Text

Improved Technical Specification Conversion License Amendment Request, Revision 4
	ML063000088
Person / Time
Site:	Beaver Valley
Issue date:	10/24/2006
From:	; FirstEnergy Nuclear Operating Co
To:	; Office of Nuclear Reactor Regulation
References
	L-06-149
	Download: ML063000088 (217)
	v • d • e

BEAVER VALLEY POWER STATION UNITS 1& 2 IMPROVED TECHNICAL SPECIFICATION CONVERSION LICENSE AMENDMENT REQUEST REVISION 4 CHANGES Affected Pages Organized by Change Number The Revision 4 Pages In This Volume Are Organized By Individual Change Number With All Affected Pages For Each Change Grouped Together To Facilitate The Review Of Each Change. The Enclosed Revision 4 Pages May Also Be Used To Replace The Affected Pages In The Original BVPS 10 Volume ITS Conversion Submittal.

FENOC FirstEnergy Nuclear Operating Company Richard G. Mende 724-682-7773 Director, Site Operations October 24, 2006 L-06-149 U. S. Nuclear Regulatory Commission Attention:

Document Control Desk Washington, DC 20555-0001

Subject:

Beaver Valley Power Station, Unit Nos. 1 and 2 BV-1 Docket No. 50-334, License No. DPR-66 BV-2 Docket No. 50-412, License No. NPF-73 Supplement to License Amendment Request Nos. 296 and 169, Improved Standard Technical Specification Conversion This letter provides updated pages (Revision

4) to the FirstEnergy Nuclear Operating Company (FENOC) License Amendment Request (LAR) Nos. 296 and 169 to convert the Beaver Valley Power Station (BVPS) Unit Nos. 1 and 2 Technical Specifications to the Improved Technical Specifications (ITS) for Westinghouse Plants, NUREG- 1431.The BVPS ITS conversion LAR was originally submitted by FENOC letter L-05-027 dated February 25, 2005.The purpose of this supplement is to update the BVPS ITS conversion documentation contained in LAR Nos. 296 and 169 (ITS conversion) to incorporate the following:

Approved License Amendments,* Resolution of NRC comments, and* Other changes identified during the NRC review process.This submittal also includes hardcopy pages of the joint NRC-BVPS conversion website which is maintained (electronically) by the Excel Services Corporation.

The attached hardcopy pages contain NRC questions and associated BVPS responses regarding the BVPS ITS conversion LAR. The hardcopy pages of the website are current as of September 2006 and show all the website question and response entries as closed.As part of the ITS conversion License Amendment, FENOC requests that two new License Conditions be added to the BVPS Unit 1 Operating License and to the BVPS Unit 2 Operating License. The two License Conditions requested by FENOC are consistent with the Licensing Conditions approved by the NRC for other utilities that have recently converted to the ITS. The first License Condition addresses the Beaver Valley Power Station, Unit Nos. 1 and 2 Supplement to License Amendment Request Nos. 296 and 169, Improved Standard Technical Specification Conversion L-06-149 Page 2 performance requirements for new or revised Technical Specification Surveillances.

The second License Condition addresses the disposition of Technical Specification requirements relocated to other documents as part of the ITS conversion.

The following License Conditions are requested:

1. Schedule for New and Revised Surveillance Requirements (SRs)The schedule for performing SRs that are new or revised in Amendment No. XXX shall be as follows: For SRs that are new in this amendment, the first performance is due at the end of the first surveillance interval, which begins on the date of implementation of this amendment.

For SRs that existed prior to this amendment, whose intervals of performance are being reduced, the first reduced surveillance interval begins upon completion of the first surveillance performed after implementation of this amendment.

For SRs that existed prior to this amendment, whose intervals of performance are being extended, the first extended surveillance interval begins upon completion of the last surveillance performed prior to implementation of this amendment.

For SRs that existed prior to this amendment that have modified acceptance criteria, the first performance subject to the modified acceptance criteria is due at the end of the first surveillance interval that began on the date the surveillance was last performed prior to the implementation of this amendment.

2. Relocation of Certain Technical Specification Requirements License Amendment No. XXX authorizes the relocation of certain Technical Specifications to other licensee-controlled documents.

Implementation of this amendment shall include relocation of the requirements to the specified documents, as described in (1)Section XX of the NRC staffs Safety Evaluation, and (2) Table LA, Removed Detail Changes, and Table R, Relocated Specifications, attached to the NRC staffs Safety Evaluation, which is enclosed in this amendment.

Beaver Valley Power Station, Unit Nos. 1 and 2 Supplement to License Amendment Request Nos. 296 and 169, Improved Standard Technical Specification Conversion L-06-149 Page 3 In addition to the request for new License Conditions, FENOC requests that Unit 1 License Condition C. (9), "Steam Generator Surveillance Interval Extension" and Unit 2 License Condition (C. 12), "Steam Generator Surveillance Interval Extension" be deleted.These existing License Conditions provided a limited surveillance extension for the Unit 1 and Unit 2 Steam Generators based on specific dates which have expired.Therefore, Unit 1 License Condition C. (9) and Unit 2 License Condition (C.12) are no longer applicable.

Attachment 1 of this supplement contains the revised pages organized by individual changes, such that all the affected pages for each change are grouped together by a unique change number. The purpose of Attachment 1 is to facilitate the review of each change by providing all the affected pages for that change in one place. In addition, the revised pages included in Attachment 1 may be used to update the affected pages in the original 10 volume BVPS ITS conversion submittal.

Attachment 2 of this supplement contains the hardcopy pages of the joint NRC-BVPS conversion website.This supplement completes the update of the BVPS ITS Conversion License Amendment documentation.

With this supplement, all outstanding License Amendments affecting the BVPS ITS conversion LAR have been incorporated into the BVPS ITS Conversion documentation and all identified NRC comments resolved.FENOC requests an implementation period of 150 days for the ITS conversion License Amendment after it is approved by the NRC.The information provided with this submittal does not change the evaluations or conclusions of the No Significant Hazards Consideration provided with the ITS conversion LAR. No new regulatory commitments are contained in this submittal.

If there are any questions or if additional information is required, please contact Mr. Gregory A. Dunn, Manager, FENOC Fleet Licensing, at (330) 315-7243.I declare under penalty of perjury that the foregoing is true and correct. Executed on October Z4 , 2006.Sincerely, Richard G. Mende Beaver Valley Power Station, Unit Nos. 1 and 2 Supplement to License Amendment Request Nos. 296 and 169, Improved Standard Technical Specification Conversion L-06-149 Page 4 Attachments:

1. BVPS ITS Conversion (LARs 296 and 169) Revision 4 pages sorted by change number.2. Hardcopy of BVPS ITS Conversion Database entries.c: Mr. T. G. Colburn, NRR Senior Project Manager (*) (2 hardcopies)

Mr. P. C. Cataldo, NRC Senior Resident Inspector

(*)Mr. S. J. Collins, NRC Region I Administrator

(*)Mr. D. A. Allard, Director BRP/DEP (*)Mr. L. E. Ryan (BRP/DEP)

(*)(*) Electronic Copy BVPS UNITS I & 2 ITS CONVERSION LICENSE AMENDMENT REQIUEST (LAR)Nos. 296 (UNIT 1) & 169 (UNIT 2)REVISION 4 CHANGES This volume identifies each Revision 4 change by a unique numeric designation.

The tabbed sections of this volume are labeled with the change numbers. Each tabbed section of this volume includes the following information: " A description of the Revision 4 change,* If applicable, the following information is also included; the name of the associated NRC Reviewer(s), and the Beyond Scope Issue (BSI) number,* A separate cover sheet and page number index is included for the affected pages of each ITS section, and" A copy of each revised page with revision bars to show the associated change.Depending on which pages are affected by each change, the pages for each change are presented in the following order; ITS markups and associated Justifications for Deviation (JFDs), ITS Bases Markups and associated JFDs, Current Technical Specification (CTS)markups and associated Discussion of Change (DOC).Each affected page is identified as a Revision 4 page. In addition, each affected page is identified with the associated change number(s) for that page. The Revision 4 changes made to each page are further identified by revision bars.The page numbers referenced in the page number indexes associated with each ITS Section affected by a change are the ITS section specific sequential numbers added to the bottom right hand corner of each page.In most cases, the BVPS ITS Conversion documentation can be updated to Revision 4 by simply replacing the existing page with the corresponding Revision 4 page. However, in order to add pages and avoid excessive repagination, one or more alpha-numeric numbered pages (e.g., 129A) were created for some changes. When updating the original BVPS submittal document with Revision 4 pages, the alpha-numeric numbered pages are inserted in alpha order after the page with the same number (e.g., page 129B follows page 129A, which in turn follows page 129).

BVPS UNITS I & 2 IMPROVED TECHNICAL SPECIFICATION (ITS) CONVERSION LICENSE AMENDMENT REIQJUEST (LAR)NOS. 296 (UNIT 1) & 169 (UNIT 2)REVISION 4 CHANGE 1 Implementation of Amendments 275 (Unit 1) and 156 (Unit 2) -Extended Power Uprate (EPU)NOTE: As the EPU changes were anticipated in the original ITS conversion documentation, this change does not impact the final BVPS ITS. The significant impact of this change is the implementation of the revised License Amendment numbers on the affected Current Technical Specifications (CTS) pages used in the ITS conversion documentation.

Description This change updates the affected pages to the latest version of LAR numbers 302 Unit 1) and 173 Unit 2) as approved in Amendments 275 (Unit 1) and 156 (Unit 2).Amendments 275 (Unit 1) and 156 (Unit 2) were issued by the NRC on 7/19/06 and correspond to License Amendment Request (LAR) numbers 302 (Unit 1) and 173 (Unit 2). These amendments revise the Technical Specifications to authorize an approximately 8% increase in the licensed rated thermal power from 2,689 megawatts thermal (MWt) to 2,900 MWt. This increase is considered an extended power uprate. Additionally, these amendments approve full implementation of an alternative source term in accordance with 10 CFR 50.67, and using Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors." The NRC also approved other Technical Specification changes which are not directly related to the EPU, including deletion of the power range neutron-flux high-negative rate trip for both BVPS Units, and the following changes for BVPS Unit 1: removal of the boron injection tank boron concentration, renaming the boron injection flow path Technical Specification, addition of a footnote addressing time constants to Table 3.3-3, and correction of an inconsistency regarding a referenced permissive in an Action on Table 3.3-1.Affected Pages: The affected pages are presented in ITS Section order. The Table(s) preceding each ITS section list the affected pages by type (i.e., ITS markup, CTS markup, etc.) in that Section of the ITS. The affected page numbers listed are the ITS section specific consecutive numbers found in the lower right corner of each page.Note: Because the affected page(s) for each change were extracted from a complete ITS section electronic file, the electronic hyperlinks (created in the complete ITS section file) do not work in the collection of affected pages that follow this cover page.

BVPS UNITS 1 & 2 ITS CONVERSION LICENSE AMENDMENT REQUEST (LAR)Nos. 296 (UNIT 1) & 169 (UNIT 2)REVISION 4 CHANGE 1 AFFECTED PAGES FOR ITS SECTION 1.0 (USE AND APPLICATION)

ITS SECTION 1.0 (USE & APPLICATION)

INDEX OF AFFECTED PAGES ITS MARKUPS NONE ITS JFDS NONE ITS BASES MARKUPS NONE ITS BASES JFDS NONE CTS MARKUPS PAGES: 39 CTS DOGS NONE 1:0 Use and Application (A.~ DEFINITIONS

1.1 Definitions

Rev._4,_Change_1 NOE DEFINED TERMS defined terms _and Bases_!.-- The DEFINED TERMS of this section appear in capitali dtype and are applicable throughout these Technical Specifications.

THERMAL POWER 1.2 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.RATED THERMAL POWER)1.3 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant as specified in the Licensing Requtirements' Matnttl .2nd nnt exee 2900 MWI-I with fuel in the reactor vessel OPEPATIONAL MODE ,a d reactor vessel head closure bolt tensioning 1.4 An OPERATIONA MODE shall correspond to any o e inclus- A.2 combination of core react vity condition, power leve , average reactor coolant temperature specified in Table 1.1. -Sthat part of a Specification that prescribes Required Actions to be taken under designated Conditions ACTION within specified Completion Times.5 ACTION shall be additional rcquircmonts specifiod as corollary statomcnts

t. .a.h prin.ipal and shall be part of the spocifications.

A(.OPERABLE -OPERABILITY safetyanwad A system, system, tr n, component, or devic shall be PERABLE or ve OPERABILI when it is capable of pe forming its-asýýw'e that- all necessary attendant instrumentati n, controls,/ normal~h~1 emergency electric power seurees, cooling seal water, lubrication other auxiliary equipment that are required for the system, subsys em, train, component or device to perform its, A.5 function(s) are also capable of performing their related function(s) and support specifie REPORTABLE EVEN A.3 1.7 A REPORTABLE EVENT shall bo any of thoec conditions spocifiod in Scctian 50.73 to 10 CFR Part 50.IMENT INTEGRITY 1.8 CONTAINME TEGRITY shall exist when: 1.8. A toatio i e s osed during accident conditions areet A .6 BEAVER VALLEY -UNIT 2 1-1 Amendment No. 156 39 BVPS UNITS I & 2 ITS CONVERSION LICENSE AMENDMENT REQUEST (LAR)Nos. 296 (UNIT 1) & 169 (UNIT 2)REVISION 4 CHANGE 1 AFFECTED PAGES FOR ITS SECTION 2.0 (SAFETY LIMITS)ITS SECTION 2.0 (SAFETY LIMITS) INDEX OF AFFECTED PAGES ITS MARKUPS NONE ITS JFDS NONE ITS BASES MARKUPS NONE ITS BASES JFDS NONE CTS MARKUPS PAGES: 17 CTS DOCS NONE

2.0 Safety

Limits (SLs Rev. 4, Change I 2.0 SAFETY LIMITS p 2.1 SAFETY LIMITS Reactor Coolant System (RCS) highest loop average temperature, andCOREIn MODES I and 2, SLs 12. he combination of THERMAL POWER, pressurizer pressurea---nd/

M ighcst opcrating loop coolant tcmpcraturc (TW shall not exceed the limits specified in the COLR; and the following Lini-s shall not be exceeded: 2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained

> 1.17 for WRB-I DNB correlation for Vantage 5H (V5H) fuel assemblies, and > 1.14 for WRB-2M DNB correlation for Robust Fuel Assemblies (RFA).2.1.1.2 The peak fuel centerline temperature shall be maintained

< 4700 0 F.APPLIGABILITY.

MO)DES 1: and 2.2.2 Safety Limit Violations I MODE 3 If Safety Limit 2.1.1 is violated, restore compliance and be in HGT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . SL REACTOR COOLANT SYSTEM PRESSURE 2.1.2 .he Reactor Coolant System pressure shall net e..eed 2735 psig n-I 3,4,a-d5-L,-cD-LITY.

MODE 1, 2, ,4 and 5.22___ )a-4 ------ 2.2.2.1 In MODE I or 2, restore!2.2.2 If SL 2.1.2 is violated:

compliance and be in MODE 3 Whenever the Reatctor Coolant SySte t pre ssure has .......2735 psig, be in HOT- ST-ANDBY with the Reactor Coolant Systeii pressure within its lim~it within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .MOGDES 3, 4, and 5 Whenever the Reactor Coolant System pressure has dxeeed 2735 psig, reduce the Reactor Coolant System pressure to withi-n is-4i-t1t-within 5 minutes.2.2.2.2 In MODE 3, 4, or 5, restore compliance BEAVER VALLEY -UNIT 2 2-1 Amendment No.156 17 BVPS UNITS 1 & 2 ITS CONVERSION LICENSE AMENDMENT REQUIEST (LAR)Nos. 296 (UNIT 1) & 1 69 (UNIT 2)REVISION 4 CHANGE 1 AFFECTED PAGES FOR ITS SECTION 3.1 (REACTIVITY CONTROL SYSTEMS)ITS SECTION 3.1 (REACTIVITY CONTROL SYSTEMS) INDEX OF AFFECTED PAGES ITS MARKUPS NONE ITS JFDS NONE ITS BASES MARKUPS NONE ITS BASES JFDS NONE CTS MARKUPS PAGES: 146 CTS DOCS NONE REACTIVITY CONTROL SYSTEMS REFUELING WATER STORAGE TANK (RWST)Rev. 4, Change I LIMITING CONDITION FOR OPERATION 3.1.2.8 The RWST shall be OPERABLE.APPLICABILITY:

MODES 1, 2, 3 & 4.ACTION: With the refueling water storage tank inoperable, restore the tank to OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .SURVEILLANCE REQUIREMENTS 4.1.2.8 The RWST shall be verified OPERABLE: a. At least once per 7 days by: 1. Verifying the boron concentration is between 2,400 and 2,600 ppm, and 2. Verifying a minimum usable volume of 859,248 gallons.b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST solution temperature is 45°F and 65 0 F when the RWST ambient air temperature is < 45 0 F or > 65 0 F.I Moved into Section 3.5. Changes to this specification are described in Section 3.5 BEAVER VALLEY -UNIT 2 3/4 1-15 (Next Page is 3/4 1-17)Amendment No. 156 146 BVPS UNITS 1 & 2 ITS CONVERSION LICENSE AMENDMENT REQIUEST (LAR)Nos. 296 (UNIT 1) & 1 69 (UNIT 2)REVISION 4 CHANGE 1 AFFECTED PAGES FOR ITS SECTION 3.3A (REACTOR TRIP SYSTEM (RTS)INSTRUMENTATION)

ITS SECTION 3.3A (RTS INSTRUMENTATION)

INDEX OF AFFECTED PAGES ITS MARKUPS NONE ITS JFDS NONE ITS BASES MARKUPS NONE ITS BASES JFDS NONE CTS MARKUPS PAGES: 102, 113, 117, 125, 126 CTS DOCS NONE Rev. 4, Change 1 UNIT 2 OR OTHER SPECIFIED CONDITIONS

1. Manual Reactor Trip 2. Power Range, Neutron Flux a. High Setpoint b. Low Setpoint 3. Power Range, Neutron Flux0 High Positive Rate F_4~. DGLETED 1K Intermediate Range, Neutron Flux 4+-Source Range , Neutron Flux All ___________

_b. With All Rods Fully Inserted and Without Rod Withdrawal Capability

-4 (f) Above the P-6 (Intermediate Range Neutron Flux) interlocks.

iiTi i'iJi7 (8) Alternate detectors may only be used for monitoring purposes Without Rod Withdrawal Capability until detector functions are modified to permit equivalent alarm and trip functions.

BEAVER VALLEY -UNIT 2 3/14 3-2 Amendment No.156-.X.O)

BEAVER VALLEY -UNIT 2 3/4 3-10 Amendment No. 156 CA)

BEAVER VALLEY -UNIT 1 3/4 3-2 Amendment No.275 I Rev. 4, Change I TABLE 3.3-1 (Continued)

Changes to this Unit I material are addressed in the Unit 2 markup.AC-O With the number of OPERABLE channels' one an tan he Total Number of Channels, STARTUP Tor POWER TION may proceed provided .e following conditi r aisfied : a l ahe inoperable channeld in the tripped conditipe w in w i ona inueounti posrerforance ofstin o othe reqire bCe- Ntpplificable.

4.3.1.1.CTION 0 -Nooappue ACTION 1 With e nnuNmber of OPERABLE channels TOP , r oper hatinel may cn prvie thTEMA OER lee bv -einoperable channelispae in thetrp odin t require L1 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .ACTION 11 -it umber of channels OPERABLE one les Srequired by-ti-Miimum Channels OP rquirement, Unit 1 only. Replace with restore the inopera e to OPERABLE status ITS Conditions 0 and P within 48 e in HOT within the next 6 for RTS Interlocks and/or open the reactor trip brea e oor.Changes to this Unit I material are addressed in the Unit 2 markup.BEAVER VALLEY -UNIT 1 3/4 3-7 Amendment No. 275 125 Changes to this Unit I material are addressed in the Unit 2 markup. TABLE 4.3-1 Rev. 4, Changel REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS Channel M s in Which Channel Channel Functional urveillance ctional Unit Check Calibration Test Required 1. Manua ea to rip N.A. N.A.R 0S/U(1 --A-. 1---- 12,3 P),4(",5al'5(2. PowermRange, tron Fluxg a. High Setpoint D (2)'an Q (3) Q 1, 2 b.LwSetpoint S R()S/U (1)2.Power ange, Neutron FluN.A. R Q , 2 High Positive Rate 4. DELETED 5. Intermediate Range, R(6) S/U(t h 3t(14)Neutron Flux 5(1)6. Source Range(1) NeutronFl a. With Rod Withdrawal S R (6) Q (8) 2, 3 (14)4 4(14)Capability and 5 (1 b. With All Rods Fxly S R (6) (8) 3, 4 and 5 Inserted and Rod Withdr al Capability

7. Overtemper ure AT S R(6) Q 1, 8. Overpo r AT S R Q 2 9. Pr surizer Pressure-Low S R Q 1, 2 10. ressurizer Pressure-High S R Q 1, 2 Pressurizer Water S R Q 1, 2 Level-High (a) With Rod Control System capable of rod withdrawal or one or more rods not fully inserted.(b) Below the P-10 (Power Range Neutron Flux) interlocks.

BEAVER VALLEY -UNIT 1 3/4 3-11 Amendment No. 275 N)

BVPS UNITS 1 & 2 ITS CONVERSION LICENSE AMENDMENT REQUEST (LAR)Nos. 296 (UNIT 1) & 1 69 (UNIT 2)REVISION 4 CHANGE 1 AFFECTED PAGES FOR ITS SECTION 3.3B (INSTRUMENTATION OTHER THAN REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION)

ITS SECTION 3.3B (OTHER INSTRUMENTATION)

INDEX OF AFFECTED PAGES ITS MARKUPS NONE ITS JFDS NONE ITS BASES MARKUPS NONE ITS BASES JFDS NONE CTS MARKUPS PAGES: 197, 200 CTS DOCS NONE New ITS 3.3.8 Boron Dilution Detection Instrumentatioi Flux New ITS 3.3.8 Boron Dilution Detection Instrumentation

[Rev. 4, Change I Functional Un.1. Manual Reactor 2. Power Range, NE TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS Channel -ces in Which Channel Channel Functional Surveillance it Check Calibration Test' Required Trip N.A..A. S( , i, 2, 3 "", 4 autron Flux a. High Note: Changes to these requirements are addressed in the documentation

b. Lowl and markups associated with the Reactor Trip System Technical Specification (ITS 3.3.1). These requirements are not part of ITS 3.3.8.3. Power RE High Positive Rate 4. DELETED 5. I mediate Range, Neutron S R 6 S/U(1)1lux 1, 2 i(7)1 , 2 1, 2 14) (14)5(14)6. Source Range , Neutron Flux-1 a. With Ro Wi ('An~bi iL]-. With All Rods Inserted and Without Rod Withdrawa Capability

7. Ove rature AT kA~1 ITS 3.3.8 =ApplIcablI ty Q -" 8. Overpower AT Q 1, 2 ITS 3.3.8 Applicability

9. Pressurizer Pressure-Low S R Q 1 2 (Above P-7 )10. Pressurizer Pressu s R Q 2 11. Pr er Water Level-High S R Q 1-i- Abov P-7ýBEAVER VALLEY -UNIT 2/3/4 3-10 Amendment No. 156 Note: Changes to these requirements are addressed in the documentation and markups associated with the Reactor Trip System Technical Specification (ITS 3.3.1). These requirements are not part of ITS 3.3.8.200 BVPS UNITS 1 & 2 ITS CONVERSION LICENSE AMENDMENT REIQUEST (LAR)Nos. 296 (UNIT 1) & 1 69 (UNIT 2)REVISION 4 CHANGE 1 AFFECTED PAGES FOR ITS SECTION 3.3C (ENGINEERED SAFETY FEATURES ACTUATION SYSTEM (ESFAS) INSTRUMENTATION)

ITS SECTION 3.3C (ESFAS INSTRUMENTATION)

INDEX OF AFFECTED PAGES ITS MARKUPS NONE ITS JFDS NONE ITS BASES MARKUPS NONE ITS BASES JFDS NONE CTS MARKUPS PAGES: 105, 108 CTS DOCS NONE Time constants utilized in the lead-lag controllers for Steam line Pressure-Low are T1 2 50 seconds and T 2 5 seconds. CHANNEL CALIBRATION shall ensure that these time constants are adjusted to these values.Note in Channel Calibration SR. A9 BEAVER VALLEY -UNIT 1 3/4 3-1S Amendment No. 275 C.0x Time constants utilized in the lead-lag controllers for Steam line Pressure-Low are Ti >- 50 seconds and-T2 -5 seconds. [CHANNEL CALIBRATION shall ensure that these time constants are adjusted to these values.BEAVER VALLEY -UNIT 1 3/4 3-18 Amendment No. 275 Note in Channel Calibration SR. A9 0 02 BVPS UNITS I & 2 ITS CONVERSION LICENSE AMENDMENT REQUEST (LAR)Nos. 296 (UNIT 1) & 1 69 (UNIT 2)REVISION 4 CHANGE 1 AFFECTED PAGES FOR ITS SECTION 3.4 (REACTOR COOLANT SYSTEM)ITS SECTION 3.4 (REACTOR COOLANT SYSTEM) INDEX OF AFFECTED PAGES ITS MARKUPS NONE ITS JFDS NONE ITS BASES MARKUPS NONE ITS BASES JFDS NONE CTS MARKUPS PAGES: 244, 245, 274, 275, 276, 286 CTS DOCS NONE Rev. 4, Change I REACTOR COOLANT SYSTEM 3/4-.4-,-3,---'SAFETY VALVES Ii S 3.4.10 I LPressurizer C F LIMITING CONDITION FOR OPERATION Al 3. All prei1mrizer codes afet~v lves shall be OPERABLE with a lift setting- of 2485 21 pg L 2t 2410.5 psig and!-. 2524.7 psig APPLICABILITY:

MODES 1, 2, and 3, I .. .Mode 4 with all RCS cold leg temperatures

> the LI INSER]ACTION: T1 enable temperature specified in the PTLR.D 24 L2 With one pressurizer code safety valve inoperabl , either restore the inoperable valve to OPERABLE stat s within 15 minutes be in HOT SHUTDOWN with any RC cold leg perature <ý the enable temperature specified in the PTLR L A accordancc with Spccificatien 3.4.9.3 within hours.-61-ý týer any pressurizer cýdý sa.fet.y valve lift, as in .. ed by th-?ý e ' ty valve position indicator, invo i oop se 1 or water dl7s ýe; be in at least ANDBY within t e next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and 1 T SH with any RCS cold le temperature

-< the e temp re specified in the PTLR and apply overpressure protec requirements in acco e with Specification 3.4.9.3 within following ours.k LRM LA2 A4 INSERT 2 SURVEILLANCE REQUIREMENTS 4.4.3 No addition ose required by iS eci---SR 3.4.10.1 Verify each pressurizer safety valve is OPERABLE in accordance with the In accordance with Inservice Testing Program. Following testing, lift settings shall be within +/- 1%. the Inservice Testing Program The lift setting shall correspond to ambient conditions of the valve at nominal oDeratinq temperature and pressure.

I-- Within +/- 1% following pressurizer code safety val SR 3.4.10.1 ve testing.LAI Bases BEAVER VALLEY -UNIT 2 3/4 4-9 Amendment No. 156 244 Rev. 4, Change1I REACTOR COOLANT SYSTEM UNIT I PAGE I R 3/4.4.3 SAFETY VALVES -OPERATING ITS3.410j LIMITING CONDITION FOR OPERATION I.All pressurizer code safety valves shall be OPERABLE with a lift setting* of[ lift ettig* of .** 2410.5 psig and 2559.5 psig S ABILITY: MODES 1, 2 and 3, MODE 4 with all RCS cold leg temper > the able temperature specified in the R.ACTION: Changes to this Unit I material are addressed in the Unit 2 markup a. With one pressuriz ode ety valve inoperable, either restore the ' erable valve OPERABLE status within 15 minut or be in HOT SHUTDOWN w any RCS cold leg t rature < the enable temperature speci in the PTLR and apply RCS overpressure protection requir ts in accordance with Specification 3.4.9.3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .th a pressurizer code safety valve having dis ed liqui er from a water solid pressurizer itigate an overpressure be in at least TANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , an in WN with any RCS cold leg temperature

! the e temper specified in the PTLR and appl overpressure protectio e uirements in a ance with Specification 3.4.9.3 within t llowing 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .SURVEILLANCE REQUIREMENTS

4. No additional requirements other than those requi by Specifi ion 4.0.5.Changes to this Unit 1 material are addressed in the Unit 2 markup* The Lift Setting press e shall car ond to ambient conditions

of :the valve t n hal operating tempe -r n pressure.Within : following pressurizer code safety va ting.BEAVER VALLEY -UNIT 1 3/4 4-6 Amendment No. 275 245 Rev. 4, Change I REACTOR COOLANT SYSTEM?A- SPECIFIC ACTIVITY ITS 3.4.16 LIMITING CONDITION FOR OPERATION F ithn limits.3.4.8 The specific activity of the reactor coolant shall be to:0.35 laCi/gram DOSE EQUIVALENT 1-131, and SR 3.4.16.2 b,- : 100/E laCi/g~r~am SR 3.4.16.1I L Mwith RCS average temperature (Tavg) 5001F.APPLICABILITY:

MODES 1, 2, ,4 n ACTION: LIg MODES 1, 2 and 3*.a-. With the specific activity of the primary coolant > 0.35 gCi/gram DOSE EQUIVALENT 1-131'- for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in HOT STANDBY with Tavg< 500 0 F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .b,. With the specific activity of the primary coolant > 100/E&C.1 gCi/gram, be in HOT STANDBY with Tavg < 500°F within-hours.L1 0!L=4ITS Action B.1 MGDLI L42 , ,aa a--. With the specific activity of th primary coolant > 0.35 pCi/gram DOSE EQUIVALENT 1-131 er ->1004-/-

i/ga, perform the sampling analysis requirement of item 4a of Table 4.4-12 until the specific activity of the primary coolant is restored to within its limits.SURVEILLANCE REQUIREMENTS

4. 4.8 The spccific activity of the primfary coolant shall be dctcrmincd to bc within thc pcrformancc limits of the sampling and analysis program of Table 4.4 12.SR 3.4.16.1, SR 3.4.16.2, SR 3.4.16.3* With Tavg !- 500°1: .EMove to Applicability

-(--)- Specification 3.0.4.c is applicable.

MoventonNoteinRequiredActionsfor Condition A BEAVER VALLEY -UNIT 2 3/4 4-27 Amendment No. 144 274 Rev. 4, Change I Gross Activity times per 7 days Determination with a maximum tim.of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months bctwce SR 3.4.16.1 Once samiples.L3* Isotopic Analysis for DOSE EQUIVALENT 1-131 Concentration F-R 3.4-16-21 3. Radiochemical for SR 3.4.16.3]

]E Determination 47-.- Isotopic Analysis for Iodine including 1-131 1-133, and 1-135 ITS Action A.1 I L 1 per 14 days Ml I Note: Only required to be performed in Mode I I 1 per 6 months a) Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , whenever the specific activity exceeds 0.35 JiCi/gram DOSE EQUIVALENT 1-131 er 10o0/B ýke~i/gr-aomi, a-nd 1#, 2#, 3# 401S with RCS average temoerature (Taval > 500'F.:ir Note: Not required to be performed until 31 days after a minimum of 2 effective full power days and 20 days of Mode I operation have elapsed since the reactor was last subcritical for> 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .One sample between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following a THERMAL POWER change exceeding 15 percent of the RATED THERMAL POWER within a 1-hour period.I Second Frequency of SR 3.4.16.2 A3 ITS Action A.1#Until the specific activity of the primary coolant system is restored to within its limits.BEAVER VALLEY -UNIT 2 3/4 4-28 Amendment No. 101 275 Rev. 4, Change I-t E 250 M In-05 0 100 cc z I-._4:>C-F-4: 0... I'I--UNA CCEPTABI.E OPERATION N E OPERAT ACCEPTABL ION 0 20 30 40 50 60 70 80 PERCENT OF RATED THERMAL POWER 90 10)3.4,16-1 FIGURE 3.4 1 DOSE EQUIVALENT 1-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Geelant Spccific Aetiva-uy

> .2 e tti/6iUA IJQXA. Eqi--BEAVER VALLEY -UNIT 2 3/4 4-29 Amendment No. 101 276 UNIT IPA=GE I TS 3.4.12 I ECORE COOLIN EMS 3/4.5.F Rev. 4, Change 1 Al Note: This is a Unit I Specific TS moved from Section 3.5 to be incorporated into ITS 3.4.12.Unit 2 does not have a corresponding TS.LIMITING CONDITION FOR OPERATION

---ý ITS3"4"12LCitd}

3The ECCS automatic high head safety injection (HHSI)flow path shall be isolated.APPLICABILITY:

MODE 4 when any RCS cold leg temperature is less than or equal to the enable temperature specified in the PTLR, MODE 5, MODE 6 when the reactor vessel head is on.3 3ITS 3.4.12 Applicabilit With the ECCS automatic HHSI flow path not isolated, isolate the flow path within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .SURVEILLANCE REQUIREMENTS

-S- The ECCS automatic H14T flow path qhall he verified/.isolated at- least once per 7 dayd except for purposes of flow testingI lor valve stroke testing.BEAVER VALLEY -UNIT 1 3/4 5-7 Amendment No. 275 286 BVPS UNITS I & 2 ITS CONVERSION LICENSE AMENDMENT REIQUEST (LAR)Nos. 296 (UNIT 1) & 1 69 (UNIT 2)REVISION 4 CHANGE 1 AFFECTED PAGES FOR ITS SECTION 3.5 (EMERGENCY CORE COOLING SYSTEMS)ITS SECTION 3.5 (EMERGENCY CORE COOLING SYSTEMS) INDEX OF AFFECTED PAGES ITS MARKUPS NONE ITS JFDS NONE ITS BASES MARKUPS NONE ITS BASES JFDS NONE CTS MARKUPS PAGES: 76, 83, 88, 89, 90, 91, 93 CTS DOCS NONE Rev. 4, Change 1 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)ACCUMULATORS ITS 3.5.1 ()LIMITING CONDITION FOR OPERATION' ThreeECCS S-1 3.5.1 Each Rcacter Coolant. Sst accumulatof shall be OPERABLE A2 S The isolation valve open, SR3.5.1.2 Between 6898 gallons and 8019 gallons of usable borated water, R .Between 2300 and 2600 ppm of boron, and B ;SR 3.5.1.3 .A nitrogen cover-pressure of between 611 and 685 psig.APPLICABILITY:

MODES 1, 2 and 3.*ACTION: CONDA COND B COND C CONDD0 SURVE a-. With one accumulator inoperable due to boron concentration not within limits, restore the inoperable accumulator to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .b-. With one accumulator inoperable for reasons other than Action a, restore the inoperable accumulator to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . MODE3 KWith either Action a or b not being completed within the specified completion time, be in at least t within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce prcssurizcr pressure to K 1000 psig within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . 7Ie )ILLANCE REOUIREMENTSn A4. S.Each accumulator shall be demonstrated OPERABLE: a-. At I SSR 3-5.1.2]_1 SR 3.5.1.3 S3 =.SR 3.5.1.1 I least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by: is > 6898 gallons and < 8019 gallons Verifying the sable borated water volumepnd nitrogen cover-pressure in the tanks arc within limits, and A2 Verifying that ach accumula r isolation valve isýcpn.Bases LAI accumulator is 611 psig and ý685 psig*PrPssuriz e ur a e 1000 psig.BEAVER VALLEY -UNIT 2 3/4 5-1 Amendment No. 156 76 I Changes to this Unit I material are addressed in the Markup and DOCs associated with the corresponding Unit 2 text.EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 ECCS SUBSYSTEMS

-Tavg > 350-F (Ul)LIMITING CONDITION FOR OPERATION Rev. 4, Change 1I Th~i sP epasaesu systems shall b-e[a-- One OPERABLE eentrifuga-l charging pump, b-. One OPERABLE low head safety injection pump, and e-- An OPERABLE flow path capable of taking suction from the refueling water storage tank on a safety injection signal and transferring suction to the containment sump during the recirculation phase of operation.

MODES 1, 2 and 3.ACTINperable su Changes to this Unit 1 material are addressed in the Markup be in HOTand DOCs associated with the corresponding Unit 2 text.b. In the event_ th6ECCS is actuated a njects water into the Reactor olant System, a Special Repor -all be prepared a ubmitted in accordance with 10 CFR 50.4 in 30 days escribing the circumstances of the actuation and ttal accumulated actuation cycles to date.Changes to this Unit I material are addressed in the Markup and DOCs associated with the corresponding Unit 2 text.i 3 one of the required centrifugal chaarýlinng.piimpe--m-aa-y-bb-ee made incapa injecting to support tr h into or from p ic il 0 f t 9 0 p to rs or s x ýs 0 ic v r the Applicability of at .4. .33 for up to 44 hours5.092593e-4 days 0.0122 hours 7.275132e-5 weeks 1.6742e-5 months or until the temperatur RC ee s exceeds thee OPPS ur sp ci fj t PTL plus w OPPS f r u ou port 0 t 0 or ry enable ure specified in thee PTLRR plus whichever es first.In MODE 3, the ECCS automatic HHSI flow path may be isolated to support transition into or from the Applicability of piIIcfs' t on d.5.4 for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or until the temperature of a CS cold legs exceeds the OPPS enable temperature speci fed in the PTLR plus 25°F, whichever comes first./LCO 3.4.12, "Overpressure Protection Systems (OPPS)"-ýý ITS 3.5.2 Unit I Specific LCO Note 3 BEAVER VALLEY -UNIT 1 3/4 5-3 Amendment No. 275 83 Changes to this Unit I material are addressed in the Markup and DOCs associated with the corresponding Unit 2 text.EMERGENCY CORE COOLING SYSTEMS ITS 3.5.3 ECCS SUBSYSTEMS

-Tavg < 350-F (Ul) ( U LIMITING CONDITION FOR OPERATION 4DI Rev. 4, Change I 3.5.3 As a minimu s s em comprised of the-I be OPERABLE: a.b.C.LAI One OPER-BL ee-:rr--f-a&

charging pump, Bases One OPERABLE Low Head Safety Injection Pump, and An OPERABLE flow path capable of taking suction from the refueling water storage tank upon being manually realigned and transferring suction to the containment sump during the recirculation phase of operation.

.LICABILITY:

MODE 4.--------------------

GENERAL NOTE-------------------

Specificatio 3.0.4.b is not applicable to CCS centrifugal

a. With no ECCS subsy em OP BLE because of the mnoper-ability of either the trifugal charging pump or the f low nath f rom the- r I P..c water storaae tank, restore at Changes to this Unit I material are addressed in the Markup within 1 and DOCs associated with the corresponding Unit 2 text. urs.b.l In ce -v -bi d Q; d U t2 U Ct+/-I I C -L1vu , ater into the Re tor Coolant System, a Special eport shall be pre ed and submitted in accordance wi 10 CFR 50.4 whin 30 days describing the circumstanc of the actuation and the total accumulated actuation cle to date.SURVEILLANCE REQUIREMENTS, I train The ECCS sui]tc shall be demonstrated OPERABLE by the performance of each ef tho Surveillance Requirements of 4.5.2 cxcpt SR R53.5r.qu2r.nts

.5.R .S.2.f.2 and 4.5.2.S.32 BEAVER VALLEY -UNIT 1 3/4 5-6 Amendment No. 275 88 Rev. 4, Change 1 EMERGENCY CORE COOLING SYSTEMS 3/4.5.4 SEAL INJECTION FLOW ITS3.5.5 LIMITING CONDITION FOR OPERATION 3.5.4 Reactor coolant pump seal injection flow shall be less than or equal to 28 gpm with the charging pump discharge pressure greater than or equal to 2457 psig and the seal injection flow control valve full open.APPLICABILITY:

MODES 1, 2, and 3.ACTION: I I LAI Action Bases I SCONDA -. With the seal injection flow not within the limit, adjust manual seal injection throttle valves to give a flow within the limit -ithin the limngi wt n preu greater than or equal to 2457 psig and the seal injection inflow control valve full open 1 within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at ONeast 7hy within te next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months an28in SHU9TDO0WN within the hours.[MODE 3 1 JMDE S3.5.5.....

LANCE REQUIREMENTS A E aulsa neto trtl 4.-. Verify at least once per 31 days that the vales, are adjusted to give a within the limit with the charging pump discharge at greater th n or equal to 2457 psig and the seal injection flow control va ve full open.ý1)ýý28 gpm SR 3.5.5.1 Note--- Not required to be performed until 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the Reactor Coolant System pressure stabilizes at greater than or equal to 2215 psig and less than or equal to 2255 psig.BEAVER VALLEY -UNIT 2 3/4 5-7 Amendment No. 156 89 I Rev. 4, Change I REACTIVITY CONTROL SYSTEMS Refueling Water Storage Tank (RWST)LIMITING CONDITION FOR OPERATION I.2.8 The RWST shall be OPERABLE.3.5.4 PLICABILITY:

MODES 1, 2, 3 & 4.{ t Ifor reasons/other than Condition A ACTION: CONDA Inset , CONDB With the refueling water storage tank inoperabl , restore the tank to OPERABLE status within one hour or be in at least HGT COND within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in wOLD SHUTEithin the Sfoll1 ing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .SURVEILLANCE REQUIREMENTS 4.1.2.8 The RWST shall be verified OPERABLE:

LAI a-.- At least once per 7 days by: Bases SR 3.5.4.3 -Verifying the boron concen ration is between 2400 and 2600 ppm, and SR3.5.4.2 Verifying a minimu usable volume of 859,248 gallons.SR3.5.4.1 -At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST solution temperature is 45°F and 65 0 F when the RWST ambient air temperature is < 45 0 F or > 65 0 F.BEAVER VALLEY -UNIT 2 3/4 1-15 (Next page is 3/4 1-17)Amendment No. 156 90 IVUnitY L REACTIVITY CONTROL SYSTEMS I ITS 3.5.4 Rev. 4, Change I I REFUELING WATER STORAGE TANK(RWST)

LIMITING CONDITION FOR OPERATION 3.":2. TheRWSýT shall be OPERABLE.APPLICABI TY: MODES 1, 2, 3 & 4.ACTION: Changes to this Unit I material are addressed in the MarkuplWith the and DOCs associated with the corresponding Unit 2text. ore the tank to ( Bast HOT STANDBY within the next hours Pr in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .SURVEILLANCE REQUIREMENT 4.1.2.8 The RWS shall be verified OPERABLE: a. least once per 7 days by: A2 1. Verifying the boron conce tra___________2,600 ppm, and usbe Bas es 14-7'Verifying a be.tween gallons and 44 ,00 gallons of borated water.BEAVER VALLEY UNIT 1 3/4 1-16 (Next page is 3/4 1-17a)Amendment No. 275 91 Unit I Only_EMERGENCY CORE COOLING SYSTEMS Rev. 4, 3 /4.. 4 HHSI Flow Path Note: This is a Unit I Specific TS. Unit 2 does not have a corresponding TS.hange I LIMITING COND ION FOR OPERATION 3.5.4 The CCS automatic high head safe injection (HHSI)flow path shall be is ated.APPLICABILITY:

MODE 4 wh any RCS cold 1 temperature is less than or equal to he enable emperature specified in the PTLR, MODE 5, _ _ .MODE 6 when the r tor vessel head is on.ACTION: With the ECCS automat HHSI flow path not iso ted, isolate the flow path within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .SURVE ILLANCE QURMET X4.5.4 The ECCS automatic HHSI flow path shall be rified oralve stroke testing.The requirements of this Unit 1 TS are related to low temperature overpressure protection of the RCS. Therefore, the requirements of this TS are moved to Section 3.4 (ITS 3.4.12 Overpressure Protection Systems (OPPS)). Changes to the requirements of this TS are shown in Section 3.4.BEAVER VALLEY -UNIT 1 3/4 5-7 Amendment No. 275 93 BVPS UNITS 1 & 2 ITS CONVERSION LICENSE AMENDMENT REQUEST (LAR)Nos. 296 (UNIT 1) & 169 (UNIT 2)REVISION 4 CHANGE 1 AFFECTED PAGES FOR ITS SECTION 3.7 (PLANT SYSTEMS)ITS SECTION 3.7 (PLANT SYSTEMS) INDEX OF AFFECTED PAGES ITS MARKUPS NONE ITS JFDS NONE ITS BASES MARKUPS NONE ITS BASES JFDS NONE CTS MARKUPS PAGES: 218, 219, 220, 221,226, 227 CTS DOCS NONE 3/4.7 PLANT SYSTEMS 3 /4. 7.1 FTURBINT~E' G].TfLE Rev. 4, Change I AlK MAIN STEAM SAFETY VALVES (MSSVs)LIMITING CONDITION FOR OPERATION 3.7.1.1 Five MSSVs per steam generator shall be OPERABLE.APPLICABILITY:

MODES 1, 2 and 3.ACTION: I ACTIONS NOTE--- --------------

-- GENERAL NOTE -Separate ACTION entry is allowed for each MSSV.CONDa-With one or more steam generators with one MSSV inoperable and the Moderator Temperature Coefficient (MTC) zero or negative at all power levels, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months reduce THERMAL POWER to less than or equal to 57% RTP; otherwise, be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .-]a-. With one or more steam generators with two or more MSSVs inoperable, or with one or more steam generators with one MSSV inoperable and the MTC positive at any power level, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months reduce THERMAL POWER to less than or equal to the Maximum Allowable

% RTP specified in Table 3.7-1 for the number of OPERABLE MSSVs, and reduce the Power Range Neutron Flux-High reactor trip setpoint to less than or equal to the Maximum Allowable

% RTP specified in Table 3.7-1 for the number of OPERABLE MSSVs within the next 32 hour3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months s+1+;

otherwise, be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .SCONDC 1e-:-With one or more steam generators with four or more MSSVs inoperable, within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months be in HOT STANDBY and in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .SURVEILLANCE REQUIREMENTS R .7.1. Verify-2+

each required MSSV lift setpoint per Table 3.7-2 in 3.71.1 ccordance with the Inservice Testing Program. Following testing, lift settings shall be within + 1 percent. I___________

COND B.2 NOTE 4-+ Required to be performed only in MODE 1.-(-2+ Required to be performed only in MODES 1 and 2.- R37.1.1NOTE BEAVER VALLEY -UNIT 2 3/4 7-1 Amendment No. 156 218 Rev. 4, Change I TABLE 33-"1-OPERABLE Main Steam Safety Valves versus Maximum Allowable Power NUMBER OF OPERABLE MSSVs MAXIMUM ALLOWABLE POWER PER STEAM GENERATOR

(% RTP)4 < 50 3 < 34 2< 19 BEAVER VALLEY -UNIT 2 3/4 7-2 Amendment No. 156 219 ITS 3.7.1 Rev. 4, Change I I r 3.7.1-2b TABLE -lni!'i ",., Main Steam Safety Valve Lift Settings I STEAM! LINE SAFETY VALVES PER LOOP T LIFT SETTING-*LIFT SETTING TOLERANCES VALVE NUMBER a. 2MSS-SV101A, B & C b. 2MSS-SV102A, B & C c. 2MSS-SV103A, B & C d. 2MSS-SV104A, B & C e. 2MSS-SVI0SA, B & C 1075 psig 1085 psig 1095 psig 1110 psig 1125 psig+/-3%+/-3%+/-3%+3%LIL LAI*The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.Bases for SR 3.7.1.1 BEAVER VALLEY -UNIT 2 3/4 7-3 Amendment No. 156 220 Unit I Page 3.7.1-2a;ettingsTABLE 3.-7-2 Rev. 4, Change I Unit 1 Main Steam Safety Valve Lift S LIFT SETTING'---

LIFT SETTING TOLERANCES VALVE NUMBER a. SV-MS101A, B & C b. SV-MS102A, B & C C. SV-MSI03A, B & C d. SV-MSI04A, B & C e. SV-MSI05A, B & C 1075 psig 1085 psig 1095 psig 1110 psig 1125 psig+/-3%+/-3%+/-3%+/-3%LAI*** The Lift Setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.Bases\for SR 3.7.1.1 BEAVER VALLEY -UNIT 1 3/4 7-4 Amendment No. 275 221 Rev. 4, Change I APPLICA BILITSY:TE M ODES 1,. 2 aStorage Tank (P PDST)]PRIMARY PLANT DEMINERALIZED WATER LIMITING CONDITION FOR OPERATION R 3... PPDWST..... .7I SR 3.7.6.1 ý,,- ý ,7 3.7.1. The primfary Bj R jn-ýzedwt- OPERABLElwith a minimum~lusablel volumel of ga lons. R ase APPLICABILITY:

MODES 1, 2 and CONDA ION: MODE 4 when steam generator is relied upon for heat removal With the' =AD. .tr.g. tank water volume not within the limit, within 4hourl e4:ý adocpe12 hours thereafter M2ii MI a. Restoro the water volumo to within tho limit or be in IIO LIin Verify by administrative means a the OPERABILITY of the Iservice water s stem a backup supply to the auxiliary feedwater pumps and restore the eank water volume to wi in its LAI CONDB limit withi 7 days or be in HOT SHUTDOWN within the next hours. MODE3within L2 24 PPDWST 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and without reliance on steam generator for heat removal-SURVEILLANCE REQUIREMENTS M3 4.... The PPD.W st.rag. t.an. shall be demonstrated OPERABLE at once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying the water level.SR 3.7.6.1 least BEAVER VALLEY -UNIT 2 3/4 7-6 Amendment No. 156 226 Rev. 4, Change I PLANT SYSTEMS I[ITS3.7.13

/ACTIVITY Secondary Specific Ic4 LIMITING CONDITION FOR OPERATION 3.4.1.4 The specific activity of the secondary coolant system shall be _ 0.10 pCi/gram DOSE EQUIVALENT 1-131.APPLICABILITY:

MODES 1, 2, 3, and 4 ACTION: CONDA With the specific-activity of>0.10 i/gra.. DOSE EQUIVALENT I within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN not within limit the secondary coolant system 131- be in at least HOT STANDBY within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .fTT.1PMTFT1LANCE REQUIREMENTS I fe specific activity of the secondary coolant system shall be te be within the 4im-4i by performanco of the sampling and analysTh -r ---- of Table 4." 2.SR 37.13.1 is 10 Ci/m DOSE EQUIVALENT 1-131 I / "/ I ...... I ... "J4 ,4, ... I Li (Unlit 2) 1eve y 12 day I Mi BEAVER VALLEY -UNIT 2 3/4 7-7 227 BVPS UNITS I & 2 ITS CONVERSION LICENSE AMENDMENT REQUEST (LAR)Nos. 296 (UNIT 1) & 1 69 (UNIT 2)REVISION 4 CHANGE 1 AFFECTED PAGES FOR ITS SECTION 5.0 (ADMINISTRATIVE CONTROLS)ITS SECTION 5.0 (ADMINISTRATIVE CONTROLS)

INDEX OF AFFECTED PAGES ITS MARKUPS NONE ITS JFDS NONE ITS BASES MARKUPS NONE ITS BASES JFDS NONE CTS MARKUPS PAGES: 79, 87B, 87F CTS DOCS NONE I ITSi5.6 1 Rev. 4, Change I ADMINISTRATIVE CONTROLS REPORTING REQUIREMENTS (Continued)

WCAP-8745-P-A, "Design Bases for the Thermal Overtemperatfe AT .nd Thermal Overpower AT Trip~Functions," ýSeptember 19861.*WCAP 12945--,10 eZ 1 (Revision 2)1 and Volumes 2 WCthrough 5 A (Revision M), "Code Qualification Document for CBest Estimate LOCA Analysis," March 1998 (Westinghouse roprietary).

COLR WCAP-10216-P-A "Revision T AE "Relaxation of Constant Axial Offset Control-F Surveillance Technical Specification," 4 February 1994 5 WCAP-14565-P-A, "IVIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety COR Analysisl, October 19991.WCAP-12610-P-A, "VANTAGE+

Fuel Assembl Reference Core specifApril 1995 (Westinghouse Prov metaryh0~WCAP-15025-P-A, "Modified WRB-2 Correlation, WRB-2M, for Predicating Critical Heat Flux in 1x7Rod Bundles with Modified LPD Mixing Vane Grids," April 1999.As described in reference documents listed above, when an initial assumed power level of 102% of rated thermal power is specified in a previously approved method, 100.6% of rated thermal power may be used when input for reactor thermal power measurement of feedwater flow is by the leading edge flow meter (LEFM).Caldon, Inc. Engineering Report-80P, "Improving Thermal BEVR Power Accuracy and Plant Safety While Irai nnor.15n6Level Using the LEFM4- System," 'Revis ion 6, Marc_/J Caldon, Inc. Engineering Report-160P, "Supplement to L,^4 Topical Report ER-80P: Basis for a Power Uprate With the BEAVER VALLEY -UNIT 2 6-20 Amendment No. 156 79 DITS 5I Rev. 4, Change 1, 10,&11 ADMINISTRATIVE CONTROLS STEAM GENERATOR PROGRAM (Continued)

2. Tubes found by inservice inspection to contain a flaw 11 in a sleeve (excluding the sleeve to tube joint) with a depth equal to or exceeding the following percentages 10 of the nominal sleeve wall thickness, shall be plugged: ABB Combustion Engineering TIG welded sleeves 27%Westinghouse laser welded sleeves 25%3. Tubes with a flaw in a sleeve to tube joint shall be 10 plugged.,5.5.5.2

4. Tube support plate voltage-based repa eria may be 11 applied as an alternative to the 40% depth based criteria of Teehniea!

Specification

-.--i-9.c.l.Tube Support Plate Plugging Limit is used for the disposition of an Alloy 600 steam generator tube for 10 continued service that is experiencing predominantly axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates. At tube support plate intersections, the plugging (repair) limit is described below: a) Steam generator tubes, with degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with bobbin voltages less than or equal to 2.0 volts will be allowed to remain in service.b) Steam generator tubes, with degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts will be repaired or plugged, except as noted in 6-19.c.4.c below. 5.55.c) Steam generator tubes, with indications of potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts but less than or equal to the upper voltage repair limit (calculated according to the methodology in Generic Letter 95-05 as supplemented) may remain in service if a rotating pancake coil or acceptable alternative inspection does not detect degradation.

BEAVER VALLEY -UNIT 2 6-29 Amendment No. 160 87B LI~i~iIII

~iAlli Rev. 4, Change I &1I1 ADMINISTRATIVE CONTROLS (11I for pagination only)STEAM GENERATOR PROGRAM (Continued)

f. Provisions for SG Tube Repair Methods Steam generator tube repair methods shall provide the means to reestablish the RCS pressure boundary integrity of SG tubes without removing the tube from service. For the purposes of these Specifications, tube plugging is not a repair. All acceptable tube repair methods are listed below.1. ABB Combustion Engineering TIG welded sleeves, CEN-629-P, Revision 02 and CEN-629-P Addendum 1.2. Westinghouse Revision 2.laser welded sleeves, WCAP-13483, Add ITS 5.5.3, Component Cyclic or Transient Limit -1 Insert 1 Add ITS 5.5.11, Safety Function Determination Program 2 [AddlITS 5.5.13, Battery Monitoring and Maintenance 1:rogýi:ram I net Add ITS 5.5.9, Diesel Fuel Oil Testing Program Insert 5A24 Add ITS 5.6.5, Post Accident Monitoring Report BEAVER VALLEY -UNIT 2 6-33 Amendment No. 160 87F BVPS UNITS 1 & 2 IMPROVED TECHNICAL SPECIFICATION (ITS) CONVERSION LICENSE AMENDMENT REQUEST (LAR)NOS. 296 (UNIT 1) & 1 69 (UNIT 2)REVISION 4 CHANGE 2 Partial withdrawal of Beyond Scope Issue (BSI) number 26 NOTE: The changes involved with the partial withdrawal of BSI-26 were previously submitted to the NRC by FENOC letter number L-06-137 dated 9/1/2006 for a separate review by the NRC. The affected pages included in this Revision 4 change are the same (except for a minor clarification to more accurately show the location of inserted text on page 17) as the pages submitted by FENOC letter number L-06-137.

However, as the affected pages are also part of Revision 4 to the BVPS ITS conversion LAR documentation they have been retained within the Revision 4 documentation.

Description This change addresses the partial withdrawal of BSI-26. The changes introduced by BSI-26 were based on changes proposed in Technical Specification Task Force (TSTF) traveler number 412 Revision 0. The change is intended to resolve the NRC concern regarding the reliance on the turbine-driven Auxiliary Feedwater (AFW) pump with only one operable steam supply line as the sole safety related means of cooldown when placing the plant in Mode 4. The change eliminates the proposed Required Action to place the plant in Mode 4 relying only on the turbine-driven AFW pump with one of the two required steam supply lines operable.

The change affects the Actions for 3 inoperable AFW trains and results in requiring one AFW train to be restored to full operability (e.g., two operable steam supply lines for the turbine-driven pump or an operable motor-driven pump) prior to placing the plant in Mode 4.Affected Pages: The affected pages are presented in ITS Section order. The Table(s) preceding each ITS section list the affected pages by type (i.e., ITS markup, CTS markup, etc.) in that Section of the ITS. The affected page numbers listed are the ITS section specific consecutive numbers found in the lower right corner of each page.Note: Because the affected page(s) for each change were extracted from a complete ITS section electronic file, the electronic hyperlinks (created in the complete ITS section file) do not work in the collection of affected pages that follow this cover page.

BVPS UNITS 1 & 2 ITS CONVERSION LICENSE AMENDMENT REQUEST (LAR)NOs. 296 (UNIT 1) & 169 (UNIT 2)REVISION 4 CHANGE 2 AFFECTED PAGES FOR ITS SECTION 3.7 (PLANT SYSTEMS)ITS SECTION 3.7 (PLANT SYSTEMS) INDEX OF AFFECTED PAGES ITS MARKUPS PAGES: 17, 20 ITS JFDS NONE ITS BASES MARKUPS PAGES: 106, 107 ITS BASES JFDS NONE CTS MARKUPS PAGES: 223, 225 CTS DOCS PAGES: 274, 275 Rev. 4, Change 2 AFW System 3.7.5 Insert 1 3 ACTIONS (continued)

CONDITION Required Action and associated Completion Time for Condition A not met.[OR B,orC Two AFW trains inoperable in MODE 1, 2,""--.ýor 3. 1 WOG STS 3.7.5-2 Rev. 2, 04/30/01 17 Rev. 4, Change 2 INSERTS FOR ITS 3.7.5 Auxiliary Feedwater (AFW) System CONDITION C (From TSTF-412)C. Turbine driven AFW train inoperable C.1 Restore the steam supply to the 1241491 hours due to one required steam supply turbine driven train to inoperable in MODE 1, 2 or 3. OPERABLE status.AND OR One motor driven AFW train C.2 Restore the motor driven AFW [244-[48 hours inoperable in MODE 1, 2 or 3. train to OPERABLE status.(For Information Only.) TSTF-412 Condition C Completion Time Reviewers Note: The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time is applicable to plants that can no longer meet the safety analysis requirement of 100% AFW flow to the SG(s) assuming no single active failure and a FLB or MSLB results in the loss of the remaining steam supply to the turbine driven AFW pump.The 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months Completion Time is applicable to plants that can still meet the safety analysis requirement of 100% AFW flow to the SG(s) assuming no single active failure and a FLB or MSLB results in the loss of the remaining steam supply to the turbine driven AFW pump.2. Condition E Three AFW trains inoperable in MODE 1, 2, or 3.20 AFW System B 3.7.5 I Rev. 4, Change 2 1 BASES ACTIONS (continued) for an inoperable turbine-driven AFW pump in MODE 3 realign OPERABLE AFW pumps to separate train supply headers within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (if both train supply headers are OPERABLE) and to restore the AFW train to Required Action B.1 to realign the OPERABLE pumps to separate supply headers preserves train separation and enhances system reliability.

The two hours allowed for this action is reasonable based on operating experience to perform the specified task.that both Completion Times apply simultaneously, and the more restricti\

e must be met.Conditioi A is modified by a Note which limits the applicability of the Conditio' to when the unit has not entered MODE 2 following a refueling.

Condition A allows one AFW train to be inoperable for 7 days vice the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time in Condition B. This longer Completion Time is based on the reduced decay heat following refueling and prior to the reactor being critical.

] Required Action B.1 is modified by a Note indicating that the Required Action is only B. andB 2 required applicable if both supply headers are OPERABLE.With one of the required AFtrains (pump or flow path) inoperable in MODE 1, 2, or 3 [for other than Condition A], action must be taken t restere OPERAEE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . This Condition includes the loss of twotteam supply lines to the turbine driven AFW pump The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is reasonable, based on redundant ca ilities afforded by the AFW System, time needed for repairs, and Wie low probability of a DBA occurring during this time period.The second Completion Time for Required Action B. establishes a limit on the maximum time allowed for any combination of (-nditions to be inoperable during any continuous failure to meet this LCO. 01 The 10 day Completion Time provides a limitation time allowed in this specified Condition after discovery of failure to meet the LCO. This limit is considered reasonable for situations in which Conditions A and B are entered concurrently.

The AND connector between 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and 10 days dictates that both Completion Times apply simultaneously, and the more restrictive must be met.,,C.1 and C.2 D.1 and D.2 B.1, B.2, C.1, or C. 2 hen Required Action A.1 annot be completed within the required Completion Time, or if two AFW trains are inoperable in-,ODE 1, 2, or 3, the unit must be placed in a MODE in which the 0 LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 4 within f181 hours.The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.O IInsert6* If two AFW trains are inoperable in MODE 1, 2, or 3 for reasons other than Condition C, or If one or two feedwater injection headers are inoperable in MODE 1, 2, or 3.B 3.7.5 -5 Rev. 2, 04/30/01 106 Rev. 4, Change 2 If a motor-driven AFW pump is not available in MODE 4 and In MODE 4, with one or two feedwater injection the SG(s) are relied on for headers inoperable, operation is allowed to decay heat removal then AFW System continue because the remaining OPERABLE Condition F is applicable.

B 3. r injection header(s) provide a flow path to the SG(s) However, in MODE 4, two RHR loops may be used for decay subjected relied on for decay heat removal. Additionally, in heat removal in lieu of the to a BASES Mode 4, the RHR loops may be used in lieu of or to SG(s) consistent with the reduction in supplement the SG(s) for decay heat removal requirements of LCO 3.4.6, MODE that consistent with the requirements of LCO 3.4.6, "RCS Loops -MODE 4." could ACTIOI "RS Loops -MODE 4. increase the In MODE 4 with two AFW trains inoperable, operati n is allowed to likelihood of continue because only one motor driven pump AF train is required in the AFW accordance with the Note that modifies the LCO. Although not required, system the unit may continue to cool down and .initiate RHR. being required to 0.1 (7' I / or if all three feedwater injection headers support/ heat If all [three] AFW train are inoperable in MODE 1, 2, or 3, the unit is in removal.E[f seriously degraded condition with no safety related means for conducting a cooldown, and only limited means for conducting a cooldown with related equipment,.,/nsuch a condition, the u nitshtould ,nt 6 Uip."seriousnes's of this condition requires'that actio'n'be" startedimmediat to, restore one AFW train to OPERABLE Required Actionft.1 is modified by a Note indicating that all required I MODE changes or power reductions are suspended until one AFW train Q with the capability of is restored to OPERABLE statu In this case, LCO 3.0.3 is not () providing flow to the applicable because it could force the unit into a less safe condition.

steam generator(s).

II " steam geneator with the required feedwater injection header(s)"F In MODE 4, either the reactor coolant pumps or the RH oops can be used to provide forced circulation.

This is address in LCO 3.4.6, '"RCS Loops -MODE 4." With one required AFW trai noperable, action must be taken to immediately restore the inoperable train to OPERABLE statu% The immediate Completion Time is consistent with LCO 3.4.6.4 Insert 7 SURVEILLANCE SR 3.7.5.1 REQUIREMENTSVerifying the correct alignment for manual, power operated, and automatic valves in the AFW System water and steam supply flow paths Completing verification provides assurance that the proper flow paths will exist for AFW includes re-verifying these operation.

This SR does not apply to valves that are locked, sealed, or requirements by a second otherwise secured in position, since they are verified to be in the correct and independent operator.

position prior to locking, sealing, or securing.

This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This Surveillance does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position.The SR is modified by a Note that states one or more AFW trains may be considered OPERABLE during alignment and operation for steam WOG STS B 3.7.5 -6 Rev. 2, 04/30/01 107 ITS 3.7.5 Rev. 4, Change 2 PLANT SYSTEMS LIMITING CONDITION FOR OPERATION ACTION (Continued)

CONDC Inset2 L e-. With one AFW train inoperable in MODE 1, 2, or 3 for reasons other than one of the two steam supplies or one CONDB feedwater injection header inoperable, realign the two AFW pumps to separate train supply headers within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months if both train supply headers are operable, and restore the AFW train to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and within 10 days CONDD from discovery of failure to meet the LCO or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following

+2++hours.

or one or two feedwater injection headers CONDD :- With two AFW tra.ns inoperable in MODE 1, 2, or 3, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN wi hin the following hours. 12 orthree feedwater injection headers A4 C OND E L3-e-. With three AFW trains inoperable in MODE 1, 2, r 3, immediately initiate action to restore one AFW tra n to OPERABLE status- or required feedwater injection header COND -f With the re ired AFW train inoperable in MODE 4, immediately ini iate action to restore one AFW train to OPERABLE status Sith the capability of providr"ing A4B SURVEILLANCE REQUIREMENTS flow to the steam generator.LA

GENERAL NOTE- --------------

Establish and maintain constant communications between the control room and the auxiliary feed pump room while any normal AFW pump discharge valve is closed during surveillance testing.Note: Only applicable if MODE 2 has not been entered following refueling.

-(-2 This time period may be extended for up to h hours for the turbine driven AFW pump provided that the plant has not entered ODE 2 following a refueling outage.4-3+/- LCO 3.0.3 and all other LCO ACTION statements requiring MODE changes are suspended until one AFW train is restored to OPERABLE status.COND. A Actions & NOTE& COND. D Actions ,Required Action BEAVERVALLE -UNIT 2en NOTE 8 BEAVER VALLEY -UNIT 2 3/4 7-5 Amendment No. 85 223 Rev. 4, Change 2 INSERTS FOR CTS 3.7.1.2 MARKUP 1.-NOTE -AFW train(s) may be considered OPERABLE during alignment and operation for steam generator level control, if it is capable of being manually realigned to the AFW mode of operation.

2. CONDITION C C. Turbine driven AFW train C.A Restore the steam supply 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months inoperable due to one to the turbine driven train inoperable steam supply in to OPERABLE status.MODE 1, 2 or 3.OR AND C.2 Restore the motor driven 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months One motor driven AFW train AFW train to OPERABLE inoperable in MODE 1, 2 or 3. status.3.4.225 I Rev. 4, Change 2 1 BVPS ISTS Conversion

3.7 Plant

Systems Enclosure 3 Changes to CTS time for restoration of the turbine-driven pump in Mode 3 prior to requiring the plant to be placed in Mode 4. Therefore, the CTS Action c cumulative time is considered to be a total of 162 hours0.00188 days 0.045 hours 2.678571e-4 weeks 6.1641e-5 months (72+90) from entry into CTS Action c until entry into Mode 4 is required.

The corresponding ITS 3.7.5 Condition A allows 7 days (168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months )restoration time and ITS Condition D allows an additional 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months to transition from Mode 3 to Mode 4 whenever the turbine driven AFW pump is inoperable in Mode 3 prior to entering Mode 2 following a refueling outage. As such, the ITS cumulative time from entry into Condition A until entry into Mode 4 is required (by Condition D)is a total of 186 hours0.00215 days 0.0517 hours 3.075397e-4 weeks 7.0773e-5 months (168+18).

Therefore, the ITS provides an additional 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (186-162) beyond the time allowed in the CTS. The CTS has been revised to incorporate and extend the 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months provided in footnote (2) for restoring an inoperable turbine-driven AFW pump by an additional 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for a total of 114 hours0.00132 days 0.0317 hours 1.884921e-4 weeks 4.3377e-5 months (90+24).The actual format and presentation of the allowed Completion times is also revised to be consistent with the ISTS. This results in the Completion Times being included in ITS 3.7.5 Condition A (for the 7 day restoration time) and ITS Condition D (for the 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months to place the plant in Mode 4). This DOC is only intended to address the fact that the CTS time is being extended.

The format changes associated with adopting the presentation of ISTS Action Conditions for these times are addressed by DOC A.I. This change is being made so the BVPS ITS is consistent with the corresponding requirements of the ISTS.The purpose of CTS 3.7.1.2 Action c and Footnote (2) is to provide additional time to complete any necessary repairs and testing of the turbine driven AFW pump prior to initiating a plant cool down to Mode 4. Corresponding ITS 3.7.5, Condition A, provides additional time for repairs and testing prior to requiring entry into Mode 4 and presents the allowed time in a substantially different format than the CTS.The additional time provided by the ISTS reduces the number of unnecessary MODE changes and requests for enforcement discretion by providing added flexibility in Mode 3 to repair and test the turbine driven AFW pump following a refueling outage. This change is acceptable based upon the redundant capabilities afforded by the AFW system, the time needed to perform repairs and testing of the turbine driven pump, the reduced decay heat load following a refueling outage, and the low probability of a DBA occurring during this period that would require the operation of the turbine driven pump. This change is designated as less restrictive because additional time is allowed in the ITS to restore equipment to within the LCO limits prior to exiting the Mode of Applicability than was allowed in the CTS.More Restrictive Changies (M)M.1 Not used.BVPS Units 1 & 2 Page 6 Revision 4, 11 0/06 274 I Rev. 4, Change 2 BVPS ISTS Conversion

3.7 Plant

Systems Enclosure 3 Changes to CTS M.2 CTS surveillance 4.7.1.2.7 is revised by CTS Note 7 that states: "This surveillance is required to be performed prior to entry into MODE 2 whenever the plant has been in MODES 5 or 6 for greater than 30 continuous days." The corresponding ISTS surveillance contains a similar frequency for performance with the exception that the ISTS specifies that the surveillance is applicable after the plant has been in a"defueled" condition as well as in Modes 5 and 6. The CTS is revised to conform to the ISTS. This changes the CTS by expanding the operating conditions for which the cumulative time is tracked by CTS surveillance 4.7.1.2.7.

As such, the proposed change requires the time spent in a defueled condition be accounted for as well as the time spent in Modes 5 and 6 for determining when surveillance 4.7.1.2.7 must be performed.

The purpose of CTS surveillance 4.7.1.2.7 (and the corresponding ISTS surveillance) is to verify the normal AFW flow Path to the steam generators.

Both the ISTS and CTS surveillances require this surveillance to be performed based on how long the plant has been outside the applicable Modes for the AFW (i.e., in a condition where AFW was not required operable and may have been misaligned from its normal standby condition).

However, the ISTS surveillance includes the time spent in a defueled condition as well as in Modes 5 and 6. The proposed change is acceptable because it continues to assure the required surveillance is BVPS Units 1 & 2 Page 7 Revision 4, 10/06 275 BVPS UNITS 1 & 2 IMPROVED TECHNICAL SPECIFICATION (ITS) CONVERSION LICENSE AMENDMENT REQUEST (LAR)Nos. 296 (UNIT 1) & 1 69 (UNIT 2)REVISION 4 CHANGE 3 NRC Comment Resolution NRC Reviewer:

R. Clark Description During the review of the BVPS ITS Conversion Revision 3, BVPS received NRC comments regarding Changes 4G and 4M of Revision 3.Revision 3 Change 4G addressed a proposed clarification to the bases of ITS 3.8.9, Distribution Systems -Operating, which described a worst case loss of AC power scenario in the Actions Section of the bases. This Revision 3 change simplified the bases description to simply state that the worst case is an entire AC train de-energized.

The comment regarding this Revision 3 change was that the NUREG-1431 wording should be followed more closely (i.e., the applicable BVPS AC sources lost during the worst case accident should be listed). Therefore, the necessary bases change to conform more closely to NUREG-1431 (i.e., listing the BVPS AC sources) is included in this revision of the BVPS ITS conversion documentation.

Revision 3 Change 4M addressed changes to the Bases of ITS 3.2.1, FQ(Z). The changes updated the bases references to the W(Z) information provided in the COLR. This included changes to reduce the percent of upper and lower core regions discussed in the bases from 15% to 10%. It was identified in the review that an additional bases change (15% to 10%) was necessary due to the previously described changes to the percent of upper and lower core regions referenced in the bases. The required bases change is included in this revision of the BVPS ITS conversion documentation.

Affected Pages: The affected pages are presented in ITS Section order. The Table(s) preceding each ITS section list the affected pages by type (i.e., ITS markup, CTS markup, etc.) in that Section of the ITS. The affected page numbers listed are the ITS section specific consecutive numbers found in the lower right corner of each page.Note: Because the affected page(s) for each change were extracted from a complete ITS section electronic file, the electronic hyperlinks (created in the complete ITS section file) do not work in the collection of affected pages that follow this cover page.

BVPS UNITS 1 & 2 ITS CONVERSION LICENSE AMENDMENT REQIUEST (LAR)Nos. 296 (UNIT 1) & 169 (UNIT 2)REVISION 4 CHANGE 3 AFFECTED PAGES FOR ITS SECTION 3.2 (POWER DISTRIBUTION LIMITS)ITS SECTION 3.2 (POWER DISTRIBUTION LIMITS) INDEX OF AFFECTED PAGES ITS MARKUPS NONE ITS JFDS NONE ITS BASES MARKUPS PAGES: 49 ITS BASES JFDS NONE CTS MARKUPS NONE CTS DOCS NONE FQ(Z) (RAOC ^V(Z) Methodology)

B 3.2.1B Rev. 4, Change 3 BASES SURVEILLANCE REQUIREMENTS (continued)

Performing this Surveillance in MODE 1 prior to exceeding 75% RTP ensures that the FC(Z) limit is met when RTP is achieved, because peaking factors generally decrease as power level is increased.

The SR"If meas the maxi K(Z)l ha statemer the fact functions At each elevation/K(Z) is determin (maximu maximun since the evaluati modifyin additiona must be 2 If THERMAL POWER has been increased by > 10% RTP since the last determination of FC(Z), another evaluation of this factor is required 1121 hours0.013 days 0.311 hours 0.00185 weeks 4.265405e-4 months after achieving equilibrium conditions at this higher power level (to ensure that FC(Z) values are being reduced sufficiently with power increase to stay within the LCO limits).The Frequency of 31 EFPD is adequate to monitor the change of power distribution with core burnup because such changes are slow and well controlled when the plant is operated in accordance with the Technical Specifications (TS).Note specifies in part SR 3.2.1.2 urements indicate that mum over z of [F&Q(Z)/ The nuclear design process includes calculations performed to determine s increased

.". This that the core can be operated within the FQ(Z) limits. Because flux maps nt in the Note refers to are taken in steady state conditions, the variations in power distribution that both F~o and K are s of the axial height. resulting from normal operational maneuvers are not present in the flux applicable core map data. These variations are, however, conservatively calculated by n the ratio of F&Z) considering a wide range of unit maneuvers in normal operation.

The calculated to maximum peaking factor increase over steady state values, calculated as ne the maximum ratio im over z). If this a function of core elevation, Z, is called W(Z). Multiplying the measured m ratio has increased total peaking factor, FC(Z), by W(Z) gives the maximum FQ(Z) calculated e last set of to occur in normal operation, FWQ(Z).ons, then the Note g this SR specifies The limit with which FWQ(Z) is compared varies inversely with power above al verifications that 50% RTP and directly with the function K(Z) provided in the COLR.performed.

The W(Z is provided in the COLR for discrete core elevations.

Flux map data are typically taken for 30 to 75 core elevations.

FWQ(Z)Table evaluations are not applicable for the following axial core regions, measured in percent of core height: 10 a. Lower core region, from 0 to 'inclusive and b. Upper core region, from 100% inclusive.

The top and bottorn -l% of the core are excluded from the evaluation because of the I0 probability that these regions would be more limiting WOG STS B 3.2.1B -8 Rev. 2, 04/30/01 49 BVPS UNITS I & 2 ITS CONVERSION LICENSE AMENDMENT REQUEST (LAR)Nos. 296 (UNIT 1) & 169 (UNIT 2)REVISION 4 CHANGE 3 AFFECTED PAGES FOR ITS SECTION 3.8 (ELECTRICAL POWER SYSTEMS)ITS SECTION 3.8 (ELECTRICAL POWER LIMITS) INDEX OF AFFECTED PAGES ITS MARKUPS NONE ITS JFDS NONE ITS BASES MARKUPS PAGES: 150 ITS BASES JFDS PAGES: 171 CTS MARKUPS NONE CTS DOCS NONE Distribution Systems -Operating Rev. 4, Change 3 B 3.8.9 BASES no power from the unit and system /and station service transformers to the train ACTIONS (continued) 7 Zt and the associated DG inoperable buses, load centers, mnotor contro! centers, and distrib'-tion panels must be restored to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . '.. iv /Condition A worst scenario is one train without AC power (i.e., power to the train and the associated DG inoperable).

In this Condition, the unit is more vulnerable to a complete loss of AC power. It is, therefore, imperative that the unit operator's attention be focused on minimizing the potential for loss of power to the remaining train by stabilizing the unit, and on restoring power to the affected train. The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months time limit before requiring a unit shutdown in this Condition is acceptable because of: a. The potential for decreased safety if the unit operator's attention is diverted from the evaluations and actions necessary to restore power to the affected train, to the actions associated with taking the unit to shutdown within this time limit and b. The potential for an event in conjunction with a single failure of a redundant component in the train with AC power.The second Completion Time for Required Action A.1 establishes a limit on the maximum time allowed for any combination of required distribution subsystems to be inoperable during any single contiguous occurrence of failing to meet the LCO. If Condition A is entered while, for instance, a DC bus is inoperable and subsequently restored OPERABLE, the LCO may already have been not met for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . This could lead to a total of 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months , since initial failure of the LCO, to restore the AC distribution system. At this time, a DC circuit could again become inoperable, and AC distribution restored OPERABLE.

This could continue indefinitely.

The Completion Time allows for an exception to the normal "time zero" for beginning the allowed outage time "clock." This will result in establishing the "time zero" at the time the LCO was initially not met, instead of the time Condition A was entered. The 16 hour1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months Completion Time is an acceptable limitation on this potential to fail to meet the LCO indefinitely.

Required Action A.1 is modified by a Note that requires the applicable Conditions and Required Actions of LCO 3.8.4, "DC Sources -Operating," to be entered for DC trains made inoperable by inoperable power distribution subsystems.

This is an exception to LCO 3.0.6 and ensures the proper actions are taken for these components.

Inoperability WOG STS B 3.8.9 -4 Rev. 2, 04/30/01 150 Rev. 4, Change 3 BVPS ISTS Conversion

3.8 Electrical

Power Systems Enclosure 2 Changes to The ISTS Bases ITS 3.8.9 Distribution Systems -Operating Bases JUSTIFICATION FOR DEVIATION (JFD)1. Changes are made (additions, deletion, and or changes) to the ISTS, which reflect the plant specific nomenclature, number reference, system description, analysis, or licensing basis description.

2. Section / Chapter references are changed to reflect a unit specific reference (i.e., Accident analysis for Unit1 is Chapter 14 and for Unit 2 is Chapter 15), if applicable.

3. Editorial change made to be consistent with the ISTS writers' guide.4. Specific bus nomenclature is moved from the CTS requirements to the Bases.5. Changes are made to reflect specific listings in ITS 3.8.9 -1 Table.6. Changes to the ITS Bases are made to reflect changes in the ITS Specifications.

7. Editorial change made to the description of worse case scenario (i.e., the loss of all AC power to one electrical train). The change lists the BVPS specific AC sources for an electrical power train.BVPS Units 1 & 2 Page 10 Revision 4, 10/06 171 BVPS UNITS 1 & 2 IMPROVED TECHNICAL SPECIFICATION (ITS) CONVERSION LICENSE AMENDMENT REQUEST (LAR)NOS. 296 (UNIT 1) & 1 69 (UNIT 2)REVISION 4 CHANGE 4 Bases change.Text clarifying the Bases description of the required BVPS Unit 2 Digital Rod Position Indication (DRPI) was added to the standard Bases text.Description This change only affects the Bases of Unit 2 ITS 3.1.7.2, "Unit 2 Rod Position Indication." The LCO Bases section of ITS 3.1.7.2 states that one DRPI system must be operable to meet the LCO. This change provides an enhanced Bases description regarding the DRPI design consisting of two systems (data system A and B). In addition, the change enhances the existing description regarding the fact that either data system is capable of providing the required indication accuracy and that either data system can be used to meet the LCO requirement.

The proposed change is consistent with the ITS bases statement that one DRPI system is required operable and is consistent with the application of the current BVPS Unit 2 technical specifications for DRPI. The proposed change helps to avoid potential confusion regarding the term "DRPI System" as used in the bases for ITS 3.1.7.2.Affected Pages: The affected pages are presented in ITS Section order. The Table(s) preceding each ITS section list the affected pages by type (i.e., ITS markup, CTS markup, etc.) in that Section of the ITS. The affected page numbers listed are the ITS section specific consecutive numbers found in the lower right corner of each page.Note: Because the affected page(s) for each change were extracted from a complete ITS section electronic file, the electronic hyperlinks (created in the complete ITS section file) do not work in the collection of affected pages that follow this cover page.

BVPS UNITS I & 2 ITS CONVERSION LICENSE AMENDMENT REQUEST (LAR)NOs. 296 (UNIT 1) & 169 (UNIT 2)REVISION 4 CHANGE 4 AFFECTED PAGES FOR ITS SECTION 3.1 (REACTIVITY CONTROL SYSTEMS)ITS SECTION 3.1 (REACTIVITY CONTROL SYSTEMS) INDEX OF AFFECTED PAGES ITS MARKUPS NONE ITS JFDS NONE ITS BASES MARKUPS PAGES: 97, 98 ITS BASES JFDS NONE CTS MARKUPS NONE CTS DOCS NONE I Rev. 4, Change4 4 Rod Position Indication Unit 2 3.1 BASES BACKGROUND (continued)

+/- 4 steps, for full accuracy, and +4, -10 steps at half accuracy with data system A, and +10,-4 steps at half accuracy with data system B.each group of rods. Individual rods in a group all receive the same signal to move and should, therefore, all be at the same position indicated by the group step counter for that group. The Bank Demand Position Indication System is considered highly precise (+/- 1 step or +/- 5 / 8 inch).If a rod does not move one step for each demand pulse, the step counter will still count the pulse and incorrectly reflect the position of the rod.The fDIRPI System provides a highly accurate indication of actual control rod position, but at a lower precision than the step counters.

This system is based on inductive analog signals from a series of coils spaced along a hollow tube with a center to center distance of 3.75 inches, which is 6 steps. To increase the reliability of the system, the inductive coils are connected alternately to data system A or B. Thus, if one system fails, the [DIRPI will go on half accuracy with an effective coil spacing of 7.5 inches. which is 12 steos. Therefore, the normal indication accuracy of the fDIR'PI System is +/- 6 step, (.. 3.7-5 inches), andi the mai... imu Rco.rtaRt" is +- 12 steps (+/-- 7.5 inches).,With an indicated deviation of A2L2112 steps between the group step counterand iDeRPI, the maximum SFToeviation between actual rod position and th demand position could be ANASteps, or inches. As such onD y one data system (A or B) is required for an Ish d OPERA LE DRPI System indicating within 12 steps of the group 13.75 t step counter demand position indicator.

APPLICABLE Control and shutdown rod position accuracy is essential during power SAFETY operation.

Power peaking, ejected rod worth, or SDM limits may be ANALYSES violated in the event of a Design Basis Accident (Ref. 2), with control or shutdown rods operating outside their limits undetected.

Therefore, the acceptance criteria for rod position indication is that rod positions must be known with sufficient accuracy in order to verify the core is operating within the group sequence, overlap, design peaking limits, ejected rod worth, and with minimum SDMV (LCO 3.1.5, "Shutdown Bank Insertion Limits," and LCO 3.1.6, "Control Bank Insertion Limits").

The rod positions must also be known in order to verify the alignment limits are preserved (LCO 3.1.4, "Rod Group Alignment Limits").

Control rod positions are continuously monitored to provide operators with information that ensures the plant is operating within the bounds of the accident analysis assumptions.

The control rod position indicator channels satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).

The control rod position indicators monitor control rod position, which is an initial condition of the acciden analyses WOG STS B 3.1.7-2 Rev. 2, 04/30/01 97 Rev. 4, Change 4 Rod Position Indication (data system A or B)BASES LCO LCO 3.1 .70sspecifies that one IDIRPI Systemý/nd one Bank Demand Position Indication System be OPERABLE for each control rod. For the control rod position indicators to be OPERABLE requires meeting the SR of the LCO and the following:

a. Th System indicates within 12 steps of the group step;1unter demand position as required by LCO 3.1.4, "Rod Group required Alignment Limits," b. Forth IDIRPI System there are no failed coils, and c. The Bank Demand Indication System has been calibrated either in the fully inserted position or to the fD]RPI System.The 12 step agreement limit between the Bank Demand Position Indication System and the IDIRPI System indicates that the Bank Demand Position Indication System is adequately calibrated, and can be used for indication of the measurement of control rod bank position.A deviation of less than the allowable limit, given in LCO 3.1.4, in position indication for a single control rod, ensures high confidence that the position uncertainty of the corresponding control rod group is within the assumed values used in theonalysis (that specified control rod group insertion limits). safet These requirements ensure that control rod position indication during power operation and PHYSICS TESTS is accurate, and that design assumptions are not challenged.

OPERABILITY of the position indicator channels ensures that inoperable, misaligned, or mispositioned control rods can be detected.

Therefore, power peaking, ejected rod worth, and SDM can be controlled within acceptable limits.APPLICABILITY The requirements on the [D]RPI and step counters are only applicable in MODES 1 and 2 (consistent with LCO 3.1.4, LCO 3.1.5, and LCO 3.1.6), because these are the only MODES in which power is generated, and the OPERABILITY and alignment of rods have the potential to affect the safety of the plant. In the shutdown MODES, the OPERABILITY of the shutdown and control banks has the potential to affect the required SDM, but this effect can be compensated for by an increase in the boron concentration of the Reactor Coolant System.WOG STS B 3.1. -3 Rev. 2, 04/30/01 98 BVPS UNITS 1 & 2 IMPROVED TECHNICAL SPECIFICATION (ITS) CONVERSION LICENSE AMENDMENT REQUEST (LAR)Nos. 296 (UNIT 1) & 1 69 (UNIT 2)REVISION 4 CHANGE 5 Bases change.Proposed modifications to the standard NUREG-1431 Bases text are removed by this change.Description This change includes the deletion of three ITS Bases changes (inserted text) proposed in the BVPS ITS conversion.

In each case, the deletion of the proposed (added) Bases text results in the BVPS ITS Bases conforming more closely to the NUREG-1431 Bases text. In order to more clearly show the elimination of the proposed Bases additions, the inserted text is left intact and marked-up to show it is being deleted.In ITS 3.6.7, "Recirculation Spray System", the Bases text for Required Action E.1 was enhanced with an additional explanation of Required Action E.1. The inserted text was later determined to be unnecessary.

The original NUREG-1431 text was found to be adequate to describe the Required Action without the additional enhancement.

Therefore, the inserted text is being deleted.The Required Action A.1 bases text for ITS 3.7.7, "Component Cooling Water System", and ITS 3.7.8, "Service Water System", were original modified with additional text to enhance the explanation of the Note modifying the Required Action. The added text was later determined to be unnecessary.

The original NUREG-1431 text is adequate to describe the Action Notes.Therefore, the inserted text is being deleted.Affected Pages: The affected pages are presented in ITS Section order. The Table(s) preceding each ITS section list the affected pages by type (i.e., ITS markup, CTS markup, etc.) in that Section of the ITS. The affected page numbers listed are the ITS section specific consecutive numbers found in the lower right corner of each page.Note: Because the affected page(s) for each change were extracted from a complete ITS section electronic file, the electronic hyperlinks (created in the complete ITS section file) do not work in the collection of affected pages that follow this cover page.

BVPS UNITS I & 2 ITS CONVERSION LICENSE AMENDMENT REQUEST (LAR)Nos. 296 (UNIT 1) & 1 69 (UNIT 2)REVISION 4 CHANGE 5 AFFECTED PAGES FOR ITS SECTION 3.6 (CONTAINMENT SYSTEMS)ITS SECTION 3.6 (CONTAINMENT SYSTEMS) INDEX OF AFFECTED PAGES ITS MARKUPS NONE ITS JFDS NONE ITS BASES MARKUPS PAGES: 131 ITS BASES JFDS PAGES: 153 CTS MARKUPS NONE CTS DOCS NONE Rev. 4, Change 5 RS System (Subatmespherii)

B 3.6, 7 2 Condition E is ne due to the potential for on Conditions A Iand B to be applidin uha ean would be permitted for up ACTIONS (continued) to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months wihtre or Ssioperable in a single With three or more RS subsystems inoperable, the unit is in a condition outside the accident analysis.

Therefore, LCO 3.0.3 must be entered immediately.

I SURVEILLANCE REQUIREMENTS R 3.6.6E.1 Veri *ng that the casing cooling tank solution temperature is within e specifi tolerances provides assurance that the water injected i the suction o e outside RS pumps will increase the NPSH avail e as per design. The,4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Frequency of this SR was developedd nsidering operating expe rince related to the parameter variations d instrument drift during the ap cable MODES. Furthermore, the hour Frequency is considered adequ in view of other indications ailable in the control room, including alarms, alert the operator to a abnormal condition.

SR 3.6.6E.2 Verifying the casing cooling tan ontain borated water volume provides assurance that sufficient r is available to support the outside RS subsystem pumps during the ti e ey arerequired to operate. The 7 day Frequency of this SR was evelop d considering operating experience related to the par eter variati s and instrument drift during the applicable MODES. F ermore, the 7 d Frequency is considered adequate in view of oth indications available in e control room, including alarms, to rt the operator to an abnorm condition.

SR 3.6.6E.3 Verifying the oron concentration of the solution in the casi cooling tank provides surance that borated water added from the casing oling tank to S subsystems will not dilute the solution being recirculat in the cont nment sump. The 7 day Frequency of this SR was dvelope c sidering the known stability of stored borated water and the low robability of any sour iluting pure water.S R 3.6.6 F 7.1 Verifying the correct alignment of manual, power operated, and automatic valves, excluding check valves, in the RS System and cGaS;i tank provides assurance that the proper flow path exists for operation of the RS System. This SR does not apply to valves that are locked, sealed, or WOG STS 7-6 Rev. 2, 04/30/01 131 Rev. 4 Change 5 BVPS ISTS Conversion

3.6 Containment

Systems Enclosure 2 Changes to The ISTS Bases ITS 3.6.7 Recirculation Spray System Bases JUSTIFICATION FOR DEVIATION (JFD)1. The ISTS Bases text is revised to incorporate BVPS specific system design and safety analysis information.

For example, the QS system does not start until after the peak containment pressure is reached after the DBA and therefore this system does not affect the peak containment temperature.

2. The ISTS Bases Text for Actions and Surveillances is revised to incorporate changes made to the Actions and Surveillances in corresponding Technical Specification.

The changes made to the generic ISTS Technical Specification for the RS System are discussed in the associated JFDs in Enclosure 1.3. Not used.4. The ISTS surveillance bases discussion regarding "the need to perform the surveillance under conditions that apply during a plant outage and the potential for an unplanned transient if the surveillance were performed with the reactor at power" is revised to clarify the intent of the ISTS. The purpose of the ISTS bases discussion is to assure the surveillance is performed consistent with safe plant operation.

However, the ISTS bases text could be interpreted to require all performances of the surveillance be conducted during shutdown conditions.

The proposed change to the ISTS bases text is consistent with the NRC conclusions regarding shutdown restrictions on TS surveillances stated in Generic Letter 91-04. In Generic Letter 91-04, the NRC stated, "This restriction

[performance only during shutdown]

ensures that a surveillance would only be performed when it is consistent with safe plant operation." The Generic Letter further stated that "The staff concludes that the TS need not restrict surveillances as only being performed during shutdown.

Nevertheless, safety dictates that when refueling interval surveillances are performed during power operation, licensees give proper regard for their effect on the safe operation of the plant." As such, the proposed change to the ISTS bases incorporates a clarification to the bases that reflects the NRC guidance stated in Generic letter 91-04.BVPS Units 1 & 2 Page 11 Revision 4, 10/06 153 BVPS UNITS I & 2 ITS CONVERSION LICENSE AMENDMENT REIQUEST (LAR)Nos. 296 (UNIT 1) & 1 69 (UNIT 2)REVISION 4 CHANGE 5 AFFECTED PAGES FOR ITS SECTION 3.7 (PLANT SYSTEMS)ITS SECTION 3.7 (PLANT SYSTEMS) INDEX OF AFFECTED PAGES ITS MARKUPS NONE ITS JFDS NONE ITS BASES MARKUPS PAGES: 121,126 ITS BASES JFDS NONE CTS MARKUPS NONE CTS DOCS NONE CCW System B 3.7.7 Rev. 4, Change 5 BASES ACTIONS long as adequate CC flowto support e require ecay h removal fu ti of the RHR loop i ilable, an inopera CC in doe ot result in an i perable RHR loop.A.1 Required Action A.1 is modified by a Note indicating that the applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops -MODE 4," be entered if an inoperable CCW train results in an inoperable RHR loop.*This is an exception to LCO 3.0.6 and ensures the proper actions are taken for these components.

it to If one CCW train is inoperable, action must be taken to restor OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . In this Condition, the remaining OPERABLE CCW train is adequate to perform the heat removal function.

2" The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is reasonable based on the redundant capabilities afforded by the OPERABLE train, and the low probability of a 3 DBA occUrring during this period.z B.1 and B.2 40 Insert 3 If the CCW train cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in MODE 5 within 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, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.to the RHR heat exchangers SURVEILLANCE REQUIREMENTS CD '1-7 7 4 This SR is modified by a Note indicating that t isolation of the CCW flow to individual components may render th se components inoperable but does not affect the OPERABILITY of th CCW System.Verifying the correct alignment for man 1, power operated, and automatic valves in the CCW flow pat rovides assurance that the proper flow paths exist for CCW operation.

This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves are verified to be in the correct position prior to locking, sealing, or securing.

This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This Surveillance does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position.3 WOG STS B 3.7.7 -3 Rev. 2, 04/30/01 121 Rev. 4, Change 5 SWS B 3.7.8 BASES APPLICABLE SAFETY ANALYSES (continued)

TaS sf Cperaturer 3f C95]°F The SWS satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO Two SWS trains are required to be OPERABLE to provide the required redundancy to ensure that the system functions to remove post accident heat loads, assuming that the worst case single active failure occurs coincident with the loss of offsite power.An SWS train is considered OPERABLE during MODES 1, 2, 3, and 4 when: a. The pump is OPERABLE and_b. The associated piping, valves, j and instrumentation and controls required to perform the safety related function are OPERABLE.APPLICABILITY In MODES 1, 2, 3, and 4, the SWS is a normally operating system that is required to support the OPERABILITY of the equipment serviced by the SWS and required to be OPERABLE in these MODES.In MODES 5 and 6, the OPERABILITY requirements of the SWS are determined by the systems it supports.ACTIONS Aong as adequate SW flow support the requir heat remov function r the R te system or t ergency diesel gener r is available, n in erable SWS tr 'an does n result m'Xa operable e rgency diesel/enerator or RHR loop.WOG STS A.1 it to If one SWS train is inoperable, action must be taken to restor OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . In this Condition, the remaining 2 OPERABLE SWS train is adequate to perform the heat removal function.However, the overall reliability is reduced because a single failure in the OPERABLE SWS train could result in loss of SWS function.

Required Action A.1 is modified by two Notes. The first Note indicates that the applicable Conditions and Required Actions of LCO 3.8.1, "AC Sources -Operating," should be entered if an inoperable SWS train results in an inoperable emergency diesel generator.

The second Note indicates that the applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops -MODE 4," should be entered if an inoperable SWS train results in an inoperable decay heat removal train. This is an exception to LCO 3.0.6 and ensures the proper actions are taken for these componentslThe 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is based on the redundant I B 3.7.8 -2 Rev. 2, 04/30/01 126 BVPS UNITS 1 & 2 IMPROVED TECHNICAL SPECIFICATION (ITS) CONVERSION LICENSE AMENDMENT REQUEST (LAR)Nos. 296 (UNIT 1) & 1 69 (UNIT 2)REVISION 4 CHANGE 6 Implementation of Amendments 277 (Unit 1) and 159 (Unit 2)- Control Room Habitability Changes.NOTE: As the changes introduced by these License Amendments were anticipated in the original ITS conversion documentation, this change does not impact the final BVPS ITS. The significant impact of this change is the implementation of the revised License Amendment numbers on the affected Current Technical Specifications (CTS) pages used in the ITS conversion documentation.

Description This change updates the affected pages to the latest version of LAR numbers 325 (Unit 1) and 195 (Unit 2) as approved in License Amendments 277 (Unit 1) and 159 (Unit 2). License Amendment Numbers 277 (Unit 1) and 159 (Unit 2) were issued by the NRC on 9/25/06.Control Room Habitability License Amendment Request (LAR) Numbers 325 and 195 were submitted by FENOC letter L-05-015 dated 2/17/05. The changes requested in these LARs divided the Unit 1 and 2 Control Room Habitability Technical Specifications into two separate specifications for each unit (control room emergency ventilation system and control room emergency cooling system). In addition the changes proposed in these LARs improved consistency with the Standard Technical Specifications (NUREG-1431) and between the BVPS units.Affected Pages: The affected pages are presented in ITS Section order. The Table(s) preceding each ITS section list the affected pages by type (i.e., ITS markup, CTS markup, etc.) in that Section of the ITS. The affected page numbers listed are the ITS section specific consecutive numbers found in the lower right corner of each page.Note: Because the affected page(s) for each change were extracted from a complete ITS section electronic file, the electronic hyperlinks (created in the complete ITS section file) do not work in the collection of affected pages that follow this cover page.

BVPS UNITS I & 2 ITS CONVERSION LICENSE AMENDMENT REQ:UEST (LAR)Nos. 296 (UNIT 1) & 1 69 (UNIT 2)REVISION 4 CHANGE 6 AFFECTED PAGES FOR ITS SECTION 3.3B (INSTRUMENTATION OTHER THAN REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION)

ITS SECTION 3.3B (OTHER INSTRUMENTATION)

INDEX OF AFFECTED PAGES ITS MARKUPS NONE ITS JFDS NONE ITS BASES MARKUPS NONE ITS BASES JFDS NONE CTS MARKUPS PAGES: 157-159, 183-185, 193-195 CTS DOCS NONE ITS 3.3.3 1 Rev. 4, Change6 6 UNIT I PAGE T;5 ....3 .3 6 ( nP TABLE NOTATIONS (Not used)(2 During movement of recently irradiated fuel assemblies wit n the containment and during movement of fuel assemblies ver ecently irradiated fuel assemblies within the contain nt.(3) ye background.

(4) Dur g movement of recently irradiated fuel ass lies and durin movement of fuel assemblies over recent irradiated fuel as mblies.Changes to this Unit I material are addressed in the corresponding Unit 2 markups and DOCs ]ACTION 20 -With the nu er of channels 0 RABLE less than required by the Minim Channels 0 RABLE requirement, comply with the ACTION equiremes of Specification 3.4.6.1.ACTION 21 -This Action is not u ACTION 22 -With the number chann s OPERABLE less than required by the Minimu Channels ERABLE requirement, comply with the ACT requirements f Specification 3.9.9.ACTION 35 -With the umber of OPERABLE chan ls less than required by th Minimum Channels OPERABLE equirement, either rest e the inoperable Channel(s)

OPERABLE status wi in 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or: a) Initiate the preplanned alternate method of monitoring the appropriate parameter(s), nd b) Return the channel to OPERABLE status wit *n 30 days, or, explain in the next Annual Radioac *ve Effluent Release Report why the inoperability w not corrected in a timely manner.A 1- a) With the number of Unit 1 OPERABLE channels o~a less than the Minimum Channels 0 BLE reuirement:

1. .rify the respective t 2 control .room rad. ion monitr an is OPERABLE within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months a at ast once per 31 days.Changes to these Actions are addressed in the markups and DOCs associated with ITS 3.3.7, Control Room Emergency Ventilation Isolation BEAVER VALLEY -UNIT 1 3/4 3-35 Amendment No. 277 157 ITS 3.3.3 Rev. 4, Change 6 UNIT 1 PAGE: ACTION STATEMENT-S CTION 41 (Continued)

2. With the respective Unit 2 control r om radiation monitor train inoperable, su end all operations involving movement of r ently irradiated fuel assemblies and mov ent of fuel assemblies over recently irrad ted fuel assemblies within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and r store the Unit 1 control room radiation monitor to OPERABLE status within 7 days isolate the control room from the outsid atmosphere by closing all series air in ake and exhaust isolation dampers, unles the respective Unit 2 control room rad tion monitor train is restored to OPERABLE tatus within 7 days.b) With o Unit 1 control room radiation monitors Changes to these Actions areaaddressedeinsthe

1. Verif both Un 2 control room radiation markups and DOCs monito are ERABLE within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at associated with ITS 3.3.7, lato ý 3 as Control Room Emergency least on 31 days.Ventilation Isolation 2. With eit Unit 2 control room radiation monitor mtop able, suspend all operations 1 involvw g mov ent of recently irradiated fuel assemblie and movement of fuel ass lies over ecently irradiated fuel a emblies within hour and restore the espective Unit 1 ontrol room radiation monitor train to OPE BLE status within 7 days or isolate the c trol room f rom the outside atmosphere by dlo ing all series air intake and exhaust isolati dampers, unless the respective Unit 2 contro room radiation monitor train is restored to ERABLE status within 7 days.3. With no Unit 2 control room radiation monitors OPERABLE, immediately iso ate the combined control room by closing all series air intake and exhaust isolation dampe and be in at least HOT STANDBY within tPhe ext 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within e following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .BEAVER VALLEY -UNIT 1 3/4 3-35a Amendment No. 277 158 ITS 3.3. UNIT 1 PAGE TABLE 4.3 3 F Rev. 4, Change 6 1 7R'DIATIN

' M .. ..T f. .RI .................

SURVEIftLfCyEr R'R.E. TS......CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE TRUENT CHECK CALIBRATION TEST REQUIRED 1. ARE OITORS a. De ted Changes to this Unit I material are addressed in the corresponding Unit 2 markups and DOCs b. Contairm t i. Purge & xhaust Isolation S R M**ii. Area (RM-RM-219

& B)S R 1,2,3,& 4 c. Control Room Isolation S R M### 1,2,3,4, (RM-RM-218 A & B) and ##2. PROCESS MONITORS a. Containment

i. Gaseous Activity RCS Leak- R# M 1,2,3 & 4 age Detection (RM 215B)ii. Particulate Activity RC S R M 1,2,3 & 4 Leakage Detection (R 15A)b. Deleted** During movement recently irradiated fuel assemblies within the conta ment and durin oveme of fuel assemblies over recently irradiated fuel assemblie within the cognmtai __nt.# Surveill ce interval may be extended to the upcoming refueling outage if the i erval betwe refueling outages is greater than 18 months.## Du ng movement of recently irradiated fuel assemblies and during movement of fuel as mblies recently irradiated fuel assemblies.

AControl Room intake and exhaust isolation dampers are not actuated.BEAVER VALLEY -UNIT 1 3/4 3-36 Amendment No. 277 159 I UNIT I PAGE I Rev. 4, Change 6 New Unit 2 ITS 3.3.6 Containment Purge and Exhaust Isolation Instrumentation TABLE 3.3-6 (Continued)

TABLE NOTATIONS Unit I Licensing Requirements Manual (LRM)(I ) (ISSR# ;]-;PFi)During movement of recently irradiated fuel assemblies within the containment and during movement of fuel assemblies over recently irradiated fuel assemblies within the containment.

ent of recently irradiated fuel during movement o u over recently irradiated-es.Wit .of channels ess than required by the Min e E requirement, comply mSp 6.1.--l - 2 -ý -- 2 2 ---- ---ACTION 22 -With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.9.9.-With the number of OPERABLE channels less t EO -required by the Minimum Channels OPERABLE reuI m--ent, either restore the inoperable Channel(s o OPERABLE st- within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or: Note: These portions of the requirements are addressed in ITS 3.3.3, PAM Instrumentation, consistent with the location of this information in the ISTS. All changes to these portions of the requirements will be discussed and documented in the markups and DOCs associated with ITS 3.3.3.b) Re the channel to RABLE status within 0 days, or, explain in next Annual Radioactive Effluent Release Rep why the inoperability was not corrected in a ý mely ma nn er.I ACTIO -a With the number of Unit 1 OPERABLE c s one less than the Minimum e s OPERABLE req ent: 1. Ver he re ive Unit 2 control room radiation monitor tra s OPERABLE within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per s.Note: These portions of the requirements are addressed in ITS 3.3.7, Control Room Emergency Ventilation System (CREVS) Instrumentation consistent with the location of this information in the ISTS. All changes to these portions of the requirements will be discussed and documented in the markups and DOCs associated with ITS 3.3.7.BEAVER VALLEY -UNIT 1 3/4 3-35 Amendment No. 277** Note: These portions of the requirements are addressed in ITS 3.4.15, RCS Leakage Detection Instrumentation consistent with the location of this information in the ISTS. All changes to these portions of 183 the requirements will be discussed and documented in the markups and DOCs associated with ITS 3.4.15.

New Unit 2 ITS 3.3.6 Containment Purge and Exhaust Isolation Instrumentation r7\DT.V I 2'Z C UNIT1I PAGE Rev. 4, Change 6 ACTION STATEMENTS CTION 41 (Continued)

2. With the respective Unit 2 control om radiation monitor train inoperable, s pend all operations involving movement of r cently irradiated fuel assemblies and mov ent of fuel assemblies over recently irradj.ated fuel assemblies within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and store the Unit 1 control room radiation monitor to OPERABLE status within 7 days r isolate the control room from the outsi atmosphere by closing all series air i ake and exhaust isolation dampers, unle the respective Unit 2 control room rad'ation monitor train i3S restored to OPERABLstatus within 7 days.b) With ko Unit 1 control room radiation monitors~OPERABL : Note: These portions of the requirements are addressed in ITS 3.3.7, Control Room Emergency Ventilation System (CREVS) Instrumentation consistent with the location of this information in the ISTS. All changes to these portions of the requirements will be discussed and documented in the markups and DOCs associated with ITS 3.3.7.2. With eit er Unit 2 control room radiation monitor inop able, suspend all operations invol ng move ent of recently irradiated fuel assemblie and movement of fuel as mblies over ecently irradiated fuel semblies within hour and restore the espective Unit 1 ontrol room radiation monitor train to OPE BLE status within 7 days or isolate the co trol room from the outside atmosphere by clo ing all series air intake and exhaust isolati dampers, unless the respective Unit 2 contro room radiation monitor train is restored to ERABLE status within 7 days.3. With no Unit 2 control room radiation monitors OPERABLE, immediately iso te the combined control room by closing all series air intake and exhaust isolation damper and be in at least HOT STANDBY within the ext 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within te following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .BEAVER VALLEY -UNIT 1 3/4 3-35a Amendment No. 277 184 New Unit 2 ITS 3.3.6 Containment Purge and Exhaust Isolation Instrumentation TABLE 4.3-3 Rev. 4, Change 6 UNIT 1 PAGE ] RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS INSTRUMENT CHANNEL CHECK CHANNEL CALIBRATION CHANNEL FUNCTIONAL TEST MODES IN WHICH SURVEILLANCE REQUIRED Note: These portions of the requirements are addressed in ITS 3.3.3 PAM & ITS 3.3.7, Control Room Emergency Ventilation System (CREVS) Instrumentation consistent with the location of this information in the ISTS. All changes to these portions of the requirements will be discussed and documented in the markups and DOCs associated with ITS 3.3.3 or ITS 3.3.7.185 New ITS 3.3.7 Control Room Emergency Ventilation System (CREVS) Actuation Instrumentation Rev. 4, Change 6 UNIT TA L PA E3. .3. .6 ...........

I PNote: These portions of the requirements are addressed in ITS 3.3.6, Containment Purge and Exhaust Isolation Instrumentation consistent with the location of this information in the ISTS. All changes to these portions of the requirements will be discussed and documented in the markups and DOCs associated with ITS 3.3.6.(2 eetof recently irradiated ulwti S recent e. ...seble containment.l

., ou t- h a -kq -Trn ]in ] l-(-4-- During movement of recently irradiated fuel assemblies and Iduring movement of fuel assemblies over recently irradiated fuel assemblies.

ITS 3.3.7 Applicability

]ACTION STATEMENTS ACTION 2-1 This A4tion is net used.by the Min'i requirement, comply C 35 With the number of OPERABLE channels less an required by the Minimum Channels OPERABLE r rement, Note: These portions of the requirements are addressed in ITS 3.3.3, PAM Instrumentation, consistent with the 7 location of this information in the ISTS. All changes to these portions of the requirements will be discussed and documented in the markups and DOCs associated with ITS 3.3.3.monitoring t ropriate parameter(s), and b) R n the channel to RABLE status within 30 days, or, explain in -next Annual Radioactive Effluent Release Rep-OTt..

why the noperability was not corrected in imely manner.ACTGN- -a-)- With number of Unit 1 OPERABLE channe ne less than the Minimum Channe OPERABLE Condition A requirement:

1. Verify the r ctive 2 control room radiati monitor train is 0 RELE within ForFuelMovement ur and at least once per 31 days.RI M For Modes 1-4 only BEAVER VALLEY -UNIT 1 3/4 3-35 Amendment No. 277 Note: These portions of the requirements are addressed in ITS 3.4.15, RCS Leakage Detection Instrumentation consistent with the location of this information in the ISTS. All changes to these portions of the requirements will be discussed and documented in the markups and DOCs associated with ITS 3.4.15.193 New ITS 3.3.7 Control Room Emergency Ventilation System (CREVS) Actuation Instrumentation UNIT 1 PAGE TABLrE 3. 6 (Gntn td Rev. 4, Change 6 M9 iued)ACTION GTATEMENT L:3 7days MB ACTION 41 (Contii Action Condition A & D OEA he respective Unn t 2 controla r ol r onitor train inoperable, t spend aloe osinvolving movement irradiated 1i assembl es and movement of fuel assemblies ser rec ntl rradiated fuel assemblies withind e e and restore the Unit 1 control room adiation monitor to OPERABLE status withn 7 s or isolate the control room f/r the outs' atmosphere by closing al Jseries air intake d exhaust isolatio dampruless the r 4ectiveýnit coto omrdainmn r ain For Fuel Movement[Action Condition A &D 1 Uth no unit 1 control room radiation monitor L3 1. Ver both Unit 2 control room iation monito are OPERABLE within 1 h r and at least on per 31 days.2. With either it 2 contr room radiat'monitor inopera e, su end all ope ions involving movemen recently radiated fuel assemblies move of fuel assemblies over recen irradiated fuel assemblies wi 1in 4and restore the respective nit 1 control oom radiation monitor ain to OPERABLE sta us within 7 days isolate the control roo from the Out e atmosphere by closing all s ies air i ke and exhaust isolation dampers, less he respective Unit 2 control room radia *on monitor train is restored to OPERABLE statu within 7 days.MB no Unit 2 control room radia monitor--

PERABLE, immediately is e the combined cont room by clo g all series air intake and exha ation dampers and be in at least STAN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months a in COLD SHUTDO ithin the foll 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .R1 For Modes 1-4 only Am9 Amendment No. 277 BEAVER VALLEY -UNIT 1 3/4 3-35a 194 New ITS 3.3.7 Control Room Emergency Ventilation System (CREVS) Actuation Instrumentation

ý TABLE 4. 3 3 Rev. 4, Change 6_______RAD A+/-ON P()? F HHI+/--- O T NC RNP TATI N UNIT 1 PAGE CHANNEL CHANNEL CHANNEL FUNCTIONAL CHECK CALIBRATION TEST MODES IN WHICH SURVEILLANCE REQUIRED INSTRUMENT

1. AREA MONITORS a. Delcted Note: These portions of the requirements are addressed in ITS 3.3.3, PAM Instrumentation, consistent with the location of this information in the ISTS. All changes to these portions of the requirements will be discussed and documented in the markups and DOCs associated with ITS 3.3.3.Note; These portions of the requirements are addressed in ITS 3.3.6, Containment Purge and Exhaust Isolation Instrumentation consistent with the location of this information in the ISTS. All changes to these portions of the requirements will be discussed and documented in the markups and DOCs associated with ITS 3.3.6.I -- -- --- m7 7 i nJ 1 4 A r, R )----------

Iii. Area 0R-RM0 -A& B)-1 c. Control Room Isolation S R M## 1,2,3,4, 1(RM-RM-218 A & B) and #2. PROCESS MONITORS a a-n 'ment I Note: These portions of the requirements are addressed in ITS 3.4.15, RCS Leakage Detection Instrumentation consistent with the location of this information in the ISTS. All changes to these portions of the requirements will be discussed and documented in the markups and DOCs associated with ITS 3.4.15.ii. Part'rlt Activity RCS L~eakage Detection (KM 215A)b. Delet-ed S R# l2 ,3 & 4/** During mc QfrcntlV IIrradiated fuel assemblies wihi ., ýainme~ntan~d during movement of fuel ass radiated fuel assemblies within t h-.Surveillance interval may be extended to the upcomi between ref1iili.g

't rea er than 18 months.I## During movement of recently irradiated fuel assemblies and during movement of fuel KI assemblies over recently irradiated fuel assemblies.

1. 1. Control Room intake and exhaust isolation dampers arc nut actuated.

-- A9 BEAVER VALLEY -UNIT 1 3/4 3-36 Amendment No. 277/ITS......7 requirements.are Note: These portions of the requirements are addressed In ITS 3.3.6, Containment Purge and Exhaust Changes to these Unit I 3 ..r Isolation Instrumentation consistent with the location of this Information in the ISTS. All changes to addressed in the corresponding Unit 2 markup these portions of the requirements will be discussed and documented in the markups and DOCs associated with ITS 3.3.6.195 BVPS UNITS I & 2 ITS CONVERSION LICENSE AMENDMENT REQUEST (LAR)Nos. 296 (UNIT 1) & 1 69 (UNIT 2)REVISION 4 CHANGE 6 AFFECTED PAGES FOR ITS SECTION 3.7 (PLANT SYSTEMS)ITS SECTION 3.7 (PLANT SYSTEMS) INDEX OF AFFECTED PAGES ITS MARKUPS NONE ITS JFDS NONE ITS BASES MARKUPS NONE ITS BASES JFDS NONE CTS MARKUPS PAGES: 239-244, 244A, 245, 246 CTS DOCS NONE M Am A PLANT SYSTEMS Rev. 4, Cnang 3.7.11 1:;3/4.7.r, CONTROL ROOM EMERGENCY AIR COOLING SYSTEM (CREACS)ie LIMITING CONDITION FOR OPERATION Two CREACS trains shall be OPERABLE-.

APPLICABILITY:

MODES 1, 2, 3 and 4, and During movement assemblies, and of recently irradiated fuel" " J During movement of fuel assemblies over recently irradiated fuel assemblies.

ACTION: MODES 1, 2, 3 and 4: a.1 With one CREACS train inoperable, restore the CREACS train to ACnd OPERABLE status within 30 days or be in HOT STANDBY within A&B] the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .a-.- With two CREACS trains inoperable, enter Specification 3.0.3 Cond E immediately.

During movement of recently irradiated fuel assemblies and during movement of fuel. assemblies over recently irradiated fuel assemblies:

b-4I With one CREACS train inoperable, restore the CREACS train to COPERABLE status within 30 days or immediately place the OPERABLE CREACS train in operation or immediately suspend movement of irradiated fuel assemblies and movement of fuel assemblies ver --'-- irradiated fuel assemblies.

b--_- With two CREACS t ins inoperable, immediately suspend Cnndfl movement of i a aated fuel assemblies and movement of fuel assemblies ove irradiated fuel assemblies.

SR 3.7.11.1' SR3..11.1 A3 Unit 2 s~p~ecific Note in Cond C and D for __SURVEILLANCE REQUIREMENTS movement of recently irradiated fuel only.-.-- CREACS shall be demonstrated OPERABLE at least once per 18 months by verifying each CREACS train has the capability to remove A2 the required heat load and purge the control room atmosphere at the -assumed flow rate.* Emergency backup powe CS train is required in MODES 5, 6 a uel assemb ie eactor pressure BEAVER VALLEY -UNIT 2 3/4 7-14 Amendment No. 159 239 3.7.11IT 3..14Chne

]I IT,37.1 Rev. 4 PLANT SY__ STEM__SS/.. CONTROL ROOM EMERGENCY AIR COOLING SYSTEM (CREACS)F IIING CONDITION FOR OPERATION* Two CREACS trains shall be OPERABLE-*-.

Unit I -General Note -Specific The heat removal function of CREACS is not required OPERABLE to LCONote support fuel movement involving non-recently irradiated fuel.APPLICABILITY:

MODES 1, 2, 3 and 4, and Unit1Specific During movement of irradiated fuel assemblies, and Applicability During movement of fuel assemblies over irradiated fuel assemblies.

ACTION: Unit I specific Note in Cond CandDfor A3DE 1,2movement of irradiated fuel.MODES 1, 2, 3 and 4: a.! With one CREACS train inoperable, restore the CREACS train to OPERABLE status within 30 days or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .a-.2 With two CREACs trains inoperable, enter Specification 3.0.3 Cond E immediately.

During movement of irradiated fuel assemblies and during movement of fuel assemblies over irradiated fuel assemblies:

b--1 With one CREACS train inoperable, restore the CREACS train to OPERABLE status within 30 days or immediately place the Cond OPERABLE CREACS train in operation or immediately suspend movement of irradiated fuel assemblies and movement of fuel assemblies over irradiated fuel assemblies.

b-D -a With two CREACS trains inoperable, immediately suspend movement of irradiated fuel assemblies and movement of fuel CondD assemblies over irradiated fuel assemblies.

SURVEILLANCEREQUIREMENTSR 711.1 4.:-. CREACS shall be demonstrated OPERABLE at least once per 18 months by verifying each CREACS train has the capability to removAC the required heat load and purge the control room atmosphere at th required flow rate. J*Emergency

ýbac onl on aln is required in MODES 5, 6 and wit ass the reactor pressure-BEAVER VALLEY -UNIT 1 3/4 7-15 Amendment No. 277__Note: For Unitl1, the verification ofthe heat removal function of the CREACS L 1is not required to supportthe movement of non-recently irradiatedfuel

.240 SYSTEMS I ITS 3.7.10 A.1 Rev. 4, Change 6 Two CREVS trains shall be OPERABLE*.

General Note---------------------------

The control room boundary may be opened intermittently under administrative control.APPLICABILITY:

MODES 1, 2, 3 and 4, and During movement of recently irradiated fuel assemblies, and During movement of fuel assemblies over recently irradiated fuel assemblies.

ACTION: MODES 1, 2, 3 and 4: Cond A&C With one required CREVS train inoperable, restore the CREVS train to OPERABLE status within 7 days or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .With two required CREVS trains inoperable due to an inoperable control room boundary, restore the control room boundary to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .a-3 With two required CREVS trains inoperable for reasons other Cond than described in ACTION a.2, enter Specification 3.0.3 immediately.

7 BEAVER VALLEY -UNIT 2 3/4 7-15 Amendment No. 159 241 Rev. 4, Change 6 PLANT SYSTEMS I ITS 3.7.10 ]LIMITING CONDITION FOR OPERATION (continued)

ACTION (Continued)

During movement of recently irradiated fuel assemblies and during movement of fuel assemblies over recently irradiated fuel assemblies:

Cond A&D With one required CREVS train inoperable, restore the CREVS train to OPERABLE status within 7 days, or immediately place the OPERABLE CREVS train in the emergency pressurization mode of operation, or immediately suspend movement of recently irradiated fuel assemblies and movement of fuel assemblies over recently irradiated fuel assemblies.

With two required CREVS trains inoperable, immediately suspend movement of recently irradiated fuel assemblies and movement of fuel assemblies over recently irradiated fuel assemblies.

SURVEILLANCE REQUIREMENTS 4.7.7.1 The CREVS shall be demonstrated OPERABLE: a. Deleted.SR3.7.10.1 b- At least once per 31 days by verifying that each CREVS train operates for 2 15 minutes with the heaters in operation.

c. At least once per 18 months or (1) after each complete r rtial replacement of a HEPA filter or charcoal a orber ban or (2) after any structural maintenance o he HEPA filter r charcoal adsorber housings by: 1. Verifyin that the charcoal ads er satisfies the inplace pe tration and by- ss leakage testing acceptance cri ia of less an 0.05% when tested in accordance with I N 0-1980 while operating the CREVS train at a flo te of 800 to 1000 cfm.2. Verifying tha the HEPA ter bank satisfies the inplace p etration and by- s leakage testing accepta e criteria of less than 0. 5%6 when tested in acco ance with ANSI NS10-1980 whil operating the VS train at a flow rate of 800 to 1000 M.A.3 .Verifying a system flow rate of 800 to 10 cfm during operation of each CREVS train.BAVER VALLEY -UNIT 2 3/4 7-16 Amendment No. 159 rNOTE: The CREVS filtration requirements are moved to the Ventilation Filter Testing Program (VFTP) in the Administrative Controls Section (5.0) of the Tech Specs consistent with the location of these requirements in the ISTS.Any changes to this information will be discussed and documented in Section 5.0 of the Tech Specs.242 SUREILANC SYS EMS RMNSCniud ITS 3.7.10 I Rev. 4, Change 6 SURVEILLANCE REQUIREMENTS (Continued)

At least once per 18 months or (1) after 720 hou of trem operation, or (2) following painting, ire or chemi release in the vicinity of contro oom outside air intake while the system is operati , within 31 days after removal, bjecting the carbo ontained in at least one test canister at least o carbon samples removed f rom one of the charcoa o6rbers to a laboratory carbon sample analysis and ye yin removal efficiency of ! 99%for radioactive m yl iodide a n air f low velocity of 0.7 ft/se~c wi an inlet methyl io e concentration of 1.75 mg/ >,ý 70% relative humidity, and 0 C; other test cond' ons including test parameter tolerances all be in cordance with ASTM D3803-1989.

The carbon samp not obtained from test canisters shall be prepared by either.SR 3.7.10.2 Perform required CREVS filter testing in accordance with the Ventilation Filter Testing Program (VFTP).a) tying one entire bed from a removed adsorb~eý

ýy,!mixin th adsorbent thoroughly, ýa obtaining samples at t two inches in e~mter and with a length equal to t ikns the bed, or b) Empyn o nlsmple mo an adsorber tray, mi i g t -s r e t t oro g ly , and .n ing samp les at st two inches in diameter and wit length equal to the thickness of the bed.At least once per 18 months by: V eriin he pessre drop fo om mned HEPA filters and charcoa er banks is less than 5.6 inches Gauge while op each CREVS* at a flow rate of 800 to 1000 cfm.SR 3.7.10.3 Verifyii-g that each CREVS train actuates on a simulated or actual actuation signal.~-3. Deleted 4. Delet-ed eý. Ven the heaters dissi a s 3.87 kw and e n n tested in accordance wi -1980. -a--'

at lcast oncc every months, on a STAGGERED TEST BASIS, that each CREVS .in can maintain the control SR 3.7.10.4 room at a positive pre re of _> 1/8 inch Water Gauge re3710 relative to the outs e atmos here during operation at a flow rate of 800 1000 [BEAVER VALLEY -UNIT 2 3/4 7-17 Amendment No. 159 e NTE The CREVS filtration requirements are moved to the Ventilation Filter Testing Program (VFTP) in the Administrative Controls Section (5.0) of the Tech Specs consistent with the location of these requirements in the ISTS. 43 Any changes to this information will be discussed and documented in Section 5.0 of the Tech Specs. 1 I Rev. 4, Change 6 PLANT SYSTEMS I ITS 3.7.10 ]4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEMS (CREVS)LI TING CONDITION FOR OPERATION 3.7.7 Two CREVS trains shall be OPERABLE*:

General Note -------------------

The co trol room boundary may be opened intermi ently under administ ative control.APPLICABILITY:

MO S 1, 2, 3 and 4, and Durin movement of recent irradiated fuel assemb *es, and During m ement of fuel ssemblies over recently irradiated uel assemblies ACTION: MODES 1, 2, 3 and 4: a.l With one required CREVS in inoperable, restore the CREVS train to OPERABLE statu wit in 7 days or be in HOT STANDBY within the next 6 h rs an in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .a.2 With two requl d CREVS train inoperable due to an inoperable cont o room boundary store the control room boundary to 0 RABLE status within hours or be in HOT STANDBY with' the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the followi g 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .a.3 With tw required CREVS trains inoperable r reasons other than scribed in ACTION a.2, enter Speci ication 3.0.3 imme ately.Changes to these Unit I requirements are discussed corresponding Unit 2 TS markup Emergency power for only one CREVS train is required in MODES 6 and with no fuel assemblies in the reactor pressure vessel.BEAVER VALLEY -UNIT 1 3/4 7-16 Amendment No. 277 244 I ITS 3.7 I 0Rev. 4, Change 6 PLANT SYSTEMSCONDITION FOR OPERATION (continued)

During moveme of recently irradiated fuel assembli and during movement of fuel emblies over recently irradiate uel assemblies:

b.l With one require REVS train inope le, restore the CREVS train to OPERABLE wi in 7 da , or immediately place the OPERABLE CREVS train in ergency pressurization mode of operation, or immediat uspend movement of recently irradiated fuel asse ies and vement of fuel assemblies over recently irr ated fuel assem s.Changes to these Unit I requirements are discussed corresponding Unit 2 TS markup.2 With -o required CREVS trains inoperabe,-immediately susj nd movement of recently irradiated fuel asse ies and-ovement of fuel assemblies over recently irradiate fuel assemblies.

BEAVER VALLEY -UNIT I 3/4 7-16a Amendment No. 277 244A ITS 3.7.10 ]Rev. 4, Change 6 SURVEILLANCE REQUIREMENTS 4.7.7.1 The CREVS shall be demonstrated OPERABLE:

each A6 a. Deleted.b-. At least once per 31 days by verifying that rCREVS train SR3.7.10.-1 operates for 15 minutes with the heaters in operation.

At least once per 18 months or after every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months o system operation or (1) after each complete or part 1 replacement of a HEPA filter or charcoal adsorber ban , or after any structural maintenance on the HEPA fil er or coal adsorber housing or (3) following painti , fire or emical release in any ventilation zone com nicating with t system by: 1. Ver ying that the filtration system satisfies the in-place penetration and by-pass 1 akage testing accepta ce criteria of less than 0.0 .- when tested in accordan with ANSI N510-1980 w le operating the CREVS trai at a flow rate of 800- 1000 cfm.2. Within 31 da s after removal, subjecting the carbon contained in a least one t st canister or at least two carbon samp s removed from one of the charcoal adsorbers to a la rator carbon sample analysis and verifying a remo 1 efficiency of 99% for radioactive methyl io ne at an air flow velocity of.68 ft/sec wih an 1 methyl iodide concentration of 1.75 mg/i , 0% r ative humidity, and 30°C;other test co itions cluding test parameter tolerances sha be in acc dance with ASTM D3803-1989. The arbon samples t obtained from test canisters s 11 be prepared by ether: a) Empty g one entire bed from a removed adsorber tra mixing the adsorbent oroughly, and ob ining a sample volume equivale to at least o inches in diameter and with a le th equal to the thickness of the bed, or b) Removing a longitudinal sample from an dsorber tray using a slotted-tube sampler, mixin the adsorbent thoroughly, and obtaining a sample v ume equivalent to at least two inches in diameter d with length equal to the thickness of the bed.A.3 NOTE: The CREVS filtration requirements are moved to the Ventilation Filter Testing Program (VFTP) in the Administrative Controls Section (5.0) of the Tech Specs consistent with the location of these requirements in the ISTS.Any changes to this information will be discussed and documented in Section 5.0 of the Tech Specs.BEAVER VALLEY -UNIT 1 3/4 7-17 Amendment No. 277 245 I ITS 3.7.10 1 I Rev. 4, Change 6 1 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (continued)

BEAVER VALLEY -UNIT 1 3/4 7-18 Amendment No. 277 246 BVPS UNITS I & 2 ITS CONVERSION LICENSE AMENDMENT REIQUEST (LAR)Nos. 296 (UNIT 1) & 1 69 (UNIT 2)REVISION 4 CHANGE 6 AFFECTED PAGES FOR ITS SECTION 5.0 (ADMINISTRATIVE CONTROLS)ITS SECTION 5.0 (ADMINISTRATIVE CONTROLS)

INDEX OF AFFECTED PAGES ITS MARKUPS NONE ITS JFDS NONE ITS BASES MARKUPS NONE ITS BASES JFDS NONE CTS MARKUPS PAGES: 112-115 CTS DOCS NONE N(of Unit 2 CTS Page for ITS 5.5.7 1 Rev. 4, Change 6 PLANT SYSTEMS I ITS5.5 1 tITING CONDITION FOR OPERATION (continued)

ACTI ON Continued))TE: These requirements are contained in the Plant Systems section (3.7.10) of the Tech Specs consistent with the location these requirements in the ISTS. Changes to this information is discussed and documented in Section 3.7 of the TS.b.l With one re ired CREVS train inoperable, restore the CREVS train to OPE LE status within 7 days, or immediately place the OPERABLE CR train in emergenc pressurization mode of operation, or im d~iat ely susped movement of recently irradiated fuel asse lies and 2ovement of fuel over recently irradiate ul semblies.b.2 With two required C S ains inoperable, immediately Ees and A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) fuel filter ventilation systems for the Control Room Emergency Ventilation System (CREVS) and the Supplemental Leak Collection and Release System (SLCRS). Tests described in Specifications 5.5.7.a and 5.5.7.b shall be performed " "\ _ ,and following significant painting, fire, or chemical release in the vicinity of control room outside air intake Soperation.

[7]e- At least once per 18 months or (1) a ter each complete or L___J partial replacement of a HEPA filter/or charcoal adsorber bank, or (2) after any structural mointenance on the HEPA filter or charcoal adsorber housings hvby:-Verifying that the charcoal adsorber satisfies the inplace penetration and by-pass leakage testing acceptance criteria of less than 0.05% when tested in accordance with ANSI N510-1980 while operating the CREVS train at a flow rate of 800 to 1000 cfm.9-. Verifying that the HEPA filter bank satisfies the inplace penetration and by-pass leakage testing acceptance criteria of less than 0.05% when tested in accordance with ANSI N51r-1980 while operating the CREVS train at a flow rate of 800 to 1000 cfm.f n sy Verifying a system flow rate of 800 to 1000 cfm during operation of each CREVS train.BEAVER VALLEY r UNIT 2 3/4 7-16 Amendment No. 159 s I 112 Rev. 4, Change 6 SITS 5.5 I Unit 2 CTS Page for ITS 5.5.7 I I PLANT SYSTEMS S SURVEILLANCE REQUIREMENTS (Continued) sinfcn or :atraysrcurlmitnneo A'\ , the charcoal adsorber bank housing d-. At least once per 18 months or (1) af ter 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation, or (2) following

pai ting, fire or"- Ichemical release in the vicinity of cont ol room outside air intakes while the system is operatin , within 31 days cafter removal, inud ing tes paraer contained in at least accordae wither or at least two carbon samples removedt obtane f the charcoal adsorbers ts a laboratory carbon sml niss and verifying a removal efficiency of ý! 99%fo aiatve methyl iodide at an air flow velocity of VFP 07f/e ith an inlet methyl iodide concentration of 1.75 mg/mt 3 > 70% relative humidity, and 30 C; other test including test parameter tolerances shall be in faccordance with ASTM D3803-1989.

I The carbon samples not 0obtained from test canisters shall be prepared by either: Sample obtained in accordance with Regulatory Guide 1.52, Revision 2, or using slotted tube samples in accordance with ANSI N509-1980.

a) Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed, or b) Emptying a longitudinal sample from an adsorber tra mixing the adsorbent thoroughly, and obtaining sampl s at least two inches in diameter and with a len th equal to the thickness of the bed.At least once per 18 months by: 26 usingaslotedtubesampler

1. Verifying that the pressure drop for the combined HEPA filters and charcoal adsorber banks is less than 5.6 inches Water Gauge while operating each CREVS train at a flow rate of 800 to 1000 cfm.2. Ve that each CREVS train ac s on a ti si i 0 c simulated o al actuation si mul ate tua on 3. Deleted:4. eted.5-. Verifying that the heaters dissipate at least 3.87 kw and not exceeding 5.50 kw when tested in accordance with ANSI N510-1980.

f. By ye at least once every 36 months AGGERED TEST BASIS, tha REVS tra' maintain the control room at a positi ssure > 1/8 inch Water Gaug relati e outside atmo sphere dunin ation at ow rate of 800 to 1000 cfm.The provisions of SR 3.0.2 and SR 3.0.3 are applicable.

A17 p 1 I Amendment No. 159\I BEAVER VALLEY -UNIT 2 3/4 7-17 NOTE: These requirements are contained in the Plant Systems section (3.7.10) of the Tech Specs consistent with the location of these requirements in the ISTS. Changes to this information is discussed and documented in Section 3.7 of the Tech Specs.1 113 ITI LITS5.5i Unit 1 TS Page for ITS 5.5.7 PLANT SYSTEMS Rev. 4, Change 6 At least once per 18 mon s or after every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation or (1) after each complete or partial replacement of a HEPA fi er or charcoal adsorber bank, or (2) after any structural maintenance on the HEPA filter or charcoal adsorber housi g or (3) followingainting, tire A19t or 'hemicl releae in i. ' t C I------- ------- --1 ........by...a..I with the sys tem by: significant 5-aI-5 Verifying that the filtration system satisfies t e in-55.7aand.5.7.b place penetration and by-pass leakage testing acceptance criteria of less than 0.05% when tested in Aaccordance with ANSI N510-1980 while operating the CREVS train at a flow rate of 800 -1000 cfm.w" VFTP LA1I [E.68 Wihn 1days after removal, Isubjecting the carbo contained in at least one test canister' or at leas two car'boon Asamples removed from one of the charcoal adsorbers to a laboratory carbon sample analysis and verifying a removal efficiency of 2: 99% f or radioactive methyl iodine at an air f low velocity offt/sec with an inlet methyl iodide concentration of 1.75 mg/m , 70% relative humidity, and 300C;other test conditions including test parameter toeane shall be in accordance with ASTM D3803-A17 tolerances shall be in accordance with ASTM D3803-1989. j The carbon samples not obtained from test canisters shall be prepared by either: a) Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly, and obtaining a sample volume equivalent to at least two inches in diameter and with a length equal to the thickness of the bed, or b) Removing a longitudinal sample from an adsorber tray using a slotted-tube sampler, mixing the adsorbent thoroughly, and obtaining a sample volume equivalent to at least two inches in diameter and with length equal to the thickness of the bed.BEAVER VALLEY -UNIT 1 3/4 7-17 Amendment No. 277 114 Rev. 4, Change 6 PLANT SYSTEMS UnitEIRCTS Page for ITS 5.5.7 SURVEILLANCE REQUIREMENTS (continued) 5.5.7.a and 5.57.b Verifying a system flow rate of 800 -1000 cfm during operation of the CREVS train.d, At least once per 18 months by:.4 Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is < 6 inches Water Gauge while operating the CREVS train at a flow rate of 800 -1000 cfm.2. that each CREVS train act n /a simulated or a tuatio .4-.7.Verifying that the heaters dissipate at least 3.87 kw and not exceeding 5.50 kw when tested in accordance with ANSI N510-1980.

B-e y verifying at least once every 36 months onaRE ASIS, each CREVS train can maintain ontrol room at a pos .e ressure of ý!1/ n aer Gauge relative of

... _W'r A7The provisions of SR 3.0.2 and SR 3.0.3 are applicable.

1 NOTE: These requirements are contained in the Plant Systems section (3.7.10) of the Tech Specs consistent with the location of these requirements in the ISTS. Changes to this information is discussed and documented in Section 3.7 of the Tech Specs.BEAVER VALLEY -UNIT 1 3/4 7-18 Amendment No. 277 115 BVPS UNITS 1 & 2 IMPROVED TECHNICAL SPECIFICATION (ITS) CONVERSION LICENSE AMENDMENT REQ1UEST (LAR)Nos. 296 (UNIT 1) & 1 69 (UNIT 2)REVISION 4 CHANGE 7 NRC Comment Resolution NRC Reviewer:

R. Clark/K. Wood Description This change addresses the NRC comment regarding the retention of the BVPS current Technical Specification (CTS) requirements for Residual Heat Removal (RHR) flow in the BVPS specific ITS. As such, this change incorporates the BVPS CTS 3000 gpm and 1000 gpm flow requirements for RHR in ITS 3.9.4, "RHR and Coolant Circulation

-High Water Level" and ITS 3.9.5, "RHR and Coolant Circulation

-Low Water Level". This change requires the addition of (1) new Surveillance Requirement (SR) in ITS 3.9.4 and (2) new SRs in ITS 3.9.5. The existing ITS SRs are renumbered to accommodate the new SRs. Additionally, new Bases text is added to the ITS to discuss the new SR requirements.

In addition, this change addresses the NRC reviewer concern regarding a BVPS specific addition to the Bases explanation of ITS 3.9.5 LCO Note 2. The additional text added by BVPS to enhance the description of the LCO Note 2 is deleted from the ITS Bases. The deletion of this additional text makes the resulting BVPS ITS 3.9.5 LCO Bases text more consistent with the corresponding NUREG-1431 Bases text.Affected Pages: The affected pages are presented in ITS Section order. The Table(s) preceding each ITS section list the affected pages by type (i.e., ITS markup, CTS markup, etc.) in that Section of the ITS. The affected page numbers listed are the ITS section specific consecutive numbers found in the lower right corner of each page.Note: Because the affected page(s) for each change were extracted from a complete ITS section electronic file, the electronic hyperlinks (created in the complete ITS section file) do not work in the collection of affected pages that follow this cover page.

BVPS UNITS 1 & 2 ITS CONVERSION LICENSE AMENDMENT REQUEST (LAR)Nos. 296 (UNIT 1) & 169 (UNIT 2)REVISION 4 CHANGE 7 AFFECTED PAGES FOR ITS SECTION 3.9 (REFUELING OPERATIONS)

ITS SECTION 3.9 (REFUELING OPERATIONS)

INDEX OF AFFECTED PAGES ITS MARKUPS PAGES: 12,16 ITS JFDS NONE ITS BASES MARKUPS PAGES: 55, 58, 60, 61, 73 ITS BASES JFDS NONE CTS MARKUPS PAGES: 88, 90, 91 CTS DOCS PAGES: 123, 124,127-130, 132 I Rev. 4, Change 7 1 RHR and Coolant Circulation

-High Water Level 3.9. 5 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME A.4 Close equipment hatch 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and secure with [four]bolts.AND A.5 Close one door in each air 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months lock.AND A.6.1 Close each penetration 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months providing direct access from the containment atmosphere to the outside atmosphere with a manual or automatic isolation valve, blind flange, or equivalent.

OR A.6.2 Verify each penetration is 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months capable of being closed by an OPERABLE Containment Purge and Exhaust Isolation System.SURVEILLANCE REQUIREMENTS (SURVEILLANCE FREQUENCY SR 3.9.5.4* Verify one RHR loop is in operation

..... 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months reactor coolant at a flow rate of ! [2800] gpm.SR 3.9-4.1 ---------

.-------NOTE-Only required to be met prior to the start ofand during operations that cause the introduction of coolant into the RCS with boron concentration less than that required to meet the minimum required boron concentration of LCO 3.9.1.Verify one RHR loop is circulating reactor coolant at a flow rate of > 3000 gpm. I hour c'WOG STS From CTS 3)2 4l Rev. 2, 04/30/01 12 Rev. 4 Change7 RHR and Coolant Circulation

-Low Water Level 3.9.6 wxz ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B.5.2 Verify each penetration is 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months capable of being closed by an OPERABLE Containment Purge and Exhaust Isolation System.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9 ý. Verify one RHR loop is in operation and GicGu!ating 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 5 ~~reac-tor c-oolant at a flowf mate of ý [2800] gpm.SR 3 ..6\X*f Vericorrect breaker alignment and indicated power 7 days ava, ble to the required RHR pump that is not in opration.

2 I I From CTS 11 *NOTE Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a required pump is not in operation.

.............................................................................................................

SR 3.9.5.2-NOTE-Only required to be met when RCS water level is >three feet below the reactor vessel flange.Verify one RHR loop is circulating reactor coolant at a 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months flow rate of > 1000 gpm.SR 3.9.5.1-NOTE-Only required to be met prior to the start of and during operations that cause the introduction of coolant into the RCS with boron concentration less than that required to meet the minimum required boron concentration of LCO 3.9.1.Verify one RHR loop is circulating reactor coolant at a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months flow rate of _ 3000 gpm.WOG STS From CTS 3 5Li Rev. 2, 04/30/01 16 Rev. 4, Change 7 RHR and Coolant Circulation

-High Water Level BE3.BASES ACTIONS (continued) penetrations are either closed or can be closed so that the dose limits are not exceeded.The Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months allows fixing of most RHR problems and is reasonable, based on the low probability of the coolant boiling in that time.zJ41 SURVEILLANCE REQUIREMENT 6 R;EFERENCES SR 3 This Surveillance demonstrates that the RHR loop is in operation and circulating reactor coolant. ow rate is determined by the ;e necessary to provide sufficient deca t remov I ity and to prevent thermal and boron stratificatio re. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient, con e flow, tempera ump control, and alarm indi available to the operator in the con m for m g the RHR System.L 1. ýFSAR, Seet*(-)R 5.5.7-1 t Unit 1 UFSAR, Appendix 1A, "1971 AEC General Design Criteria Conformance".

Unit 2 UFSAR, Section 3.1, "Conformance with NRC General Design Criteria".

Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal and to prevent thermal and boron stratification in the core. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient considering other indications and alarms available to the operator in the control room to monitor RHR loop performance.

SR 3.9.4.1 This surveillance verifies that the RHR loop is circulating reactor coolant at the specified flow rate of > 3,000 gpm.The verification of the specified flow rate provides additional assurance of adequate forced circulation and mixing of the RCS during operations involving the addition of coolant into the RCS with a boron concentration that is less than required to maintain the required SHUTDOWN MARGIN.The Surveillance is modified by a Note that specifies the conditions under which the surveillance is required to be met. The Note states that the Surveillance is only required to be met prior to the start of (i.e., within an hour before)and during operations that cause the introduction of coolant into the RCS with boron concentration less than that required to meet the minimum required boron concentration of LCO 3.9.1. The Frequency of one hour ensures the required RHR flow is maintained during the specified operations and has been shown to be adequate by operating experience.

1CTS WOG STS ,-4 Rev. 2, 04/30/01 55 RHR and Coolant Circulation

-Low Water Level Rev. 4, Change 7 BASES LCO (continued)

TSTF-21 1 The circumstances for stopping both RHR pumps are to be limited to situations when the outage time is short [and the core outlet temperature is maintained

> 10 degrees F below saturation temperature].

The Note prohibits boron dilution or draining operations when RHR forced flow is stopped.~_~__________-___

_ 'hikS ILG"CO is modified by a oIlte that allows one RHR loop to be n the testing results in inoperable for a period of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months provided the other loop is OPERABLE the r uired RHR loop ing and in operation.

Prior to declaring the loop inoperable, consideration rendere -operab The should be given to the existing plant configuration.

This consideration remaining 1BLE RHR should include that the core time to boil is short, there is no draining loop is ade at provide operation to further reduce RCS water level and that the capability exists the e allowed by No to inject borated water into the reactor vessel. This permits surveillance tests to be performed n the inoperable loop during a time when these te ase pi normal recirculation TSTF-21 & NUREG 1431, Rev.:: 3 e r n il. RSnra eiclý: An OPERABLE RHR loop consist of an RHR pump, heat exchanger, valves, piping, instruments and ontrols to ensure an PERABLE flow path and to determine the temperature.

Th flow path starts in INSERT one of the RCS hot legs and is returned to the RCS cold legs.APPLICABILITY Two RHR loops are required to be OPERABLE, and one RHR loop must be in operation in MODE 6, with the water level < 23 ft above the top of the reactor vessel flange, to provide decay heat removal. Requirements for the RHR System in other MODES are covered by LCOs in Section 3.4, Reactor Coolant System (RCS), and S ,,tion 3.5, Er G,,c.Core Cooling Systems (ECGS). RHR loop requirements in MODE 6 with the water level > 23 ft are located in LCO 3.9.* Residual Heat Removal (RHR) and Coolant Circulation

-High Water Level.ACTIONS A.1 and A.2 If less than the required number of RHR loops are OPERABLE, action shall be immediately initiated and continued until the RHR loop is restored to OPERABLE status and to operation or until > 23 ft of water level is established above the reactor vessel flange. When the water level is> 23 ft above the reactor vessel flange, the Applicability changes to that of LCO 3.9.k, and only one RHR loop is required to be OPERABLE and in operation.

An immediate Completion Time is necessary for an operator to initiate c.rective actions.r WOG STS B3 -2 Rev. 2, 04/30/01 58 I Rev. 4 Change 7 1 RHR a INSERT SR 3.9.5.1 & SR 3.9.5.2 BASES CTS info nd Coolant Circulation

-Low Water Levelf SURVEILLANCE REQUIREMENTS SR 3.9.~A..J3 This Surveillance demonstrates that one RHR loop is in operation and circulating reactor coolant. rY6-ýw rate is determine neesr opoid ufc eoal capalbility and to and oronstraifiatio in he In addition, during operation of the RHR loop with the water level in the vicinity of the reactor vessel nozzles, the RHR pump suction requirements must be met. The I Frequenzy fo' s sufficient, considering the flo I v.;tpm in thp rnntrnl rnnm Verification that the required pump is OPERABLE ensures that an additional RCS r) RHR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation.

Verification is performed by verifying proper breaker alignment and power available to the required pump. The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience.

REFERENCES 1.A FSAR, Section 15.5.7].3 The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient considering other indications and alarms available to the operator in the control room to monitor RHR loop performance.

This SR is modified by a Note that states the SR is not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a required pump is not in operation.

5 Unit 1 UFSAR, Appendix 1A, "1971 AEC General Design Criteria Conformance".

Unit 2 UFSAR, Section 3.1, "Conformance with NRC General Design Criteria".

Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal and to prevent thermal and boron stratification in the core.WOG STS B 3.Qr -4 5 Rev. 2, 04/30/01 60 Rev. 4, Change 7 INSERT ITS 3.9.5 BASES ADDITION (FROM TSTF-21 & NUREG1431, Rev. 3)Both RHR pumps may be aligned to the Refueling Water Storage Tank to support -/draining the refueling cavity or for performance of required testing.INSERT ITS 3.9.5 BASES FOR NEW SRs 3.9.5.1 & 3.9.5.2 (Rev. 4 Change 7)SR 3.9.5.1 This surveillance verifies that the RHR loop is circulating reactor coolant at the specified flow rate of > 3,000 gpm. The verification of the specified flow rate provides additional assurance of adequate forced circulation and mixing of the RCS during operations involving the addition of coolant into the RCS with a boron concentration that is less than required to maintain the required SHUTDOWN MARGIN.The Surveillance is modified by a Note that specifies the conditions under which the surveillance is required to be met. The Note states that the Surveillance is only required to be met prior to the start of (i.e., within an hour before) and during operations that cause the introduction of coolant into the RCS with boron concentration less than that required to meet the minimum required boron concentration of LCO 3.9.1. The Frequency of one hour ensures the required RHR flow is maintained during the specified operations and has been shown to be adequate by operating experience.

SR 3.9.5.2 This surveillance verifies that the RHR loop is circulating reactor coolant at the specified flow rate of >_ 1,000 gpm. The verification of the specified flow rate provides additional assurance of adequate forced circulation of the RCS when the RCS water level is more than three feet below the reactor vessel flange.The Surveillance is modified by a Note that specifies the conditions under which the surveillance is required to be met. The Note states that the Surveillance is only required to be met when RCS water level is > three feet below the reactor vessel flange. The Frequency of six hours ensures the required RHR flow is maintained during low water level conditions and has been shown to be adequate by operating experience.

61 I Rev. 4, Change 7 BVPS ISTS Conversion

3.9 Refueling

Operations-nclosure 2 Changes to The ISTS Bases ITS 3.9.5 RHR and Coolant Circulation

-Low Water Level Bases JUSTIFICATION FOR DEVIATION (JFD)1. Not used.2. The purpose of the ISTS Action B.5.2 is to ensure the capability to close the containment purge and exhaust penetrations is available to minimize the release of radioactive material should the RHR requirements continue to not be met and boiling occurs in the core. The proposed change recognizes that under these circumstances the closure of one valve in the purge and exhaust penetrations is sufficient and may be accomplished automatically by radiation monitor actuation or by manual action from the control room.Manual isolation of one valve in the purge and exhaust penetrations is acceptable considering that the potential release from heating up the RCS is not the same as the immediate and large release assumed in a design basis fuel handling accident.Therefore, considering the nature of the potential release from heating the RCS, the heightened awareness of the operations staff during a loss of RHR and the radiation monitor indications available to the control room, sufficient information and time BVPS Units 1 & 2 Page 8 Revision 4, 10/06 73 REFUELIN OPEATON Re.4,Cane7High Water LevelI Rev. 4, Change7 344-.-9-.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATIO LCO 39 A2 LIMITING CONDITION FOR OPERATION 1-3.9.8.1 At least one residual heat removal (RHR) loop shall be OPERABLE and in operation.Notes ] /with the water level _> 23 ft above the top of reactor vessel flange.APPLICABILITY:

MODE 6.A3 LI MI ACTION: Replace with Action A.2 Replace with Action A.1 Add Action A.3 A4, a--. With less than one residual heat removal loop i operation, Replace with an increase in tH -ereactor decay heat loaaor reduction ctadspAer oE in boron concentration of the Reactor Coolant System.ose all containment penetrations providing direct access i A orom the containment atmosphere to the outside atmosphere b L2 : 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . LI(A9 4Vrb-. oThe residual heat removal loop may be removed from* )I operation for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months uring the/ performance of CORE ALTERATIONS in the vicin ty of thet bA o reactor pressure vessel hf t legs. ange SThe residual heat removal loop may be removed from q I operation for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period du~ring theIperformance of Ultrasonic In-service Inspection insi e the] reactor vessel nozzles% ... there is at leas... .t VA7 eify above the tneRH lopiiclne rearter vesto f !anc/ S R 3.9 .4 .2 V A L L E Y ...I.23 / 9 -8 A m e n d m n t N .. 9,rov ided no operations are permitted that would cause introduction of coolant into the RCS]w'-ýiith bo ron concentration less than that required to meet the minimum required boron C___EL3 E REQUIREMENTS raio of LO 3..1 (A9 )...:\Verify at least one residual heat removal loop iýs ýin.operation nd circuilating reactor= coolant a: ----eey1husl'SR 3.9.5.2 A flow rate _> 1000 gpm tw 4prs~hift when the Reactorl iCoolant System is inarde netr adte*SR 3.94. flow rate >! 3000 gpm prAt Qh tr f and oncepr SR3.9.5.--l-G. r.. C .. ... 1an, syster ...... r l- el...- is -than three feet e beow the reactor vessel flange.Verfy oeRHR loop is circulating reactor coolant at B2EAVER VALLEY -UNIT 2 3/4 9-8 Amendment No. 97 Note: Only required to be met prior to the start of and during operations that cause the introduction of coolant into the RCS with boron concentration less than that required to meet the minimum required boron concentration of LCO 3.9.1.

ITS 3.9.4 UNIT 1 PAGE REFUELING OPERATIONS Rev. 4, Change 7 3/4.9.8 RESIDUAL HEAT REMOVAL A]L ITING CONDITION FOR OPERATION ND COOLANT CIRC ULATION M4 O -PERABLE and 3.9.8. At least one residual heat removal (RHR) loop sha be in operati.APPLICABIL TY: MODE 6#.Note that changes to this portion of the Unit I TS are addressed in the markup of the Unit 2 TS.a. With ss than one residual heat remova loop in operation, except provided below, suspend al operations involving an increa e in the reactor decay he load or a reduction**

in boron oncentration of the eactor Coolant System.Close all c tainment penetratio s providing direct access from the cont inment atmospher to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .b. The residual he remov loop may be removed from operation for up t 1 ht r per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period during the performance of CORE ERATIONS in the vicinity of the reactor pressure vesse (hot) legs.c. The residual heat remo al loop may be removed from operation for up o 4 hou per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period during the performance of U trasonic In service Inspection inside the reactor vessel ozzles provid there is at least 23 feet of water abov the top of the r ctor vessel flange.d. The provisons of Specification

3. .3 are not applicable.

SURVEILLANCE REQ REMENTS i 4.9.8.1 Verify at least one residual heat rem val loop is in operationad circulating reactor coolant at: a. A flow rate 1000 gpm twice per shift when the Reactor Coolant System is in a reduced inventory conditI n*.A flow rate 3000 gpm prior to the start of and nce per hour during a reduction**

in the Reactor Coolant System boron concentration.

_______________

SR 3.9.4.1 & SR 3.9.5.1/

The Reactor Coolant System water level is lower than three fe t below the reactor vessel flange.I* urposes of this specification, the addition of water to tT e s-ý not constitute a r or dilution in RCS boron concentration r e boron concentration of the A8borated wat a ded is greater t inimum required isfy the requirements of Specification 3.9.1 for# With fuel in the vessel. Definitionof~l A9 BEAVER VALLEY -UNIT 1 3/4 9-8 Amendment No. 150 Note: Only required to be met prior to the start of and during operations that cause the introduction of coolant into the RCS with boron concentration less than that required to meet the minimum required boron concentration of LCO 3.9.1. 90 ITS 3.9.5 REFUELING OPERATIONS Rev. 4, Change7 A.2 LOW WATER LEVEL and Coolant Circulation

-LCO 3.9.5 In LIMITING CONDITION FOR OPERATION 3.9.8.2 -- Q TwualHeat Removal (RHR)loops shall be OPERABL/*APPLICABILITY:

MODE 6 when the water level above the top of the reactor pressure vessel flange is less than 23 feet.ACTION: immediately OR A.2 Initiate action to establish

_> 23 ft of water above the top of the reactor vessel flange immediately.

FAction A-l --less than the rqui~e RR loops OPERABLE, initiate corrective actio to return the re ed RHR loops A3[ to OPERABLE status a-s soa b-.--Tho provisions of Specification 3.0.3 are not applicable.

Insert Condition B "No RHR Loop in operation" and Actions B. I through B.5.1 I- nsert Action B.5.2 L3 UTRVEPITTT.T, ANKC'E PRTETIR.MENThIST' A2.L2-he ic e pFU 4 re-' P.-;q i Fi 11;4 1 U.- A # -, PrAF=1;rA I I F:)F:ýnq nnL!nAnTy

4-P4--f-4--

A a C SR 3.9.5.1 Note: Only required to be met prior to the start of and during operations that cause the introduction of coolant into the RCS with boron concentration less than that required to meet the minimum required boron concentration of LCO 3.9.1.Verify one RHR loop is circulating reactor coolant at a flow rate of >_ 3000 gpm every hour.SR 3.9.5.2-NOTE-Only required to be met when RCS water level is > three feet below the reactor vessel flange.1 Verify one RHR loop is circulating reacto a flow rate of > 1000 gpm every 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .r coolant at SR 3.9.5.3 Verify one RHR loop is in operation every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .Fror n CTS 3.9.8.1 Mjio NOTE Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the required pump is not in operation.

SR 3.9.5.4 Verify correct breaker alignment and indicated power available to the required RHR pump that is not in operation every 7 days.1 1-- 4 ------1 ýP-- ---I, nun leep.A5 BEAVER VALLEY -UNIT 2 3/4 9-9 91 I Rev. 4, Change 7 BVPS ISTS Conversion

3.9 Refueling

Operations Enclosure 3 Changes to CTS minimize any potential radioactive release. This change is acceptable because the Actions continue to provide for containment closure. Closure of the Purge and Exhaust System isolation valves may be accomplished automatically by radiation monitor actuation or by manual action from the control room.Manual isolation of the purge and exhaust penetrations is acceptable considering that the potential release from heating up the RCS is not the same as the immediate and large release assumed in a design basis fuel handling accident.

Therefore, considering the nature of the potential release from heating the RCS, the heightened awareness of the operations staff during a loss of RHR and the radiation monitor indications available to the control room, sufficient information and time would be available to enable the operators to manually isolate the purge and exhaust penetrations if it becomes necessary to prevent any significant radioactive release. Although not specified as an Action in the TS, the BVPS purge exhaust may also be lined up to the filtration system in the Supplemental Leak Collection and Release System (SLCRS) which could provide a defense in depth capability to mitigate any release.The proposed change only allows for a delay in isolating the containment purge and exhaust system. This delay may be necessary for continued habitability of the containment and restoration of RHR (BVPS RHR pumps are inside containment).

As such, the proposed change continues to provide adequate assurance that the containment will be closed and that the release of radioactive material would be minimized should boiling occur in the core. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS.L.3 Not used.BVPS Units 1 & 2 Page 22 Revision 4, 10/06 123 Rev. 4, Change 7 1 BVPS ISTS Conversion

3.9 Refueling

Operations Enclosure 3 Changes to CTS LA4 Not used.BVPS Units 1 & 2 Page 23 Revision 4, 10/06 124 Rev. 4, Change 7 BVPS ISTS Conversion

3.9 Refueling

Operations Enclosure 3 Changes to CTS M.4 Unit 1 only. Unit 1 CTS 3.9.8.1 LCO only specifies that an RHR loop be in operation.

The corresponding ITS 3.9.4 also requires that the RHR loop be operable.

The CTS LCO is revised to conform to the ITS LCO. This changes the CTS LCO by adding the requirement for the RHR loop to be operable as well as in operation.

The proposed change is acceptable because it provides additional assurance that the required cooling function of the RHR system is available when necessary to assure adequate core cooling. The additional requirement of being operable ensures the required RHR is capable of performing its intended safety function.Therefore, the proposed change continues to assure the plant is operated in a safe manner without adversely impacting equipment availability or operational resources.

This change is designated as more restrictive because it imposes a new LCO requirement that must be met in Mode 6.Removed Detail Changes (LA)LA. 1 (Type 3 -Removing Procedural Details for Meeting TS Requirements and Related Reporting Requirements)

CTS 3.9.8.1 Action b states that "the residual heat removal loop may be removed from operation for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period during the performance of CORE ALTERATIONS in the vicinity of the reactor pressure vessel hot legs. The corresponding requirement in the ISTS states that"the residual heat removal loop may be not in operation for < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period....

The CTS is revised to conform to the ISTS. This changes the CTS by moving the procedural details describing what may be accomplished during the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months exception to the Bases.The removal of these details for performing this CTS Action from the TS is acceptable because this type of information is not necessary to be included in the TS to provide adequate protection of public health and safety. The TS still retain the fundamental requirement that the RHR loop must be maintained operable and in operation.

The affected CTS Action provides a 1-hour exception to the LCO requirement that may be used once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The CTS exception remains unchanged and continues to limit the time the RHR loop may not be in operation.

Therefore, the removal of the information describing how the approved exception will be used does not reduce the CTS requirements.

Also, this change is acceptable because these types of procedural details will be adequately controlled in the TS Bases. The TS Bases Control Program specified in Section 5 of the TS controls changes to the Bases. This program provides for the evaluation of Bases changes in accordance with 10 CFR 50.59 to ensure the Bases are properly controlled.

This change is designated as a less restrictive removal of detail change because procedural details describing the TS requirements are being removed from the Technical Specifications.

LA.2 Not used.BVPS Units 1 & 2 Page 26 Revision 4, 10/06 127 I Rev. 4, Change 7 BVPS ISTS Conversion

3.9 Refueling

Operations Enclosure 3 Changes to CTS LA.3 (Type 2 -Removing Descriptions of System Operation)

CTS 4.9.8.1 requires that the RHR loop be verified in operation and circulating reactor coolant. The corresponding ISTS surveillance only requires that the RHR loop be verified in operation.

This changes the CTS by moving the descriptive detail of "circulating reactor coolant" from the CTS surveillance to the Bases.The removal of these details, which are related to system operation, from the TS is acceptable because this type of information is not necessary to be included in the BVPS Units 1 & 2 Page 27 Revision 4, 10/06 128 I Rev. 4, Change7 7 BVPS ISTS Conversion

3.9 Refueling

Operations Enclosure 3 Changes to CTS TS to provide adequate protection of public health and safety. The ITS still retains the requirement to verify the RHR loop is in operation.

The ITS Bases document associated with the TS requirement contains an adequate description of the systems required operable and provides sufficient background information to explain why the TS requirements are necessary.

As such, the descriptive detail in the CTS is no longer required.

Also, this change is acceptable because the removed information will be adequately controlled in the TS Bases. Changes to the Bases are controlled by the TS Bases Control Program in Section 5 of the TS. This program provides for the evaluation of Bases changes in accordance with 10 CFR 50.59 to ensure the Bases are properly controlled and that prior NRC review and approval is obtained when required.

This change is designated as a less restrictive removal of detail change because information relating to system operation is being removed from the Technical Specifications.

LA.4 Not used.BVPS Units 1 & 2 Page 28 Revision 4, 10/06 129 Rev. 4, Change 7 BVPS ISTS Conversion

3.9 Refueling

Operations Enclosure 3 Changes to CTS Administrative Changes (A)A.1 In the conversion of the Beaver Valley Power Station current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering or order, etc.) are made to obtain consistency with NUREG-1431, Rev.2, "Standard Technical Specifications-Westinghouse Plants" (ISTS).Due to the large number of such changes, A.1 changes may not always be marked on each CTS page. Marked or unmarked, all A.1 changes are identified by a single annotation of A.1 at the top of the first page of each CTS. These changes include all non-technical modifications of requirements to provide consistency with the ISTS, including all significant format changes made to update the older NUREG-0452 Technical Specification presentation to the ISTS format. This type of change is also associated with the movement of requirements within the Technical Specifications and with changes made to the presentation of Technical Specifications requirements to combine the Unit 1 and 2 Technical Specifications into one document and highlight the differences between the Unit 1 and 2 requirements.

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS requirements.

A.2 The title and applicability of CTS 3/4.9.8.1 are revised consistent with the ISTS. The phrase "high water level" replaces "all water levels" in the title and the applicability is revised to add "with the water level greater than or equal to 23 feet above the reactor vessel flange" to Mode 6. This revision is consistent with the bases for the CTS. When the water level is equal to or greater than 23 feet above the reactor vessel flange a large heat sink is available for core cooling and adequate time exists to restore cooling if the single required RHR loop fails. Since CTS 3.9.8.2 (Low water level) is applicable when the water level is less than 23 feet above the reactor vessel flange it requires two operable RHR loops. As such, the appropriate applicability for CTS 3.9.8.1 (one RHR Loop required) is with the water level equal to or greater than 23 feet above the reactor vessel flange.The incorporation of this change provides a clear separation between the Applicabilities of the two RHR Loop TS (CTS 3.9.8.1 and 3.9.8.2) based on water level. The change requires that some of the CTS 3.9.8.1 requirements previously applicable at all water levels be repeated in CTS 3.9.8.2 (low water level) but does result in more clear RHR TS requirements for each water level. In addition, the proposed change does not result in a technical change to the RHR requirements for each water level. The proposed change is acceptable because it conforms to the ISTS, is consistent with the CTS Bases, and does not introduce a technical change.As such, this change is considered administrative.

A.3 CTS 3.9.8.1, Action a, states, in part, that with less than one RHR loop in operation, suspend all operations involving an increase in the reactor decay heat load. The BVPS Units 1 & 2 Page 29 Revision 4, 10/06 130 SRev. 4, Change 7 BVPS ISTS Conversion

3.9 Refueling

Operations Enclosure 3 Changes to CTS A.6 CTS 3.9.8.1 Action c states "The residual heat removal loop may be removed from operation for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period during the performance of Ultrasonic In-service Inspection inside the reactor vessel nozzles provided there is at least 23 feet of water above the top of the reactor vessel flange." There is no corresponding ISTS requirement for this BVPS specific allowance.

However, when moved into the corresponding ISTS, this CTS requirement is changed by deleting "provided there is at least 23 feet of water above the top of the reactor vessel flange" from the requirement.

This change is acceptable based on the ISTS organization of the RHR/Coolant Circulation TS. The corresponding ISTS to CTS 3.9.8.1 is only applicable when the water level is _ 23 feet above the reactor vessel flange. Therefore, the limitation in CTS Action c regarding the water level is no longer required.

The proposed change represents a change in the organization of the TS requirements that does not introduce a technical change to the CTS requirements.

As this change is one of format and presentation it is considered administrative.

A.7 CTS 3.9.8.1 Action d states, "The provisions of Specification 3.0.3 are not applicable".

The corresponding ISTS does not contain this provision.

ISTS LCO 3.0.3 states, "LCO 3.0.3 is only applicable in MODES 1, 2, 3, and 4." Therefore, in the ISTS, an exception to the provisions of LCO 3.0.3 in Mode 6 is not required.This changes the CTS by deleting the exception to the provisions of 3.0.3 in CTS 3.9.8.1.This change is acceptable because the ISTS LCO 3.0.3 applicability (Modes 1-4) is consistent with the CTS 3.9.8.1 exception to 3.0.3. This change is designated as administrative because it does not result in technical changes to the CTS.A.8 Unit 1 only. A note that specifies "with fuel in the vessel" modifies the Unit 1 CTS 3.9.8.1 applicability of Mode 6. The corresponding ISTS applicability does not contain this note. The CTS footnote modifying the applicability is deleted consistent with the ISTS.In the ISTS the definition of Mode in Section 1.0 includes the requirement that fuel is in the vessel. Therefore, the Unit 1 CTS footnote specifying "with fuel in the vessel" is no longer required to modify the applicable Mode. Given the ISTS definition of Mode includes the requirement for fuel in the vessel, the deletion of the footnote does not introduce a technical change to the CTS. As such this change is considered administrative.

A.9 CTS Surveillances 4.9.8.1a and b verify a specific RHR flow under certain operating conditions (i.e., low inventory and during dilution operations).

The CTS surveillances are reformatted to ITS standards and retained in the ITS as SR 3.9.4.1, SR 3.9.5.1, and SR 3.9.5.2 (CTS Surveillance 4.9.8.1.b is moved to both ITS SR 3.9.4.1 and SR 3.9.5.1).

The ISTS does not have surveillances that correspond to the CTS surveillances being retained.The proposed change includes modifying the frequency of CTS 4.9.8.1 a from twice per shift to once every 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and converting parts of the affected CTS surveillance text into surveillance notes consistent with the typical notes used in the ITS. In addition the proposed change includes modifying the Unit 1 CTS ** footnote to surveillance 4.9.8.1 .b to be consistent with the corresponding ITS footnotes and adopting the modified footnote for Unit 2. The Unit 1 footnote provides a clarification of the term "dilution operations" that is consistent with the ITS. The proposed changes retain the CTS requirements in the ITS and do not introduce a technical change to the CTS. Therefore, this change is designated as an administrative change.BVPS Units 1 & 2 Page 31 Revision 4, 10/06 132 BVPS UNITS 1 & 2 IMPROVED TECHNICAL SPECIFICATION (ITS) CONVERSION LICENSE AMENDMENT REQUEST (LAR)Nos. 296 (UNIT 1) & 169 (UNIT 2)REVISION 4 CHANGE 8 Implementation of Unit 2 Amendment 157 -Battery Charger Upgrades NOTE: As the changes from this License Amendment were anticipated in the original ITS conversion documentation, this change does not impact the final BVPS ITS. The significant impact of this change is the implementation of the revised License Amendment numbers on the affected Current Technical Specifications (CTS) pages used in the ITS conversion documentation.

Description This change updates the affected pages to the latest version of LAR number 202 as approved in Unit 2 Amendment number 157.Unit 2 License Amendment number 157 issued by the NRC on 8/28/06 incorporates changes requested in Unit 2 LAR number 202 submitted by FENOC letter L-05-157 dated 10/14/05.

The requested changes were necessary to implement station battery charger upgrades, including the installation of new battery chargers to replace the rectifiers currently referenced in the Unit 2 Technical Specifications.

The amendment affects current technical specifications 3/4.8.2.3 and 3/4.8.2.4.

The primary effect of this change is to replace the word "rectifier' with the word"charger" in the affected Unit 2 technical specifications.

Affected Pages: The affected pages are presented in ITS Section order. The Table(s) preceding each ITS section list the affected pages by type (i.e., ITS markup, CTS markup, etc.) in that Section of the ITS. The affected page numbers listed are the ITS section specific consecutive numbers found in the lower right corner of each page.Note: Because the affected page(s) for each change were extracted from a complete ITS section electronic file, the electronic hyperlinks (created in the complete ITS section file) do not work in the collection of affected pages that follow this cover page.

BVPS UNITS 1 & 2 ITS CONVERSION LICENSE AMENDMENT REQUEST (LAR)Nos. 296 (UNIT 1) & 169 (UNIT 2)REVISION 4 CHANGE 8 AFFECTED PAGES FOR ITS SECTION 3.8 (ELECTRICAL POWER SYSTEMS)ITS SECTION 1.0 (USE & APPLICATION)

INDEX OF AFFECTED PAGES ITS MARKUPS NONE ITS JFDS NONE ITS BASES MARKUPS NONE ITS BASES JFDS NONE CTS MARKUPS PAGES: 202, 210 CTS DOCS NONE Rev. 4, Change 8 ECO LECTRICAL POWER SYSTEMS App -C. DISTRIBUTION

-OPERATING ITS 3.8.4, 3.8.6, NoteC arer 2-2 & 2-4.38.LIMITING CONDITY: -ES ,P,3ATINd L4 ai A4 C :L Insert LCO, Applicability, and Action Note for"L T ---I -3.8.4 Bases CO3.8.2.3 The folnlin D.C. reus trainr shall be energized eand3.8.9 AbOPERABLE:

st s LwI ti Bases/V~f power subsystems LC& hoRAIN "As (or beiaconsisting oT 125-voS t D thbusses No u L(2 ig D.C. J attery aCnks -n LCO chargers 2-1 8.CO3.8.4 i"B" opurple) consistA:rP9_of Jnl LCO 3.8.9 c rgers 2-2 & 2-4. APPLICABILITY

/XQýES 1, 2, 3 and ortwo. I -,I]-4L4 "[on the same train MI_:-: q~~~nsert LCO 3.8.9 ConditionC

/scnCopein LCO 3.8.4 1a. one the Irequired btey banks~inoperabie, restore ctions the inoperable battery bank to OPERABLE status within 2 B, C, &u D hours or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ' Iand in COLD.SýýTKWNýwiti 1 h olwn 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . -PActions C & D With one I 'qui'e~ fui capacity, ie ,cor .0.i ae s are charger Action A atlate pr8husteefe.i n BEAVER VALLEY -UNIT 2 202 Rev. 4 Change 8 ELECTRIC POWER SYSTEMS P.C. DISTRIBUTION

-SHUTDOWN 3 8.5 LIMITING CONDITION FOR OPERATION ITS 3.8.5 &3.8.10 I I A2 I .-Y-'L~ 1 3.8.2.4 As a minimum, one of e following trains of D.C.cetrieal cidipmcnt-a.. -------shal be OPERABLE and energized

-LAI the peefiedmaner.LI a.fA3 LCO 3.8.10 1. 125 volt D.C. busses Ne. 2 2 & 2 3, and 2. 12S velt D.C. battery banks 2 2: & 2 4 and eharger'2 i* & 2 3+.Train ,,lhp,, le eesstn .... the ... el-l e w in... : b.APPLICABILITY:

MODES 5 and 6, and Action Note: During movement of recently irradiated fuel LCO 3.0.3 is assemblies, and not applicable.

During movement of fuel assemblies over recently irradiated fuel assemblies.

irradiated fuel assemblies.

dditions that could result in loss of ACTION/ L31 requilred SDM or boron concentration.

e othre above required train of D.C. elec ical equipment and busses not fully OPERABLE, immediately u s d operation involving CORE ATRATIONS, positive c e~js, moveent of!rcnty irradiated fuel assemblies and mo'llment of fuel assemblies irradiated fuel assemblies.1 nitiate corrective action'to/

restore roe required train of D.C. el'ectr qal equipment and busses to A OPERABLE status as soon as possible. SR3810.1 1 L2 Declare affected required SR 3.8.10.1 M1 feature(s) inoperable, or~ T IP ? T T.TI XMt]C 'T' p TTT Dt'M pQ correct voltage.8.2.4.1 The above required 125-volt D.C. bus train shall be T eetermined OPERABLE and -n-rgized at least once per 7 days by erifying correct breaker a ignment and indicat-d pawor availability.

4.8.2.4.2 The above requi ed 125-volt battery bank and chargersbe demonstrated OP RABLE per Surveillance Requirement LA 1 L4 A7 eharger romoeved from sorvieo faer maintonaneo.t SRs 3.8.4.1, 3.8.4.2, 3.8.4.3 Note to SR 3.8.5.1.The following SRs are not required to be performed:

SR 3.8.4.2 and 3.8.4.3.BEAVER VALLEY -UNIT 2 3/4 8-12 Amendment No. 157 210 BVPS UNITS 1 & 2 IMPROVED TECHNICAL SPECIFICATION (ITS) CONVERSION LICENSE AMENDMENT REQUEST (LAR)Nos. 296 (UNIT 1) & 169 (UNIT 2)REVISION 4 CHANGE 9 Minor format and editorial revisions made to be more consistent with the standard Technical Specification (NUR EG-1431) format.Description This change includes the revision of several Applicability statements to eliminate the word "and" from the applicability consistent with the NUREG-1431 Applicability format. In addition, this change includes the correction of a minor editorial error in NUREG-1431.

The Applicability format of NUREG-1431 requires that each separate applicability statement be listed as a separate line item in the Applicability section of the ITS without the use of the word"and" to connect multiple Applicability statements.

In some cases, the BVPS ITS included the word "and" in applicability statements to connect multiple Applicability statements.

This change revises those Applicability statements to delete the word "and" consistent with the format of NUREG-1431.

In NUREG-1431, ISTS 3.7.8, Service Water System, the word "Conditions" was inadvertently left out of Note 1 in Required Action A.1. Similar to Note 2 in Required Action A.1, this change inserts the word "Conditions" in Note 1. This change is necessary to make the Action Note consistent with other NUREG-1431 Action Notes and for the Note to be complete and unambiguous.

Affected Pages: The affected pages are presented in ITS Section order. The Table(s) preceding each ITS section list the affected pages by type (i.e., ITS markup, CTS markup, etc.) in that Section of the ITS. The affected page numbers listed are the ITS section specific consecutive numbers found in the lower right corner of each page.Note: Because the affected page(s) for each change were extracted from a complete ITS section electronic file, the electronic hyperlinks (created in the complete ITS section file) do not work in the collection of affected pages that follow this cover page.

BVPS UNITS I & 2 ITS CONVERSION LICENSE AMENDMENT REQUEST (LAR)Nos. 296 (UNIT 1) & 1 69 (UNIT 2)REVISION 4 CHANGE 9 AFFECTED PAGES FOR ITS SECTION 3.3B (INSTRUMENTATION OTHER THAN REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION)

ITS SECTION 3.3B (OTHER INSTRUMENTATION)

INDEX OF AFFECTED PAGES ITS MARKUPS PAGES: 14 ITS JFDS NONE ITS BASES MARKUPS NONE ITS BASES JFDS NONE CTS MARKUPS NONE CTS DOCS NONE Rev. 4, Change 9 ,.Containment Purge and Exhaust Isolation Instrumentation U n it 2 ---N O T E ----- --------------3. 3 .6 3.3 INSTRUMENTATION 3.3.6 *Containment Purge and Exhaust Isolation Unit2 t LCO 3.3.6 The Containment Purge and Exhaust Isolation Function in Table 3.3.6-1 shall be OPERABLE.During movement of recently irra within the containment, ad During movement of fuel assemb APPLICABILITY:

..irradiated fuel assemblies wi 2 ation for each diated fuel assemblies lies over recently thin the containment.

I ACTIONS-NOTE -Separate Condition entry is allowed for each Function.CONDITION REQUIRED ACTION COMPLETION TIME A. One radiation monitoring A.1 Restore the affected 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months channel inoperable, channel to OPERABLE status.-NOTE -Ony a i ale in M D 1, , or 4.One or more with one or more manual automatic actuation trains inoperable.

OR Two or more radiation monitoring channels iinoperable.

OR Required tin and asso ;ied Completion T"eof Condition A not B.1 Enter applicable Conditions and Required Actions of LCO 3.6.3, "Containment Isolation Valves," for containment purge and exhaust isolation valvesl-ýmaeinoperable b,ýisoation instrur Immediately 7* 3 NýWOG STS 3.3.6 -1 Rev. 2, 04/30/01 14 BVPS UNITS 1 & 2 ITS CONVERSION LICENSE AMENDMENT REIQUEST (LAR)Nos. 296 (UNIT 1) & 1 69 (UNIT 2)REVISION 4 CHANGE 9 AFFECTED PAGES FOR ITS SECTION 3.7 (PLANT SYSTEMS)ITS SECTION 3.7 (PLANT SYSTEMS) INDEX OF AFFECTED PAGES ITS MARKUPS PAGES: 24, 29 ITS JFDS NONE ITS BASES MARKUPS NONE ITS BASES JFDS NONE CTS MARKUPS NONE CTS DOCS NONE SWS 3.7.8 Rev. 4 Change 9 3.7 PLANT SYSTEMS 3.7.8 Service Water System (SWS)LCO 3.7.8 Two SWS trains shall be OPERABLE.APPLICABILITY:

MODES 1, 2, 3, and 4.ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SWS train A.1 ------------

inoperable.

-NOTES -Conditions

1. Enter applicabl and Required Actions of LCO 3.8.1, "AC Sources -Operating," for emergency diesel generator made inoperable by SWS.2. Enter applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops -MODE 4," for residual heat removal loops made inoperable by SWS.Restore SWS train to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OPERABLE status.B. Required Action and B.1 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. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months WOG STS 3.7.8- 1 Rev. 2, 04/30/01 24 Rev. 4, Change9 9 [ CREVS F 3.7.10 SVentilation

3.7 PLANT

SYSTEMS 3.7.10 Control Room Emergency.Flitation System E )LCO 3.7.10 Two trains shall be OPERABLE.-NOTE -The control room boundary may be opened intermittently under administrative control.------- -an-------

------------------------

MODES 1, 2, , 4,[5--ap~d-6I, During movement of [recently]

irradiated fuel assemblie

..< During movement of fuel assemblies over recently irradiated fuel assemblies.

APPLICABILITY:

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One --rain A1 train to 7 days inoperable.

required CREVS B Two trains B.1 Restore control room 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months inoperable due to boundary to OPERABLE inoperable control room status.boundary in MODE 1, 2, 3, or 4.C. Required Action and C.1 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 or B AND not met in MODE 1, 2, 3, or 4. C.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months WOG STS 3.7.10 -1 Rev. 2, 04/30/01 29 BVPS UNITS 1 & 2 ITS CONVERSION LICENSE AMENDMENT REQUEST (LAR)Nos. 296 (UNIT 1) & 1 69 (UNIT 2)REVISION 4 CHANGE 9 AFFECTED PAGES FOR ITS SECTION 3.9 (REFUELING OPERATIONS)

ITS SECTION 3.7 (PLANT SYSTEMS) INDEX OF AFFECTED PAGES ITS MARKUPS PAGES: 9, 18 ITS JFDS NONE ITS BASES MARKUPS NONE ITS BASES JFDS NONE CTS MARKUPS NONE CTS DOGS NONE Rev. 4, Change 9]Containment Penetrations 13.0 3.9 REFUELING OPERATIONS 3.9.4+v Containment Penetrations LCO 3.9. The containment penetrations shall be in the following status: a. The equipment is hatch closed and held in place by [four] bots ý2 b. One door in each air lock is closed, and c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere is either: 1. Closed by a manual or automatic isolation valve, blind flange, or equivalent or 2. Capable of being closed by an OPERABLE Containment Purge and Exhaust Isolation System. ,*,,,,'1 and 3. n&ly. The Containment Purge and Exhaust System penetrations may be open when the system airflow is exhausted to an OPERABLE filtered Supplemental Leak Collection and Release System train./0-- NOTE -Penetration flow path(s) pro irect om the containment atmosphere to the outside ere m isolated under administrative APPLICABILITY:

During movement of [recently]

irradiated fuel assemblies within containment-.

During movement of fuel assemblies over recently irradiated fuel assemblies within containment-I ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Suspend movement of Immediately containment trecently]-irradiated fuel penetrations not in assemblies within required status. containment.

I I WOG STS 394-1 Rev. 2, 04/30/01 9 Refueling Cavity Water Level Rev. 4, Change 9 39-6 3.9 REFUELING OPERATIONS 3.9.* VRefueling LCO 3.9.7-Cavity Water Level 1 Refueling cavity water level shall be maintained

_ 23 ft above the top of reactor vessel flange. During nad movement of fuel assemblies During movement of irradiated fuel assemblies within con tainment over irradiated fuel assemblies within the REQUIRED ACTION COMPLETION containment APPLICABILITY:

ACTIONS CONDITION A. Refueling cavity water level not within limit.A.1 Suspend movement of irradiated fuel assemblies within containment.

Immediately 9 AND A.2 Suspend movement of fuel assemblies over irradiated fuel assemblies within containment.

Immediately I SURVEILLANCE REQUIREMENTS

_ _ _k SURVEILLANCE FREQUENCY SR 3.911 Verify refueling cavity water level is > 23 ft above the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months top of reactor vessel flange.WOG STS 3.9.ý6 Rev. 2, 04/30/01 18 BVPS UNITS 1 & 2 IMPROVED TECHNICAL SPECIFICATION (ITS) CONVERSION LICENSE AMENDMENT REQUEST (LAR)Nos. 296 (UNIT 1) & 1 69 (UNIT 2)REVISION 4 CHANGE 10 Implementation of Amendments 276 (Unit 1) and 158 (Unit 2) -Technical Specification Task Force (TSTF) Number 449, Steam Generator Program.Description This change updates the affected ITS conversion documentation pages to the latest version of LAR numbers 324 (Unit 1) and196 (unit 2) as approved in License Amendments 276 (Unit 1)and 158 (Unit 2).License Amendment numbers 276 (Unit 1) and 158 (Unit 2) were issued 9/7/06. These license amendments approve changes requested in LARs 324 (unit 1) and 196 (Unit 2) which were submitted by FENOC Letter L-05-144 dated 11/7/05. License Amendments 276 and 158 implement approved TSTF-449.

TSTF 449 revises the definition of Leakage, introduces a new ITS LCO 3.4.20, Steam Generator Tube Integrity, revises ITS 3.4.13, Operational Leakage, revises Specification 5.5.5, SG Program, and Revises 5.6.6, SG Tube Inspection Report among other changes. The ITS conversion documentation was revised to be as close as possible to the approved current BVPS Technical Specifications (CTS) and Bases implementing License Amendments 276 and 158. As such, this change includes the addition of specific BVPS CTS information to the Bases as well as minor editorial (e.g., punctuation) changes to conform to the corresponding approved CTS.Affected Pages: The affected pages are presented in ITS Section order. The Table(s) preceding each ITS section list the affected pages by type (i.e., ITS markup, CTS markup, etc.) in that Section of the ITS. The affected page numbers listed are the ITS section specific consecutive numbers found in the lower right corner of each page.Note: Because the affected page(s) for each change were extracted from a complete ITS section electronic file, the electronic hyperlinks (created in the complete ITS section file) do not work in the collection of affected pages that follow this cover page.

BVPS UNITS 1 & 2 ITS CONVERSION LICENSE AMENDMENT (LAR)NDS. 296 (UNIT 1) & 169 (UNIT 2)REVISION 4 CHANGE 10 AFFECTED PAGES FOR ITS SECTION 1.0 (USE AND APPLICATION)

ITS SECTION 1.0 (USE & APPLICATION)

INDEX OF AFFECTED PAGES ITS MARKUPS NONE ITS JFDS NONE ITS BASES MARKUPS NONE ITS BASES JFDS NONE CTS MARKUPS PAGES: 41 CTS DOCS NONE 1.0 Use and Application

1.1 Definitions

DEFINITIONS Rev. 4 Change 10 1 CORE ALTERATION 1.12 CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe positionD SHUTDOWN MARGIN [:aonrl.RCs conro subeC CtAeal 4.43 SHUTDOWN MARGIh shal be the instantaneous

/amount of re lctivity by which the react or isior would be subcriti al from itV present condition assuming all full lenefgth rod clusterlassemblies

-(shut-dewn and control) are fully inserted except for the single red clustr asseembly of highest reactivity worth which is assumed to be fully withdrawn.

I ith all RCCAs verified fully inserted by two independent means, it is not RCCA LEAKAGE Lnecessary to account for a stuck RCCA in the SDM calculation.

With any RCCA not capable of being fully inserted, the reactivity

.:-14 LEAKAGE shall be: worth of the RCCA must be accounted for in the determination of A.9 SDM. and a. Identified LEAKAGE b. In MODES I and 2, the fuel and moderator tcmpcraturcs A.O10[ are changed to the nominal zero power dlesi~gn lIevel .C=1. LEAKAGE, atfrom pump seals or valve packing (except reactor coolant pu seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank-, 4u 2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be aressure goundary LEAKAGE, or (RCS)3. Reactor Coolant System LEAKAGE through a steam generator to the gecondary

ystem (primary to secondary LEAKAGE).b. Unidentified LEAKAGE IC Unidentified LEAKACE shall bc ,ii LEAKAGE (except Feactor colant pum.p seal water injection or leakoff) that is not identified LEAKAGE--c. Pressure Boundary LEAKAGE RCS Prcssurc Boundary LEAKAGE shall bc LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in a Reactor Coolant Systcm component body, pipe wall or vessel wall.BEAVER VALLEY -UNIT 2 1-3 Amendment No. 158 41 BVPS UNITS I & 2 ITS CONVERSION LICENSE AMENDMENT REIQUEST (LAR)NoS. 296 (UNIT 1) & 169 (UNIT 2)REVISION 4 CHANGE 10 AFFECTED PAGES FOR ITS SECTION 3.4 (REACTOR COOLANT SYSTEM)ITS SECTION 3.4 (REACTOR COOLANT SYSTEM) INDEX OF AFFECTED PAGES ITS MARKUPS NONE ITS JFDS NONE ITS BASES MARKUPS PAGES: 163, 166, 167, 196A, 196H ITS BASES JFDS NONE CTS MARKUPS PAGES: 248, 266, 267 CTS DOCS NONE Rev. 4, Change 10 RCS Operational LEAKAGE B 3.4.13 BASES APPLICABLE SAFETY ANALYSES Except for primary to secondary LEAKAGE, the safety analyses do not address operational LEAKAGE. However, other operational LEAKAGE is related to the safety analyses for LOCA; the amount of leakage can affect the probability of such an event. The sGafety analysis for an event"re, ulting in steam discharge to the atmosphere assumes a 1 gpm prim to s ndary LEAKAGE as the initial condition.

Primary to ondary LEAKAGE is a factor in the dose rele s outside containment re ting from a steam line break (SLB) ac ident. To a lesser extent, other cidents or transients involve s ondary steam release to the atmosph such as a steam ge ator tube rupture (SGTR). The leakage cont inates the se dary fluid.The FSAR (Ref. 3) analysis for S ssumes the contaminated secondary fluid is only briefly rel sed safety valves and the majority is steamed to the condenser he 1 gpm pi ary to secondary LEAKAGE is relatively i nsequential.

The SLBis more iting for site radiation releases.

T safety analysis forthe ident assumes 1 gpm primary to seconda AKAGE in one gen or as an initial condition.

The dose consequences sulting from SLB accident are well within the limits defined in 10 CFR 0 or tstaff approved licensing basis (i.e., a small fraction of these limits).The RCS operational LEAKAGE satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO From CTS Bases Should pressure boundary LEAKAGE occur through a component which can be isolated from the balance of the Reactor Coolant System, plant operation may continue provided the leaking component is promptly isolated from the Reactor Coolant System since isolation removes the source of potential failure.RCS operational LEAKAGE shall be limited to: a. Pressure Boundary LEAKAGE No pressure boundary LEAKAGE is allowed, being indicative of material deterioration.

LEAKAGE of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher LEAKAGE. Violation of this LCO could result in continued degradation of the RCPB. LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. ,,\b. Unidentified LEAKAGE One gallon per minute (gpm) of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment sump level monitoring equipment can detect within a reasonable time period. Violation of this LCO could WOG STS B 3.4.13 -2 Rev. 2, 04/30/01 163 Rev. 4 change 10 & 12 RCS Operational LEAKAGE I B 3.4.13 SURVEILLANCE REQUIREMENTS (continued)

The Surveillanceis modified 1states bytwo Notes TSTF-449, Rev.4 ,The RCS water inventory b nceLmu be met with the reacto at steady' Rev 4 state operating conditions.

Note isa that this SR is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishing steady------_ state operation.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance provides sufficient time to collect 10 112 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.and process all necessary data after stable plant conditions are established.

Steady state operation is required to perform a proper inventory balance since calculations during maneuvering are not useful. For RCS operational LEAKAGE determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeu and letdown, and RCP seal injection and return flows. s, instrumentation An early warning of pressure bounn ry LEAKAGE or unidentified LEAKAGE is provided by the a systems that monitor the containment atmosphere radioactivity and the containment sump level. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. These leakage detection systems are specified in LCO 3.4.15, "RCS Leakage Detection Instrumentation."[TSTF-449, Rev. 4 The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months INSERT 3 recognizes th accidents.

SR 3.4.13.2 Frequency is a reasonable interval to trend LEAKAGE and ie importance of early leakage detection in the prevention of This ides the means necessary to determine SG OP Y in an operational E he requirement to rate SG tube integrity in accordance withjth nerator Tube Surveillance Program emphasi importance of S~tu "t , even though this .ance cannot be performed at normal operating condtos REFERENCES !0FR 5,.A ppfenailX , -'.911 UFSAR Section 4.2.7.1 (Unit 1) and UFSAR Section 5.2.5 (Unit 2).NRC Generic Letter 95-05: Voltage-Based Repair 3. F , Criteria For Westinghouse Steam Generator Tubes Affected By Outside Diameter Stress Corrosion Cracking Unit I UFSAR Appendix 1A, "1971 AEC General Design Criteria Conformance" and Unit 2 UFSAR Section 3.1, "Conformance with U.S. Nuclear Regulatory Commission General Design Criteria" NEI 97-06, "Steam Generator Program Guidelines." EPRI, "Pressurized Water Reactor Primary-to-Secondary Leak Guidelines." 3 WOG STS / B 3.4.13 -5 TSTF-449, Rev. 4 ]Rev. 2, 04/30/01 166 I Rev. 4, Change 10 & 12 BASES INSERTS FOR 3.4.13 1. Primary to secondary LEAKAGE is a factor in the dose assessment of accidents or 10 transients that involve secondary steam release to the atmosphere, such as a main steam line break (MSLB), a locked rotor accident (LRA), a Loss of AC Power (LACP), a Control Rod Ejection Accident (CREA) and to a lesser extent, a Steam Generator Tube Rupture (SGTR). The leakage contaminates the secondary fluid. The limit on the primary to secondary LEAKAGE ensures that the dose contribution at the site boundary from tube 10 leakage following such accidents are limited to appropriate fractions of the 10 CFR 50.67 limit of 25 Rem TEDE as allowed by Regulatory Guide 1.183. The limit on the primary to secondary leakage also ensures that the dose contribution from tube leakage in the control room is limited to the 10 CFR 50.67 limit of 5 Rem TEDE. Among all of the analyses that release primary side activity to the environment via tube leakage, the MSLB is of particular concern because the ruptured main steam line provides a pathway to release the primary to secondary leakage directly to the environment without dilution in the secondary fluid.For Unit 1, the safety analysis for an event resulting in steam discharge to the atmosphere conservatively assumes that primary to secondary LEAKAGE from all steam generators is 450 gallons per day (gpd) (i.e., 150 gpd per steam generator) or increases to 450 gpd as a 12 result of accident induced conditions.

Currently, the Unit 1 safety analyses do not specifically assume additional primary to secondary LEAKAGE due to accident induced conditions.

For Unit 2, due to adoption of the voltage based steam generator tube repair criteria per guidance provided by Generic Letter 95-05 (Reference 3), the safety analysis for an event resulting in steam discharge to the atmosphere conservatively assumes that primary to secondary LEAKAGE from all steam generators is 450 gallons per day (gpd) (i.e., 150 gpd per steam generator) or increases to 450 gpd as a result of accident induced conditions for 12 all accidents other than the MSLB. Currently, the Unit 2 MSLB safety analysis is the only analysis that specifically assumes additional primary to secondary LEAKAGE due to accident induced conditions.

The Unit 2 dose consequences associated with the MSLB addresses an additional 2.1 gpm 10 leakage, which, per GL 95-05, is postulated to occur (via pre-existing tube defects) as a 49, Iresult of the rapid depressurization of the secondary side due to the MLSB, and the consequent high differential pressure across the faulted steam generator.

The maximum allowed Unit 2 total accident induced leakage is 2.4 gpm. 10 2. 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.3. This SR verifies that primary to secondary LEAKAGE is less than or equal to 150 gallons 10 per day through any one SG. Satisfying the primary to secondary LEAKAGE limit ensures that the operational LEAKAGE performance criterion in the Steam Generator Program is met. If this SR is not met, compliance with LCO 3.4.20, "Steam Generator Tube Integrity," 167 Rev. 4, Change 10 SG Tube Integrity B 3.4.20 B 3.4 REACTOR COOLANT SYSTEM (RCS)B 3.4.20 Steam Generator (SG) Tube Integrity BASES BACKGROUND Steam generator (SG) tubes are small diameter, thin walled tubes that carry primary coolant through the primary to secondary heat exchangers.

The SG tubes have a number of important safety functions.

Steam generator tubes are an integral part of the reactor coolant pressure boundary (RCPB) and, as such, are relied on to maintain the primary system's pressure and inventory.

The SG tubes isolate the radioactive fission products in the primary coolant from the secondary system. In addition, as part of the RCPB, the SG tubes are unique in that they act as the heat transfer surface between the primary and secondary systems to remove heat from the primary system. This Specification addresses only the RCPB integrity function of the SG. The SG heat removal function is addressed by LCO 3.4.4, "RCS Loops -MODES 1 and 2," LCO 3.4.5,"RCS Loops -MODE 3," LCO 3.4.6, "RCS Loops -MODE 4," and LCO 3.4.7, "RCS Loops -MODE 5, Loops Filled." SG tube integrity means that the tubes are capable of performing their intended RCPB safety function consistent with the licensing basis, pending upon including applicable regulatory requirements.

terials and design, Steam geneator tubing is subject to a variety of degradation mechanisms.

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

The SG performance criteria are used to manage SG tube degradation.

Spcfcton."ta Generator (SG) Program," requires that a p ram be established a ted to ensure that SG tube integrity is maained. Pursuant to Specification 57", tube integrity is maintai d when the SG performance criteria are met. There are three SG perfor ce criteria:

structural integrity, accident induced leakage, and operation LEAKAGE. The SG performance criteria are described in Specification 5.5. .Meeting the SG performance criteria provides reasonable assurance of maintaining tube integrity at normal and accident conditions.

The processes used to meet the SG performance criteria are defined by the Steam Generator Program Guidelines (Ref. 1).WOG STS B 3.4.20-1 Rev. 3.1, 12/01/05 196A Rev. 4, Change 10 & 12 BASES INSERTS FOR ITS 3.4.20 1. For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT 1-131 is assumed to be equal to the LCO 3.4.16, "RCS Specific Activity," limits. Pre-accident and concurrent iodine spikes are assumed in accordance with applicable regulatory guidance.

For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaged fuel. The dose consequences of these events are within the limits of 10 CFR 50.67 (Ref. 2) as supplemented by Regulatory Guide 1.183 (Ref. 3) and within GDC-19 (Ref. 4) values.Unit 1: The analysis for design basis accidents and transients other than a SGTR assume the SG tubes retain their structural integrity (i.e., they are assumed not to rupture.)

In these analyses, the steam discharge to the atmosphere is conservatively assumed to include the total primary to secondary LEAKAGE from all SGs of 450 gpd (i.e., 150 gpd per 12 steam generator) or is assumed to increase to 450 gpd as a result of accident induced conditions.

Currently, the Unit 1 safety analyses do not specifically assume additional primary to secondary LEAKAGE due to accident induced conditions.

Unit 2: The analysis for design basis accidents and transients other than a SGTR assume the SG tubes retain their structural integrity (i.e., they are assumed not to rupture).

In these analyses, the steam discharge to the atmosphere is conservatively assumed to include the total primary to secondary LEAKAGE from all SGs of 450 gpd (i.e., 150 gpd per 12 steam generator) or is assumed to increase to 450 gpd as a result of accident induced conditions for all accidents other than the Unit 2 main steam line break (MSLB).Currently, the Unit 2 MSLB safety analysis is the only analysis that specifically assumes additional primary to secondary LEAKAGE due to accident induced conditions.

For the Unit 2 main steam line break (MSLB) analysis, an increased leakage assumption 12 is applied. In support of voltage based repair criteria pursuant to Generic Letter 95-05 (Ref. 5) analyses were performed to determine the maximum MSLB induced primary to secondary leak rate that could occur without offsite doses exceeding the limits of 10 CFR 50.67 (Ref. 2) as supplemented by Regulatory Guide 1.183 (Ref. 3) and without control room doses exceeding GDC-19 (Ref. 4). An additional 2.1 gpm leakage is assumed in the Unit 2 MSLB analysis resulting from accident conditions.

Therefore, in the MSLB analysis, the steam discharge to the atmosphere includes primary to secondary LEAKAGE equivalent to the operational leakage limit of 150 gpd per SG and an additional 2.1 gpm which results in a total assumed accident induced leakage of 2.4 10 gpm.The combined projected leak rate from all alternate repair criteria (i.e., voltage based repair criteria and application of F*) must be less than the maximum allowable steam line break leak rate limit in any one steam generator in order to maintain doses within the limits of 10 CFR 50.67 (Ref. 2) as supplemented by Regulatory Guide 1.183 (Ref. 3) and within GDC-19 (Ref. 4) values during a postulated steam line break event.196H I Rev. 4, Change 10 REACTOR COOLANT SYSTEM IT.20-I TSG 3 .4 .r3#--4-~STEAM GENERATOR (SG) TUBE INTEGRITY Al 3.4.20=LI MITIN CONDITION FOR OPERATION CONDITION FOR OPERATION SG tube integrity shall be maintained AND All SG tubes satisfying the tube repair criteria shall be plugged or repaire in accordance with the Steam Generator Program.APPLICABILITY:

MODES 1, 2, 3, and 4.ACTION:--- -----------

GENERAL NOTE- -- .........-...---------

Separate action statement entry is allo ed for each SG tube.CojJa--. With one or more SG tubes atisfying the tube repair criteria and not plugged or r paired in accordance with the Steam Generator Program: (1)Action A.1 0--. Verify within 7 ys that tube integrity of the affected tube(s) is maintained until the next refueling outage or SG tube inspection.

ActionA.2

-Plug or repai the affected tube(s) in accordance with the Steam Generator Program prior to entering MODE 4 following the next refueling outage or SG tube inspection.

Cond. B b-.Actions B.1 & B.2 With Action a not being completed within the specified completion time or if SG tube integrity is not being maintained, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .4.4.-S.1-Verify SG tube integrity in accor ance with the Steam Generator Program.44.4..2 Verify that each inspected S tube that satisfies the tube repair criteria is plugged or repaired in accordance with the Steam Generator Program prior to entering MODE 4 followin a SG tube spectio.(1)

SG Tube repair is only applicable to Unit 2.BEAVER VALLEY -UNIT 2 3/4 4-11 Amendment No. 158 SR3.4.20.2 Next Page is 3/4 4-17 248 Rev. 4, Change 10 REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE 3.4.13 LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System operational LEAKAGE shall be limited to: a. No pressure boundary LEAKAGE, b. 1 gpm unidentified LEAKAGE, E 150 gallons per day primary to secondary LEAKAGE through any one steam generator, and 1 0 gpm identified LEAKAGE.APPLICABILITY:

MODES 1, 2, 3, and 4.ACTION: Cond. A a.L on. Yb.With any Reactor Coolant System operational LEAKAGE not within limits for reasons other then pressure boundary LEAKAGE or primary to secondary LEAKAGE, reduce the LEAKAGE to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .With the required action and associated completion time of Action a not met, or with pressure boundary LEAKAGE, or with primary to secondary leakage not within limit, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .SURVEILLANCE REQUIREMENTS

4.6.2 Reactor

Coolant System operational LEAKAGES shall be Ll demonstrated to be within each of the above limits by: Moitoring the following leakage detection instrumen a t nepr1 ours: (1) Only leakage detection instrumentation require LCO BEAVER VALLEY -UNIT 2 3/4 4-19 Amendment No. 158 266 Rev. 4, Change 10 REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE 2. 0 ent atmosphere particulate radioact'3. Containment sump dis monitor.4. ment sump narrow range level monitor.a Performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .(3 y Verifying primary to secondary LEAKAGE is less than or equal to 150 gallons per day through any one steam generator at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .+2+SR 3.4.13.2 Z Note in SR 3.4.13.1 & SR 3.4.13.2-(-2+ Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment

+3INot applicable to primary to secondary LEAKAGE. I\ý Note 2 in SR 3.4.13.1 BEAVER VALLEY -UNIT 2 3/4 4-20 Amendment No. 158 267 BVPS UNITS 1 & 2 ITS CONVERSION LICENSE AMENDMENT REQUEST (LAR)Nos. 296 (UNIT 1) & 1 69 (UNIT 2)REVISION 4 CHANGE 10 AFFECTED PAGES FOR ITS SECTION 5.0 (ADMINISTRATIVE CONTROLS)ITS SECTION 3.4 (REACTOR COOLANT SYSTEM) INDEX OF AFFECTED PAGES ITS MARKUPS PAGES: 33-36, 39 ITS JFDS NONE ITS BASES MARKUPS NONE ITS BASES JFDS NONE CTS MARKUPS PAGES: 81,87-87C, 91-94 CTS DOCS NONE Rev. 4, Change 10 & 12 Section 5.0 Inserts Insert I for Section 5.3.1 Each member of the unit and radiation protection staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for the following: " the operations manager as specified in Specification 5.2.2.e,* the radiation protection manager who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975, and* the technical advisory engineering representative who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design and response analysis of the plant for transients and accidents.

Insert 2 for Section 5.5.5 (from CTS requirements)

A Steam Generator Program for Unit 1 and Unit 2 shall be established and implemented to ensure that SG tube integrity is maintained.

In addition, the Steam Generator Program for Unit 1 shall include the provisions of Specification 5.5.5.1 and the Steam Generator Program for Unit 2 shall include the provisions of Specification 5.5.5.2.5.5.5.1 Unit 1 Steam Generator Program a. Provisions For Condition Monitoring Assessments)( 10 Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes.Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged, to confirm that the performance criteria are being met.b. Provisions for Performance Criteria for SG Tube lntegriý( 10 SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.1. Structural integrity performance criterion:

All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, 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 10 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 Section 5.0 Inserts Page 1 Rev.4, Change 10 Section 5.0 Inserts 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 also not to exceed 1 gpm per SG, except during a SG tube rupture.3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS Operational LEAKAGE".c. Provisions for SG Tube Repair Criteri4<Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.d. Provisions for SG Tube Inspections\r" 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.

A degradation assessment shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

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.During each period 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 intervals between refueling outages (whichever is less) without being inspected.

3. If crack indications are found in any SG tube, then the next inspection for Section 5.0 Inserts Page 2 34 Rev. 4, Change 10 Section 5.0 Inserts each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one interval between refueling outages (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.e. Provisions for monitoring operational primary to secondary LEAKAGE, 5.5.5.2 Unit 2 Steam G_*enuragtorPrbam

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

SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.1. Structural integrity performance criterion:

All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, 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, except for flaws addressed through application of the alternate repair criteria discussed in Specification 5.5.5.2.c.4, 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.When alternate repair criteria discussed in Specification 5.5.5.2.c.4 are applied to axially oriented outside diameter stress corrosion cracking indications at tube support plate locations, the probability that one or more of these indications in a SG will burst under postulated main steam line break conditions shall be less than lx10-2 Section 5.0 Inserts Page 3 Rev. 4, Change 10 & 11 Section 5.0 Inserts 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. Except during a SG tube rupture, leakage from all sources excluding the leakage attributed to the degradation described in 10 Specification 5.5.5.2.c.4 is also not to exceed 1 gpm per SG.3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS Operational Leakage".c. Provisions for SG Tube Repair Criteria 1. Tubes found by inservice inspection to contain a flaw in a non-sleeved region with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged or repaired except if permitted to remain in service through application of the alternate repair criteria discussed in Specification 5.5.5.2.c.4 or 5.5.5.2.c.5.

1 2. Tubes found by inservice inspection to contain a flaw in a sleeve 11 (excluding the sleeve to tube joint) with a depth equal to or exceeding the following percentages of the nominal sleeve wall thickness, shall be 10 plugged: ABB Combustion Engineering TIG welded sleeves 27% 10 Westinghouse laser welded sleeves 25% 10 3. Tubes with a flaw in a sleeve to tube joint shall be plugged. 10 4. Tube support plate voltage-based repair criteria may be applied as an alternative to the 40% depth based criteria of Specification 5.5.5.2.c.1.

Tube Support Plate Plugging Limit is used for the disposition of an Alloy 10 600 steam generator tube for continued service that is experiencing predominantly axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates. At tube support plate intersections, the plugging (repair) limit is described below: a) Steam generator tubes, with degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with bobbin voltages less than or equal to 2.0 volts will be allowed to remain in service.b) Steam generator tubes, with degradation attributed to outside Section 5.0 Inserts Page 4 36 Rev. 4, Change 10 & 11 Section 5.0 Inserts refueling outages (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.4. Indications left in service as a result of application of the tube support plate voltage-based repair criteria (Specification 5.5.5.2.c.4) shall be inspected by bobbin coil probe during all future refueling outages.Implementation of the steam generator tube-to-tube support plate repair criteria requires a 100-percent bobbin coil inspection for hot-leg and cold-leg tube support plate intersections down to the lowest cold-leg tube support plate with known outside diameter stress corrosion cracking (ODSCC) indications.

The determination of the lowest cold-leg tube support plate intersections having ODSCC indications shall be based on the performance of at least a 20-percent random sampling of tubes inspected over their full length.5. When the F* methodology has been implemented, inspect 100% of the inservice tubes in the hot-leg tubesheet region with the objective of 11 detecting flaws that may satisfy the applicable tube repair criteria of Specification 5.5.5.2.c.5 every 24 effective full power months or one interval between refueling outages (whichever is less).e. Provisions for monitoring operational primary to secondary LEAKAGEy 7 [70 f. Provisions for SG Tube Repair Methods Steam generator tube repair methods shall provide the means to reestablish the RCS pressure boundary integrity of SG tubes without removing the tube from service. For the purposes of these Specifications, tube plugging is not a repair.All acceptable tube repair methods are listed below.1. ABB Combustion Engineering TIG welded sleeves, CEN-629-P, Revision 02 and CEN-629-P Addendum 1.2. Westinghouse laser welded sleeves, WCAP-13483, Revision 2.n 5.0 Inserts Page 7 Section ITS 5.64 ADMINISTRATIVE CONTROLS Rev. 4, Change 10 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (Continued) 5 e-. The PTLR shall be provided to the NRC upon issuance for each reactor "luence period and for any revision or supplement theret v STEAM GENERATOR TUBE INSPECTION REPORT 1.5.6.6.1 Unit 1 SG Tube Inspection Report Link to Unit I Repot 5.6.6.2 Unit 2 SG Tube Inspection Report 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 Steam Generator (SG) Program. The report shall include: : .., a. The scope of inspections performed on each SG, b. Active degradation mechanisms found, c.Nondestructive examination techniques utilized for each degradation mechanism, d. Location, orientation (if linear), and measured sizes (if available) of service-induced indications, e. Number of tubes plugged or repaired during the inspection outage for each active degradation mechanism, f. Total number and percentage of tubes plugged or repaired to date, g. The results of condition monitoring, including the results of tube pulls and in-situ testing, h. The effective plugging percentage for all plugging and tube repairs in each SG, and i. Repair method utilized and the number of tubes repaired by each repair method.2. A report shall be submitted within 90 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.5.2 ., Steam Generator Program, when voltage-based alternate repair criteria have been applied. The report shall include information described in Section 6.b of Attachment 1 to Generic Letter 95-05, "Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking." BEAVER VALLEY -UNIT 2 6-22 Amendment No. 158 81 D ITS 5.5 ADMINISTRATIVE CONTROLS I Rev. 4, Change 10 & 12 1 TECHNICAL SPECIFICATIONS (TS) BASES CONTROL PROGRAM (Continued)

2. a change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR. ' 5 -5.10.b.1 and 5.5.10.b.2 I d. Proposed/changes that meet the criteria of Specification 6.18.b.l & 2 above shall be reviewed and approved by the NRC prior to implementation.

Changes to the Bases implemented without prior NRC approval shall be provided to the NRr nn A r-ancy consistent with 10 CFR 50.71(e).forIn dditin the nteam Generato r n I Init 1 shall include the provisions of Specification 5.5.5.1 and the Steam Generator Program 12 for Unit 2 shall include the provisions of Specification 5.5.5.2.A Steam Generator Program~shall be established and implemented to ensure that SG tube integrity is maintained.

In addition, the Steam Cenerater Pregramf shall ineludde the following previsiens:

_,--a. Provisions for Condition Monitoring Assessments 5.5.5.1 Unit I SGProgram Condition monitoring assessment means an evaluation of the"as found" condition of the tubing with respect to the Unit 1pages performance criteria for structural integrity and accident follow. LINK induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by 5.5.5.2UUnit2SG other means, prior to the plugging or repair of tubes.Program Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected, plugged, or repaired to confirm that the performance criteria are being met.b. Provisions for Performance Criteria for SG Tube Integrity SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.1. Structural integrity performance criterion:

All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, 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, except for flaws addressed 10 10 87 BEAVER VALLEY -UNIT 2 6-27 Amendment No. 158 F ITS5.5 Rev. 4, Change 10 & 11 ADMINISTRATIVE CONTROLS STEAM GENERATOR PROGRAM (Continued) 5-5.5.2 through application of the alternate repair criteria 10 discussed in Specification 6.-9.c.4, a safety factor of 1.4 against burst applied to the design basis accident primary to secondary pressure differentials.

Apart 10 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.5.5.5.2 When alternate

/ repair criteria discussed in Specification 6-.-9.c.4 are applied to axially oriented outside diameter stress corrosion cracking indications at tube support plate locations, the probability that 10 one or more of these indications in a SG will burst under postulated main steam line break conditions shall be less than 1xl0-2.2. 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.Except during a steam generator tube rupture, leakage from all sources excluding the leakage attributed to 10 the degradation described in TS Section 6.3-9.c.4 is also not to exceed 1 gpm per SG. 5"- .3. The operational LEAKAGE performance criterion is specified in LCO 3.4. 13 c. Provisions for SG Tube Repair Criteria 1. Tubes found by inservice inspection to contain a flaw in a non-sleeved region with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged or repaired except if permitted to remain in service through application of the alternate repair criteria discussed in Specification c.4 or .c.5.5.5.5.2 BEAVER VALLEY -UNIT 2 6-28 Amendment No. 160 I11 87A iITS5"5 I Rev. 4, Change 1, 10, &11 ADMINISTRATIVE CONTROLS STEAM GENERATOR PROGRAM (Continued)

2. Tubes found by inservice inspection to contain a flaw 11 in a sleeve (excluding the sleeve to tube joint) with a depth equal to or exceeding the following percentages 10 of the nominal sleeve wall thickness, shall be plugged: ABB Combustion Engineering TIG welded sleeves 27%Westinghouse laser welded sleeves 25%3. Tubes with a flaw in a sleeve to tube joint shall be 10 plugged. 5.5.5.2 4. Tube support plate voltage-based repair teria may be 11 applied as an alternative to the 740% depth based criteria of Teehniea!

Specification 6-49.c.l.Tube Support Plate Plugging Limit is used for the disposition of an Alloy 600 steam generator tube for 10 continued service that is experiencing predominantly axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates. At tube support plate intersections, the plugging (repair) limit is described below: a) Steam generator tubes, with degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with bobbin voltages less than or equal to 2.0 volts will be allowed to remain in service.b) Steam generator tubes, with degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts will be repaired or plugged, except as noted in 6-19.c.4.c below. 5552 c) Steam generator tubes, with indications of potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts but less than or equal to the upper voltage repair limit (calculated according to the methodology in Generic Letter 95-05 as supplemented) may remain in service if a rotating pancake coil or acceptable alternative inspection does not detect degradation.

BEAVER VALLEY -UNIT 2 6-29 Amendment No. 160 87B i ~~iTSi.] [~i) Rev. 4, Change 10 ADMINISTRATIVE CONTROLS STEAM GENERATOR PROGRAM (Continued) d) Steam generator tubes, with indications of potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than the upper voltage repair limit (calculated according to the methodology in Generic Letter 95-05 as supplemented) will be plugged or repaired.e) If an unscheduled mid-cycle inspection is performed, the following mid-cycle repair limits apply instead of the limits specified in 5..52 .1 .c.4.a, PF .-c.4.b, Y ý .-c.4.c and 4. c.4.d.The mid-cycle repair limits are determined from the following equations:

V v MR = SL C A MURL I.O+NDE+Gr (CL-At)"CL)(CL -Atx VMLRL = VMU -(VURL- VLRL )C)where: VURL = upper voltage repair limit VLRL = lower voltage repair limit VMURL = mid-cycle upper voltage repair limit based on time into cycle VMLRL = mid-cycle lower voltage repair limit based on VMURL and time into cycle At = length of time since last scheduled inspection during which VURL and VLRL were implemented CL = cycle length (the time between two scheduled steam generator inspections)

VSL = structural limit voltage Gr = average growth rate per cycle length NDE = 95-percent cumulative probability allowance for nondestructive examination uncertainty (i.e., a value of 20 percent has been approved by NRC). The NDE is the value provided by the NRC in GL 95-05 as supplemented.

Implementation of these mid-cycle repair limits should follow the same approach as in Specifications

.... c.4.d.BEAVER VALLEY -UNIT 2 6-30 Amendment No. 158 5.6.6 Unit I Page I (ADMINISTRATIVE CONTROLS Rev. 4, Change 10 PR RE AND TEMPERATURE LIMITS REPORT (PTLR) (Continued)

T ethodology listed in WCAP-14040-NP-A was with two a) Use of A Code Case N-64 "Alternative Reference F1 Changes to this Unit I material are addressed in P-T Limits for S1the corresponding Unit 2 marked-up page.b) Use ot metroJ ogy ot tne iaslk version of ASME Section XI, A dix G, "Fracture T hness Criteria for Pr/ t"Ection Against Failure'".

n s it ri o c The PTLR shall be provided to the NRC upon issua for 5.6.6.1 Unitl each reactor fluence period and for any revision-supplement thereto.

STEAM GENERATOR TUBE INSPECTION REPORT 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 6., Steam Generator (SG) Program.The report shall include: 5.5.5.1, a. The scope of inspections performed on each SG, b. Active degradation mechanisms found, c. Nondestructive examination techniques utilized for each degradation mechanism, d. Location, orientation (if linear), and measured sizes (if available) of service-induced indications, e. Number of tubes plugged during the inspection outage for each active degradation mechanism, f. Total number and percentage of tubes plugged to date, g. The results of condition monitoring, including the results of tube pulls and in-situ testing, and h. The effective plugging percentage for all plugging in each SG.6.1RDAINPOET OGRAM Procedures for Changes to this Unit I material are addressed in 1 be prepared consistent with the corresponding Unit 2 marked-up page. and shall be approved, mai CLUIltli Ut i p Iinvolving p o adiation exposure.BEAVER VALLEY -UNIT 1 6-21 (next page is 6-23)Amendment No. 276 91 5.5.5 Unit I Page I Rev. 4, Change 10 ADMINISTRATIVE CONTRC'&,-

bntainment Leakage Rate Testing Program (Continued)

Air Lock testing acceptance criteria and required a ion are as stated in Specification 3.6.1.3 titled "Cont nment Air Locks." The provis ns of Specification 4.0.2 do not apply the test frequencies ecified in the Containment Leakage ate Testing Program.The provisions o Specification 4.0.3 are plicable to the Containment Leakage te Testing Program.6.18 Technical Specifica ions (TS) Bases Con ol Program This program provides a mea for process g changes to the Bases of these Technical Specification

a. Chang Changes to this Unit I material are addressed be made under appro° in the corresponding Unit 2 marked-up page. Fvaws" b. Licensees may ma e canges o 3ases without prior NRC approval provided e change do not require either of the following:

1. a changen the TS incorporate in the license; or 9/2. a ch ge to the updated FSAR or ases that requires NRCapproval pursuant to 10 CFR c. The ases Control Program shall contain rovisions to en re that the Bases are maintained consist t with the d. Proposed changes that meet the criteria of Speci cation 6.18.b.l & 2 above shall be reviewed and approved b the_ NRC prior to implementation.

Changes to the B es implemented without prior NRC approval shall be provided 5.5.5.1 Unit I the NRC on a frequency consistent with 10 CFR 50.71(e)." ---9 Steam Generator (SG) Program Generata inldthfolwn rvs a. Provisions for Condition Monitoring Assessments Condition monitoring assessment means an evaluation of the"as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by BEAVER VALLEY -UNIT 1 6-26 Amendment No. 276 92 Unit I Page __ Rev. 4, Change 10 ADMINISTRATIVE CONTROLS Steam Generator Program (Continued) other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged, to confirm that the performance criteria are being met.b. Provisions for Performance Criteria for SG Tube Integrity SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.1. Structural integrity performance criterion:

All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, 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 also not to exceed 1 gpm per SG, except during a SG tube rupture.3. The operational LEAKAGE performance criterion is specified in LCO 3.4. 13, 'Operational Leakage" c. Provisions for SG Tube Repair Criteria Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.BEAVER VALLEY -UNIT 1 6-27 Amendment No. 276 13 I Unit I Page (i I Rev. 4, Change 10 ADMINISTRATIVE CONTROLS Steam Generator Program (Continued)

d. Provisions for SG Tube Inspections Periodic SG tube inspections shall be performed.

The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria.

The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.l, 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.

A degradation assessment shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

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. During each period 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 intervals between 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 interval between refueling outages (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.e. Provisions for monitoring operational primary to secondary LEAKAGE BEAVER VALLEY -UNIT 1 6-28 Amendment No. 276 94 BVPS UNITS 1 & 2 IMPROVED TECHNICAL SPECIFICATION (ITS) CONVERSION LICENSE AMENDMENT REQUEST (LAR)Nos. 296 (UNIT 1) & 169 (UNIT 2)REVISION 4 CHANGE 11 Implementation of BVPS Unit 2 Amendment 260 -Revised Steam Generator (SG) inspection and repair scope using the F* methodology.

Description This change updates the affected ITS conversion documentation pages to the latest version of LAR No. 183 as approved in Unit 2 License Amendment 260.Unit 2 License Amendment 260 was issued by the NRC on 9/27/06 and corresponds to Unit 2 LAR No. 183 submitted by FENOC Letter L-05-061 dated 4/11/05. This License Amendment implements the F* Tube repair criteria for U2 SG tubes with degradation in the tubesheet roll expansion region (in accordance with WCAP-16385-NP, Rev. 1). The ITS conversion documentation is updated to reflect the NRC approved text as incorporated into the current Technical Specifications by License Amendment 260.Affected Pages: The affected pages are presented in ITS Section order. The Table(s) preceding each ITS section list the affected pages by type (i.e., ITS markup, CTS markup, etc.) in that Section of the ITS. The affected page numbers listed are the ITS section specific consecutive numbers found in the lower right corner of each page.Note: Because the affected page(s) for each change were extracted from a complete ITS section electronic file, the electronic hyperlinks (created in the complete ITS section file) do not work in the collection of affected pages that follow this cover page.

BVPS UNITS I & 2 ITS CONVERSION LICENSE AMENDMENT (LAR)Nos. 296 (UNIT 1) & 1 69 (UNIT 2)REVISION 4 CHANGE 11 AFFECTED PAGES FOR ITS SECTION 5.0 (ADMINISTRATIVE CONTROLS)ITS SECTION 5.0 (ADMINISTRATIVE CONTROLS)

INDEX OF AFFECTED PAGES ITS MARKUPS PAGES: 36, 38, 39, 53 ITS JFDS NONE ITS BASES MARKUPS NONE ITS BASES JFDS NONE CTS MARKUPS PAGES: 81A, 82, 87A, 87B, 87D, 87E, 87F CTS DOCS NONE I A Rev. 4, Change 10 & 11 Section j.0 Ineits 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. Except during a SG tube rupture, leakage from all sources excluding the leakage attributed to the degradation described in 10 Specification 5.5.5.2.c.4 is also not to exceed 1 gpm per SG.3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS Operational Leakage".c. Provisions for SG Tube Repair Criteria 1 .Tubes found by inservice inspection to contain a flaw in a non-sleeved region with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged or repaired except if permitted to remain in service through application of the alternate repair criteria discussed in Specification 5.5.5.2.c.4 or 5.5.5.2.c.5.

11 2. Tubes found by inservice inspection to contain a flaw in a sleeve 11 (excluding the sleeve to tube joint) with a depth equal to or exceeding the following percentages of the nominal sleeve wall thickness, shall be 10 plugged: ABB Combustion Engineering TIG welded sleeves 27% 10 Westinghouse laser welded sleeves 25% 10 3. Tubes with a flaw in a sleeve to tube joint shall be plugged. 10 4. Tube support plate voltage-based repair criteria may be applied as an 11 alternative to the 40% depth based criteria of Specification 5.5.5.2.c.1.

I Tube Support Plate Plugging Limit is used for the disposition of an Alloy 10 600 steam generator tube for continued service that is experiencing predominantly axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates. At tube support plate intersections, the plugging (repair) limit is described below: a) Steam generator tubes, with degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with bobbin voltages less than or equal to 2.0 volts will be allowed to remain in service.b) Steam generator tubes, with degradation attributed to outside n 5.0 Inserts Page 4 36 Section I Rev. 4, Change 11 Section 5.0 Inserts the value provided by the NRC in GL 95-05 as supplemented.

Implementation of these mid-cycle repair limits should follow the same approach as in Specifications 5.5.5.2.c.4.a through 5.5.5.2.c.4.d.

5. The F* methodology, as described below, may be applied to the expanded portion of the tube in the hot-leg tubesheet region as an alternative to the 40% depth based criteria of Specification 5.5.5.2.c.1:

a) Tubes with no portion of a lower sleeve joint in the hot-leg tubesheet region shall be repaired or plugged upon detection of any flaw identified within 3.0 inches below the top of the tubesheet or within 2.2 inches below the bottom of roll transition, whichever elevation is lower. Flaws located below this elevation may remain in service regardless of size.b) Tubes which have any portion of a sleeve joint in the hot-leg tubesheet region shall be plugged upon detection of any flaw identified within 3.0 inches below the lower end of the lower sleeve joint. Flaws located greater than 3.0 inches below the lower end of the lower sleeve joint may remain in service regardless of size.d. Provisions for SG Tube Inspections Periodic SG tube inspections shall be performed.

The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria.

The tube-to-tubesheet weld is not part of the tube. In tubes repaired by sleeving, the portion of the original tube wall between the sleeve's joints is not an area requiring re-inspection.

In addition to meeting the requirements of d.1, d.2, d.3, d.4, and d.5 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection.

A degradation assessment shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

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 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. No SG shall operate for more than 24 effective full power months or one interval between 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 interval between Section 5.0 Inserts Page 6 38 Rev. 4, Change 10 & 11 Section 5.0 Inserts refueling outages (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.4. Indications left in service as a result of application of the tube support plate voltage-based repair criteria (Specification 5.5.5.2.c.4) shall be inspected by bobbin coil probe during all future refueling outages.Implementation of the steam generator tube-to-tube support plate repair criteria requires a 100-percent bobbin coil inspection for hot-leg and cold-leg tube support plate intersections down to the lowest cold-leg tube support plate with known outside diameter stress corrosion cracking (ODSCC) indications.

The determination of the lowest cold-leg tube support plate intersections having ODSCC indications shall be based on the performance of at least a 20-percent random sampling of tubes inspected over their full length.5. When the F* methodology has been implemented, inspect 100% of the inservice tubes in the hot-leg tubesheet region with the objective of detecting flaws that may satisfy the applicable tube repair criteria of Specification 5.5.5.2.c.5 every 24 effective full power months or one interval between refueling outages (whichever is less).e. Provisions for monitoring operational primary to secondary LEAKAGES f. Provisions for SG Tube Repair Methods Steam generator tube repair methods shall provide the means to reestablish the RCS pressure boundary integrity of SG tubes without removing the tube from service. For the purposes of these Specifications, tube plugging is not a repair.All acceptable tube repair methods are listed below.1. ABB Combustion Engineering TIG welded sleeves, CEN-629-P, Revision 02 and CEN-629-P Addendum 1.2. Westinghouse laser welded sleeves, WCAP-1 3483, Revision 2.n 5.0 Inserts Page 7 Sectior Rev. 4, Change 11 & 16 Section 5.0 Inserts in each SG, and i. Repair method utilized and the number of tubes repaired by each repair method.2. A report shall be submitted within 90 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.5.2, Steam Generator Program, when voltage-based alternate repair criteria have been applied. The report shall include information described in Section 6.b of Attachment 1 to Generic Letter 95-05, "Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking." 3. For implementation of the voltage-based repair criteria to tube support plate intersections, notify the Commission prior to returning the steam generators to service (MODE 4) should any of the following conditions arise: a. If circumferential crack-like indications are detected at the tube support plate intersections.

b. If indications are identified that extend beyond the confines of the tube support plate.c. If indications are identified at the tube support plate elevations that are attributable to primary water stress corrosion cracking.4. Report the following information to the NRC within 90 days after achieving ,IZ MODE 4 following an outage in which the F* methodology was applied: 16 a. Total number of indications, location of each indication, orientation of 11 each indication, severity of each indication, and whether the indications initiated from the inside or outside surface.b. The cumulative number of indications detected in the tubesheet region as a function of elevation within the tubesheet.

c. The projected end-of-cycle accident-induced leakage from tubesheet indications.

serts Page 21 53 Section 5.0 Ins ITS 5.6 (Iii. I Rev. 4, Change 11 ADMINISTRATIVE CONTROLS STEAM GENERATOR TUBE INSPECTION REPORT (Continued)

3. For implementation of the voltage-based repair criteria to tube support plate intersections, notify the Commission prior to returning the steam generators to service (MODE 4)should any of the following conditions arise: a. If circumferential crack-like indications are detected at the tube support plate intersections.

b. If indications are identified that extend beyond the confines of the tube support plate.c. If indications are identified at the tube support plate elevations that are attributable to primary water stress corrosion cracking.4. Report the following information to the NRC within 90 days after achieving Mode 4 following an outage in which the F*methodology was applied: a. Total number of indications, location of each indication, orientation of each indication, severity of each indication, and whether the indications initiated from the inside or outside surface.b. The cumulative number of indications detected in the tubesheet region as a function of elevation within the tubesheet.

c. The projected end-of-cycle accident-induced leakage from tubesheet indications.

6.10 DELETED 6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared Iconsistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.[ -i-5 HIGH RADIATION AREA 6In lieu of the "control device" or "alarm signal" required by paragraph 20.1601 of 10 CFR 20, each high radiation area in which the intensity of radiation is greater than 100 mrem/hr but less than 1000 mrem/hr shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring BEAVER VALLEY -UNIT 2 6-22a Amendment No. 160 1A ITS 5.7 ] IRev. 4, Change 11 ADMINISTRATIVE CONTROLS IR.4--'F HIGH RADIATION AREA (Continued) issuance of a Radiological Work Permit +1. Any individual or group of individuals permitted to enter such areas all be provided with or accompanied by one or more of the following:

a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received.

Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them.c. An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device. This individual shall be responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by a faeility radiation protection 5.7.2I supervisor in the Radiological Work Permit.e requirements of .-_ , above, also apply to each high radiatio area in which the inte sity of radiation is greater than 1000 mr /hr. In addition, locke doors shall be provided to prevent unauth ized entry into such area and the keys shall be maintained under the administrative control of the shift supervisor on duty an a fa-i-lity radiation protect on supervisor.

In addition to the 5.//Insert Note (1) below, directly into text above as marked. -escorted by th approved.WP issuance n protection n protection it No. 160 82 Radiation protection personnel, or personnel radiation protection personnel in accordance wi emergency procedures, shall be exempt from the R requirement during the performance of their radiatioj duties, provided they comply with approved radiatioi procedures for entry into high radiation areas.BEAVER VALLEY -UNIT 2 6-23 ITS5.5 Rev. 4, Change 10 & 11 ADMINISTRATIVE CONTROLS STEAM GENERATOR PROGRAM (Continued) through application of the alternate repair criteria 10 discussed in Specification .c.4, a safety factor of 1.4 against burst applied to the design basis accident primary to secondary pressure differentials.

Apart 10 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.5.5.5.2 When alternate

/ repair criteria discussed in Specification 6-I-9.c.4 are applied to axially oriented outside diameter stress corrosion cracking indications at tube support plate locations, the probability that 10 one or more of these indications in a SG will burst under postulated main steam line break conditions shall be less than 1x1lO 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.Except during a steam generator tube rupture, leakage from all sources excluding the leakage attributed to 10 the degradation described in TS Section 6-19.c.4 is also not to exceed 1 gpm per SG. *"- 2 3. The operational LEAKAGE performance criterion is specified in LCO 2.-.-. 3.4.13 ]c. Provisions for SG Tube Repair Criteria 1. Tubes found by inservice inspection to contain a flaw in a non-sleeved region with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged or repaired except if permitted to remain in service through application of the alternate repair criteria discussed in Specification c.4 or .c.5.5.5.5.2 BEAVER VALLEY -UNIT 2 6-28 Amendment No. 160 11 SITS 5.5 Iý I Rev. 4, Changel, 1O,&1 ADMINISTRATIVE CONTROLS STEAM GENERATOR PROGRAM (Continued)

2. Tubes found by inservice inspection to contain a flaw 11 in a sleeve (excluding the sleeve to tube joint) with a depth equal to or exceeding the following percentages 10 of the nominal sleeve wall thickness, shall be plugged: ABB Combustion Engineering TIG welded sleeves 27%Westinghouse laser welded sleeves 25%3. Tubes with a flaw in a sleeve to tube joint shall be 10 plugged. 5.4. Tube support plate voltage-based repair cieria may be 11 applied as an alternative to the 40% depth based criteria of Teehnieal Specification 6-.1-9.c.l.

Tube Support Plate Plugging Limit is used for the disposition of an Alloy 600 steam generator tube for 10 continued service that is experiencing predominantly axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates. At tube support plate intersections, the plugging (repair) limit is described below: a) Steam generator tubes, with degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with bobbin voltages less than or equal to 2.0 volts will be allowed to remain in service.b) Steam generator tubes, with degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts will be repaired or plugged, except as noted in 6-49*.c.4.c below. 5.55.c) Steam generator tubes, with indications of potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts but less than or equal to the upper voltage repair limit (calculated according to the methodology in Generic Letter 95-05 as supplemented) may remain in service if a rotating pancake coil or acceptable alternative inspection does not detect degradation.

BEAVER VALLEY -UNIT 2 6-29 Amendment No. 160 87B ITS5.5 _ I Rev. 4, Change 11 ADMINISTRATIVE CONTROLS STEAM GENERATOR PROGRAM (Continued) 5.5.5.2 5. The F* methodology, as described belo may be applied to the expanded portion of the tube *n the hot-leg tubesheet region as an alternative to the 40% depth based criteria of Teehnieal Specification 4-6.-9.c.1:

a) Tubes with no portion of a lower sleeve joint in the hot-leg tubesheet region shall be repaired or plugged upon detection of any flaw identified within 3.0 inches below the top of the tubesheet or within 2.2 inches below the bottom of roll transition, whichever elevation is lower. Flaws located below this elevation may remain in service regardless of size.b) Tubes which have any portion of a sleeve joint in the hot-leg tubesheet region shall be plugged upon detection of any flaw identified within 3.0 inches below the lower end of the lower sleeve joint.Flaws located greater than 3.0 inches below the lower end of the lower sleeve joint may remain in service regardless of size.d. Provisions for SG Tube Inspections Periodic SG tube inspections shall be performed.

The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria.

The tube-to-tubesheet weld is not part of the tube. In tubes repaired by sleeving, the portion of the original tube wall between the sleeve's joints is not an area requiring re-inspection.

In addition to meeting the requirements of d.l, d.2, d.3, d.4, and d.5 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection.

A degradation assessment shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

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

BEAVER VALLEY -UNIT 2 6-31 Amendment No. 160 87D (i) IRev. 4, Change 11 ADMINISTRATIVE CONTROLS STEAM GENERATOR PROGRAM (Continued)

2. Inspect 100% of the tubes at sequential periods of 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. No SG shall operate for more than 24 effective full power months or one interval between 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 interval between refueling outages (whichever is less) .If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.4. Indications left in service as a result of application of the tube support plate voltage-based repair criteria c.4) shall be inspected by bobbin coil probe ring all future refueling outages.Implementation of the steam generator tube-to-tube support plate repair criteria requires a 100-percent bobbin coil inspection for hot-leg and cold-leg tube support plate intersections down to the lowest cold-leg tube support plate with known outside diameter stress corrosion cracking (ODSCC) indications.

The determination of the lowest cold-leg tube support plate intersections having ODSCC indications shall be based on the performance of at least a 20-percent random sampling of tubes inspected over their full length.5. When the F* methodology has been implemented, inspect 100% of the inservice tubes in the hot-leg tubesheet region with the objective of detecting flaws that may satisfy the applicable tube repair criteria of Teehniceal Specification .c.5 every 24 effective full power months or one in erval between refueling outages (whichever is less).5.5.5.2 e. Provisions for monitoring operational primary to secondary LEAKAGE BEAVER VALLEY -UNIT 2 6-32 Amendment No. 160 87E IS 5 ADMINISTRATIVE CONTROLS Rev. 4, Change I & 11 (11 for pagination only)STEAM GENERATOR PROGRAM (Continued)

f. Provisions for SG Tube Repair Methods Steam generator tube repair methods shall provide the means to reestablish the RCS pressure boundary integrity of SG tubes without removing the tube from service. For the purposes of these Specifications, tube plugging is not a repair. All acceptable tube repair methods are listed below.1. ABB Combustion Engineering TIG welded sleeves, CEN-629-P, Revision 02 and CEN-629-P Addendum 1.2. Westinghouse Revision 2.laser welded sleeves, WCAP-13483, Add ITS 5.5.3, Component Cyclic or Transient Limit Po1g r Insert 1 SAdd ITS 5.5.11, Safety Function Determination Program [Insert 2 M3 Add ITS 5.5.13, Battery Monitoring and Maintenance Program E Insert 3 Add ITS 5.5.9, Diesel Fuel Oil Testing Program]Iýnsert 4A3 Add ITS 5.6.5, Post Accident Monitoring Report : Ine I A24-BEAVER VALLEY -UNIT 2 6-33 Amendment No. 160 87F BVPS UNITS I & 2 IMPROVED TECHNICAL SPECIFICATION (ITS) CONVERSION LICENSE AMENDMENT REQUEST (LAR)Nos. 296 (UNIT 1) & 1 69 (UNIT 2)REVISION 4 CHANGE 12 NRC Comment Resolution NRC Reviewer:

R. Clark/K. Karwoski Description This change addresses the NRC reviewer's comments on the Revision 3, Change 2 to the ITS conversion documentation.

Revision 3 Change 2 of the ITS conversion documentation incorporated the initial version of LARs 324 (Unit 1) and 196 (Unit 2). These BVPS LARs proposed changes to implement Technical Specification Task Force (TSTF) 449 (the steam generator (SG) program).

Subsequent to the NRC Reviewer's comments, LARs 324 and 196 were approved by the NRC in License Amendments 276 (Unit 1) and 158 (Unit 2). The final approved pages resulting from License Amendments 276 and 156 were incorporated into the ITS conversion documentation in Revision 4, Change 10. The following NRC comments (summarized for brevity) resulted in a change in the ITS conversion documentation:

Comment 1 addressed the placement of a Bases insert (Note 2 description) from TSTF# 449 on Bases page B 3.4.13-5.

BVPS agreed to move the insert on this page to be consistent with the TSTF.Comment 2 addressed the text of Insert 1 to the Bases for ITS 3.4.13. BVPS agreed to update the Unit 2 safety analysis discussion to include the total accident induced leakage rate of 2.4 gpm.Comment 3 addressed the deletion of a sentence on ISTS Bases Page B 3.4.20-3.BVPS agreed to restore the affected sentence.Comment 4 addressed the text of Insert 1 to the Bases of ITS 3.4.20. BVPS agreed to update the Unit 2 safety analysis discussion in Insert 1 to clarify that structural integrity is maintained for all accidents except for a SG tube rupture and that leakage from all sources must be less than the maximum allowable.

Comment 5 addressed the opening paragraph of ITS 5.5.5. The reviewer sugested a change to this text which was agreed to by BVPS.

BVPS UNITS I & 2 IMPROVED TECHNICAL SPECIFICATION (ITS) CONVERSION LICENSE AMENDMENT REQUEST (LAR)Nos. 296 (UNIT 1) & 1 69 (UNIT 2)CHANGE 12 (continued)

A subsequent comment from the NRC expressed an concern that additional wording from TSTF-449 should be incorporated into the BVPS ITS conversion documentation.

TSTF-449 includes text that expressed the maximum allowed leakage in different terms than the proposed ITS Bases. BVPS agreed to add the following sentence derived from TSTF-449 text into the ITS Bases discussion of SG tube leakage; "The safety analysis for an event resulting in steam discharge to the atmosphere conservatively assumes that primary to secondary LEAKAGE from all steam generators is 450 gallons per day (gpd) (i.e., 150 gpd per steam generator) or increases to 450 gpd as a result of accident induced conditions.

The addition this text into the bases discussion supports the TSTF concept that the accident induced leakage limit not only includes leakage that is induced during the accident but also pre-existing primary-to-secondary leakage. It also is important because it illustrates that it may be unacceptable to operate with primary-to-secondary leakage near the normal operating leakage limit of 150 gpd (since the leakage may increase as a result of the accident induced conditions).

BVPS agreed to modify the text of Insert 1 to both the Bases of ITS 3.4.13 and ITS 3.4.20 to include the corresponding text discussed above. The following ITS conversion documentation pages contain the resolution of the comments discussed above.Affected Pages: The affected pages are presented in ITS Section order. The Table(s) preceding each ITS section list the affected pages by type (i.e., ITS markup, CTS markup, etc.) in that Section of the ITS. The affected page numbers listed are the ITS section specific consecutive numbers found in the lower right corner of each page.Note: Because the affected page(s) for each change were extracted from a complete ITS section electronic file, the electronic hyperlinks (created in the complete ITS section file) do not work in the collection of affected pages that follow this cover page.

BVPS UNITS 1 & 2 ITS CONVERSION LICENSE AMENDMENT REQUEST (LAR)Nos. 296 (UNIT 1) & 1669 (UNIT 2)REVISION 4 CHANGE 12 AFFECTED PAGES FOR ITS SECTION 3.4 (REACTOR COOLANT SYSTEM)ITS SECTION 3.4 (REACTOR COOLANT SYSTEM) INDEX OF AFFECTED PAGES ITS MARKUPS NONE ITS JFDS NONE ITS BASES MARKUPS PAGES: 166, 167, 196C, 196H ITS BASES JFDS NONE CTS MARKUPS NONE CTS DOCS NONE Rev. 4 change 10 & 12 RCS Operational LEAKAGE-B 3.4.13 (stable temperature, power level, presurizer and makeup tank levels, makeup and letLown and RCP seal injection and return flows)7 v 10 SURVEILLANCE REQUIREMENTS (continued)

The Surveillance ismodified 1 states..../: .... ....by two Notes-TSTF-449, Rev. 4 The RCS water inventory b nce mu, be met with the reacto at steady state operating conditions.

Noters added Note -that this SR is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishing steady state operation.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance provides sufficient time to collect 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.and process all necessary data after stable plant conditions are established.

1 12 Steady state operation is required to perform a proper inventory balance since calculations during maneuvering are not useful. For RCS operational LEAKAGE determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeu and letdown, and RCP seal injection and return flows. instrumentation An early warning of pressure boun ry LEAKAGE or unidentified LEAKAGE is provided by the systems that monitor the containment atmosphere radioactivity and the containment sump level. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. These leakage detection systems are specified in LCO 3.4.15, "RCS Leakage Detection Instrumentation." I TSTF-449, Rev. 4 The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months INSERT recognizes tl" accidents.

SR 3.4.13.2 Frequency is a reasonable interval to trend LEAKAGE and te importance of early leakage detection in the prevention of This ides the means necessary to determine SG OP Y in an operational ODF-The requirement to rate SG tube integrity in accordance with thee neerator Tube Surveillance Program emphasi Importance of SG tu i ,even though thi ance cannot be performed at normal operating co UFSAR Section 4.2.7.1 (Unit 1) and UFSAR Section 5.2.5 (Unit 2).I I NRC Generic Letter 95-05: Voltage-Based Repair 3. FSAR, SeGtiGR+I, ý. ý Criteria For Westinghouse Steam Generator Tubes I Affected By Outside Diameter Stress Corrosion Cracking Unit I UFSAR Appendix 1A, "'1971 AEC General Design Criteria Conformance" and Unit 2 UFSAR Section 3.1, "Conformance with U.S. Nuclear Regulatory Commission General Design Criteria" 6===T-4.5.NEI 97-06, "Steam Generator Program Guidelines." EPRI, "Pressurized Water Reactor Primary-to-Secondary Leak Guidelines." 3 WOG STS I TSTF-449, Rev. 4 B 3.4.13 -5 Rev. 2, 04/30/01 166 Rev. 4, Change 10 & 12 BASES INSERTS FOR 3.4.13 1. Primary to secondary LEAKAGE is a factor in the dose assessment of accidents or 10 transients that involve secondary steam release to the atmosphere, such as a main steam line break (MSLB), a locked rotor accident (LRA), a Loss of AC Power (LACP), a Control Rod Ejection Accident (CREA) and to a lesser extent, a Steam Generator Tube Rupture (SGTR). The leakage contaminates the secondary fluid. The limit on the primary to secondary LEAKAGE ensures that the dose contribution at the site boundary from tube 10 leakage following such accidents are limited to appropriate fractions of the 10 CFR 50.67 limit of 25 Rem TEDE as allowed by Regulatory Guide 1.183. The limit on the primary to secondary leakage also ensures that the dose contribution from tube leakage in the control room is limited to the 10 CFR 50.67 limit of 5 Rem TEDE. Among all of the analyses that release primary side activity to the environment via tube leakage, the MSLB is of particular concern because the ruptured main steam line provides a pathway to release the primary to secondary leakage directly to the environment without dilution in the secondary fluid.For Unit 1, the safety analysis for an event resulting in steam discharge to the atmosphere conservatively assumes that primary to secondary LEAKAGE from all steam generators is 450 gallons per day (gpd) (i.e., 150 gpd per steam generator) or increases to 450 gpd as a 12 result of accident induced conditions.

Currently, the Unit 1 safety analyses do not specifically assume additional primary to secondary LEAKAGE due to accident induced conditions.

For Unit 2, due to adoption of the voltage based steam generator tube repair criteria per guidance provided by Generic Letter 95-05 (Reference 3), the safety analysis for an event resulting in steam discharge to the atmosphere conservatively assumes that primary to secondary LEAKAGE from all steam generators is 450 gallons per day (gpd) (i.e., 150 gpd per steam generator) or increases to 450 gpd as a result of accident induced conditions for 12 all accidents other than the MSLB. Currently, the Unit 2 MSLB safety analysis is the only analysis that specifically assumes additional primary to secondary LEAKAGE due to accident induced conditions.

The Unit 2 dose consequences associated with the MSLB addresses an additional 2.1 gpm 10 leakage, which, per GL 95-05, is postulated to occur (via pre-existing tube defects) as a 49, result of the rapid depressurization of the secondary side due to the MLSB, and the consequent high differential pressure across the faulted steam generator.

The maximum allowed Unit 2 total accident induced leakage is 2.4 gpm. 10 2. 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.3. This SR verifies that primary to secondary LEAKAGE is less than or equal to 150 gallons 10 per day through any one SG. Satisfying the primary to secondary LEAKAGE limit ensures that the operational LEAKAGE performance criterion in the Steam Generator Program is met. If this SR is not met, compliance with LCO 3.4.20, "Steam Generator Tube Integrity," 167 SG Tube Integrity B 3.4.20 Rev. 4, Change 12 BASES LCO (continued)

There are three SG performance criteria:

structural integrity, accident induced leakage, and operational LEAKAGE. Failure to meet any one of these criteria is considered failure to meet the LCO.The structural integrity performance criterion provides a margin of safety against tube burst or collapse under normal and accident conditions, and ensures structural integrity of the SG tubes under all anticipated transients included in the design specification.

Tube burst is defined as,"The gross structural failure of the tube wall. The condition typically corresponds to an unstable opening displacement (e.g., opening area increased in response to constant pressure) accompanied by ductile (plastic) tearing of the tube material at the ends of the degradation." Tube collapse is defined as, "For the load displacement curve for a given structure, collapse occurs at the top of the load versus displacement curve where the slope of the curve becomes zero." The structural integrity performance criterion provides guidance on assessing loads that have a significant effect on burst or collapse.

In that context, the term"significant" is defined as "An accident loading condition other than differential pressure is considered significant when the addition of such loads in the assessment of the structural integrity performance criterion could cause a lower structural limit or limiting burst/collapse condition to be established." For tube integrity evaluations, except for circumferential degradation, axial thermal loads are classified as secondary loads. For circumferential degradation, the classification of axial thermal loads as primary or secondary loads will be evaluated on a case-by-case basis.The division between primary and secondary classifications will be based on detailed analysis and/or testing. 7 6 Structural integrity requires that the pri ary membran stress intensity in a tube not exceed the yield strengt or all ASME Co e,Section III, Service Level A (normal operati conditions) and ervice Level B (upset or abnormal conditions) tran ents included in the esign specification.

This includes safety fact and applicable desi basis loads based on ASME Code, Section , Subsection NB (Ref. and Draft Regulatory Guide 1.121 (Ref. as described in the Applicable Safety Analyses section of this Bases.I -.The accident induced leakage performance criterion ensures)Xat the primary to secondary LEAKAGE caused by a design basis Acident, c wher~e the NRC has approved greater accident ind-u-ed-leakage.]accident induced leakage rate includes any primary to secondary LEAKAGE existing prior to the accident in addition to primary to secondary LEAKAGE induced during the accident.WOG STS B 3.4.20-3 Rev. 3.1, 12/01/05 196C Rev. 4, Change 10 & 12 BASES INSERTS FOR ITS 3.4.20 1. For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT 1-131 is assumed to be equal to the LCO 3.4.16, "RCS Specific Activity," limits. Pre-accident and concurrent iodine spikes are assumed in accordance with applicable regulatory guidance.

For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaged fuel. The dose consequences of these events are within the limits of 10 CFR 50.67 (Ref. 2) as supplemented by Regulatory Guide 1.183 (Ref. 3) and within GDC-19 (Ref. 4) values.Unit 1: The analysis for design basis accidents and transients other than a SGTR assume the SG tubes retain their structural integrity (i.e., they are assumed not to rupture.)

In these analyses, the steam discharge to the atmosphere is conservatively assumed to include the total primary to secondary LEAKAGE from all SGs of 450 gpd (i.e., 150 gpd per 12 steam generator) or is assumed to increase to 450 gpd as a result of accident induced conditions.

Currently, the Unit 1 safety analyses do not specifically assume additional primary to secondary LEAKAGE due to accident induced conditions.

Unit 2: The analysis for design basis accidents and transients other than a SGTR assume the SG tubes retain their structural integrity (i.e., they are assumed not to rupture).

In these analyses, the steam discharge to the atmosphere is conservatively assumed to include the total primary to secondary LEAKAGE from all SGs of 450 gpd (i.e., 150 gpd per 12 steam generator) or is assumed to increase to 450 gpd as a result of accident induced conditions for all accidents other than the Unit 2 main steam line break (MSLB).Currently, the Unit 2 MSLB safety analysis is the only analysis that specifically assumes additional primary to secondary LEAKAGE due to accident induced conditions.

For the Unit 2 main steam line break (MSLB) analysis, an increased leakage assumption 12 is applied. In support of voltage based repair criteria pursuant to Generic Letter 95-05 (Ref. 5) analyses were performed to determine the maximum MSLB induced primary to secondary leak rate that could occur without offsite doses exceeding the limits of 10 CFR 50.67 (Ref. 2) as supplemented by Regulatory Guide 1.183 (Ref. 3) and without control room doses exceeding GDC-19 (Ref. 4). An additional 2.1 gpm leakage is assumed in the Unit 2 MSLB analysis resulting from accident conditions.

Therefore, in the MSLB analysis, the steam discharge to the atmosphere includes primary to secondary LEAKAGE equivalent to the operational leakage limit of 150 gpd per SG and an additional 2.1 gpm which results in a total assumed accident induced leakage of 2.4 10 gpm.The combined projected leak rate from all alternate repair criteria (i.e., voltage based repair criteria and application of F*) must be less than the maximum allowable steam line break leak rate limit in any one steam generator in order to maintain doses within the limits of 10 CFR 50.67 (Ref. 2) as supplemented by Regulatory Guide 1.183 (Ref. 3) and within GDC-1 9 (Ref. 4) values during a postulated steam line break event.196H BVPS UNITS 1 & 2 ITS CONVERSION LICENSE AMENDMENT REQUEST (LAR)Nos. 296 (UNIT 1) & 1 69 (UNIT 2)REVISION 4 CHANGE 12 AFFECTED PAGES FOR ITS SECTION 5.0 (ADMINISTRATIVE CONTROLS)ITS SECTION 5.0 (ADMINISTRATIVE CONTROLS)

INDEX OF AFFECTED PAGES ITS MARKUPS PAGES: 33 ITS JFDS NONE ITS BASES MARKUPS NONE ITS BASES JFDS NONE CTS MARKUPS PAGES: 87 CTS DOCS NONE Rev. 4, Change 10 & 12 S Section 5.0 Inserts insert I for Section 5.3.1 Each member of the unit and radiation protection staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for the following: " the operations manager as specified in Specification 5.2.2.e," the radiation protection manager who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975, and" the technical advisory engineering representative who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design and response analysis of the plant for transients and accidents.

Insert 2 for Section 5.5.5 (from CTS requirements)

A Steam Generator Program for Unit 1 and Unit 2 shall be established and implemented to ensure that SG tube integrity is maintained.

In addition, the Steam Generator Program for Unit 1 shall include the provisions of Specification 5.5.5.1 and the Steam Generator Program for Unit 2 shall include the provisions of Specification 5.5.5.2.5.5.5.1 U`in iti w S eneraor Pqoram'a. Provisions For Condition Monitoring Assessments)( 10 Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes.Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged, to confirm that the performance criteria are being met.b. Provisions for Performance Criteria for SG Tube Integritj c 10 SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.1. Structural integrity performance criterion:

All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, 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 10 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 Section 5.0 Inserts Page 1 33 S ITS55 Rev. 4, Change 10 & 12 ADMINISTRATIVE CONTROLS TECHNICAL SPECIFICATIONS (TS) BASES CONTROL PROGRAM (Continued)

2. a change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR. 'h 5.5.10.b.1 and 5.5.10.b.2

d. Proposed changes that meet the criteria of Specification 6.18.b.1 & 2 above shall be reviewed and approved by the NRC prior to implementation.

Changes to the Bases 5 implemented without prior NRC approval shall be provided to the N e consistent with 10 CFR 50.71(e).Sand I In addition, the Steam Generator Program for Unit 1 shall include the STEAM GENERATOR (SG) \PROGRAM provisions of Specification 5.5.5.1 and the Steam Generator Program 12 for Unit 2 shall include the provisions of Specification 5.5.5.2.A Steam Generator Program-shall be establi hed and implemented to ensure that SG tube integrity is maintained.

4In addition, the Stca Conerator Pregramf shall inelude the following previsiens:Provisions for Condition Monitoring Assessments 5.5.5.1 Unit I SGProgram Condition monitoring assessment means an evaluation of the"as found" condition of the tubing with respect to the Unit 1pages performance criteria for structural integrity and accident follow- LINK induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by 5.5.5.2Unit2SG other means, prior to the plugging or repair of tubes.Program Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected, plugged, or repaired to confirm that the performance criteria are being met.b. Provisions for Performance Criteria for SG Tube Integrity SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.1. Structural integrity performance criterion:

All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, 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 1 pressure differential and, except for flaws addressed BEAVER VALLEY -UNIT 2 6-27 Amendment No. 158 10 87 BVPS UNITS 1 & 2 IMPROVED TECHNICAL SPECIFICATION (ITS) CONVERSION LICENSE AMENDMENT REQUEST (LAR)Nos. 296 (UNIT 1) & 169 (UNIT 2)REVISION 4 CHANGE 13 Description ITS 3.8.3, "Diesel Fuel Oil, Lube Oil, and Starting Air", contains new requirements for the BVPS technical specifications that include the requirement for a minimum emergency diesel generator (EDG) air start pressure to support 5 EDG start attempts.

This change revises the 394 psig minimum pressure originally proposed for the Unit 2 (EDGs). The current proposed air pressure of 394 psig in the new BVPS ITS 3.8.3 has been revised to 380 psig to increase the starting air pressure normal operating band when instrument uncertainty is considered.

The original air pressure proposed for this new ITS requirement (394 psig) was determined to be overly restrictive and resulted in an insufficient operating band. The revised air pressure requirement of 380 psig is adequate to support the required 5 EDG air start attempts consistent with the air start pressure data provided in the Unit 2 UFSAR. In addition, the revised pressure of 380 psig, when combined with instrument uncertainty, results in an acceptable operating band for the required air pressure.Affected Pages: The affected pages are presented in ITS Section order. The Table(s) preceding each ITS section list the affected pages by type (i.e., ITS markup, CTS markup, etc.) in that Section of the ITS. The affected page numbers listed are the ITS section specific consecutive numbers found in the lower right corner of each page.Note: Because the affected page(s) for each change were extracted from a complete ITS section electronic file, the electronic hyperlinks (created in the complete ITS section file) do not work in the collection of affected pages that follow this cover page.

BVPS UNITS I & 2 ITS CONVERSION LICENSE AMENDMENT REQUEST (LAR)Nos. 296 (UNIT 1) & 1 69 (UNIT 2)REVISION 4 CHANGE 13 AFFECTED PAGES FOR ITS SECTION 3.8 (ELECTRICAL POWER SYSTEMS)ITS SECTION 3.8 (ELECTRICAL POWER SYSTEMS) INDEX OF AFFECTED PAGES ITS MARKUPS PAGES: 25, 26 ITS JFDS NONE ITS BASES MARKUPS PAGES: 109 ITS BASES JFDS NONE CTS MARKUPS PAGES: 190 CTS DOCS PAGES: 229 Rev. 4, Change 13 Diesel Fuel Oil, Lube Oil, and Starting Air 3.8.3 3.8 ELECTRICAL POWER SYSTEMS 3.8.3 Diesel Fuel Oil, Lube Oil, and Starting Air LCO 3.8.3 The stored diesel fuel oil, lube oil, and starting air subsystem shall be within limits for each required diesel generator (DG).APPLICABILITY:

When associated DG is required to be OPERABLE.ACTIONS----....-------

-.2 ---- -----.t NOTE -inventory Separate C ition entry is allowed for each DG.---- ------- -------- ----------------

--------- ----- -- -------- ------------------

inventory CONDITION A. One or more DGs with fuel level -[33,000] gal and ihu>- [28,285 -gaI in storage tank.REQUIRED ACTION/COMPLETION TIME 0w A.1 Restore fuel oil I to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ', within limits. (Unit<53,225 galand45)25 g(Unit-2)B. One or more DGs with lube B.1 Restore lube oil inventory 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months oil inventory

<r 500]-,gal to within limits.ad .- [42 l.- < 330 gal and _ 283 gal C. One or more DGs with C.1 Restore fuel oil total 7 days stored fuel oil total particulates to within limits.particulates not within limit.D. One or more DGs with new D.1 Restore stored fuel oil 30 days fuel oil properties not within properties to within limits.limits.E. One or more DGs with E.1 Restore starting air 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months starting air receiver receiver pressure to pressure < [22.51 .r g[425]_1 165 psig and,- 125.psigj (Unit 1).~3Tpsig~ang

__ '2854 (Unit24 165 ýpsig' (U-nit 1)380 psitgj(Unit 2)WOG STS 3.8.3 -1 Rev. 2, 04/30/01 25 I Rev. 4, Change 13 1 Diesel Fuel Oil, Lube Oil, and Starting Air 3.8.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME F. Required Action and F.1 Declare associated DG Immediately associated Completion inoperable.

Time not met.OR One or more DGs with diesel fuel oil, lube oil, or starting air subsystem not within limits for reasons CTS other than Condition A, B, C, D, or E..17,500 g l ffuiel o fil (Unhit 1)SURVEILLANCE REQUIREMENTS SURVEIL F FREQUENCY SIR 3.8.3.1 Verify e~acrlol storage tank contains_

[33L,0,00]

galu 31 days el-fuel. l-330 gal S R 3.8.3.2 Verify lubricating oil inventory is >-450]-ga*ý[:

31 days SR 3.8.3.3 Verify fuel oil properties of new and stored fuel oil are In accordance tested in accordance with, and maintained within the with the Dieselof, the Diesel Fuel Oil Testing Program. Fuel Oil Testing Program SR 3.8.3.4 Verify each DG air start receiver pressure is 31 days SR 3.8.3.5 Check for and remov ccumulated water from each 3 ays fuel oil storage tank. 92 CTS Value 65 psig (Unit 1)ý80 p iý (u it 1 WOG STS 3.8.3 -2 Rev. 2, 04/30/01 26 Rev. 4 Change 13 1 Diesel Fuel Oil, Lube Oil, and Starting Air B 3.8.3 BASES ACTIONS (continued) 92 4 performance has been recently demonstrated (within 34 days), it is prudent to allow a brief period prior to declaring the associated DG inoperable.

The 7 day Completion Time allows for further evaluation, resampling and re-analysis of the DG fuel oil. 5 D.1 3.8.3.3 With the new fuel oil properties defined in the Bases for SR ...not within the required limits, a period of 30 days is allowed for restoring the stored fuel oil properties.

This period provides sufficient time to test the stored fuel oil to determine that the new fuel oil, when mixed with previously stored fuel oil, remains acceptable, or to restore the stored fuel oil properties.

This restoration may involve feed and bleed procedures, filtering, or combinations of these procedures.

Even if a DG start and load was required during this time interval and the fuel oil properties were outside limits, there is a high likelihood that the DG would still be capable E.1 165 psig for Unit 1, and 380 psig for Unit 2 With starting air receiver pressure < , sufficient capacity for five successive DG start attempts does not exist. However, as long as the receiver pressýureý , there is adequate capacity for at least 25 psig for Unit 1, asempt, and the DG can be considered OPERABLE while the id > 285 psig for air receiver pressure is restored to the required limit. A period of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months nits considered sufficient to complete restoration to the required pressure prior to declaring the DG inoperable.

This period is acceptable based on the remaining air start capacity, the fact that most DG starts are accomplished on the first attempt, and the low probability of an event during this brief period.F._1 E 2 With a Required Action and associated Completion Time not et, orrone or more DG's fuel oil, lube oil, or starting air subsystem not ithin limits for reasons other than addressed by Conditions A through , the associated DG may be incapable of performing its intended function and must be immediately declared inoperable.

WOG STS B 3.8.3 -4 Rev. 2, 04/30/01 109 Rev. 4, Change 13 Inserts 3.8.1 and 3.8.3 (continued)

Lube oil LCO, Condition, and SR for Unit 1 LCO The lube oil subsystem shall be within limits for each required diesel generator(DG).

Condition Required Action Completion Time B. One or more DGs B.1 Restore lube oil inventory 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months lube oil inventory within limits.< 330 gal and_> 283 gal SR 3.8.3.3 surveillance and frequency Surveillance Requirement Frequency SR 3.8.3.3 Verify fuel oil properties of new and stored fuel oil are In accordance tested in accordance with, and maintained within the with the Diesel limits of, the Diesel Fuel Oil Testing Program. Fuel Oil Testing Program Starting Air LCO, Condition, and SR LCO The starting air subsystem shall be within limits for each required diesel generator (DG).Condition Required Action Completion Time E. One or more DGs E.1 Restore starting air 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months starting air receiver receiver pressure to pressure < 1665,PSig

> 165 pSig(Unit i1), and ---125 ýsiq gjU -~Unit 2).'Surveillance Requirement Frequency SR 3.8.3.4 Verify DG air start receiver pressure is _165 psig 31 days i(J nit 1), 38 pi -j(Unit 2)-190 Rev. 4 Change 13 BVPS ISTS Conversion

3.8 Electrical

Power Systems Enclosure 3 Changes to CTS M.14 CTS 3.8.1.1 does not require specific limits or requirements for starting air system for DG OPERABILITY.

ITS LCO 3.8.3 adds a requirement for starting air system to be OPERABLE when an associated DG is required to be OPERABLE.

ITS 3.8.3 Condition E is added. Condition E specifies one or more DGs with starting air receiver pressure< 165 psig and > 125 psig, the receiver pressure must be restored to >_ 165 psig within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . These requirements are for Unit 1. For Unit 2, the Condition specifies one or more DGs with starting air receiver pressure < 380 psig and _> 285 psig restore the receiver pressure to > 380 within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . Failure to comply with the specified Actions results in declaring the associated DG inoperable.

ITS SR 3.8.3.4 is added and requires the verification that DG air start receiver pressure is 2! 165 psig for Unit 1 and >_ 380 psig for Unit 2 every 31 days. This changes the CTS by adding the appropriate requirements for starting air system to ensure DG OPERABILITY.

The purpose of the ITS LCO, Action, and surveillance requirements are to ensure the DG air start capacity is maintained within the design requirements for 5 start attempts.The addition of these requirements is acceptable because they provide additional assurance the DG is capable of starting within the time limit assumed by the safety analysis for analyzed events. The proposed ITS requirements are consistent with the ISTS for these requirements.

This change is designated as more restrictive because it adds additional technical specification requirements that the CTS does not specify.M.15 CTS surveillance requirement 4.8.1.1.2.b.2 requires a verification that the generator (DG) is capable of rejecting a load _ 825 kw without tripping and without exceeding 64.4 Hz for Unit 2. Unit 1 surveillance requirement states; verify the generator capability to reject a load of > 615 kw without tripping and without exceeding 66.2 Hz. ITS SR 3.8.1.8 states Verify each DG rejects a load greater than or equal to its associated single largest post-accident load, and following load rejection, the frequency is <_ 66.2 Hz (Unit 1) or _64.4 Hz (Unit 2). The SR additionally requires that within 3 seconds following load rejection, the voltage is > 4106 V and < 4368 V for (Unit 1), or > 3994 V and < 4368 V for (Unit 2), and within 4 seconds following load rejection, the frequency is > 58.8 Hz and <61.2 Hz (Unit 1) or -> 59.9 Hz and < 60.3 Hz (Unit 2). This changes the CTS by adding additional requirements to the surveillance requirement.

The purpose of the ITS SR 3.8.1.8 limitation on voltage and frequency after the transient is to ensure the response of the DG is within a specific band. The addition of these requirements is acceptable because they ensure the DG is capable of responding to a transient within specified limits. The ITS requirements are consistent with the applicable Regulatory Guide 1.9 requirements.

This change is designated as more restrictive because it adds additional surveillance requirement that the CTS does not require.M.16 CTS Action a requires with one offsite circuit inoperable, it must be restored to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . CTS Action b states with one diesel generator inoperable, it must be restored to OPERABLE status within 14 days. The corresponding ITS Action A requires with one offsite circuit inoperable, the circuit must be restored to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and within 17 days from discovery of failure to meet the LCO. Corresponding ITS Action B states with one DG inoperable, the DG must be restored to OPERABLE status within 14 days and within 17 days from discovery of failure to meet the LCO. The CTS is revised to conform to the ITS. This changes the CTS by adding an additional restriction for an inoperable offsite circuit or DG that limits the total time for not meeting the LCO.BVPS Units 1 & 2 Page 17 Revision 4, 10/06 229 BVPS UNITS 1 & 2 IMPROVED TECHNICAL SPECIFICATION (ITS) CONVERSION LICENSE AMENDMENT REQUEST (LAR)Nos. 296 (UNIT 1) & 1 69 (UNIT 2)REVISION 4 CHANGE 14 Description This change addresses revisions made to the current Technical Specification (CTS) discussion of changes (DOCs) in the ITS conversion documentation.

The changes are made to improve the completeness and accuracy of the DOCs.DOC L.1 for CTS 3.1.1.5, "Minimum Temperature for Criticality", is revised to change a >symbol to a > symbol consistent with the symbol used in the CTS surveillance referenced in the DOC. The DOC is also revised to discuss the inconsistency of the greater than and greater than or equal to symbols used in the CTS LCO requirement and the associated Surveillance Requirement (SR). The change is necessary to make the DOC more accurately reflect the referenced surveillance and proposed change. The change does not affect the ITS.DOC L.3 for CTS 3.7.1.2, "Auxiliary Feedwater (AFW) System", to include CTS Actions a and b as well as Actions c and d consistent with the markup of CTS 3.7.1.2. The change is necessary to make the DOC more accurately reflect the associated CTS markup. This change also does not affect the ITS.Affected Pages: The affected pages are presented in ITS Section order. The Table(s) preceding each ITS section list the affected pages by type (i.e., ITS markup, CTS markup, etc.) in that Section of the ITS. The affected page numbers listed are the ITS section specific consecutive numbers found in the lower right corner of each page.Note: Because the affected page(s) for each change were extracted from a complete ITS section electronic file, the electronic hyperlinks (created in the complete ITS section file) do not work in the collection of affected pages that follow this cover page.

BVPS UNITS 1 & 2 ITS CONVERSION LICENSE AMENDMENT REIQUEST (LAR)Nos. 296 (UNIT 1) & 1169 (UNIT 2)REVISION 4 CHANGE 14 AFFECTED PAGES FOR ITS SECTION 3.4 (REACTOR COOLANT SYSTEM)ITS SECTION 3.4 (REACTOR COOLANT SYSTEM) INDEX OF AFFECTED PAGES ITS MARKUPS NONE ITS JFDS NONE ITS BASES MARKUPS NONE ITS BASES JFDS NONE CTS MARKUPS NONE CTS DOCS PAGES: 298 Rev. 4, Change 14 BVPS ISTS Conversion

3.4 Reactor

Coolant System Enclosure 3 Changes to CTS CTS 3.1.1.5 Minimum Temperature For Criticality ITS 3.4.2 RCS Minimum Temperature for Criticality DISCUSSION OF CHANGE (DOC)Less Restrictive Changes (L)L.1 (Category 7- Relaxation Of Surveillance Frequency)

CTS Surveillance 4.1.1.5 states that the RCS Tavg shall be determined to be > 541 °F within 15 minutes prior to achieving reactor criticality and every 30 minutes when the RCS Tavg < 551°F and the Tavg deviation alarm not reset. The corresponding ISTS SR 3.4.2.1 requires RCS Tavg in each loop to be verified to be _> 541 OF every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The CTS is revised to conform to the ISTS. This changes the CTS Surveillance Frequency by requiring that the RCS Tavg for each loop be verified every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months which would include one verification within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months prior to achieving criticality.

In addition, this change makes the greater than symbol used in the surveillance consistent with the required greater than or equal to symbol specified in the LCO.In accordance with the general rules of TS usage as required by ISTS SR 3.0.4, a Surveillance must be performed within the specified Frequency prior to entering the MODE or other specified condition in the Applicability of the TS. The Applicability of ITS 3.4.2 starts in Mode 2 with Keff _ 1 (criticality).

Therefore, the revised surveillance would be required to be performed at least once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months prior to achieving criticality and every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while in the Mode of applicability.

The proposed change is acceptable because the new Surveillance Frequency continues to provide adequate assurance that the RCS temperature is maintained above the minimum requirement specified in the TS. The conditional CTS surveillances are replaced with a single continuous surveillance that requires the RCS temperature be verified every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Given the availability of indications in the control room, including alarms, and the relatively constant RCS temperatures during normal operation, the 12-hour interval is frequent enough to prevent inadvertent violation of the LCO and trend operating conditions while not distracting the operators with unnecessarily repetitive verifications.

In addition, the approach to criticality is a carefully controlled evolution during which RCS temperature is closely monitored.

As such, more frequent temperature verifications during an approach to criticality (e.g., every 15 minutes as in the CTS) are not necessary to assure the LCO requirements are met and may be an unnecessary distraction to operations personnel.

This change is designated as less restrictive because Surveillances will be performed less frequently under the ITS than under the CTS.More Restrictive Changes (M)None Removed Detail.Changes (LA)None BVPS Units 1 & 2 Page 5 Revision 4, 10/06 298 BVPS UNITS I & 2 ITS CONVERSION LICENSE AMENDMENT REQUEST (LAR)Nos. 296 (UNIT 1) & 1 69 (UNIT 2)REVISION 4 CHANGE 14 AFFECTED PAGES FOR ITS SECTION 3.7 (PLANT SYSTEMS)ITS SECTION 3.7 (PLANT SYSTEMS) INDEX OF AFFECTED PAGES ITS MARKUPS NONE ITS JFDS NONE ITS BASES MARKUPS NONE ITS BASES JFDS NONE CTS MARKUPS NONE CTS DOCS PAGES: 272 Rev. 4, Change 14 BVPS ISTS Conversion

3.7 Plant

Systems Enclosure 3 Changes to CTS AFW pump testing will continue on a specified IST Frequency but not necessarily on an equally staggered basis. Planned maintenance and testing will typically ensure the AFW pumps are not tested at the same time. The change does not affect the method of AFW pump testing or the capability of the pumps to perform their safety function as assumed in the safety analyses.

This change is designated as less restrictive because specific details of testing on a STB have been eliminated.

L.3 (Category 3 -Relaxation of Completion Time) CTS 3.7.1.2, Actions a, b, c, and d include the requirement that the plant must be in HOT SHUTDOWN "within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ." This is a total shutdown Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to MODE 4. The corresponding ITS 3.7.5 shutdown Action Condition D states the plant must be in MODE 4 within 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months . This changes the time to be in MODE 4 from 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months . The CTS has been revised to incorporate and extend the Completion Time by an additional 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to be in MODE 4. This change is being made so the BVPS ITS is consistent as possible with NUREG-1431.

The purpose of the affected portions of CTS 3.7.1.2 Actions a through d is to place the unit in a condition where the LCO does not apply. ITS 3.7.5, Condition D, requires placing the unit in MODE 4. In MODE 4, only one of the three AFW trains are required OPERABLE.

This change is acceptable because the Completion Time is reasonable, based upon operating experience, to reach the required unit conditions from full power conditions without challenging unit systems. This change is designated as less restrictive because additional time is allowed to restore parameters to within the LCO limits prior to exiting the Mode of Applicability than was allowed in the CTS.L.4 (Category 5 -Deletion of Surveillance Requirement)

CTS surveillance 4.7.1.2.2 requires cycling each manual Service Water to Auxiliary Feedwater System valve through at least one complete cycle at least once per 31 days. In addition, CTS surveillance 4.7.1.2.6 requires cycling each power operated (excluding automatic) valve in the AFW flowpath through at least one complete cycle at least once per 18 months during shutdown.

The corresponding ITS 3.7.5 does not contain a similar surveillance requirement for either 4.7.1.2.2 or 4.7.1.2.6.

The CTS is revised to conform to the ISTS. This changes the CTS by deleting surveillance requirements 4.7.1.2.2 and 4.7.1.2.6.

The purpose of a TS surveillance is to confirm the associated system is capable of performing its intended safety function (i.e., the system is operable).

The proposed change is acceptable because the CTS surveillances being deleted are not required to confirm the capability of the AFW system to perform its required safety function.The affected CTS surveillances verify the capability of the non-automatic power-operated and manual valves to be cycled. However, the remaining AFW surveillances assure non-automatic power operated and manual valves are pre-positioned and verified in the required position to support the AFW system operation.

ITS SR 3.7.5.1 requires each manual, power operated, and automatic valve in the flow path that is not locked sealed or otherwise secured in position is in the correct position (note that this surveillance includes the position of non-automatic power operated valves). ITS 3.7.5.3 verifies that each automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to BVPS Units 1 & 2 Page 4 Revision 4, 10/06 272 BVPS UNITS 1 & 2 IMPROVED TECHNICAL SPECIFICATION (ITS) CONVERSION LICENSE AMENDMENT REQUEST (LAR)Nos. 296 (UNIT 1) & 1 69 (UNIT 2)REVISION 4 CHANGE 15 NRC Comment Resolution NRC Reviewer:

R. Clark Description The Bases for ITS SR 3.8.1.14 discusses the required diesel generator testing. The ITS bases text is revised to resolve an NRC concern regarding the ITS description of the allowance to perform testing in any series of sequential, overlapping, or total steps so that the entire diesel generator connection and loading sequence is verified.

The additional text added by this change clarifies the intent of the ITS allowance for testing in sequential, overlapping, or total steps and places it in better context.Affected Pages: The affected pages are presented in ITS Section order. The Table(s) preceding each ITS section list the affected pages by type (i.e., ITS markup, CTS markup, etc.) in that Section of the ITS. The affected page numbers listed are the ITS section specific consecutive numbers found in the lower right corner of each page.Note: Because the affected page(s) for each change were extracted from a complete ITS section electronic file, the electronic hyperlinks (created in the complete ITS section file) do not work in the collection of affected pages that follow this cover page.

BVPS UNITS 1 & 2 ITS CONVERSION LICENSE AMENDMENT REQUEST (LAR)Nos. 296 (UNIT 1) & 1 69 (UNIT 2)REVISION 4 CHANGE 15 AFFECTED PAGES FOR ITS SECTION 3.8 (ELECTRICAL POWER SYSTEM)ITS SECTION 3.8 (ELECTRICAL POWER SYSTEMS) INDEX OF AFFECTED PAGES ITS MARKUPS NONE ITS JFDS NONE ITS BASES MARKUPS PAGES: 95 ITS BASES JFDS NONE CTS MARKUPS NONE CTS DOCS NONE AC Sources -Operating B 3.8.1 Rev. 4 Change 15 BASES SURVEILLANCE REQUIREMENTS (continued) 2 The requirement to verify the connection and power supply of permanent and autoconnected loads is intended to satisfactorily show the relationship of these loads to the DG loading logic.In certain circumstances, many of these loads cannot actually be connected or loaded without undue hardship or potential for undesired operation.

For instance, Emergency Core Cooling Systems (ECCS)injection valves are not desired to be stroked open, or high pressure injection systems are not capable of being operated at full flow.b. ance of the SR will not cause perturbations to an e electrical is n systems that could result allenge to steady state operation ant s ystems, and c. Performance , or failure of the , ot cause, or result in with attendant challenge to plant safetys SR 3.8.1.4-9 1j4 (In the event of a DBA coincident with a loss of offsite power, the DGs are required to supply the necessary power to ESF systems so that the fuel, RCS, and containment design limits are not exceeded.This Surveillance demonstrates the D operation, as discu.id-the Bases for SR 3.8.1.11, during a loss of o ite power actuation test signal in conjunction with an ESF actuation signal. In lieu of actual demonstration of connection and loading of loads, testing that adequately shows the capability of the DG system to perform these functions is acceptable.

This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified.The Frequency of [18 months] takes into consideration unit conditions required to perform the Surveillance and is intended to be consistent with an expected fuel cycle length of [18 months].The 10-second start requirement supports the assumptions of the design basis accident analyses described in the UFSAR (Ref. 5). The 10-second timing requirement begins when the DG start signal is received by the DG start circuit and the time it instrumen loss of vo emergen d does not include This SR is modified by two Notes. The reason for Note 1 is to minimize ttakes the we on the DGs during testing. For the purpose of this testing, ntation to detect a CO g ntamd)i...'

Itage on the 9 e DGs must be started from standby conditions, that is, with the engine ltage on toh ee c a .., , cy busses. coolant and oil continuously circulated and temperature maintained consistent with manufacturer recommendations for DGs. The reason for cNote 2 is that the performance of the Surveillance would rem- a required offsite circuit from service, perturb the electrical distribution 1,2,3, or4 system, and challenge safety systems. This restriction from normally performing the Surveillance in MODE-1:-2 is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines plant safety is maintained or enhanced.

This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system Barring of the engine may be performed prior to DG start without invalidating the requiremen t for starting from standby conditions.

WOG STS B 3.8.1 -31 Rev. 2, 04/30/01 95 BVPS UNITS 1 & 2 IMPROVED TECHNICAL SPECIFICATION (ITS) CONVERSION LICENSE AMENDMENT REQUEST (LAR)Nos. 296 (UNIT 1) & 1 69 (UNIT 2)REVISION 4 CHANGE 16 Description This change provides various minor editorial and format revisions.

These revisions do not introduce any technical changes to the proposed ITS and are only intended to make the BVPS ITS conform more closely with the standard technical specifications (NUREG-1431) format and presentation.

The proposed changes include improving the clarity of markups, spelling out acronyms on first use, adding an acronym to complete an LCO title, inserting quotation marks in LCO titles, and consistently capitalizing the words OPERABILITY and MODE in the ITS.Affected Pages: The affected pages are presented in ITS Section order. The Table(s) preceding each ITS section list the affected pages by type (i.e., ITS markup, CTS markup, etc.) in that Section of the ITS. The affected page numbers listed are the ITS section specific consecutive numbers found in the lower right corner of each page.Note: Because the affected page(s) for each change were extracted from a complete ITS section electronic file, the electronic hyperlinks (created in the complete ITS section file) do not work in the collection of affected pages that follow this cover page.

BVPS UNITS I & 2 ITS CONVERSION LICENSE AMENDMENT REQUEST (LAR)Nos. 296 (UNIT 1) & 1 69 (UNIT 2)REVISION 4 CHANGE 16 AFFECTED PAGES FOR ITS SECTION 3.1 (REACTIVITY CONTROL SYSTEMS)ITS SECTION 3.1 (REACTIVITY CONTROL SYSTEMS) INDEX OF AFFECTED PAGES ITS MARKUPS PAGES: 12,28 ITS JFDS NONE ITS BASES MARKUPS NONE ITS BASES JFDS NONE CTS MARKUPS NONE CTS DOCS NONE Rev. 4, Change 16 1 Rod Group Alignment Limits 3.1.4 3.1 REACTIVITY CO 3.1.4 Rod Group?LCO 3.1.4 APPLICABILITY:

NTROL SYSTEMS Limits Ul & U2 CTS LCOs I I ,1(as determined in"1 accordance with Specification 3.1.7, Rod All shutdown and control rods shall be OPERABLE.

Position Indication3 AND U1 CTS Individual indicated rod positions shall be within 12 stepsof their group step counter demand position.MODES 1 and 2.NOTE For Unit4 o, verification of rod OPERABILITY and that the individual indicated rod positions are within the 12 step limit is not required during rod motion and for the first hour following rod motion.ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more rod(s) A.1.1 Verify SDM to be within the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months inoperable, limits specified in the COLR.OR A.1.2 Initiate boration to restore 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months SDM to within limit.AND A.2 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months B. One rod not within B.1 Restore rod to within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months alignment limits, alignment limits.OR B.2.1.1 Verify SDM to be within the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months limits specified in the COLR.OR WOG STS 3.1.4-1 Rev. 2, 04/30/01 12 Rev. 4 Change 16 PHYSICS TESTS Exceptions

-MODE 2 3.1 REACTIVITY CONTROL SYSTEMS 3.1.% PHYSICS TESTS Exceptions

-MODE 2 LCO .1. During the performance of PHYSICS TESTS, the requirements of: LCO 3.1.3, "Moderator Temperature Coefficient," TSTF-315 LCO 3.1.4, "Rod Group Alignment Limits," LCO 3.1.5, "Shutdown Bank Insertion Limits," LCO 3.1.6, "Control Bank Insertion Limits," and required channels LCO 3.4.2, "RCS *mum Temperature for Criticality" may be suspende , anAe number of required channels for LCO 3.3.1,"RTS Instrumentation," Functions 2,3,-6-and

-16e, may be reduced to 3, provided that: TSTF-315 2 1 a. RCS lowest loop average temperature is > [531]°F, 1 b. SDM is within the limits specified in the COLR, and c. THERMAL POWERis " 5% RTP TSTF-14 R4 During PHYSICS TESTS initiated in MODE 2.CTS and, 2. 'or. Unit I onIy, primary detector voltage measurements may be used to determine the position of rods in shutdown banks A and B and control banks A and B for the purpose of satisfying Specification 3.1.7.1.APPLICABILITY:

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. SDM not within limit. A.1 Initiate boration to restore 15 minutes SDM to within limit.AND A.2 Suspend PHYSICS TESTS 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months exceptions.

B. THERMAL POWER not B.1 Open reactor trip breakers.

Immediately within limit.C. RCS lowest loop average C.1 Restore RCS lowest loop 15 minutes temperature not within average temperature to limit, within limit.WOG STS 3.1.t~Rev. 2, 04/30/01 28 BVPS UNITS I & 2 ITS CONVERSION LICENSE AMENDMENT REQUEST (LAR)Nos. 296 (UNIT 1) & 1 69 (UNIT 2)REVISION 4 CHANGE 16 AFFECTED PAGES FOR ITS SECTION 3.4 (REACTOR COOLANT SYSTEM)ITS SECTION 3.4 (REACTOR COOLANT SYSTEM) INDEX OF AFFECTED PAGES ITS MARKUPS PAGES: 52 ITS JFDS NONE ITS BASES MARKUPS NONE ITS BASES JFDS NONE CTS MARKUPS NONE CTS DOCS NONE Rev. 4, Change 16 RCS Isolated Loop Startup 3.4.18 3.4 REACTOR COOLANT SYSTEM (RCS)REFORMATTED BVPS CTS REQUIREMENTS FOR ISOLATED LOOP STARTUP 3.4.18 RCS Isolated Loop Startup LCO 3.4.18 Each RCS isolated loop shall remain isolated with the hot and cold leg isolation valves closed: -] (SDM)" a. If the boron concentratii n in the isolated I op is < required to satis the applicable requireme ts of LCO 3.1.1, HUTDOWN MARGI (in MODE 5) and LCO 3.9.1, Boron Concentration, in MODE 6), and b. Until the isolated portion of the loop has been drai d and refilled from the refueling water storage tank or RCS.MODES 5 and 6 when an RCS loop has been isolated > 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or drained.APPLICABILITY:

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. LCO requirement(s) not A.1 Isolate affected RCS Immediately met. loop(s) by closing the hot and cold leg isolation valves.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.18.1 Verify the isolated loop has been drained and refilled Prior to opening with water from the refueling water storage tank or the isolated loop RCS.(SDM)" hot or cold leg//, ] isolation valve SR 3.4.18.2 Verify the isolated loop boron concen tion is > the Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months required value to satisfy the applic e requirements of LCO 3.1.,* SHUTDOWN MARGI (in MODE 5) and teiol loop LCO 3[ Boron Concentration n MODE 6). hot or cold leg W isolation valve SR 3.4.18.3 Verify the isolated loop hot or cold leg isolation valve is Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months opened. following completion of refilling the isolated loop.Beaver Valley Units 1 and 2 3.4.18- 1 52 BVPS UNITS 1 & 2 ITS CONVERSION LICENSE AMENDMENT REIQUEST (LAR)NoS. 296 (UNIT 1) & 1 69 (UNIT 2)REVISION 4 CHANGE 16 AFFECTED PAGES FOR ITS SECTION 3.5 (EMERGENCY CORE COOLING SYSTEM)ITS SECTION 3.5 (EMERGENCY CORE COOLING SYSTEM) INDEX OF AFFECTED PAGES ITS MARKUPS PAGES: 6,8 ITS JFDS NONE ITS BASES MARKUPS NONE ITS BASES JFDS NONE CTS MARKUPS NONE CTS DOCS NONE Rev. 4 Change 16 1 ECCS -Operating 3.5.2 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)3.5.2 ECCS -Operating LCO 3.5.2 one of the required charging I System (OPPS I-Two ECCS trains shall be OPERABLE.-NOTES -l-1. In MODE 3, both safety. -..tiR ..Sl pump flow paths may be isolated by closing the isolation valves for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to perform pressure isolation valve testing per SR 3.4.14.1.2. In MODE 3, EGGS pumps may be made incapable of injecting to).,. ] support transition into or from the Applicability of LCO 3.4.12,"L O erre s rePrteti m& P-=!, for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or until the temperature of all RCS cold legs exceeds [3751F]1 [Low Temnperature Overpressure Protection (ILOP) aFmiiemperature specified in the PTLR plus-f251 0 F], whichever comes first.-]\the OPPS enable CTS APPLICABILIT 1 lY: MODES 1, 2, and 3.1Unit 1 CTS 3.

In MODE 3, the ECCS automatic High Head Safety Injection (HHSI) flow path may be isolated to support transition into or from the Applicability of LCO 3.4.12,"Overpressure Protection System (OPPS)" for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or until the temperature of all RCS cold legs exceeds the OPPS enable temperature specified in the PTLR plus 25°F, whichever comes first-ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more trains A.1 Restore train(s) to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable.

OPERABLE status.B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND B.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months C. Less than 100% of the C.1 Enter LCO 3.0.3. Immediately ECCS flow equivalent to a single OPERABLE ECCS train available.

WOG STS 3.5.2- 1 Rev. 2, 04/30/01 6 Rev. 4 Change 16 INSERTS FOR ITS 3.5.2 ECCS -Operating 1. ForUnlit~

1oly Number MOV-1 SI-890A MOV-1 SI-890B MOV-1SI-890C MOV-1SI-869A MOV-1SI-869B Position Closed Closed Open Closed Closed Function Low head safety injection (LHSI) to Hot Leg LHSI to Hot Leg LHSI to Cold Leg HHSI Pump to Hot Leg HHSI Pump to Hot Leg Number 2SIS*MOV8889 2SIS*MOV869A 2SIS*MOV869B 2SIS*MOV841 2CHS*MOV8132A 2CHS*MOV8132B 2CHS*MOV8133A 2CHS*MOV8133B Position Closed Closed Closed Open Open Open Open Open Function LHSI to Hot Legs HHSI to Hot Leg HHSI to Hot Leg HHSI to Cold Leg HHSI Pump Discharge Cross Connect HHSI Pump Discharge Cross Connect HHSI Pump Discharge Cross Connect HHSI Pump Discharge Cross Connect 2.SR 3.5.2.2 Verify the HHSI pump minimum flow valve is open with 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months power to the valve operator removed.8 BVPS UNITS I & 2 ITS CONVERSION LICENSE AMENDMENT REQUEST (LAR)Nos. 296 (UNIT 1) & 1 69 (UNIT 2)REVISION 4 CHANGE 16 AFFECTED PAGES FOR ITS SECTION 5.0 (ADMINISTRATIVE CONTROLS)ITS SECTION 5.0 (ADMINISTRATIVE CONTROLS)

INDEX OF AFFECTED PAGES ITS MARKUPS PAGES: 53 ITS JFDS NONE ITS BASES MARKUPS NONE ITS BASES JFDS NONE CTS MARKUPS NONE CTS DOGS NONE Rev. 4, Change 11 & 16 Section 5.0 Inserts in each SG, and Repair method utilized and the number of tubes repaired by each repair method.2. A report shall be submitted within 90 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.5.2, Steam Generator Program, when voltage-based alternate repair criteria have been applied. The report shall include information described in Section 6.b of Attachment 1 to Generic Letter 95-05, "Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking." 3. For implementation of the voltage-based repair criteria to tube support plate intersections, notify the Commission prior to returning the steam generators to service (MODE 4) should any of the following conditions arise: a. If circumferential crack-like indications are detected at the tube support plate intersections.

b. If indications are identified that extend beyond the confines of the tube support plate.c. If indications are identified at the tube support plate elevations that are attributable to primary water stress corrosion cracking.4. Report the following information to the NRC within 90 days after achieving l--MODE 4 following an outage in which the F* methodology was applied: 16 a. Total number of indications, location of each indication, orientation of F each indication, severity of each indication, and whether the indications initiated from the inside or outside surface.b. The cumulative number of indications detected in the tubesheet region as a function of elevation within the tubesheet.

c. The projected end-of-cycle accident-induced leakage from tubesheet indications.

serts Page 21 53 Section 5.0 Ins BVPS UNITS I & 2 IMPROVED TECHNICAL SPECIFICATION (ITS) CONVERSION LICENSE AMENDMENT REQUEST (LAR)Nos. 296 (UNIT 1) & 1"69 (UNIT 2)REVISION 4 CHANGE 17 Description This change revises the Bases of ITS 3.7.4, "Atmospheric Dump Valves (ADVs)." The requirements of this technical specification are new for BVPS. The change has been found to be acceptable by FENOC Design Analysis after reviewing the Westinghouse Extended Power Uprate steam generator tube rupture analysis.

The change updates the Unit 2 safety analyses discussion in the Bases to include new information regarding the SG overfill and radiological dose aspects of the steam generator tube rupture accident analysis.

As identified in FENOC Condition Report 06-04838, the capacities of the BVPS Unit 2 Atmospheric Steam Dump Valves (ASDVs) and Residual heat Release valve (RHRV) were determined to be lower than had previously been assumed as inputs to several Westinghouse design basis accident analyses.The ADVs and RHRV are installed with inlet and outlet piping that is of a smaller diameter than the nominal valve sizes. Therefore, the applicable analyses were revised to account for the lower valve capacities.

The Bases for ITS 3.7.4 have been updated to reflect these revised analyses.

Other minor changes are made to improve the clarity and accuracy of the bases discussions.

Affected Pages: The affected pages are presented in ITS Section order. The Table(s) preceding each ITS section list the affected pages by type (i.e., ITS markup, CTS markup, etc.) in that Section of the ITS. The affected page numbers listed are the ITS section specific consecutive numbers found in the lower right corner of each page.Note: Because the affected page(s) for each change were extracted from a complete ITS section electronic file, the electronic hyperlinks (created in the complete ITS section file) do not work in the collection of affected pages that follow this cover page.

BVPS UNITS 1 & 2 ITS CONVERSION LICENSE AMENDMENT REQUEST (LAR)NOS. 296 (UNIT 1) & 1 :69 (UNIT 2)REVISION 4 CHANGE 17 AFFECTED PAGES FOR ITS SECTION 3.7 (PLANT SYSTEMS)ITS SECTION 3.7 (PLANT SYSTEMS) INDEX OF AFFECTED PAGES ITS MARKUPS NONE ITS JFDS NONE ITS BASES MARKUPS PAGES: 97, 100, 101 ITS BASES JFDS NONE CTS MARKUPS NONE CTS DOCS NONE I Rev. 4,Change 17 1 ADVs B 3.7.4 BASES ACTIONS A.1 With one required ADV line inoperable, action must be taken to restore OPERABLE status within 7 days. The 7 day Completion Time allows for TSTF-359 the redundant capability afforded by the remaining OPERABLE ADV T\ lines, a nonsafety grade backup in the Steam Byp System, and MSSVs. Required..........

.db LCO 3.0.4 does net apply.1 I condenser steam dump valves T3 In this condition, the unit utilizes RHR for cooling.Therefore, operation may continue with one or more ADV lines inoperable because the RCS cooling function required to mitigate a SGTR event would be accomplished by the RHR system.The requirement to stroke the valve through the full range of operation may be accomplished by remote manual control. In addition, this surveillance must also verify the capability to locally operate each ADV. The verification of local operation does not require that the ADV be stroked through the full range of travel (i-e., if the valve is stroked full open and closed by remote manual operation, the capability to operate the ADV locally may be verified by observing valve stem movement).

The ADVs must be capable of both remote and local manual operation in order to be considered OPERABLE.DI With two or more ADV lines inoperable, action must be ken to restore all but one ADV line to OPERABLE status. Since the ock valve can be closed to isolate an ADV, some repairs may be possi e with the unit at power. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time is reasonable o repair inoperable ADV lines, based on the availability of the Steam 8Dpass System and MSSVs, and the low probability of an event occurring during this period that would require the ADV lines.C.1 and C.2 If the ADV lines cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . and in MODE 4, without reliance upon steam generator for heat removal, within 1241 hours0.0144 days 0.345 hours 0.00205 weeks 4.722005e-4 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SR 3.7.4.1 To perform a controlled cooldown of the RCS, the ADVs must be able to be opened either remetely or locally and throttled through their full range.This SR ensures that the ADVs are tested through a full control cycle at least once per fuel cycle. Performance of inservice testing or use of an ADV during a unit coolown may satisfy this requirement.

Operating experience has shown that these components usually pass the Surveillance when performed at the 1181 month Frequency.

The Frequency is acceptable from a reliability standpoint.

WOG STS B 3.7.4 -3 Rev. 2, 04/30/01 97 INSERTS FOR ITS 3.7.4 BASES Rev. 4, Change 17 Basis Accidents (DBAs). Thus, the SGTR is the limiting event for the ADVs.For Unit 1, three ADVs with associated flow paths and isolation valves are required OPERABLE.

Due to the design of the Unit 1 Residual Heat Release Valve, it can not be isolated from a SG with a ruptured tube. Therefore, the Unit 1 Residual Heat Release Valve is not used to mitigate a SGTR due to the dose requirements of the accident analysis.

The requirement for three OPERABLE ADV lines provides assurance that a single active failure of one ADV line or a single active failure of the instrument air supply will not prevent the mitigation of a SGTR accident.The Unit 1 operational assessment used to evaluate the single failures described above also assumes that one ADV is lost to the faulted SG. In the case where the instrument air supply is available and an active failure of one of the remaining ADVs is assumed, the operational assessment assumes the remaining ADV is operated from the control room to successfully mitigate the SGTR accident.

In the case where the active failure is a loss of instrument air, and ADV operation is delayed, the operational assessment assumes the two remaining ADVs are operated by local manual control to successfully mitigate the SGTR accident.

Therefore, the Unit 1 ADVs must be capable of both remote and local manual operation to be considered OPERABLE.

The Unit 1 operational assessment does not include a specific time to manually unblock an ADV. Therefore, the Unit 1 ADV block valves must remain open for the ADV lines to be considered OPERABLE.For Unit 2, four ADVs with associated flow paths and isolation valves are required OPERABLE to satisfy the SGTR accident analysis assumptions of a single active failure and loss of offsite power. Requiring four Unit 2 ADVs OPERABLE assures that two ADVs will remain OPERABLE for the SGTR analysis overfill case (i.e., one ADV lost to the faulted SG and one ADV lost to a single active failure).

Additionally, requiring four Unit 2 ADVs OPERABLE assures that three ADVs will remain OPERABLE for the SGTR radiological dose case. The radiological dose case includes the loss of one ADV as a single active failure (i.e., the ADV on the faulted SG fails open).The Unit 2 SGTR analysis requires that two ADVs (overfill case) or three ADVs (bounding dose case) remain OPERABLE to mitigate the accident within the assumed time frame.All other radiological dose cases only require two ADVs, since a longer cooldown does not have as great an impact on SGTR doses as a failed open ADV on the faulted SG.Furthermore, in order to assure the SGTR accident can be mitigated within the Unit 2 analysis requirements, the ADVs must be capable of both remote and local manual operation.

In addition, the Unit 2 safety analysis does not include additional time to manually unisolate a blocked ADV. Therefore, an ADV line with a closed block valve is considered inoperable.

The Unit 2 safety analysis does account for the time it takes to manually isolate the faulted SG from the Unit 2 Residual Heat Release Valve so that ADV line can be used to meet the accident analysis requirements.

Therefore, the individual normally open SG isolation valves associated with the Unit 2 Residual Heat Release Valve must also be maintained open with the capability of being manually closed for the Unit 2 Residual Heat Release Valve ADV line to be OPERABLE.3. INSERT 3 BVPS Specific LCO Section The LCO requires three Unit 1 ADV lines and four Unit 2 ADV lines to be OPERABLE.Page 2 100 INSERTS FOR ITS 3.7.4 BASES Rev. 4, Change 17 The ADV lines required OPERABLE include the three Atmospheric Relief Valves (one per steam generator (SG)) and the associated block (isolation) valves and for Unit 2 only, one Residual Heat Release Valve and its block valve and individual SG isolation valves. The Unit 2 Residual Heat Release Valve and all its associated isolation valves are counted as one ADV line for Unit 2. The number of ADV lines required OPERABLE is consistent with each Unit's design and the safety analyses requirements described above.An OPERABLE ADV line is capable of providing controlled relief of the main steam flow and capable of fully opening and closing. In order to be OPERABLE, the ADVs (including the Unit 2 Residual Heat Release Valve) must be capable of remote manual and local manual operation.

Also, the block valve associated with each ADV line must be open for the line to be considered OPERABLE.

In addition to the above requirements, the three individual SG isolation valves associated with Unit 2 Residual Heat Release Valve must be open and capable of being manually closed for the Residual Heat Release Valve ADV line to be considered OPERABLE.The block valves associated with each ADV line must be OPERABLE to isolate a failed open ADV line. In addition, the three individual SG isolation valves associated with the Unit 2 Residual Heat Release Valve ADV line must be OPERABLE to enable a faulted SG to be isolated from the Residual Heat Release Valve ADV line.Failure to meet the LCO could result in the inability to cool the unit under the limiting accident conditions within the time limit assumed in the applicable safety analyses described above.Page 3 101

ML063000088

Contents

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Subject:

1.1 Definitions

2.0 Safety

3.7 Plant

3.7 Plant

3.8 Electrical

3.6 Containment

3.9 Refueling

3.9 Refueling

3.9 Refueling

3.9 Refueling

3.9 Refueling

3.9 Refueling

3.9 Refueling

3.9 Refueling

3.7 PLANT

1.1 Definitions

4.6.2 Reactor

3.8 Electrical

3.4 Reactor

3.7 Plant

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ML063000088

Text

Subject:

1.1 Definitions

2.0 Safety

3.7 Plant

3.7 Plant

3.8 Electrical

3.6 Containment

3.9 Refueling

3.9 Refueling

3.9 Refueling

3.9 Refueling

3.9 Refueling

3.9 Refueling

3.9 Refueling

3.9 Refueling

3.7 PLANT

1.1 Definitions

4.6.2 Reactor

3.8 Electrical

3.4 Reactor

3.7 Plant

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