 6, LEAKAGE l i m i t s a r e n o t r e q u i r e d because t h e r e a c t o r c o o l a n t p r e s s u r e i s f a r lower, r e s u l t i n g i n l o w e r s t r e s s e s and reduced p o t e n t i a l s f o r LEAKAGE.

LC0 3.4.14, "RCS Pressure Is01 a t i o n Valve (PIV) Leakage,"

measures leakage through each i n d i v i d u a l PIV and can impact t h i s LCO. O f t h e two PIVs i n s e r i e s i n each i s o l a t e d l i n e ,

leakage measured t h r o u g h one PIV does n o t r e s u l t i n RCS LEAKAGE when t h e o t h e r i s l e a k t i g h t . I f b o t h v a l v e s l e a k and r e s u l t i n a l o s s o f mass from t h e RCS, t h e l o s s must be in c l uded i n t h e a1 1owabl e i d e n t i f i e d LEAKAGE.

ACTIONS U n i d e n t i f i e d LEAKAGE o r i d e n t i f i e d LEAKAGE i n excess o f t h e 1 LC0 l i m i t s must be reduced t o w i t h i n l i m i t s w i t h i n 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

T h i s Completion Time a1 lows t i m e t o v e r i f y leakage r a t e s and e i t h e r i d e n t i f y u n i d e n t i f i e d LEAKAGE o r reduce LEAKAGE t o w i t h i n l i m i t s b e f o r e t h e r e a c t o r must be shut down. T h i s a c t i o n i s necessary t o p r e v e n t f u r t h e r d e t e r i o r a t i o n o f t h e RCPB.

B . l and B.2 I f any p r e s s u r e boundary LEAKAGE e x i s t s , o r p r i m a r y t o secondary LEAKAGE is n o t w i t h i n 1 imi t , o r if u n i d e n t i f ied LEAKAGE, o r i d e n t i f i e d LEAKAGE, cannot be reduced t o w i t h i n l i m i t s w i t h i n 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , t h e r e a c t o r must be brought t o l o w e r p r e s s u r e c o n d i t i o n s t o reduce t h e s e v e r i t y o f t h e LEAKAGE and i t s p o t e n t i a l consequences. I t should be noted t h a t LEAKAGE p a s t s e a l s and gaskets i s n o t p r e s s u r e boundary LEAKAGE. The r e a c t o r must be brought t o MODE 3 w i t h i n

( c o n t i nued)

N o r t h Anna / I n i t s 1 and 2 B 3.4.13-4

RCS O p e r a t i onal LEAKAGE B 3.4.13 BASES ACTIONS B. 1 and 0.2 ( c o n t i n u e d )

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 w i t h i n 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . T h i s a c t i o n reduces t h e LEAKAGE and a l s o reduces t h e f a c t o r s t h a t t e n d t o degrade t h e p r e s s u r e boundary.

The a1 1owed Completion Times a r e reasonable, based on o p e r a t i n g experience, t o reach t h e r e q u i r e d u n i t c o n d i t i o n s f r o m f u l l power c o n d i t i o n s i n an o r d e r l y manner and w i t h o u t c h a l l e n g i n g u n i t systems. I n MODE 5, t h e p r e s s u r e s t r e s s e s a c t i n g on t h e RCPB a r e much lower, and f u r t h e r d e t e r i o r a t i o n i s much l e s s l i k e l y .

SURVEILLANCE SR 3.4.13.1 REQUIREMENTS V e r i f y i ng RCS LEAKAGE t o be w i t h i n t h e LC0 1 i m i t s ensures t h e i n t e g r i t y o f t h e RCPB i s m a i n t a i n e d . Pressure boundary LEAKAGE would a t f i r s t appear as u n i d e n t i f i e d LEAKAGE and can o n l y be p o s i t i v e l y i d e n t i f i e d by i n s p e c t i o n . I t should be n o t e d t h a t LEAKAGE p a s t s e a l s and gaskets i s n o t p r e s s u r e boundary LEAKAGE. U n i d e n t i f i e d LEAKAGE and i d e n t i f i e d LEAKAGE a r e determined by performance o f an RCS w a t e r i n v e n t o r y b a l ance.

The RCS w a t e r i n v e n t o r y balance must be met w i t h t h e r e a c t o r a t steady s t a t e o p e r a t i n g c o n d i t i o n s ( s t a b 1 e temperature, power 1e v e l , p r e s s u r i z e r and makeup t a n k 1e v e l s, makeup and letdown, and RCP seal i n j e c t i o n and r e t u r n f l o w s ) . The s u r v e i l l a n c e i s m o d i f i e d by two Notes. Note 1 s t a t e s t h a t t h i s SR i s n o t r e q u i r e d t o be performed u n t i l 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months a f t e r I

e s t a b l i s h i n g steady s t a t e o p e r a t i o n . The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance p r o v i d e s s u f f i c i e n t t i m e t o c o l l e c t and process a l l necessary d a t a a f t e r s t a b l e p l a n t c o n d i t i o n s a r e e s t a b l i shed.

Steady s t a t e o p e r a t i o n i s r e q u i r e d t o p e r f o r m a p r o p e r i n v e n t o r y balance s i n c e c a l c u l a t i ons d u r i n g maneuvering a r e n o t u s e f u l . F o r RCS o p e r a t i o n a l LEAKAGE d e t e r m i n a t i o n by water i n v e n t o r y balance, steady s t a t e i s d e f i n e d as s t a b l e RCS pressure, temperature, power 1e v e l , p r e s s u r i z e r and makeup t a n k l e v e l s , makeup and letdown, and RCP seal i n j e c t i o n and r e t u r n f l o w s .

An e a r l y warning o f p r e s s u r e boundary LEAKAGE o r u n i d e n t i f i e d LEAKAGE i s p r o v i d e d by t h e a u t o m a t i c systems t h a t m o n i t o r t h e containment atmosphere r a d i o a c t i v i t y and

( c o n t i nued)

N o r t h Anna U n i t s 1 and 2 B 3.4.13-5

RCS Operat i ona l LEAKAGE B 3.4.13 BASES SURVE l LLANCE SR 3.4.13.1 (continued)

REQU I REMENTS t h e conta i nment sump l eve l . I t shou l d be noted t h a t LEAKAGE past seals and gaskets i s not pressure boundary LEAKAGE.

These l eakage detect i on systems are spec i f i ed i n LC0 3.4.15, "RCS Leakage Detection Instrumentation."

Note 2 s t a t e s t h a t t h i s SR i s not applicable t o primary t o secondary LEAKAGE because LEAKAGE o f 150 gal l ons per day cannot be measured accurately by an RCS water inventory ba l ance .

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Frequency i s a reasonable i n t e r v a l t o t r e n d LEAKAGE and recognizes t h e importance o f e a r l y leakage detection i n t h e prevention o f accidents.

This SR v e r i f i e s t h a t primary t o secondary LEAKAGE i s less than o r equal t o 150 gal l ons per day through any one SG .

Sat i s f y i ng t h e p r i mary t o secondary LEAKAGE l i m i t ensures t h a t t h e operat i ona l LEAKAGE performance c r i t e r i on i n t h e Steam Generator Program i s met. I f t h i s SR i s not met, comp l i ance w i t h LC0 3.4.20, "Steam Generator Tube I ntegr i t y , " shou l d be eva l uated. The 150 gal l ons per day l i m i t i s measured a t room temperature as described i n Reference 5. The operat i ona l LEAKAGE r a t e l i m i t app l i es t o LEAKAGE through any one SG. I f i t i s not p r a c t i c a l t o assign t h e LEAKAGE t o an i nd i v i dua l SG, a l l t h e p r i mary t o secondary LEAKAGE should be conservatively assumed t o be from one SG.

The Surve i l lance i s modified by a Note, which s t a t e s t h a t t h e Surveillance i s not required t o be performed u n t i l 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months a f t e r establ i shment o f steady s t a t e operation. For RCS p r i mary t o secondary LEAKAGE determ i n a t i on, steady s t a t e i s defined as s t a b l e RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal i n j e c t i o n and r e t u r n flows.

The Surveillance Frequency o f 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months i s a reasonable i n t e r v a l t o t r e n d primary t o secondary LEAKAGE and recognizes t h e importance o f e a r l y leakage detection i n t h e prevent i on o f acci dents. The p r i mary t o secondary LEAKAGE i s determined using continuous process r a d i a t i o n monitors o r rad i ochem i ca l grab sampl i ng i n accordance w i t h t h e EPR l guidelines (Ref. 5).

North Anna U n i t s 1 and 2 B 3.4.13-6

RCS O p e r a t i o n a l LEAKAGE B 3.4.13 BASES REFERENCES 1. UFSAR, S e c t i o n 3.1.26.

2. R e g u l a t o r y Guide 1.45, May 1973.

3. UFSAR, C h a p t e r 15.

4. NEI 97-06, "Steam G e n e r a t o r Program Guide1 ines. "

5. EPRI , " P r e s s u r i z e d Water R e a c t o r P r i m a r y - t o - S e c o n d a r y Leak Gui d e l ines. "

N o r t h Anna U n i t s 1 and 2

SG Tube I n t e q r i t y B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.20 Steam Generator (SG) Tube I n t e g r i t y BASES BACKGROUND Steam g e n e r a t o r (SG) t u b e s a r e small diameter, t h i n w a l l e d tubes t h a t c a r r y p r i m a r y c o o l a n t t h r o u g h t h e p r i m a r y t o secondary heat exchangers. The SG tubes have a number o f i m p o r t a n t s a f e t y f u n c t i o n s . SG tubes a r e an i n t e g r a l p a r t o f t h e r e a c t o r c o o l a n t p r e s s u r e boundary (RCPB) and, as such, a r e r e l i e d on t o m a i n t a i n t h e p r i m a r y s y s t e m ' s p r e s s u r e and i n v e n t o r y . The SG tubes i s o l a t e t h e r a d i o a c t i v e f i s s i o n p r o d u c t s i n t h e p r i m a r y cool a n t f r o m t h e secondary system.

I n a d d i t i o n , as p a r t o f t h e RCPB, t h e SG t u b e s a r e unique i n t h a t t h e y a c t as t h e heat t r a n s f e r s u r f a c e between t h e p r i m a r y and secondary systems t o remove h e a t f r o m t h e p r i m a r y system. T h i s S p e c i f i c a t i o n addresses o n l y t h e RCPB i n t e g r i t y f u n c t i o n o f t h e SG. The SG h e a t removal f u n c t i o n i s addressed by LC0 3.4.4, "RCS Loops-MODES 1 and 2,"

LC0 3.4.5, "RCS Loops-MODE 3," LC0 3.4.6, "RCS Loops-MODE 4," and LC0 3.4.7, "RCS Loops-MODE 5, Loops Filled."

SG t u b e i n t e g r i t y means t h a t t h e tubes a r e capable o f p e r f o r m i n g t h e i r i n t e n d e d RCPB s a f e t y f u n c t i o n c o n s i s t e n t w i t h t h e 1 i c e n s i n g b a s i s , i n c l u d i n g appl i c a b l e r e g u l a t o r y r e q u i rements.

SG t u b i n g i s s u b j e c t t o a v a r i e t y o f d e g r a d a t i o n mechanisms.

SG tubes may e x p e r i e n c e t u b e d e g r a d a t i o n r e l a t e d t o c o r r o s i o n phenomena, such as wastage, p i t t i n g , i n t e r g r a n u l a r a t t a c k , and s t r e s s c o r r o s i o n c r a c k i n g , a l o n g w i t h o t h e r m e c h a n i c a l l y induced phenomena such as d e n t i n g and wear.

These d e g r a d a t i o n mechanisms can i m p a i r t u b e i n t e g r i t y i f t h e y a r e n o t managed e f f e c t i v e l y . The SG performance c r i t e r i a a r e used t o manage SG t u b e d e g r a d a t i o n .

S p e c i f i c a t i o n 5.5.8, "Steam Generator (SG) Program,"

r e q u i r e s t h a t a program be e s t a b l i s h e d and implemented t o ensure t h a t SG t u b e i n t e g r i t y i s m a i n t a i n e d . Pursuant t o S p e c i f i c a t i o n 5.5.8, t u b e i n t e g r i t y i s m a i n t a i n e d when t h e SG performance c r i t e r i a a r e met. There a r e t h r e e SG performance c r i t e r i a : s t r u c t u r a l i n t e g r i t y , a c c i d e n t induced leakage, and o p e r a t i o n a l LEAKAGE. The SG performance c r i t e r i a a r e d e s c r i b e d i n S p e c i f i c a t i o n 5.5.8. Meeting t h e

( c o n t inued)

N o r t h Anna U n i t s 1 and 2 B 3.4.20-1

SG Tube I n t e g r i t y BACKGROUND SG performance c r i t e r i a p r o v i d e s reasonable assurance o f (continued) m a i n t a i n i n g t u b e i n t e g r i t y a t normal and a c c i d e n t conditions.

The processes used t o meet t h e SG performance c r i t e r i a a r e d e f i n e d by t h e Steam Generator Program G u i d e l i n e s (Ref. 1 ) .

APPLICABLE The steam g e n e r a t o r t u b e r u p t u r e (SGTR) a c c i d e n t i s t h e SAFETY ANALYSES 1 i m i t i n g b a s i s event f o r SG tubes and a v o i d i n g a SGTR i s t h e b a s i s f o r t h i s S p e c i f i c a t i o n . The a n a l y s i s o f a SGTR e v e n t assumes a bounding p r i m a r y t o secondary LEAKAGE r a t e o f 1 gpm, which i s c o n s e r v a t i v e w i t h r e s p e c t t o t h e o p e r a t i o n a l LEAKAGE r a t e l i m i t s i n LC0 3.4.13, "RCS O p e r a t i o n a l LEAKAGE," p l u s t h e leakage r a t e a s s o c i a t e d w i t h a double-ended r u p t u r e o f a s i n g l e tube. The UFSAR a n a l y s i s f o r SGTR assumes t h e contaminated secondary f l u i d i s r e l e a s e d v i a power o p e r a t e d r e l i e f v a l v e s o r s a f e t y v a l v e s .

The source t e r m i n t h e p r i m a r y system c o o l a n t i s t r a n s p o r t e d t o t h e a f f e c t e d ( r u p t u r e d ) steam g e n e r a t o r by t h e break f l o w . The a f f e c t e d steam g e n e r a t o r d i s c h a r g e s steam t o t h e environment f o r 30 m i n u t e s u n t i l t h e g e n e r a t o r i s m a n u a l l y i s o l a t e d . The 1 gpm p r i m a r y t o secondary LEAKAGE t r a n s p o r t s t h e source t e r m t o t h e u n a f f e c t e d steam g e n e r a t o r s . Releases c o n t i n u e t h r o u g h t h e u n a f f e c t e d steam g e n e r a t o r s u n t i l t h e Residual Heat Removal System i s p l a c e d i n s e r v i c e .

The a n a l y s i s f o r d e s i g n b a s i s a c c i d e n t s and t r a n s i e n t s o t h e r t h a n a SGTR assume t h e SG tubes r e t a i n t h e i r s t r u c t u r a l i n t e g r i t y ( i .e., t h e y a r e assumed n o t t o r u p t u r e . ) I n t h e s e analyses, t h e steam d i s c h a r g e t o t h e atmosphere i s based on t h e t o t a l p r i m a r y t o secondary LEAKAGE f r o m a l l SGs o f 1 g a l l o n p e r m i n u t e o r i s assumed t o i n c r e a s e t o 1 g a l 1on p e r m i n u t e as a r e s u l t o f a c c i d e n t induced c o n d i t i o n s . F o r a c c i d e n t s t h a t do n o t i n v o l v e f u e l damage, t h e p r i m a r y c o o l a n t a c t i v i t y l e v e l o f DOSE EQUIVALENT 1-131 i s assumed t o be equal t o t h e LC0 3.4.16, "RCS S p e c i f i c A c t i v i t y , "

1 i m i t s . F o r a c c i d e n t s t h a t assume f u e l damage, t h e p r i m a r y c o o l a n t a c t i v i t y i s a f u n c t i o n o f t h e amount o f a c t i v i t y r e l e a s e d f r o m t h e damaged f u e l . The dose consequences o f t h e s e e v e n t s a r e w i t h i n t h e l i m i t s o f GDC 19 (Ref. 2),

10 CFR 50.67 (Ref. 3) o r RG 1.183 (Ref. 4), as a p p r o p r i a t e .

SG t u b e i n t e g r i t y s a t i s f i e s C r i t e r i o n 2 o f 10 CFR 50.36 (c) (2) (ii) .

N o r t h Anna U n i t s 1 and 2 B 3.4.20-2

SG Tube Inteqrity BASES The LC0 requires t h a t SG tube i n t e g r i t y be maintained. The LC0 also requires t h a t a l l SG tubes t h a t s a t i s f y the repair c r i t e r i a be plugged in accordance with the Steam Generator Program.

During an SG inspection, any inspected tube t h a t s a t i s f i e s the Steam Generator Program repair c r i t e r i a i s removed from service by plugging. If a tube was determined t o s a t i s f y the repair c r i t e r i a b u t was not plugged the tube may s t i l l have tube i n t e g r i t y .

I n the context of t h i s Specification, a SG tube i s defined as the e n t i r e length of the tube, including the tube wall between the tube-to-tubesheet weld a t the tube i n l e t and the tube-to-tubesheet weld a t the tube o u t l e t . The tube-to-tubesheet weld i s not considered part of the tube.

A SG tube has tube i n t e g r i t y when i t s a t i s f i e s the SG performance c r i t e r i a . The SG performance c r i t e r i a are defined in Specification 5.5.8, "Steam Generator Program,"

and describe acceptable SG tube performance. The Steam Generator Program also provides the evaluation process f o r determining conformance with the SG performance c r i t e r i a .

There are three SG performance c r i t e r i a : structural i n t e g r i t y , accident induced 1 eakage, and operational LEAKAGE. Failure t o meet any one of these c r i t e r i a i s considered f a i l u r e t o meet the LCO.

The structural i n t e g r i t y performance c r i t e r i o n provides a margin of safety against tube burst o r collapse under normal and accident conditions, and ensures structural i n t e g r i t y of the SG tubes under a l l anticipated t r a n s i e n t s included in the design specification. Tube burst i s defined as, "The gross structural f a i 1 ure of the tube wall. The condition typically corresponds t o an unstable opening displacement (e.g . , opening area increased in response t o constant pressure) accompanied by d u c t i l e ( p l a s t i c ) tearing of the tube material a t the ends of the degradation." Tube col lapse i s defined a s , "For the load displ acernent curve f o r a given s t r u c t u r e , collapse occurs a t the top of the load versus displacement curve where the slope of the curve becomes zero. " The structural i n t e g r i t y performance c r i t e r i o n provides guidance on assessing loads t h a t have a s i g n i f i c a n t e f f e c t on burst o r collapse. In t h a t context, the term

" s i g n i f i c a n t " i s defined as "An accident loading condition other than d i f f e r e n t i a l pressure i s considered s i g n i f i c a n t (continued)

North Anna Units 1 and 2 B 3.4.20-3

SG Tube I n t e g r i t y B 3.4.20 BASES L CO when t h e a d d i t i o n of such loads i n t h e assessment of t h e (continued) s t r u c t u r a l i n t e g r i t y performance c r i t e r i o n could cause a l ower s t r u c t u r a l l imi t o r 1 imi t i ng burst/col 1apse condition t o be e s t a b l i s h e d . " For tube i n t e g r i t y e v a l u a t i o n s , except f o r ci rcumferenti a1 degradation, axi a1 thermal 1oads a r e cl a s s i f ied a s secondary 1oads. For ci rcumferenti a1 degradation, t h e c l a s s i f i c a t i o n of a x i a l thermal loads a s primary o r secondary loads will be evaluated on a case-by-case b a s i s . The d i v i s i o n between primary and secondary c l a s s i f i c a t i o n s wi 11 be based on d e t a i l e d a n a l y s i s and/or t e s t i ng .

S t r u c t u r a l i n t e g r i t y r e q u i r e s t h a t t h e primary membrane s t r e s s i n t e n s i t y i n a tube not exceed t h e y i e l d s t r e n g t h f o r a1 1 ASME Code, Section 111, Service Level A (normal operating c o n d i t i o n s ) and Service Level B (upset o r abnormal conditions) t r a n s i e n t s included i n t h e design s p e c i f i c a t i o n .

Thi s i ncl udes s a f e t y f a c t o r s and appl i cab1 e design basi s loads based on ASME Code, Section 111, Subsection NB (Ref. 5) and Draft Regulatory Guide 1.121 (Ref. 6 ) .

The accident induced leakage performance c r i t e r i o n ensures t h a t t h e primary t o secondary LEAKAGE caused by a design b a s i s a c c i d e n t , o t h e r than a SGTR, i s within t h e accident a n a l y s i s assumptions. The accident a n a l y s i s assumes t h a t accident induced leakage does not exceed 1 gpm. The accident induced leakage r a t e includes any primary t o secondary LEAKAGE e x i s t i n g p r i o r t o t h e accident i n a d d i t i o n t o primary t o secondary LEAKAGE induced during t h e accident.

The operational LEAKAGE performance c r i t e r i o n provides an observable i n d i c a t i o n of SG tube conditions during p l a n t operation. The l i m i t on operational LEAKAGE i s contained i n LC0 3.4.13, "RCS Operational LEAKAGE," and l i m i t s primary t o secondary LEAKAGE through any one SG t o 150 g a l l o n s per day.

This l i m i t i s based on t h e assumption t h a t a s i n g l e crack leaking t h i s amount would not propagate t o a SGTR under t h e s t r e s s conditions of a LOCA o r a main steam l i n e break. I f t h i s amount of LEAKAGE i s due t o more than one crack, t h e cracks a r e very small, and t h e above assumption i s conservative.

APPLICABILITY SG tube i n t e g r i t y i s challenged when t h e p r e s s u r e d i f f e r e n t i a l a c r o s s t h e tubes i s l a r g e . Large d i f f e r e n t i a l pressures a c r o s s SG tubes can only be experienced i n MODE 1, 2, 3, o r 4.

(conti nued)

North Anna Units 1 and 2 B 3.4.20-4

SG Tube I n t e g r i t y BASES APPLICABILITY SG i n t e g r i t y l i m i t s a r e n o t p r o v i d e d i n MODES 5 and 6 s i n c e (continued) RCS c o n d i t i o n s a r e f a r l e s s c h a l l e n g i n g t h a n i n MODES 5 and 6 t h a n d u r i n g MODES 1, 2, 3, and 4. I n MODES 5 and 6, p r i m a r y t o secondary d i f f e r e n t i a l p r e s s u r e i s low, r e s u l t i n g i n l o w e r s t r e s s e s and reduced p o t e n t i a l f o r LEAKAGE.

ACTIONS The ACTIONS a r e m o d i f i e d by a Note c l a r i f y i n g t h a t s e p a r a t e C o n d i t i o n s e n t r y i s p e r m i t t e d f o r each SG tube. T h i s i s a c c e p t a b l e because t h e Required A c t i o n s p r o v i d e a p p r o p r i a t e compensatory a c t i o n s f o r each a f f e c t e d SG tube. Complying w i t h t h e Required A c t i o n s may a l l o w f o r c o n t i n u e d o p e r a t i o n ,

and subsequent a f f e c t e d SG t u b e s a r e governed by subsequent C o n d i t i o n e n t r y and a p p l i c a t i o n o f a s s o c i a t e d Required Actions.

A . l and A.2 C o n d i t i o n A a p p l i e s i f i t i s d i s c o v e r e d t h a t one o r more SG t u b e s examined i n an i n s e r v i c e i n s p e c t i o n s a t i s f y t h e t u b e r e p a i r c r i t e r i a b u t were n o t plugged i n accordance w i t h t h e Steam Generator Program as r e q u i r e d by SR 3.4.20.2. An e v a l u a t i o n o f SG t u b e i n t e g r i t y o f t h e a f f e c t e d t u b e ( s ) must be made. Steam g e n e r a t o r t u b e i n t e g r i t y i s based on m e e t i n g t h e SG performance c r i t e r i a d e s c r i b e d i n t h e Steam Generator Program. The SG r e p a i r c r i t e r i a d e f i n e l i m i t s on SG t u b e d e g r a d a t i o n t h a t a l l o w f o r f l a w g r o w t h between i n s p e c t i o n s w h i l e s t i l l p r o v i d i n g assurance t h a t t h e SG performance c r i t e r i a w i l l c o n t i n u e t o be met. I n o r d e r t o determine i f a SG t u b e t h a t should have been plugged has t u b e i n t e g r i t y , an e v a l u a t i o n must be completed t h a t demonstrates t h a t t h e SG performance c r i t e r i a w i l l c o n t i n u e t o be met u n t i l t h e n e x t r e f u e l i n g outage o r SG t u b e i n s p e c t i o n . The t u b e i n t e g r i t y d e t e r m i n a t i o n i s based on t h e e s t i m a t e d c o n d i t i o n o f t h e t u b e a t t h e t i m e t h e s i t u a t i o n i s d i s c o v e r e d and t h e e s t i m a t e d growth o f t h e d e g r a d a t i o n p r i o r t o t h e n e x t SG t u b e i n s p e c t i o n . I f i t i s determined t h a t t u b e i n t e g r i t y i s n o t b e i n g maintained, C o n d i t i o n B a p p l i e s .

A Completion Time o f 7 days i s s u f f i c i e n t t o complete t h e evaluation while minimizing the r i s k o f plant operation with a SG t u b e t h a t may n o t have t u b e i n t e g r i t y .

I f t h e e v a l u a t i o n determines t h a t t h e a f f e c t e d t u b e ( s ) have t u b e i n t e g r i t y , Required A c t i o n A.2 a l l o w s p l a n t o p e r a t i o n t o c o n t i n u e u n t i l t h e n e x t r e f u e l i n g outage o r SG i n s p e c t i o n p r o v i d e d t h e i n s p e c t i o n i n t e r v a l c o n t i n u e s t o be supported

( c o n t i nued)

N o r t h Anna U n i t s 1 and 2 B 3.4.20-5

SG Tube I n t e g r i t y B 3.4.20 BASES ACTIONS A. 1 and A.2 (continued) by an o p e r a t i o n a l assessment t h a t r e f l e c t s t h e a f f e c t e d tubes. However, t h e a f f e c t e d t u b e ( s ) must be plugged p r i o r t o e n t e r i n g MODE 4 f o l l o w i n g t h e n e x t r e f u e l i n g outage o r SG i n s p e c t i o n . T h i s Completion Time i s a c c e p t a b l e s i n c e o p e r a t i o n u n t i l t h e n e x t i n s p e c t i o n i s supported by t h e o p e r a t i o n a l assessment.

B . l and B.2 I f t h e Required A c t i o n s and a s s o c i a t e d Completion Times o f C o n d i t i o n A a r e n o t met o r i f SG t u b e i n t e g r i t y i s n o t b e i n g maintained, t h e r e a c t o r must be b r o u g h t t o MODE 3 w i t h i n 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 w i t h i n 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The a1 1 owed Completion Times a r e reasonable, based on o p e r a t i n g experience, t o reach t h e d e s i r e d p l a n t c o n d i t i o n s f r o m f u l l power c o n d i t i o n s i n an o r d e r l y manner and w i t h o u t chal 1engi ng p l a n t systems.

SURVEILLANCE SR 3.4.20.1 REQUIREMENTS D u r i n g shutdown p e r i o d s t h e SGs a r e i n s p e c t e d as r e q u i r e d by t h i s SR and t h e Steam Generator Program. N E I 97-06, Steam Generator Program Gui d e l ines (Ref. 1) , and i t s r e f e r e n c e d EPRI G u i d e l i n e s . e s t a b l i s h t h e c o n t e n t o f t h e Steam Generator program. Use o f t h e Steam Generator Program ensures t h a t t h e i n s p e c t i o n i s a p p r o p r i a t e and c o n s i s t e n t w i t h accepted i n d u s t r y p r a c t i c e s .

During SG i n s p e c t i o n s a c o n d i t i o n m o n i t o r i n g assessment o f t h e SG tubes i s performed. The c o n d i t i o n m o n i t o r i n g assessment determines t h e "as found" c o n d i t i o n o f t h e SG tubes. The purpose o f t h e c o n d i t i o n moni t o r i n g assessment i s t o ensure t h a t t h e SG performance c r i t e r i a have been met f o r the previous operating period.

The Steam Generator Program determines t h e scope o f t h e i n s p e c t i o n and t h e methods used t o determine whether t h e tubes c o n t a i n f l a w s s a t i s f y i n g t h e t u b e r e p a i r c r i t e r i a .

I n s p e c t i o n scope ( i .e., which tubes o r areas o f t u b i n g w i t h i n t h e SG a r e t o be inspected) i s a f u n c t i o n o f e x i s t i n g and p o t e n t i a l d e g r a d a t i o n l o c a t i o n s . The Steam Generator Program a l s o s p e c i f i e s t h e i n s p e c t i o n methods t o be used t o f i n d p o t e n t i a l degradation. I n s p e c t i o n methods a r e a N o r t h Anna U n i t s 1 and 2 B 3.4.20-6

SG Tube I n t e g r i t y B 314.20 BASES SURVEILLANCE SR 3.4.20.1 (continued)

REQUIREMENTS f u n c t i o n o f d e g r a d a t i o n morphology, n o n - d e s t r u c t i v e examination (NDE) t e c h n i q u e capabi 1 i t ies, and i n s p e c t i o n locations.

The Steam Generator Program d e f i n e s t h e Frequency o f SR 3.4.20.1. The Frequency i s determined by t h e o p e r a t i o n a l assessment and o t h e r 1 i m i t s i n t h e SG e x a m i n a t i o n guide1 i n e s (Ref. 7 ) . The Steam Generator Program uses i n f o r m a t i o n on e x i s t i n g d e g r a d a t i o n s and growth r a t e s t o determine an i n s p e c t i o n Frequency t h a t p r o v i d e s reasonable assurance t h a t t h e t u b i n g w i l l meet t h e SG performance c r i t e r i a a t t h e n e x t scheduled i n s p e c t i o n . I n a d d i t i o n , S p e c i f i c a t i o n 5.5.8 contains p r e s c r i p t i v e requirements concerning i n s p e c t i o n i n t e r v a l s t o p r o v i d e added assurance t h a t t h e SG performance c r i t e r i a w i l l be met between scheduled i n s p e c t i o n s .

D u r i n g an SG i n s p e c t i o n , any i n s p e c t e d t u b e t h a t s a t i s f i e s t h e Steam Generator Program r e p a i r c r i t e r i a i s removed f r o m s e r v i c e by p l u g g i n g . The t u b e r e p a i r c r i t e r i a d e l i n e a t e d i n S p e c i f i c a t i o n 5.5.8 a r e i n t e n d e d t o ensure t h a t tubes accepted f o r c o n t i nued s e r v i ce s a t i s f y t h e SG performance c r i t e r i a w i t h allowance f o r e r r o r i n t h e f l a w s i z e measurement and f o r f u t u r e f l a w growth. I n a d d i t i o n , t h e t u b e r e p a i r c r i t e r i a , i n c o n j u n c t i o n w i t h o t h e r elements o f t h e Steam Generator Program, ensure t h a t t h e SG performance c r i t e r i a w i l l c o n t i n u e t o be met u n t i l t h e n e x t i n s p e c t i o n o f t h e s u b j e c t t u b e ( s ) . Reference 1 p r o v i d e s guidance f o r p e r f o r m i n g o p e r a t i o n a l assessments t o v e r i f y t h a t t h e tubes r e m a i n i n g i n s e r v i c e w i l l c o n t i n u e t o meet t h e SG performance c r i t e r i a .

The Frequency o f p r i o r t o e n t e r i n g MODE 4 f o l l o w i n g a SG i n s p e c t i o n ensures t h a t t h e S u r v e i l l a n c e has been completed and a l l tubes m e e t i n g t h e r e p a i r c r i t e r i a a r e plugged p r i o r t o s u b j e c t i n g t h e SG tubes t o s i g n i f i c a n t p r i m a r y t o secondary p r e s s u r e d i f f e r e n t i a l .

REFERENCES 1. NEI 97-06, "Steam Generator Program G u i d e l i n e s . "

2. 10 CFR 50 Appendix A, GDC 19.

3. 10 CFR 50.67.

N o r t h Anna U n i t s 1 and 2 B 3.4.20-7

SG Tube I n t e g r i t y B 3.4.20 BASES REFERENCES 4. RG 1.183, J u l y 2000.

(continued)

5. ASME B o i l e r and Pressure Vessel Code, S e c t i o n 111, Subsection NB.

6. D r a f t R e g u l a t o r y Guide 1.121, "Basis f o r P l u g g i n g Degraded Steam Generator Tubes," August 1976.

7. EPRI, " P r e s s u r i z e d Water Reactor Steam Generator Examination G u i d e l i n e s . "

N o r t h Anna U n i t s 1 and 2

v t e Letter Sequence Request
TAC:MD1838, Steam Generator Tube Integrity (Approved, Closed)
Initiation Request Acceptance Administration Questions Answers Results Approval Other: ML062900458
MONTHYEAR2006-12-28 [Table View]

ML061430062

Text

1.0 INTRODUCTION

2.0 DESCRIPTION

3.0 BACKGROUND

5.0 TECHNICAL ANALYSIS

6.0 REGULATORY ANALYSIS

10.0 REFERENCES

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