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j' O pJ(JBeaver Valley Power StationF O Route 168

Ift-_%P0. Box 4FrstEnergy Nuclear Operating Company Shippingport. PA 15077-0004

L. Willian Pearce 724-682-5234Site Vice President Fax: 724-643-8069

October 5, 2004L-04-127

U. S. Nuclear Regulatory CommissionAttention: Document Control DeskWashington, DC 20555-0001

Subject: Beaver Valley Power Station, Unit No. 1 and No. 2BV-1 Docket No. 50-334, License No. DPR-66BV-2 Docket No. 50-412, License No. NPF-73License Amendment Request Nos. 327 and 197 onSteam Generator Level Allowable Value Setpoints

Pursuant to 10 CFR 50.90, FirstEnergy Nuclear Operating Company (FENOC) requestsan amendment to the license for Beaver Valley Power Station (BVPS) Unit No. I andUnit No. 2 in the form of changes to the Technical Specifications. The proposedamendments will modify steam generator level allowable value setpoints used in theReactor Trip System and Engineered Safety Feature Actuation System instrumentation toaddress identified non-conservative setpoints. The proposed changes address recentgeneric issues involving new steam generator level uncertainty considerations andmargins associated with Westinghouse designed steam generators.

The FENOC evaluation of the proposed changes is presented in the Enclosure. Theproposed Technical Specification changes are presented in Attachment A. Attachment Bprovides the proposed information-only changes to the Licensing Requirements Manualthat reflect the proposed license amendment. Attachment C indicates that there are nonew commitments made in this submittal.

The Beaver Valley review committees have reviewed the changes. The changes weredetermined to be safe and do not involve a significant hazard consideration as defined in10 CFR 50.92 based on the attached safety evaluation and no significant hazardevaluation.

FENOC requests approval of the proposed amendment by August 2005. Once approved,the amendment shall be implemented within 60 days.

If there are any questions concerning this matter, please contact Mr. Larry R. Freeland,Manager, Regulatory Compliance at 724-682-5284.

Beaver Valley Power Station, Unit No. 1 and No. 2License Amendment Request Nos. 327 and 197L-04-127

I declare under penalty of perjury that the foregoing is true and correct. Executed onOctober < , 2004.

Sincere

L. Iliam Parce

Enclosure: FENOC Evaluation of the Proposed Change

Attachments:A. Proposed Technical Specification Changes (mark-ups)B. Proposed Changes to Licensing Requirements Manual (Information only)C. List of Regulatory Commitments

c: Mr. T. G. Colburn, NRR Senior Project ManagerMr. P. C. Cataldo, NRC Sr. Resident InspectorMr. S. J. Collins, NRC Region I AdministratorMr. D. A. Allard, Director BRP/DEPMr. L. E. Ryan (BRP/DEP)

l

ENCLOSURE

Beaver Valley Power Station, Unit Nos. 1 and 2License Amendment Request No. 327 and 197

FENOC Evaluation of the Proposed Change

Subject: Application for Amendment of Beaver Valley Power Station TechnicalSpecifications To Revise Steam Generator Level Allowable Value Setpoints.

Table of Contents

Section Title Page

1.0 DESCRIPTION ................................ 1

2.0 PROPOSED CHANGES ............................... 1

3.0 BACKGROUND ............................... 4

4.0 TECHNICAL ANALYSIS ................................ 7

5.0 REGULATORY SAFETY ANALYSIS ............................... 22

5.1 No Significant Hazards Consideration .................. ............. 23

5.2 Applicable Regulatory Requirements/Criteria ............................... 24

6.0 ENVIRONMENTAL CONSIDERATION ............................... 25

7.0 REFERENCES ............................... 25

AttachmentsNumber Title

A Proposed Technical Specification Changes

B Proposed Licensing Requirement Manual Changes

C Commitment Summary

i

Beaver Valley Power Station Unit Nos. I & 2License Amendment Request No. 327 & 197

1.0 DESCRIPTION

FirstEnergy Nuclear Operating Company (FENOC) requests to amend OperatingLicense DPR-66 for Beaver Valley Power Station (BVPS) Unit No. I and LicenseNPF-73 for BVPS Unit No. 2. The proposed amendment would revise TechnicalSpecification 3/4.3.1, "Reactor Trip System Instrumentation" and TechnicalSpecification 3/4.3.2, "Engineered Safety Feature Actuation System Instrumentation" tomodify steam generator level allowable value setpoints. The proposed changes addressrecent generic issues involving new steam generator level uncertainty considerationsand margins associated with Westinghouse designed steam generators (SGs).

2.0 PROPOSED CHANGE

The proposed Technical Specification (TS) changes, which are submitted for NuclearRegulatory Commission (NRC) review and approval, are provided in Attachment A-1and A-2. The changes proposed to the Licensing Requirements Manual are provided inAttachment B-i and B-2. The proposed Licensing Requirements Manual (LRM)changes do not require NRC approval. Changes to the LRM are controlled through the10 CFR 50.59 process. The LRM changes are provided for information only.Attachment C provides a list of commitments associated with this License AmendmentRequest (LAR).

The proposed changes to the Technical Specifications and LRM have been preparedelectronically. Deletions are shown with a strike-through and insertions are shown by atext box insertion. This presentation allows the reviewer to readily identify theinformation that has been deleted and added.

To meet format requirements, the Index, Technical Specifications and LRM pages willbe revised and repaginated as necessary to reflect the changes being proposed by thisLAR.

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Beaver Valley Power Station Unit Nos. 1 & 2License Amendment Request No. 327 & 197

2.1 Technical Specification Proposed Changes

Changes to the following BVPS Unit No.1 and Unit No. 2 Technical Specifications(shown in Attachment A) are being proposed in this request:

Affected Technical Specifications

Change BVPS TS Section ItemNo. Unit

1 1 3.3.1.1 Reactor Trip System Instrumentation, SteamTable 3.3-1 Generator Water Level - Low-Low, Allowable ValueFunctionalUnit 14

2 1 3.3.2.1 Engineered Safety Feature Actuation SystemTable 3.3.-3 Instrumentation, Auxiliary Feedwater, SteamFunctional Generator Water Level - Low-Low, Allowable ValueUnit 7.a

3 2 3.3.1.1 Reactor Trip System Instrumentation, SteamTable 3.3-1 Generator Water Level - Low-Low, Allowable ValueFunctionalUnit 14

4 2 3.3.2.1 Engineered Safety Feature Actuation SystemTable 3.3.-3 Instrumentation, Auxiliary Feedwater, SteamFunctional Generator Water Level - Low-Low, Allowable ValueUnit 7.b

5 2 3.3.2.1 Engineered Safety Feature Actuation SystemTable 3.3.-3 Instrumentation, Turbine Trip & Feedwater Isolation,Functional Steam Generator Water Level - High-High, P-14,Unit 5.b Allowable Value

TS Proposed Change Number 1

This proposed change is a modification to BVPS Unit No. 1 Technical Specification3.3.1.1, "REACTOR TRIP SYSTEM INSTRUMENTATION." This modificationconsists of revising the Technical Specification allowable value for the SG water level-low-low reactor trip system function from 14.6% to 19.6% of narrow range instrument

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span - each steam generator. This proposed change is shown in the marked-upTechnical Specifications in Attachment A.


This proposed change is a modification to BVPS Unit No. I Technical Specification3.3.2.1, "ENGINEERED SAFETY FEATURE ACTUATION SYSTEMINSTRUMENTATION." This modification consists of revising the TechnicalSpecification allowable value for the SG water level-low-low auxiliary feedwateractuation functions from 14.6% to 19.6% of narrow range instrument span - each steamgenerator. This proposed change is shown in the marked-up Technical Specifications inAttachment A.


This proposed change is a modification to BVPS Unit No. 2 Technical Specification3.3.1.1, "REACTOR TRIP SYSTEM INSTRUMENTATION." This modificationconsists of revising the Technical Specification allowable value for the SG water level-low-low reactor trip system function from 16% to 20% of narrow range instrument span- each steam generator. This proposed change is shown in the marked-up TechnicalSpecifications in Attachment A.


This proposed change is a modification to BVPS Unit No. 2 Technical Specification3.3.2.1, "ENGINEERED SAFETY FEATURE ACTUATION SYSTEMINSTRUMENTATION." This modification consists of revising the TechnicalSpecification allowable value for the SG water level-low-low auxiliary feedwateractuation functions from 16% to 20% of narrow range instrument span - each steamgenerator. This proposed change is shown in the marked-up Technical Specifications inAttachment A.


This proposed change is a modification to BVPS Unit No. 2 Technical Specification3.3.2.1, "ENGINEERED SAFETY FEATURE ACTUATION SYSTEMINSTRUMENTATION." This modification consists of revising the TechnicalSpecification allowable value for the SG water level-high-high turbine trip andfeedwater isolation actuation function from 81.1% to 92.7% of narrow range instrumentspan - each steam generator. This proposed change is shown in the marked-upTechnical Specifications in Attachment A.

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2.2 Licensing Requirement Manual Changes

The corresponding nominal trip setpoint changes for the proposed TechnicalSpecification allowable value changes discussed above are shown in Attachment B-1and B-2. These nominal trip setpoints are maintained in the Licensing RequirementsManual, which is a licensee controlled document that is changed under the 10 CFR50.59 process. These LRM changes are being provided for information only.

3.0 BACKGROUND

3.1 Steam Generator Water Level Design

Steam generator (SG) water level-low-low is a functional unit of the Reactor TripSystem (RTS). It functions to trip the reactor and protect the reactor core from a loss ofheat sink in the event of a sustained steam/feedwater flow mismatch. This trip isactuated on two-out-of-three water level-low-low signals occurring in any SG. Thebasic function of the reactor protection circuits associated with the SG water-low-lowreactor trip is to preserve the SG heat sink for removal of long term core residual heat.Should a complete loss of feedwater occur, the reactor would be tripped on SG waterlevel-low-low.

A spurious high signal from the feedwater flow channel being used for control wouldcause a reduction in feedwater flow. The mismatch between steam demand andfeedwater flow produced by this spurious signal will actuate alarms to alert the operatorof this situation in time for manual correction or, if the condition is allowed to continue,the reactor will eventually trip on a SG water level-low-low signal independent ofindicated feedwater flow.

Steam generator water level-low-low is also a functional unit of the Engineered SafetyFeature Actuation System (ESFAS). It functions to actuate the Auxiliary Feedwater(AFW) pumps to provide auxiliary feedwater to the secondary side of the SGs in orderto maintain a heat sink. The turbine driven AFW pump is started on two-out-of-threewater level-low-low signals occurring in any one SG and the motor driven AFW pumpsare started on two-out-of-three water level-low-low signals occurring in any two SGs.

Steam generator water level-high-high is also a functional unit of the ESFAS. Itfunctions to trip the turbine and isolate feedwater to the steam generators in order to 1)protect the turbine from damage due to steam with excessive moisture carryover, 2) toavoid adverse effects of excess steam moisture on the accuracy of steam flow andsteamline pressure transmitters downstream in the steam piping, and 3) to avoid

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I


addressing the loading effects of water in the steam piping support design. Steamgenerator water level-high-high signals in two-out-of-three channels for any steamgenerator will actuate a turbine trip, trip the main feedwater pumps, and close the mainand bypass feedwater level control valves.

A spurious low signal from the feedwater flow channel being used for control wouldcause an increase in feedwater flow. The mismatch between steam flow and feedwaterflow produced by this spurious signal will actuate alarms to alert the operator of thissituation in time for manual correction or, if the condition is allowed to continue, a two-out-of-three steam generator water level-high-high signal from any steam generator,independent of the indicated feedwater flow, will trip the turbine and cause thefeedwater isolation. If the turbine trip occurs when reactor power is above the P-9permissive setpoint, the turbine trip will result in a subsequent reactor trip.

3.2 Steam Generator Water Level Setpoint Uncertainty

The BVPS Units No. 1 and 2 steam generator water level-low-low and high-highsetpoints are currently established in accordance with the methodology described inWCAP-1 1419, Rev. 2, "Setpoint Methodology for Protection Systems for BeaverValley Power Station - Unit 1," (Reference 1) and WCAP-1 1366, Rev. 4, "SetpointMethodology for Protection Systems for Beaver Valley Power Station - Unit 2,"(Reference 2). These two WCAPs were previously submitted via FENOC LetterL-00-143, dated December 27, 2000 as part of the submittal for BVPS Unit No. I LAR286 and BVPS Unit No. 2 LAR 158. These WCAPs were also used by the NRC as partof the basis for approving BVPS Unit No. 1 Technical Specification Amendment No.239, dated July 20, 2001 (Reference 3) and BVPS Unit No. 2 Technical SpecificationAmendment No. 120, dated July 20,2001 (Reference 4).

References 1 and 2 describe various generic issues and concepts that may affect waterlevel uncertainties in plants with Westinghouse-designed steam generators. Theuncertainty that must be considered is associated with two sources. These are theinstrumentation hardware itself and the non-instrumentation or process measurementaccuracy (PMA) effects. Westinghouse identified four separate PMA terms in 1992:Process Pressure Variations, Reference Leg Temperature, Fluid Velocity Effects, andDowncomer Subcooling which were addressed in References I and 2.

In February and April, 2002, Westinghouse transmitted a Nuclear Safety AdvisoryLetter (NSAL) describing issues that may affect water level uncertainties in plants withWestinghouse-designed Steam Generators. This NSAL (NSAL-02-3, Reference 5)provided guidance to utilities in assessing various issues that could affect plant-specificuncertainties in calculations of SG water level. This NSAL identified the need to

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specify additional PMA terms to address the uncertainties associated with steam flowthrough the mid-deck plate inside a steam generator. Westinghouse subsequentlyconcluded in a BVPS-specific evaluation that both BVPS Units had sufficient margin tooffset the effects of the mid-deck plate differential pressure (AP) issues described inReference 5 and that no changes were needed in the BVPS Unit Nos. 1 and 2 TechnicalSpecifications. Specifically, NSAL-02-03 concluded that for a steam generator affectedby a Feedwater Line Break (FLB), reverse flow occurs through the feedring out of thesteam generator nozzle and eventually out the break which results in a reversal in signof the mid-deck AP effect and can be conservatively ignored for that event. Thus,Westinghouse concluded in 2002 that the consideration for mid-deck AP did notadversely affect the setpoints used for the FLB analyses described in the BVPS UnitNos. 1 and 2 Updated Final Safety Analysis Reports (UFSARs) and as approved inReferences 3 and 4. It is noted that in both References 1 and 2 and in the 2002 NSALthe bounding FLB event was believed to be the large FLB (and no specific small orintermediate FLB analyses were performed). In addition, Westinghouse concluded thatthe mid-deck AP effect also did not adversely affect the other steam generator waterlevel setpoint analyses (i.e., steam line break or loss of normal feedwater).

Subsequent discussions with Westinghouse Owners Group (WOG) members and withinWestinghouse resulted in the identification of four additional elements of WestinghouseSG designs that could affect SG level measurement and should be addressed in SGuncertainty assessments. These include the intermediate deck plate d/p, the feedringd/p, the effects of steam carry under into the downcomer, and the lower deck platesupports. In addition to these design elements, four transient conditions were identifiedthat could produce transient-specific effects that should be included in assessments ofSG level uncertainties. These transients are: Single-Loop Loss of Normal Feedwater(LONF), Small Steamline Break Mass and Energy Releases Outside Containment,Small to Intermediate Feedline Break Inside Containment, and the FeedwaterMalfunction event. In October 2002, Westinghouse was authorized to complete aprogram for the WOG to provide additional generic guidance for plants in assessing theeffects of various SG design-related issues and provide guidance for evaluating SGwater level PMA effects on a plant-specific basis. As a result of this WOG program, inSeptember, 2003, Westinghouse issued WCAP-161 15-P, "Steam Generator LevelUncertainties Program" (Reference 6) and issued NSAL-03-9 (Reference 7) whichprovided additional insight into the information previously provided by NSAL-02-03(Reference 5) and on the four new design elements/transients considerations and otherinformation involving steam generator water level instrumentation uncertainties.

As a result of these Westinghouse actions, FENOC became aware in September, 2003that the uncertainty analysis performed for the steam generator water level setpoints atboth BVPS Units may not be adequate to address all credible potential conditions, as

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previously determined by Westinghouse for the BVPS Units. Specifically, the primaryissue was the new information that all plants with a mid-deck plate pressure loss mustnow consider the mid-deck AP bias, since the small to intermediate size FLB may nowbe the bounding FLB case, rather than the large FLB as previously believed.NSAL-03-09/WCAP-161 15-P recognized that there may be some small or intermediatesize of FLBs where there is no reverse flow out of the steam generator with the rupturedline attached, but also no feedwater flow entering that affected steam generator. Thus,both BVPS Units setpoints for steam generator water level-low-low would have to nowadd a new bias to their steam generator water level-low-low setpoints to addressuncertainty for the AP across the mid-deck during a FLB. It was initially estimated thataddressing the new mid-deck AP bias would require an additional five percent level inthe steam generator water level-low-low setpoint calculations for FLB. Withinsufficient margin remaining in the steam generator water level-low-low setpointcalculations, FENOC immediately implemented administrative controls toconservatively raise both Unit's steam generator water level-low-low Allowable Valuesby an additional five percent beyond the values required by Technical Specifications toaddress these new considerations and reported this event in Licensee Event Report(LER) 2003-006 submitted by FENOC letter L-03-181, dated November 12, 2003.

4.0 TECHNICAL ANALYSIS

The technical analysis conducted to support the proposed Technical Specificationchanges includes evaluation of initial condition uncertainties at power conditions, whichare provided as input to the safety analyses, and the development of the changes to thesteam generator level reactor trip and engineered safety feature actuation systemsetpoints and allowable values.

To evaluate the steam generator level uncertainty and setpoint calculations for BVPSUnit No. 1 and No. 2 which are used as the bases for these License AmendmentRequests (LARs), a complete review of the development of each individual termdocumented in the previously licensed calculation of record (References 1 and 2) wasperformed to address the new industry information on Process Measurement Accuracyterms (as described in Section 3.2) and to determine if any unnecessary conservatismcould be decreased. In the area of rack uncertainties, no adjustments could be made.Under the environmental allowance terms were two areas of potential relaxation. Theuncertainties, margins, setpoints, and Allowable Values for SG water level-low-low forboth BVPS Units and for SG water level-high-high for BVPS Unit No. 2 have beencalculated, which demonstrate positive margins are maintained with the proposedchanges.

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LAR No. 327 for BVPS Unit No. I is applicable for power levels at or below thecurrently maximum allowed rated thermal power of 2689 megawatts thermal (MWt)with the currently installed Model 51 steam generators.

LAR No. 197 for BVPS Unit No. 2 is applicable for power levels at or below 2910MWt with the currently installed Model 51M steam generators. Although 2910 MWt isabove the currently maximum allowed rated thermal power of 2689 MWt, thisevaluation bounds current licensed condition and will also support potential futurelicense actions on BVPS Unit No. 2. [Note: no change in rated thermal power level isbeing requested by this LAR.]

4.1 Basis for TS Proposed Changes 1, 2, 3, and 4: Steam Generator Water Level-Low-LowAllowable Value and Change 5: Steam Generator Water Level-High-High

Introduction:

As a result of NSAL-03-09 (Reference 7) and the WOG activities in 2002-2003 toaddress additional effects on steam generator water level uncertainties, an evaluationwas performed upon the current steam generator water level setpoints and methodologyas described in References 1 and 2. Previously, for the SG water level-low-lowparameter, separate uncertainty calculations were performed for the Steam Line Break(SLB), Feedwater Line Break (FLB), and the Loss of Normal Feedwater (LONF) designbasis accidents. As shown in Reference 1, the limiting case for SG water level-low-lowat BVPS Unit 1 was previously the LONF event. As shown in Reference 2, the limitingcase for SG water level-low-low at BVPS Unit 2 was previously the (large) FLB event.As a result of this evaluation, the SG water level-low-low setpoint analysis for bothBVPS Units needs to be revised to address new limiting events, which will requirerevision to the Technical Specification Allowable Values for SG water level-low-low.

Additionally, as a result of this evaluation, the BVPS Unit No. 2 SG water level-high-high uncertainty analysis was revised from that previously described in Reference 2 andresulted in this request to change the Technical Specification Allowable Value for SGwater level-high-high.

The evaluation of the BVPS Unit No. I SG water level-high-high PMA terms in light ofthe new Westinghouse/WOG information determined that the existing PMA termsremain bounding. The BVPS Unit No. I SG water level-high-high Allowable Valuewas evaluated and determined to remain conservative, with no changes being requested.Although the SG water level-high-high could have been altered as is being requested forBVPS Unit No. 2 to remove some of the available margin, it was determined that nochanges would be requested since the BVPS Unit No. 1 steam generators are currently

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scheduled to be replaced in the 2006 refueling outage, which will require a completenew uncertainty analysis be established in 2006 for the replacement SGs. Thus, therewas limited benefit to now revise station calculations/procedures to address a new SGwater level-high-high with only one more cycle of operation left at BVPS Unit No.1with these original steam generators. The current SG water level-high-high value atBVPS Unit No. I remains conservative.

Setpoint Analyses Changes

The process measurement accuracy (PMA) terms evaluated for the proposed TechnicalSpecification changes in these LARs were: 1) Process Pressure Variations, 2) ReferenceLeg Temperature, 3) Fluid Velocity Effects, 4) Downcomer Subcooling, 5) DynamicLosses, 6) Intermediate Deck Plate d/p, 7) Feedring d/p, and finally 8) Mid-deck plated/p. In addition, the Environmental Allowance modifier for Reference LegTemperature Effects was also modified, to be consistent with Item 2. Items 5, 6, 7 and8 are new parameters not previously described in References I and 2 and were added tothe uncertainty calculations. For the SG water level-low-low, Items 1, 2 and 4 werefound to need revision in the uncertainty calculations. For the SG water level-high-highsetpoint analysis, Items 1 and 3 were found to need revision in the uncertaintycalculation. In addition, the contribution from Item 8 was included in the uncertaintyevaluation. All the other identified PMA items either had no new additional impact oran insignificant impact on the uncertainty calculations.

The uncertainty associated with the Process Pressure effects for SG water level-low-lowpreviously calculated for both BVPS Units were incorrect and had to be increased withthe magnitude of the increase based on 1) the power level, 2) feedwater temperature forthe analyzed event (i.e., for SLB, LONF and small/intermediate/large FLB), and 3) thepressure effect on the subcooled water density in the reference leg.

For setpoint uncertainty analysis, "+" means the instruments indicates higher than actuallevel, and "-" means the instrument indicates lower than actual level.

The uncertainty associated with the Process Pressure effects for SG waterlevel-high-high previously calculated for BVPS Unit 2 was incorrect and was shown tobe slightly lower than previously assumed in Reference 2 based on the power level andfeedwater temperature for the analyzed event (i.e., feedwater malfunction) for BVPSUnit No. 2.

The uncertainty associated with the Reference Leg Temperature Heatup effect increasedfor FLB events for both BVPS Units. Previously, the FLB case only addressed the largeFLB event and limited reference leg heatup was assumed. Now, for the newly defined

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small/intermediate FLB, reference leg temperature is presumed to increase further sincethe actuation condition takes longer to reach during the transient. The uncertainty forthe reference leg temperature is modeled based on Tsat equivalent to Psat for thecontainment high setpoint. The uncertainty associated with the reference leg heatupeffect was also increased for the large FLB, but to a lesser extent.

Westinghouse issued Technical Bulletin TB-04-12, "Steam Generator Level ProcessPressure Evaluation" on June 23, 2004 (Reference 8). This indicated that very smallchanges (on the order of 0. 1% span) to the Process Pressure PMA term may need to beconsidered to address the impact of pressure on the subcooled water density in thereference leg, which had previously been considered to be an insignificant factor. Thechanges proposed by this BVPS Unit No. 1 and No. 2 LARs include the considerationsidentified by this Technical Bulletin.

The effects due to downcomer subcooling equivalent to nominal full power conditionswere applied, as a benefit reducing the previous value for this PMA parameter, for bothBVPS Units FLB analyses and the BVPS Unit No. 2 LONF and SLB analyses. [Note:Although this downcomer subcooling change could have also been applied to the SLBand LONF for BVPS Unit No. I as a benefit, it was conservatively not applied to limitthe number of changes needed for BVPS Unit No. I evaluation.]

The Fluid Velocity effects were reduced to zero for the analyzed feedwater malfunctionevent for SG water level-high-high at BVPS Unit No. 2 as determined by Reference 6.

For the mid-deck plate AP effect, the effect varies based on the accident scenario beinganalyzed. For the large FLB event, the mid-deck plate AP error is based on 100% steamflow conditions; for the SLB event, it is based on 121% steam flow conditions; and forsmall/intermediate FLB and LONF events, it is based on 112% steam flow values. Theuncertainty effect for mid-deck AP increases with higher steam generator steam flowrates. This is consistent with the recommendations of NSAL-03-9 (Reference 7).

In addition to the PMA term changes, the BVPS Unit No. I steam generator water leveltransmitter drift value was lowered to more accurately model the BVPS Unit No. 1instrument operation based upon a review of past empirical values obtained at BVPSUnit No. 1. The BVPS Unit No. 2 steam generator water level transmitter drift valuewas not changed and remains valid.

The BVPS Unit No. 2 Environmental Allowance modifier for Insulation Resistance wasrevised to reflect an upper temperature limit for a FLB causing a decrease in this biasfor SG water level-low-low for the FLB analysis. The BVPS Unit 1 EnvironmentalAllowance modifier for Insulation Resistance was not changed and remains valid.

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In addition, the Safety Analysis Limit (SAL) for the LONF event at BVPS Unit No. 2was changed from 10% to 0%. The BVPS Unit No. 2 SAL values for FLB and SLBremained unchanged at 0%. Now, the SAL for the three events analyzed for SG waterlevel-low-low (i.e., LONF, SLB, FLB) at BVPS Unit No. 2 all have a SAL value of 0%.BVPS Unit No. 2 safety analyses conclusions continue to remain valid with a newLONF SAL value of 0%. [Note: The SAL value for the LONF event at BVPS Unit Iwas not altered since LONE is not the current limiting event for the BVPS Unit No. ISG water level-low-low uncertainty analysis and the BVPS Unit No. 1 steam generatorsare currently scheduled to be replaced in the 2006 refueling outage, which will require acomplete new uncertainty analysis be established in 2006 for the replacement SGs.Thus, there was limited benefit to now revise station calculations/procedures to addressa new LONF SAL for a non-bounding event with only one more cycle of operation leftat BVPS Unit No. 1 with these original steam generators. The current LONF SAL valueof 10% at BVPS Unit No. 1 for SG water level-low-low remains conservative.]

Based on the evaluation of these changes in the BVPS Unit 1 and Unit 2 setpointanalyses, the bounding design basis accident uncertainties applicable for the SG waterlevel-low-low function now becomes the small/intermediate FLB event at both Units.

The evaluation concluded that the net impact of NSAL-03-9 and other changes was toincrease the nominal low-low level trip setpoint from 15.1% to 20.1% level with anAllowable Value of 19.6% level for BVPS Unit 1. Thus, TS Proposed Change No. 1consists of revising the BVPS Unit I Technical Specification allowable value for the SGwater level-low-low reactor trip system function to 19.6% level and TS ProposedChange No. 2 consists of revising the BVPS Unit No. 1 Technical Specificationallowable value for the SG water level-low-low auxiliary feedwater actuation functionto 19.6% level. The evaluation also indicated that the nominal low-low level tripsetpoint should be increased from 16.5% to 20.5% level with an Allowable Value of20% level for BVPS Unit 2. Thus, TS Proposed Change No. 3 consists of revising theBVPS Unit 2 Technical Specification allowable value for the SG water level-low-lowreactor trip system function to 20% level and TS Proposed Change No. 4 consists ofrevising the BVPS Unit No. 2 Technical Specification allowable value for the SG waterlevel-low-low auxiliary feedwater actuation function to 20% level.

The steam generator level control will continue to maintain steam generator water levelconstant at 44% at full power. Therefore, the steady-state full power operating marginto the low-low level reactor trip setpoint (difference between normal water level and thetrip setpoint) will decrease from 28.9% to 23.9% with the new proposed BVPS UnitNo. 1 Allowable Value and from 27.5% to 23.5% with the new proposed BVPSUnit No. 2 Allowable Value. A steady state margin of 23.9%/23.5% for BVPS Unit

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No. 1/2 is considered adequate based on field experience to avoid unnecessary reactortrips on SG water level-low-low during normal expected transients where a reactor tripis neither wanted or expected. This includes plant startups, shutdowns, steady-stateoperation and design basis load swings (± 10% load change). The other area of concernis the steam generator level during a 50% load rejection transient. The 50% loadrejection transient was analyzed for the (future anticipated) BVPS Unit 1 extendedpower uprate (EPU). This EPU analysis showed that margin to SG water level-low-lowtrip setpoint was greater than 10% with the prior trip setpoint. Therefore, with theproposed increase of the trip setpoint of 5%, the proposed setpoint will still provide atleast 5% margin. Considering that the EPU analysis is conservative with respect to thecurrent licensed power conditions, a margin of 5% is considered adequate.

The Safety Analysis Limit (SAL) limit used previously for SG water level-high-high atBVPS Unit No. 2 in Reference 2 was 86.3%, which represented the prior maximumreliable indicated steam generator water level. [The safety analysis assumed a high-high level actuation at 100%.] This SAL value initially originated from an extremelyconservative estimation of the maximum generic reliable indicated steam generator highwater level in 1992. The generic value was based upon a two loop plant with boundingvalues for all four PMA terms (in use in 1992). In Reference 2, BVPS Unit No. 2specific uncertainties (including PMA terms) were then determined and used toestablish the SG water level-high-high setpoint and allowable value based upon the SALvalue of 86.3%. This, in essence, applied.two sets of PMA considerations to the SGwater level-high-high parameter; once in the generic SAL and once in the BVPS UnitNo. 2 specific uncertainties.

Westinghouse issued NSAL-02-4 in February, 2002 (Reference 9) which providedadditional information regarding the maximum reliable indicated steam generator waterlevel and its use in steam generator water level uncertainty applications. Westinghousehas now determined that a more accurate determination of the maximum reliableindicated steam generator water level (MRIL) for BVPS Unit No. 2 is 96.7%. Thus, theSAL for BVPS Unit No. 2 steam generator water level-high-high was increased to96.7% (which still assumed an actual actuation at 100%). A SAL value of 96.7% forBVPS Unit No. 2 steam generator water level-high-high shows acceptable safetyanalysis results.

The evaluation concluded that the net impact of these changes was to increase thenominal high-high level trip setpoint from 80.6% to 92.2% level with an AllowableValue of 92.7% level for BVPS Unit 2. Thus, TS Proposed Change No. 5 consists ofrevising the BVPS Unit 2 Technical Specification allowable value for the SG waterlevel-high-high ESFAS trip function to 92.7%.

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4.2 Review of BVPS Unit No. 1 Safety Analyses Relating to the Proposed TechnicalSpecification Changes

The new proposed SG level-low-low trip and Allowable Value setpoints were evaluatedto assess their impact on the potentially impacted safety analyses for BVPS Unit No. 1,including:

* Non-Loss of Coolant Accident (LOCA) Events* SLB Mass and Energy Releases* LOCA Mass and Energy Releases* Steam Generator Tube Rupture* LOCA Events* ATWS Mitigation System Actuation Circuitry (AMSAC)

Margin to Trip* Radiological Analyses

The results of these evaluations are summarized below for BVPS-1. In this sectiondescribing safety analysis, "+' means that the actual level is higher than indicated level,and "-" means that the actual level is lower than indicated level.

Non-LOCA Events

Several non-LOCA analyses are sensitive to the SG water level-low-low trip setpoint.These are the FLB and the LONF/Loss of AC Power (LOAC) analyses. These eventsare discussed in this section.

Feedwater Line Break

(BVPS Unit No. 1 UFSAR Section: Feedwater System Pipe Break - 14.2.5.2)

The current licensing basis FLB analysis for BVPS Unit No. 1 assumes an uncertaintyof ±6% on the initial SG mass and a low-low level safety analysis limit (SAL) of 0%.The initial condition uncertainty is applied in both directions in the FLB analysis. Thefaulted loop assumes a conservatively high initial SG mass in order to delay the time ofreactor trip on SG water level-low-low and the intact loops assume a lower thannominal initial SG mass in order to minimize the SG mass available for post trip heatremoval. Based upon sensitivity studies by Westinghouse, the uncertainties applied tothe intact loops have minimal affect on the results of the safety analysis. Of these two,the timing of the reactor trip from the faulted loop has a much greater impact on theanalysis results. The acceptance criterion applied to a FLB analysis is that the hot legsremain subcooled. The current analysis of record shows significant margin (>300F) to

Page 13


boiling in the hot legs. Based on the amount of available margin, the fact that the safetyanalysis limit is unaffected, and that the initial faulted SG mass is conservative, it isconcluded that the new issues addressed by these proposed changes in this LAR wouldnot cause any safety limits to be violated.

Loss of Normal Feedwater/Loss of AC Power

(BVPS Unit No. 1 UFSAR Sections: Loss of Normal Feedwater - 14.1.8 and Loss ofOffsite Power to the Station Auxiliaries -14.1.1 1)

The current licensing basis LONF/LOAC analysis for BVPS Unit No. 1 assumes a SGwater level-low-low SAL of 10% and initial SG masses that include a +6% uncertainty.Maximum initial steam generator mass delays reactor trip on low-low steam generatorwater level. The current licensing basis analysis assumes the correct trip setpoint and anoverly conservative initial steam generator mass (nominal + 6% rather than nominal+3.5%). This results in conservatively late protection via reactor trip and auxiliaryfeedwater initiation which allows the heat-up to progress longer prior to protectiveactions occurring. Thus, current licensing basis LONF/LOAC analysis for BVPS-1remains acceptable.

Other Non-LOCA Events

No other non-LOCA event analysis credits the SG water level-low-low trip functionsfor protection. Thus, this aspect of this issue impacts no other non-LOCA event. Theinitial SG mass assumed in the non-LOCA analyses is biased high or low depending onthe direction of conservatism. If a particular event is not sensitive to the initial SGmass, then the initial mass is set to the nominal mass. Any event that is analyzed withthe initial mass set to the nominal mass is not impacted by the new changes on theinitial mass. The analysis for any event for which a high initial SG mass is conservativeremains conservative because the positive uncertainty has decreased. There are nonon-LOCA events for which a low initial SG mass -yields more limiting results than theFLB event which is discussed above.

SLB Mass and Energy Releases

(BVPS Unit No. 1 UFSAR Section: Major Secondary System Pipe Rupture 14.2.5

The analyses for the SLB mass and energy (M&E) releases inside and outsidecontainment typically use SG water level uncertainties for both the initial conditions andthe low-low reactor trip setpoint.

Page 14


The current licensing-basis analysis of record for the SLB M&Es inside containment isdocumented using the MARVEL code (UFSAR Section 14D.10.3) based onconservative generic values for the initial SG water mass. Design mass values for initialSG water mass are greater than current safety analysis values and were the standardinput to the analysis performed with MARVEL in the 1980s. Since no allowance wasmade in the MARVEL analysis for water level control, and since conservatively largeinitial SG wvater mass values were modeled, the revised uncertainties do not alter thevalidity of the analysis values at any power. The trip setpoint is not assumed in thisanalysis.

The current licensing-basis analyses of record for the SLB M&Es outside containmentare generic M&E releases documented using the LOFTRAN code (UFSAR Section14D.10.4) as part of a WOG program. The initial conditions related to the SG watermass are generic for the plants in the category for which the M&E releases werecalculated for BVPS Unit No. 1. Any increase in the level uncertainty used for thoseanalyses can be absorbed by conservatisms associated with other assumptions includedin the analyses. The reactor trip setpoint used in the generic analyses has beenevaluated (1999) and determined to be less than 0% for BVPS Unit No. 1. Thus, thetrip setpoint is conservative with respect to the increased uncertainties. Theseuncertainties do not alter the validity of the analyses values at any power.

LOCA Mass and Energy Releases

(BVPS Unit No. 1 UFSAR Section: LOCA Mass and Energy Release Safety Analysis14.3.4.2.1)

The potential effect of the NSAL on the LOCA M&E related analysis is the change tothe steam generator fluid mass. The long term LOCA M&E calculations are typicallyinitialized at 100% full power steady state conditions and the initial secondary side fluidmass is biased high. The current licensing basis long term LOCA M&E for Unit 1 wascalculated using the LOCTIC code. This is the same code which is used to calculate thecontainment response. The maximum steam generator liquid mass assumed in thisanalysis accounts for the 1.4% uprate, maximum steam generator liquid masscorresponding to 100% full power steady state conditions @ 44% NR level, with 10%uncertainty. The uncertainty used in the current safety analysis of record bounds theinitial condition instrumentation uncertainties. Therefore, the current analysis remainsbounding.

Steam Generator Tube Rupture

(BVPS Unit No. 1 UFSAR Section: Steam Generator Tube Rupture 14.2.4)

Page 15


The steam generator tube rupture analyses are not affected by the proposed changessince SG level control (initial condition uncertainties) and SG level trips are notmodeled in these analyses.

LOCA Events

(BVPS Unit No. I UFSAR Section: Loss Of Coolant Accidents 14.3 .1 and 14.3.2)

The LOCA and LOCA-related analyses are not affected by the proposed changes sinceSG level control (initial condition uncertainties) and SG level trips are not modeled inthese analyses.

AMSAC

(BVPS Unit No. 1 UFSAR Section: 7.2.1.1. 10)

The AMSAC logic in place at BVPS Unit No. 1 is actuated on a low feedwater flowcondition. This AMSAC logic is Logic 2 of the generic WOG AMSAC designsprovided in WCAP-10858-P-A (Reference 10). Since the AMSAC at Beaver ValleyUnit 1 is not actuated on a SG water level-low-low condition, there is no impact on theoperation of the BVPS-1 AMSAC by the proposed changes.

Margin to Trip

(BVPS Unit No. 1 UFSAR Section: None)

The proposed SG water level-low-low trip setpoint is being proposed to be raised from15. 1% to 20.1% and the associated Allowable Value is being proposed to be raised from14.6% to 19.6%. The steam generator level control will continue to maintain steamgenerator water level constant at 44% at full power. Therefore, the steady-state fullpower operating margin to the low-low level reactor trip setpoint (difference betweennormal water level and the trip setpoint) will decrease from 28.9% to 23.9% with thenew proposed BVPS Unit No. 1 trip setpoint and Allowable Value. A steady statemargin of 23.9% for BVPS Unit No. 1 is considered adequate based on field experienceto avoid unnecessary reactor trips on SG water level-low-low during normal expectedtransients where a reactor trip is neither wanted or expected. This includes plantstartups, shutdowns, steady-state operation and design basis load swings (± 10% loadchange).

Page 16


The other area of concern is the steam generator level during a 50% load rejectiontransient. The 50% load rejection transient was analyzed for the (future anticipated)BVPS Unit 1 extended power uprate (EPU). This EPU analysis showed that margin toSG water level-low-low trip setpoint was greater than 10% with the prior trip setpoint.Therefore, with the proposed increase of the trip setpoint of 5%, the proposed setpointwill still provide at least 5% margin. Considering that the EPU analysis is conservativewith respect to the current licensed power conditions, a margin of 5% is consideredadequate.

Radiological Analyses

The review of the post accident radiological consequences of the SG Water Level-low-low demonstrates that the offsite and control room doses associated with the accidentswill be within the acceptance criteria of 10 CFR Part 100 and 10 CFR 50.67, asapplicable. There were no new radiological safety analyses performed since thepreviously assumed low-low and high-high actuation values used in the safety analysesfor the steam generator water level setpoints were not altered.

Conclusions of BVPS Unit No. I Safety Analyses Relating to the Proposed TechnicalSpecification Changes:

In summary, the impact of the proposed changes has been evaluated for BVPS UnitNo. 1. The net impact of this evaluation confirms that the nominal SG water level-low-low trip setpoint of 20.1% and the Allowable Value of 19.6% are appropriate. There issufficient margin in safety analyses to accommodate these impacts and continue tosatisfy acceptance criteria.

4.3 Review of BVPS Unit No. 2 Safety Analyses Relating to the Proposed TechnicalSpecification Changes

The new proposed SG level-low-low trip and Allowable Value setpoints were evaluatedto assess their impact on the potentially impacted safety analyses for BVPS Unit No. 2,including:

a Non-LOCA Events* Steam Line Break (SLB) Mass and Energy Releases* LOCA Mass and Energy Releases* Steam Generator Tube Rupture* LOCA Events* AMSAC* Margin to Trip

Page 17


0 Radiological Analyses

The results of these evaluations are summarized below for BVPS-2. In this sectiondescribing safety analysis, "+" means that the actual level is higher than indicated level,and "-" means that the actual level is lower than indicated level.

Non-LOCA Events

Several non-LOCA analyses are sensitive to the SG water level-low-low and SG waterlevel-high-high trip setpoints. These are the Feedwater Malfunction (FWM), FLB andthe LONF/LOAC analyses. These events are discussed in this section.

Feedwater Malfunction:

(BVPS Unit No. 2 UFSAR Section: Excessive Heat Removal Due to Feedwater SystemMalfunctions 15.1.1 and 15.1.2)

The current licensing basis FWM analyses for BVPS Unit No. 2 assumes a SG waterlevel-high-high safety analysis limit of 100%. Thus, the current analysis remainsapplicable and includes sufficient conservatism to cover the additional uncertainties thatmust be added due to recent Westinghouse Owners Group evaluations.

Feedwater Line Break:

(BVPS Unit No. 2 UFSAR Section: Feedwater System Pipe Break - 15.2.8)

The current licensing basis FLB analysis for BVPS Unit No. 1 assumes an uncertaintyof ±6% on the initial SG mass and a low-low level safety analysis limit (SAL) of 0%.The initial condition uncertainty is applied in both directions in the FLB analysis. Thefaulted loop assumes a conservatively high initial SG mass in order to delay the time ofreactor trip on SG water level-low-low and the intact loops assume a lower thannominal initial SG mass in order to minimize the SG mass available for post trip heatremoval. Based upon sensitivity studies by Westinghouse, the uncertainties applied tothe intact loops have minimal affect on the results of the safety analysis. Of these two,the timing of the reactor trip from the faulted loop has a much greater impact on theanalysis results. The acceptance criterion applied to a FLB analysis is that the hot legsremain subcooled. The current analysis of record shows significant margin (>30'F) toboiling in the hot legs. Based on the amount of available margin, the fact that the safetyanalysis limit is unaffected, and that the initial faulted SG mass is conservative, it isconcluded that the new issues addressed by these proposed changes in this LAR wouldnot cause any safety limits to be violated.

Page 18


Loss of Normal Feedwater/Loss of AC Power

(BVPS Unit No. 2 UFSAR Section: Loss of Non-Emergency AC Power to the PlantAuxiliaries -15.2.6 and Loss of Normal Feedwater - 15.2.7)

The current licensing basis LONF/LOAC analysis for BVPS Unit No. 2 assumes a SGwater level-low-low SAL of 0% and initial SG masses that include a +6% uncertainty.Maximum initial steam generator mass delays reactor trip on low-low steam generatorwater level. The current licensing basis analysis assumes the correct trip setpoint and anoverly conservative initial steam generator mass (nominal + 6% rather than nominal+3.5%). This results in conservatively late protection via reactor trip and auxiliaryfeedwater initiation which allows the heat-up to progress longer prior to protectiveactions occurring. Thus, current licensing basis LONF/LOAC analysis for BVPS-2remains acceptable.

Other Non-LOCA Events

No other non-LOCA event analysis credits the SG water level-low-low or SG waterlevel-high-high trip functions for protection. Thus, this aspect of this issue impacts noother non-LOCA event. The initial SG mass assumed in the non-LOCA analyses isbiased high or low depending on the direction of conservatism. If a particular event isnot sensitive to the initial SG mass, then the initial mass is set to the nominal mass. Anyevent that is analyzed with the initial mass set to the nominal mass is not impacted bythe new changes on the initial mass. The analysis for any event for which a high initialSG mass is conservative remains conservative because the positive uncertainty hasdecreased. There are no non-LOCA events for which a low initial SG mass yields morelimiting results than the FLB event which is discussed above.

SLB Mass and Energy Releases

(BVPS Unit No. 2 UFSAR Section: Mass and Energy Release Analysis for PostulatedSecondary System Pipe Rupture Inside Containment - 6.2.1.4

The analyses for the SLB mass and energy (M&E) releases inside and outsidecontainment typically use SG water level uncertainties for both the initial conditions andthe low-low reactor trip setpoint.

The current licensing-basis analysis of record for the SLB M&Es inside containment isdocumented using the MARVEL code (UFSAR Section 6.2.1.4.4) based onconservative generic values for the initial SG water mass. No allowance was made in

Page 19

I.


the MARVEL analysis for water level control, and the values used are conservativelylarge. Thus these revised uncertainties do not alter the validity of the analysis values atany power. The trip setpoint is not assumed in this analysis.

The current licensing-basis analyses of record for the SLB M&Es outside containmentare generic M&E releases documented using the LOFTRAN code (UFSAR Section15.0.1) as part of a WOG program. The initial conditions related to the SG water massare generic for the plants in the category for which the M&E releases were calculatedfor BVPS Unit No. 2. Any increase in the level uncertainty used for those analyses canbe absorbed by conservatisms associated with other assumptions included in theanalyses. The reactor trip setpoint used in the generic analyses has been evaluated(1999) and determined to be less than 0% for BVPS Unit No. 2. Thus, the trip setpointis conservative with respect to the increased uncertainties. These uncertainties do notalter the validity of the analyses values at any power.

LOCA Mass and Energy Releases

(BVPS Unit No. 2 UFSAR Section: Mass and Energy Release Analyses for PostulatedLoss of Coolant Accidents - 6.2.1.3)

The potential effect of the NSAL on the LOCA M&E related analysis is the change tothe steam generator fluid mass. The long term LOCA M&E calculations are typicallyinitialized at I 00% full power steady state conditions and the initial secondary side fluidmass is biased high. The current licensing basis long term LOCA M&E for Unit 1 wascalculated using the LOCTIC code. This is the same code which is used to calculate thecontainment response. The maximum steam generator liquid mass assumed in thisanalysis accounts for the 1.4% uprate, maximum steam generator liquid masscorresponding to 100% full power steady state conditions ( 44% NR level, with 10%uncertainty. The uncertainty used in the current safety analysis of record bounds theinitial condition instrumentation uncertainties. Therefore, the current analysis remainsbounding.

Steam Generator Tube Rupture

(BVPS Unit No. 2 UFSAR Section: Steam Generator Tube Rupture - 15.6.3)

The steam generator tube rupture analyses are not affected by the proposed changessince SG level control (initial condition uncertainties) and SG level trips are notmodeled in these analyses.

Page 20


LOCA Events

(BVPS Unit No.2 UFSAR Section: Loss Of Coolant Accidents - 15.6.5)

The LOCA and LOCA-related analyses are not affected by the proposed changes sinceSG level control (initial condition uncertainties) and SG level trips are not modeled inthese analyses.

AMSAC

(BVPS Unit No. 2 UFSAR Section: Anticipated Transients Without Trip - 4.3.1.7 and15.8)

A review of Chapter 4.3.1.7 of the BVPS-2 UFSAR shows that the ATWS MitigationSystem Actuation Circuitry (AMSAC) logic in place at BVPS-2 is actuated on a lowfeedwater flow condition. This AMSAC logic is Logic 2 of the generic WOG AMSACdesigns provided in WCAP-10858-P-A (Reference 10). Since the AMSAC at BVPS-2is not actuated on a low-low steam generator water level condition, there is no impacton the operation of the BVPS-2 AMSAC.

Margin to Trip

(BVPS Unit No. 2 UFSAR Section: None)

The proposed SG water level-low-low trip setpoint is being proposed to be raised from16.5% to 20.5% and the associated Allowable Value is being proposed to be raised from16% to 20%. The steam generator level control will continue to maintain steamgenerator water level constant at 44% at full power. Therefore, the steady-state fullpower operating margin to the low-low level reactor trip setpoint (difference betweennormal water level and the trip setpoint) wvill decrease from 27.5% to 23.5% with thenew proposed BVPS Unit No.2 trip setpoint and Allowable Value. A steady statemargin of 23.5% for BVPS Unit No. 2 is considered adequate based on field experienceto avoid unnecessary reactor trips on SG water level-low-low during normal expectedtransients where a reactor trip is neither wanted or expected. This includes plantstartups, shutdowns, steady-state operation and design basis load swings (± 10% loadchange).

The other area of concern is the steam generator level during a 50% load rejectiontransient. The steam generator low low level margin to trip analysis for the 50% loadrejection transient was performed for BVPS Unit 1. Because of the differences in thefeedwater condensate pump design, BVPS Unit I was more limiting than Unit 2. As

Page 21


such, a detailed steam generator level response analysis for the 50% load rejectiontransient was performed for Unit 1. Because Unit 2 has a higher feedwater system flowcapacity than Unit 1, the results of Unit I bound Unit 2.

The proposed SG water level-high-high trip setpoint is being proposed to be raised from80.6% to 92.2% and the associated Allowable Value is being proposed to be raised from81.1% to 92.7%. The steam generator level control will continue to maintain steamgenerator water level constant at 44% at full power. Therefore, the steady-state fullpower operating margin to the high-high level trip setpoint (difference between normalwater level and the trip setpoint) will increase from 36.6% to 48.2% with the newproposed BVPS Unit No. 2 trip setpoint and Allowable Value. This increases themargin to a high level trip, which will provide additional time for operator action toaddress any unnecessary high level condition before a high-high level condition isreached.

Radiological Analyses

The review of the post accident radiological consequences of the SG Water Level-low-low demonstrates that the offsite and control room doses associated with the accidentswill be within the acceptance criteria of 10 CFR Part 100 and 10 CFR 50.67, asapplicable. There were no new radiological safety analyses performed since thepreviously assumed low-low and high-high actuation values used in the safety analysesfor the steam generator water level setpoints were not adversely altered.

Conclusions of BVPS Unit No. 2 Safety Analyses Relating to the Proposed TechnicalSpecification Changes:

In summary, the impact of the proposed has been evaluated for BVPS Unit No. 2. Thenet impact of this evaluation confirms that the nominal SG water level-low-low tripsetpoint of 20.5% and the Allowable Value of 20.0% and confirms that the nominal SGwater level-high-high turbine trip/feedwater isolation setpoint of 92.2% level and theAllowable Value of 92.7% level are appropriate. There is sufficient margin in safetyanalyses to accommodate these impacts and continue to satisfy acceptance criteria.

5.0 REGULATORY SAFETY ANALYSIS

FirstEnergy Nuclear Operating Company (FENOC) requests to amend OperatingLicense DPR-66 for Beaver Valley Power Station (BVPS) Unit No. 1 and LicenseNPF-73 for BVPS Unit No. 2. The proposed amendment would revise TechnicalSpecification 3/4.3.1, "Reactor Trip System Instrumentation" and TechnicalSpecification 3/4.3.2, "Engineered Safety Feature Actuation System Instrumentation" to

Page 22


modify steam generator level allowable value setpoints. The proposed changes addressrecent generic issues involving new steam generator level uncertainty considerationsand margins associated with Westinghouse designed steam generators (SGs).

5.1 No Significant Hazards Consideration

FENOC has evaluated whether or not a significant hazards consideration is involvedwith the proposed amendments by focusing on the three standards set forth inI OCFR50.92, "Issuance of amendment," as discussed below:

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

No. The SG water level-low-low setpoint and allowable value have been revised toaddress Westinghouse Nuclear Safety Advisory Letter NSAL-03-9 and otherconsiderations on steam generator water level uncertainties. The revised setpointand allowable value calculations continues to follow the setpoint methodologypreviously approved for BVPS Unit No.1 and No.2 while addressing newlyidentified level uncertainty considerations. The proposed changes to the SG waterlevel-low-low Allowable Value for BVPS Unit No. 1 and No. 2 and to the SG waterlevel-high-high Allowable Value for BVPS Unit No. 2 continue maintain thevalidity of the safety analysis limits used in the safety analyses that credit theactuations based on SG water level.

The proposed changes do not alter the causes for any accident described in theUpdated Final Safety Analysis Report (UFSAR) that credit the SG water levelsetpoint actuations. Therefore, they do not involve a significant increase in theprobability of an accident previously evaluated.

The proposed changes do not alter the accident analyses that credit the SG waterlevel-low-low setpoint actuation or the associated accident acceptance criteria.Therefore, they do not involve a significant increase in the consequences of anaccident previously evaluated.

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

No. The SG water level-low-low setpoint and allowable value have been revised toaddress Westinghouse Nuclear Safety Advisory Letter NSAL-03-9 and otherconsiderations on steam generator water level uncertainties. Implementation of theproposed setpoint changes have no significant effect on either the configuration of

Page 23


the plant, or the manner in which the plant is operated. The proposed changes to theSG water level-low-low allowable value for BVPS Unit No. 1 and No. 2 and to theSG water level-high-high allowable value for BVPS Unit No. 2 continue to maintainthe validity of the safety analysis limits used in the safety analyses that credit theactuations based on SG water level.

Therefore, since the plant configuration is not adversely changed and the proposedchanges do not alter the accident analyses that credit actuation based on SG waterlevel, the proposed change does not create the possibility of a new or differentaccident from any accident previously evaluated.

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

No. The Reactor Trips System and Engineered Safety Feature Actuation Systemsetpoint analysis methodology and acceptance criteria provide the margin of safety.The SG water level-low-low and SG water level-high-high actuation setpoint andallowable value have been calculated using the same methodology as previouslyapproved for BVPS Unit No. I and No. 2 while addressing newly identifiedconsiderations needed to protect the limits used in the safety analyses. Theapplicable safety analyses have been performed and show acceptable results.Therefore, the proposed change does not involve a significant reduction in a marginof safety.

Based on the above, FENOC concludes that the proposed amendments present nosignificant hazards consideration under the standards set forth in 1OCFR50.92(c), and,accordingly, a finding of "no significant hazards consideration" is justified.

5.2 Applicable Regulatory Requirements/Criteria

A review of 10 CFR 50, Appendix A, General Design Criteria for Nuclear Power Plants(Reference 11), was conducted to assess the potential impact associated with theproposed changes. General Design Criteria (GDC) 20 is potentially impacted and isassessed with respect to the need for a modification to the UFSAR description of BVPSUnit No. I and No. 2 design conformance to the GDC.

GDC 20 Protection System Functions

The protection system shall be designed (1) to initiate automatically the operation ofappropriate systems including the reactivity control systems, to assure that specifiedacceptable fuel design limits are not exceeded as a result of anticipated operational

Page 24


occurrences and (2) to sense accident conditions and to initiate the operation of systemsand components important to safety.

The proposed changes do not alter any current physical plant components. Only thesetpoint and allowable value for steam generator water level are proposed to bemodified. The proposed changes continue to maintain the underpinning safety analysesvalid. Thus, the UFSARs discussion for GDC 20 regarding the design arrangement(e.g., redundancy, isolation, single failure, physical separation, etc.) will remain valid.Thus, compliance with GDC 20 is not impacted by the proposed changes.

In conclusion, based on the considerations discussed above, (1) there is reasonableassurance that the health and safety of the public will not be endangered by operation inthe proposed manner, (2) such activities will be conducted in compliance with theCommission's regulations, and (3) the issuance of the amendment will not be inimicalto the common defense and security or to the health and safety of the public.

6.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirementwith respect to installation or use of a facility component located within the restrictedarea, as defined in 1OCFR20, or would change an inspection or surveillancerequirement. However, the proposed amendment does not involve (i) a significanthazards consideration, (ii) a significant change in the types or significant increase in theamounts of any effluent that may be released offsite, or (iii) a significant increase inindividual or cumulative occupational radiation exposure. Accordingly, the proposedamendment meets the eligibility criterion for categorical exclusion set forth in1OCFR51.22(c)(9). Therefore, pursuant to 1OCFR51.22(b), no environmental impactstatement or environmental assessment need be prepared in connection with theproposed amendment.

7.0 REFERENCES

1. WCAP-1 1419, Revision 2, "Setpoint Methodology for Protection Systems forBeaver Valley Power Station - Unit 1," dated December, 2000.

2. WCAP-1 1366, Revision 4, "Setpoint Methodology for Protection Systems forBeaver Valley Power Station - Unit 2," dated December, 2000.

3. Beaver Valley Power Station Unit No. 1 Technical Specification Amendment No.239, License No. DPR-66, dated July 30, 2001.

Page 25

;W


4. Beaver Valley Power Station Unit No. 2 Technical Specification Amendment No.120, License No. NPF-73, dated July 30, 2001.

5. Nuclear Safety Advisory Letter, NSAL-02-3, "Steam Generator Mid-Deck PlatePressure Loss Issue," Rev. 0 dated February 15, 2002, and Rev. 1 dated April 8,2002.

6. WCAP-16115-P, "Steam Generator Level Uncertainties Program," dated September,2003.

7. Nuclear Safety Advisory Letter, NSAL-03-9, "Steam Generator Water LevelUncertainties," dated September 22, 2003.

8. Technical Bulletin, TB-04-12, "Steam Generator Level Process PressureEvaluation,", dated June 23, 2004.

9. Nuclear Safety Advisory Letter, NSAL-02-4, "Maximum Reliable Indicated SteamGenerator Water Level," dated February 19, 2002

I0.WCAP-10858-P-A, "AMSAC GENERIC DESIGN PACKAGE," dated October,1986.

11.10 CFR 50, Appendix A, General Design Criteria for Nuclear Power Plants.

Page 26

Attachment A-1

Beaver Valley Power Station, Unit No. 1License Amendment Request No. 327

Proposed Technical Specification Changes

The following are the affected pages:

3/4 3-3

3/4 3-19a

"I

TABLE 3.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION

TOTAL NO.OF CHANNELS

MINIMUMCHANNELS CHANNELSTO TRIP OPERABLE

ALLOWABLEVALUE

APPLICABLEMODESFUNCTIONAL UNIT ACTION

7. Overtemperature AT

8. Overpower AT

9. Pressurizer Pressure-Low(Above P-7)

10. Pressurizer Pressure-High

11. Pressurizer Water Level- .High (Above P-7)

3 2

23

2

2

2

2

2

3 2

See TableNotation (A)

See TableNotation (B)

2 1941 psig

• 2389 psig

• 92.5% ofinstrument span

1, 2

1, 2

1, 2

1, 2

1, 2

7

7

7

7

7

3 2

3 2

12. Loss of Flow - Single Loop(Above P-8)

13. Loss of Flow - Two Loops(Above P-7 and below P-8)

3/loop

3/loop

2/loop inanyoperatingloop

2/loop intwooperatingloops

2/loop ineachoperatingloop

eachoperatingloop

Ž 89.8% ofindicated loopflow

2 89.8% ofindicated loopflow e4Z

1 7

1 7

I14. Steam Generator WaterLevel-Low-Low(Loop Stop Valves Open)

3/loop 2/loop 2/loop 2 14.60 of narrowrange instrumentspan-each steamgenerator

1, 2 7

BEAVER VALLEY - UNIT 1 3/4 3 -3 Amendment No. 213-4


ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION


MINIMUMCHANNELS CHANNELS

TO TRIP OPERABLEALLOWABLE

VALUEAPPLICABLE

MODESFUNCTIONAL UNIT ACTION

7. AUXILIARY FEEDWATER

a. Steam Gen. Water Level-Low-Low (Loop StopValves Open)

i. Start Turbine DrivenPump

ii. Start Motor DrivenPumps

3/stm. gen.

3/Stm. gen.any 2 stm.gen.

2/stm.gen. anystm. gen.

2/stm.gen. any2 stm.gen.

2/stm.gen.

2/stm.gen.

2 14.6% of narrow 1, 2, 3range ii mentspan each steagenerator 19.6%

2 narrow 1, 2, 3range instrumentspan each steamgenerator

14

14

I

I

b. Undervoltage-RCP (StartTurbine Driven Pump)

(3) -1/bus 2 2 2 71.2% rated RCPbus voltage

1 14

c. S.I. (Start AllAuxiliary FeedwaterPumps)

See 1 above (all S.I. initiating functions and requirements)

d. (Deleted)

e. Trip of Main FeedwaterPumps (Start MotorDriven Pumps)

1/pump 1 1 Not Applicable 1, 2, 3 18

BEAVER VALLEY - UNIT 1 3/4 3-19a Amendment No. 2394

Attachment A-2


Proposed Technical Specification Changes

M - - ---- M -

The following are the affected pages:

3/4 3-3

3/4 4-19

3/4 4-20


REACTOR TRIP SYSTEM INSTRUMENTATION


MINIMUMCHANNELS CHANNELSTO TRIP OPERABLE

ALLOWABLEVALUE

APPLICABLEMODESFUNCTIONAL UNIT ACTION


8. Overpower AT

9. Pressurizer Pressure-Low(Above P-7)


11. Pressurizer Water Level-High (Above P-7)

3 2

23

2

2

2

2

2

3 2

See TableNotation (A)

See TableNotation (B)

2 1941 psig**

• 2379 psig

• 92.5* ofinstrument span

1, 2

1, 2

1, 2

1, 2

1, 2

7

7

7

7

7

3 2

3 2

12. Loss of Flow - Single Loop(Above P-8)

13. Loss of Flow - Two Loops(Above P-7 and below P-8)

14. Steam Generator WaterLevel-Low-Low

3/loop

3/loop

3/loop

2/loop inanyoperatingloop

2/loop intwooperatingloops

2/loop

2/loop ineachoperatingloop

2/loopeachoperatingloop

2/loop

2 89.6* ofindicated loopflow

2 89.6% ofindicated loopflow

2 446 of narrowrange instrumentspan-each steamgenerator

1 7

1 7

1, 2 I7

** Time constants utilized in the lead-lag controller for Pressurizer Pressure-Low are > 2 seconds for lead and• 1 second for lag. Channel calibration shall ensure that these time constants are adjusted to thosevalues.

BEAVER VALLEY - UNIT 2 3/4 3 -3 Amendment No. +4-0






VALUEAPPLICABLE


5. TURBINE TRIP & FEEDWATERISOLATION

a. Automatic ActuationLogic and ActuationRelays

b. Steam Generator WaterLevel--High-High, P-14

2 1 2 N.A. 1, 2 42

14 13/loop 2/loopin anyoperatingloop

2/loopin eachoperatingloop

• 81.1- of narrowrange instrumentspan

1, 2, 3

c. Safety Injection See Item 1 above for allrequirements.

Safety Injection initiating functions and

6. LOSS OF POWER

a. 4.16kv Emergency Bus

1. Undervoltage(Trip Feed)

2. Undervoltage(Start Diesel)

2/4.16kvBus

1/4.16kvBus

2/4.16kv 2/4.16kvBus Bus

2 71.2% of ratedBus Voltage with

a 1 i 0.1 secondtime delay

2 71.2% of ratedBus Voltage,20 cycles

± 2 cycles

1, 2, 3, 4

1, 2, 3, 4

33

331/4.16kvBus

1/4kv Bus

b. 4.16kv Emergency Bus(Degraded Voltage)

2/4.16kvBus

2/Bus 2/Bus 2 93.1% of ratedBus Voltage with a

90 ± 5 secondtime delay

1, 2, 3, 4 34

BEAVER VALLEY - UNIT 2 3 /4 3 -19 Amendment No. a2Go

%I






VALUEAPPLICABLE


6. LOSS OF POWER (Continued)

c. 480 Volt Emergency Bus(Degraded Voltage)

2/480v Bus 2/Bus 2/Bus > 93.1% of ratedBus Voltage with a90 ± 5 secondtime delay

1, 2, 3, 4 34

7. AUXILIARY FEEDWATER(3)

a. Automatic ActuationLogic and ActuationRelays

b. Steam Gen. Water Level--Low-Low

2 1 2 N.A. 1, 2, 3 42

1. Start TurbineDriven Pump

2. Start MotorDriven Pumps

c. Undervoltage-RCP (StartTurbine Driven Pump)

3/stm. gen.

3/stm. gen.

(3) -1/bus

2/stm.gen. anystm. gen.

2/stm.gen. any2 stm.gen.2

2/stm.gen.

2/stm.gen.

2 of narrow 1, 2,range in~ mentspan

Ž 14-6 of narrow 1, 2,range instrumentspan

3 14

3 14

142 2 71.2% of ratedbus voltage

1, 2

(3) Manual initiation is included in Specification 3.7.1.2.

BEAVER VALLEY - UNIT 2 3/4 3 -20 Amendment No. AdG

Attachment B-i


Proposed Licensing Requirement Manual Changes

Licensing Requirement Manual changes are provided for information only.

The following are the only affected pages:

3.9-23.9-6

BVPS-ILICENSING REQUIREMENTS MANUAL

TABLE 3.9-1

Provided forInformation Only.

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

FUNCTIONAL UNIT NOMINAL TRIP SETPOINT

1. Manual Reactor Trip

2. Power Range, Neutron Flux

A. High Setpoint

B. Low Setpoint

3. Power Range, Neutron Flux, High PositiveRate

4. Power Range, Neutron Flux, High NegativeRate

5. Intermediate Range, Neutron Flux

6. Source Range, Neutron Flux

A. With Rod Withdrawal Capability

B. With All Rods Fully Inserted andWithout Rod Withdrawal Capability


8. Overpower AT

9. Pressurizer Pressure-Low


I1. Pressurizer Water Level-High

12. Loss of Flow

A. Single Loop

B. Two Loops

13. Steam Generator Water Level-Low-Low

14. Deleted

Not Applicable

109% of RATED THERMAL POWER


5% of RATED THERMAL POWER with atime constant 2 2 seconds

5% of RATED THERMAL POWER with atime constant 2 2 seconds


105 counts per second

Not Applicable

See Technical Specification Table Notation(A) on Table 3.3-1

See Technical Specification Table Notation(B) on Table 3.3-1

1945 psig

2385 psig

92% of instrument span

90.2% of indicated loop flow

90.2% of indicated loop flow

5.10% °of narrow range instrument span-each/ steam generator

I

3.9-2 Revision 24

BVPS-1_LICENSING REQUIREMENTS MANUAL

Providedfor|Information Only.


ENGINEERED SAFETY FEATURE ACTUATION SYSTEMINSTRUMENTATION TRIP SETPOINTS


6. LOSS OF POWERa. 4.16 kv Emergency Bus Undervoltage

I . Loss of Voltage (Trip Feed)

2. Loss of Voltage (Start Diesel)

b. 4.16kv Emergency Bus Undervoltage(Degraded Voltage)

c. 480v Emergency Bus Undervoltage(Degraded Voltage)

7. AUXILIARY FEEDWATER

a. Steam Generator Water Level-Low-Low

i. Start Turbine Driven Pump

ii. Start Motor Driven Pumps

b. Undervoltage - RCP (Start Turbine DrivenPump)

c. S.I. (Start All Auxiliary Feedwater Pumps)

d. (Deleted)

e. Trip of Main Feedwater Pumps (StartMotor Driven Pumps)

8. ESF INTERLOCKSa. Reactor Trip, P-4

b. Pressurizer Pressure, P-1I

c. Low-Low Tavg, P-12

75% of rated bus voltage with a 1 ± 0.1second time delay

75% of rated bus voltage with a < 0.9 secondtime delay (includes auxiliary relay times)

93.7% of rated bus voltage with a 90 ± 5second time delay

93.7% of rated bus voltage with a 90 ± 5second time delay

20.1%

- of arrow range instrument span eachsteam nerator

of narrow range instrument span eachsteam generator

75% rated RCP bus voltage

See 1 above (all SI Setpoints)

I

I

Not Applicable

Not Applicable

2000 psig

541 cF

3.9-6 Revision 49

Attachment B-2


Proposed Licensing Requirement Manual Changes

Licensing Requirement Manual changes are provided for information only.

The following are the only affected pages:

3.10-33.10-6

BVPS-2_LICENSING REQUIREMENTS MANUAL Providedfor

InforTL3t0on Only.

TABLE 3.1 10-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

20.5%FUNCTIONAL UNIT §2. NOMINAL*** TRIP SETPOINT

13. Steam Generator Water Level-Low-Low of narrow range instrument span-eachsteam generator

I

14.

15.

16.

17.

18.

19.

20.

21.

22.

DELETED.

Undervoltage - Reactor Coolant Pumps

Underfrequency-Reactor Coolant Pumps

Turbine Trip

a. Emergency Trip Header Low Pressure

b. Turbine Stop Valve Closure

Safety Injection Input from ESF

Reactor Coolant Pump Breaker Position Trip

Reactor Trip Breakers

Automatic Trip Logic

Reactor Trip System Interlocks

a. Intermediate Range Neutron Flux, P-6

b. Power Range Neutron Flux, P-8

c. Power Range Neutron Flux, P-9

d. Power Range Neutron Flux, P-b 0 (Input toP-7)

e. Turbine First Stage Pressure, P-13 (Inputto P-7)

75% of rated bus voltage-each bus

57.5 Hz-each bus

1000 psig

> 1% open

N.A.

N.A.

N.A.

N.A.

I x 10'10 amps

30% of RTP*

49% of RTP*

10% of RTP*

10% of RTP* Turbine First Stage PressureEquivalent

* = RATED THERMAL POWER

*** WVith the exception of Functional Unit number 17.b.

3.10-3 Revision l2

BVPS-2LICENSING REQUIREMENTS MANUAL


ProvidedforInformation Only.

ENGINEERED SAFETY FEATURE ACTUATION SYSTEMINSTRUMENTATION TRIP SETPOINTS


5. TURBINE TRIP & FEEDWATERISOLATION

a. Automatic Actuation Logic and ActuationRelays

b. Steam Generator Water Level - High-High, P-14

c. Safety Injection

N.A.

8A of narrow range instrument span

See Functional Unit 1. above for all SafetyInjection Trip Setpoints.

6. LOSS OF POWER

a. 4.16 kV Emergency Bus

1. Undervoltage(Trip Feed)

2. Undervoltage(Start Diesel)

b. 4.16 kV Emergency Bus(Degraded Voltage)

c. 480 Volt Emergency Bus(Degraded Voltage)

7. AUXILIARY FEEDWATER*

75% of rated Bus Voltage with aI ± 0.1 second time delay

75% of rated Bus Voltage, 20 cycles+ 2 cycles

93.4% of rated Bus Voltage with a90 ± 5 second time delay

93.4% of rated Bus Voltage with a90 ± 5 second time delay

a. Automatic Actuation Logic and ActuationRelays

b. Steam Generator Water Level-Low-Low

1. Start Turbine Driven Pump

2. Start Motor Driven Pumps

N.A.

w 20.5% i

16.5% ofnarrow range instrument span

4 % of narrow range instrument span

I

I

*Manual initiation is included in Technical Specification 3.7.1.2

3.10-6 Revision 4-9

I it

Attachment C

Beaver Valley Power Station, Unit Nos. 1 & 2License Amendment Request No. 327 & 197

Commitment Summary

The following list identifies those actions committed to by FirstEnergy Nuclear OperatingCompany (FENOC) for Beaver Valley Power Station (BVPS), Unit Nos. I & 2 in thisdocument. Any other actions discussed in the submittal represent intended or planned actionsby Beaver Valley. These other actions are described only as information and arc not regulatorycommitments. Please notify Mr. Larry R. Freeland, Manager, Regulatory Compliance, atBeaver Valley on (724) 682-5284 of any questions regarding this document or associatedregulatory commitments.

Commitment Due Date

None N/A

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