L-04-127, License Amendment Request Nos. 327 and 197 on Steam Generator Level Allowable Value Setpoints

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License Amendment Request Nos. 327 and 197 on Steam Generator Level Allowable Value Setpoints
ML042860478
Person / Time
Site: Beaver Valley
Issue date: 10/05/2004
From: Pearce L
FirstEnergy Nuclear Operating Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
L-04-127
Download: ML042860478 (43)
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Text

j' O FO pJ(JBeaver Valley Power Station Route 168 Ift-_%P0. Box 4 FrstEnergy Nuclear Operating Company Shippingport. PA 15077-0004 L. Willian Pearce 724-682-5234 Site Vice President Fax: 724-643-8069 October 5, 2004 L-04-127 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001

Subject:

Beaver Valley Power Station, Unit No. 1 and No. 2 BV-1 Docket No. 50-334, License No. DPR-66 BV-2 Docket No. 50-412, License No. NPF-73 License Amendment Request Nos. 327 and 197 on Steam Generator Level Allowable Value Setpoints Pursuant to 10 CFR 50.90, FirstEnergy Nuclear Operating Company (FENOC) requests an amendment to the license for Beaver Valley Power Station (BVPS) Unit No. I and Unit No. 2 in the form of changes to the Technical Specifications. The proposed amendments will modify steam generator level allowable value setpoints used in the Reactor Trip System and Engineered Safety Feature Actuation System instrumentation to address identified non-conservative setpoints. The proposed changes address recent generic issues involving new steam generator level uncertainty considerations and margins associated with Westinghouse designed steam generators.

The FENOC evaluation of the proposed changes is presented in the Enclosure. The proposed Technical Specification changes are presented in Attachment A. Attachment B provides the proposed information-only changes to the Licensing Requirements Manual that reflect the proposed license amendment. Attachment C indicates that there are no new commitments made in this submittal.

The Beaver Valley review committees have reviewed the changes. The changes were determined to be safe and do not involve a significant hazard consideration as defined in 10 CFR 50.92 based on the attached safety evaluation and no significant hazard evaluation.

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. 2 License Amendment Request Nos. 327 and 197 L-04-127 Page 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on October < , 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 Manager Mr. P. C. Cataldo, NRC Sr. Resident Inspector Mr. S. J. Collins, NRC Region I Administrator Mr. D. A. Allard, Director BRP/DEP Mr. L. E. Ryan (BRP/DEP)

l ENCLOSURE Beaver Valley Power Station, Unit Nos. 1 and 2 License Amendment Request No. 327 and 197 FENOC Evaluation of the Proposed Change

Subject:

Application for Amendment of Beaver Valley Power Station Technical Specifications To Revise Steam Generator Level Allowable Value Setpoints.

Table of Contents Section Title Page

1.0 DESCRIPTION

................................ 1

2.0 PROPOSED CHANGE

S ............................... 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 Attachments Number Title A Proposed Technical Specification Changes B Proposed Licensing Requirement Manual Changes C Commitment Summary i

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

1.0 DESCRIPTION

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

2.0 PROPOSED CHANGE

The proposed Technical Specification (TS) changes, which are submitted for Nuclear Regulatory Commission (NRC) review and approval, are provided in Attachment A-1 and A-2. The changes proposed to the Licensing Requirements Manual are provided in Attachment B-i and B-2. The proposed Licensing Requirements Manual (LRM) changes do not require NRC approval. Changes to the LRM are controlled through the 10 CFR 50.59 process. The LRM changes are provided for information only.

Attachment C provides a list of commitments associated with this License Amendment Request (LAR).

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

To meet format requirements, the Index, Technical Specifications and LRM pages will be revised and repaginated as necessary to reflect the changes being proposed by this LAR.

Page 1

Beaver Valley Power Station Unit Nos. 1 & 2 License 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 Item No. Unit 1 1 3.3.1.1 Reactor Trip System Instrumentation, Steam Table 3.3-1 Generator Water Level - Low-Low, Allowable Value Functional Unit 14 2 1 3.3.2.1 Engineered Safety Feature Actuation System Table 3.3.-3 Instrumentation, Auxiliary Feedwater, Steam Functional Generator Water Level - Low-Low, Allowable Value Unit 7.a 3 2 3.3.1.1 Reactor Trip System Instrumentation, Steam Table 3.3-1 Generator Water Level - Low-Low, Allowable Value Functional Unit 14 4 2 3.3.2.1 Engineered Safety Feature Actuation System Table 3.3.-3 Instrumentation, Auxiliary Feedwater, Steam Functional Generator Water Level - Low-Low, Allowable Value Unit 7.b 5 2 3.3.2.1 Engineered Safety Feature Actuation System Table 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 Specification 3.3.1.1, "REACTOR TRIP SYSTEM INSTRUMENTATION." This modification consists 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 Page 2

Beaver Valley Power Station Unit Nos. I & 2 License Amendment Request No. 327 & 197 span - each steam generator. This proposed change is shown in the marked-up Technical Specifications in Attachment A.

TS Proposed Change Number 2 This proposed change is a modification to BVPS Unit No. I Technical Specification 3.3.2.1, "ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION." This modification consists of revising the Technical Specification allowable value for the SG water level-low-low auxiliary feedwater actuation functions from 14.6% to 19.6% of narrow range instrument span - each steam generator. This proposed change is shown in the marked-up Technical Specifications in Attachment A.

TS Proposed Change Number 3 This proposed change is a modification to BVPS Unit No. 2 Technical Specification 3.3.1.1, "REACTOR TRIP SYSTEM INSTRUMENTATION." This modification consists 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 Technical Specifications in Attachment A.

TS Proposed Change Number 4 This proposed change is a modification to BVPS Unit No. 2 Technical Specification 3.3.2.1, "ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION." This modification consists of revising the Technical Specification allowable value for the SG water level-low-low auxiliary feedwater actuation functions from 16% to 20% of narrow range instrument span - each steam generator. This proposed change is shown in the marked-up Technical Specifications in Attachment A.

TS Proposed Change Number 5 This proposed change is a modification to BVPS Unit No. 2 Technical Specification 3.3.2.1, "ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION." This modification consists of revising the Technical Specification allowable value for the SG water level-high-high turbine trip and feedwater isolation actuation function from 81.1% to 92.7% of narrow range instrument span - each steam generator. This proposed change is shown in the marked-up Technical Specifications in Attachment A.

Page 3

Beaver Valley Power Station Unit Nos. 1 & 2 License Amendment Request No. 327 & 197 2.2 Licensing Requirement Manual Changes The corresponding nominal trip setpoint changes for the proposed Technical Specification allowable value changes discussed above are shown in Attachment B-1 and B-2. These nominal trip setpoints are maintained in the Licensing Requirements Manual, which is a licensee controlled document that is changed under the 10 CFR 50.59 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 Trip System (RTS). It functions to trip the reactor and protect the reactor core from a loss of heat sink in the event of a sustained steam/feedwater flow mismatch. This trip is actuated on two-out-of-three water level-low-low signals occurring in any SG. The basic function of the reactor protection circuits associated with the SG water-low-low reactor 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 water level-low-low.

A spurious high signal from the feedwater flow channel being used for control would cause a reduction in feedwater flow. The mismatch between steam demand and feedwater flow produced by this spurious signal will actuate alarms to alert the operator of 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 of indicated feedwater flow.

Steam generator water level-low-low is also a functional unit of the Engineered Safety Feature Actuation System (ESFAS). It functions to actuate the Auxiliary Feedwater (AFW) pumps to provide auxiliary feedwater to the secondary side of the SGs in order to maintain a heat sink. The turbine driven AFW pump is started on two-out-of-three water level-low-low signals occurring in any one SG and the motor driven AFW pumps are 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. It functions 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) to avoid adverse effects of excess steam moisture on the accuracy of steam flow and steamline pressure transmitters downstream in the steam piping, and 3) to avoid Page 4

I Beaver Valley Power Station Unit Nos. 1 & 2 License Amendment Request No. 327 & 197 addressing the loading effects of water in the steam piping support design. Steam generator water level-high-high signals in two-out-of-three channels for any steam generator will actuate a turbine trip, trip the main feedwater pumps, and close the main and bypass feedwater level control valves.

A spurious low signal from the feedwater flow channel being used for control would cause an increase in feedwater flow. The mismatch between steam flow and feedwater flow produced by this spurious signal will actuate alarms to alert the operator of this situation 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 the feedwater isolation. If the turbine trip occurs when reactor power is above the P-9 permissive 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-high setpoints are currently established in accordance with the methodology described in WCAP-1 1419, Rev. 2, "Setpoint Methodology for Protection Systems for Beaver Valley Power Station - Unit 1," (Reference 1) and WCAP-1 1366, Rev. 4, "Setpoint Methodology for Protection Systems for Beaver Valley Power Station - Unit 2,"

(Reference 2). These two WCAPs were previously submitted via FENOC Letter L-00-143, dated December 27, 2000 as part of the submittal for BVPS Unit No. I LAR 286 and BVPS Unit No. 2 LAR 158. These WCAPs were also used by the NRC as part of 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 Specification Amendment No. 120, dated July 20,2001 (Reference 4).

References 1 and 2 describe various generic issues and concepts that may affect water level uncertainties in plants with Westinghouse-designed steam generators. The uncertainty that must be considered is associated with two sources. These are the instrumentation hardware itself and the non-instrumentation or process measurement accuracy (PMA) effects. Westinghouse identified four separate PMA terms in 1992:

Process Pressure Variations, Reference Leg Temperature, Fluid Velocity Effects, and Downcomer Subcooling which were addressed in References I and 2.

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

Beaver Valley Power Station Unit Nos. 1 & 2 License Amendment Request No. 327 & 197 specify additional PMA terms to address the uncertainties associated with steam flow through the mid-deck plate inside a steam generator. Westinghouse subsequently concluded in a BVPS-specific evaluation that both BVPS Units had sufficient margin to offset the effects of the mid-deck plate differential pressure (AP) issues described in Reference 5 and that no changes were needed in the BVPS Unit Nos. 1 and 2 Technical Specifications. Specifically, NSAL-02-03 concluded that for a steam generator affected by a Feedwater Line Break (FLB), reverse flow occurs through the feedring out of the steam generator nozzle and eventually out the break which results in a reversal in sign of 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 not adversely affect the setpoints used for the FLB analyses described in the BVPS Unit Nos. 1 and 2 Updated Final Safety Analysis Reports (UFSARs) and as approved in References 3 and 4. It is noted that in both References 1 and 2 and in the 2002 NSAL the bounding FLB event was believed to be the large FLB (and no specific small or intermediate FLB analyses were performed). In addition, Westinghouse concluded that the mid-deck AP effect also did not adversely affect the other steam generator water level setpoint analyses (i.e., steam line break or loss of normal feedwater).

Subsequent discussions with Westinghouse Owners Group (WOG) members and within Westinghouse resulted in the identification of four additional elements of Westinghouse SG designs that could affect SG level measurement and should be addressed in SG uncertainty assessments. These include the intermediate deck plate d/p, the feedring d/p, the effects of steam carry under into the downcomer, and the lower deck plate supports. In addition to these design elements, four transient conditions were identified that could produce transient-specific effects that should be included in assessments of SG 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 Feedwater Malfunction event. In October 2002, Westinghouse was authorized to complete a program for the WOG to provide additional generic guidance for plants in assessing the effects of various SG design-related issues and provide guidance for evaluating SG water level PMA effects on a plant-specific basis. As a result of this WOG program, in September, 2003, Westinghouse issued WCAP-161 15-P, "Steam Generator Level Uncertainties Program" (Reference 6) and issued NSAL-03-9 (Reference 7) which provided additional insight into the information previously provided by NSAL-02-03 (Reference 5) and on the four new design elements/transients considerations and other information involving steam generator water level instrumentation uncertainties.

As a result of these Westinghouse actions, FENOC became aware in September, 2003 that the uncertainty analysis performed for the steam generator water level setpoints at both BVPS Units may not be adequate to address all credible potential conditions, as Page 6

Beaver Valley Power Station Unit Nos. 1 & 2 License Amendment Request No. 327 & 197 previously determined by Westinghouse for the BVPS Units. Specifically, the primary issue was the new information that all plants with a mid-deck plate pressure loss must now consider the mid-deck AP bias, since the small to intermediate size FLB may now be 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 intermediate size of FLBs where there is no reverse flow out of the steam generator with the ruptured line 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 now add a new bias to their steam generator water level-low-low setpoints to address uncertainty for the AP across the mid-deck during a FLB. It was initially estimated that addressing the new mid-deck AP bias would require an additional five percent level in the steam generator water level-low-low setpoint calculations for FLB. With insufficient margin remaining in the steam generator water level-low-low setpoint calculations, FENOC immediately implemented administrative controls to conservatively raise both Unit's steam generator water level-low-low Allowable Values by an additional five percent beyond the values required by Technical Specifications to address 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 Specification changes includes evaluation of initial condition uncertainties at power conditions, which are provided as input to the safety analyses, and the development of the changes to the steam generator level reactor trip and engineered safety feature actuation system setpoints and allowable values.

To evaluate the steam generator level uncertainty and setpoint calculations for BVPS Unit No. 1 and No. 2 which are used as the bases for these License Amendment Requests (LARs), a complete review of the development of each individual term documented in the previously licensed calculation of record (References 1 and 2) was performed to address the new industry information on Process Measurement Accuracy terms (as described in Section 3.2) and to determine if any unnecessary conservatism could be decreased. In the area of rack uncertainties, no adjustments could be made.

Under the environmental allowance terms were two areas of potential relaxation. The uncertainties, margins, setpoints, and Allowable Values for SG water level-low-low for both BVPS Units and for SG water level-high-high for BVPS Unit No. 2 have been calculated, which demonstrate positive margins are maintained with the proposed changes.

Page 7

Beaver Valley Power Station Unit Nos. 1 & 2 License Amendment Request No. 327 & 197 LAR No. 327 for BVPS Unit No. I is applicable for power levels at or below the currently 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 2910 MWt with the currently installed Model 51M steam generators. Although 2910 MWt is above the currently maximum allowed rated thermal power of 2689 MWt, this evaluation bounds current licensed condition and will also support potential future license actions on BVPS Unit No. 2. [Note: no change in rated thermal power level is being requested by this LAR.]

4.1 Basis for TS Proposed Changes 1, 2, 3, and 4: Steam Generator Water Level-Low-Low Allowable 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 to address additional effects on steam generator water level uncertainties, an evaluation was performed upon the current steam generator water level setpoints and methodology as described in References 1 and 2. Previously, for the SG water level-low-low parameter, separate uncertainty calculations were performed for the Steam Line Break (SLB), Feedwater Line Break (FLB), and the Loss of Normal Feedwater (LONF) design basis accidents. As shown in Reference 1, the limiting case for SG water level-low-low at BVPS Unit 1 was previously the LONF event. As shown in Reference 2, the limiting case 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 both BVPS Units needs to be revised to address new limiting events, which will require revision 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 and resulted in this request to change the Technical Specification Allowable Value for SG water level-high-high.

The evaluation of the BVPS Unit No. I SG water level-high-high PMA terms in light of the new Westinghouse/WOG information determined that the existing PMA terms remain bounding. The BVPS Unit No. I SG water level-high-high Allowable Value was 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 for BVPS Unit No. 2 to remove some of the available margin, it was determined that no changes would be requested since the BVPS Unit No. 1 steam generators are currently Page 8

Beaver Valley Power Station Unit Nos. 1 & 2 License Amendment Request No. 327 & 197 scheduled to be replaced in the 2006 refueling outage, which will require a complete new uncertainty analysis be established in 2006 for the replacement SGs. Thus, there was limited benefit to now revise station calculations/procedures to address a new SG water level-high-high with only one more cycle of operation left at BVPS Unit No.1 with these original steam generators. The current SG water level-high-high value at BVPS Unit No. I remains conservative.

Setpoint Analyses Changes The process measurement accuracy (PMA) terms evaluated for the proposed Technical Specification changes in these LARs were: 1) Process Pressure Variations, 2) Reference Leg Temperature, 3) Fluid Velocity Effects, 4) Downcomer Subcooling, 5) Dynamic Losses, 6) Intermediate Deck Plate d/p, 7) Feedring d/p, and finally 8) Mid-deck plate d/p. In addition, the Environmental Allowance modifier for Reference Leg Temperature Effects was also modified, to be consistent with Item 2. Items 5, 6, 7 and 8 are new parameters not previously described in References I and 2 and were added to the uncertainty calculations. For the SG water level-low-low, Items 1, 2 and 4 were found to need revision in the uncertainty calculations. For the SG water level-high-high setpoint analysis, Items 1 and 3 were found to need revision in the uncertainty calculation. In addition, the contribution from Item 8 was included in the uncertainty evaluation. All the other identified PMA items either had no new additional impact or an insignificant impact on the uncertainty calculations.

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

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

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

The uncertainty associated with the Reference Leg Temperature Heatup effect increased for FLB events for both BVPS Units. Previously, the FLB case only addressed the large FLB event and limited reference leg heatup was assumed. Now, for the newly defined Page 9

Beaver Valley Power Station Unit Nos. 1 & 2 License Amendment Request No. 327 & 197 small/intermediate FLB, reference leg temperature is presumed to increase further since the actuation condition takes longer to reach during the transient. The uncertainty for the reference leg temperature is modeled based on Tsat equivalent to Psat for the containment high setpoint. The uncertainty associated with the reference leg heatup effect was also increased for the large FLB, but to a lesser extent.

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

The effects due to downcomer subcooling equivalent to nominal full power conditions were applied, as a benefit reducing the previous value for this PMA parameter, for both BVPS 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 SLB and LONF for BVPS Unit No. I as a benefit, it was conservatively not applied to limit the number of changes needed for BVPS Unit No. I evaluation.]

The Fluid Velocity effects were reduced to zero for the analyzed feedwater malfunction event 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 being analyzed. For the large FLB event, the mid-deck plate AP error is based on 100% steam flow conditions; for the SLB event, it is based on 121% steam flow conditions; and for small/intermediate FLB and LONF events, it is based on 112% steam flow values. The uncertainty effect for mid-deck AP increases with higher steam generator steam flow rates. 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 level transmitter drift value was lowered to more accurately model the BVPS Unit No. 1 instrument operation based upon a review of past empirical values obtained at BVPS Unit No. 1. The BVPS Unit No. 2 steam generator water level transmitter drift value was not changed and remains valid.

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

Page 10

Beaver Valley Power Station Unit Nos. 1 & 2 License Amendment Request No. 327 & 197 In addition, the Safety Analysis Limit (SAL) for the LONF event at BVPS Unit No. 2 was changed from 10% to 0%. The BVPS Unit No. 2 SAL values for FLB and SLB remained unchanged at 0%. Now, the SAL for the three events analyzed for SG water level-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 new LONF SAL value of 0%. [Note: The SAL value for the LONF event at BVPS Unit I was not altered since LONE is not the current limiting event for the BVPS Unit No. I SG water level-low-low uncertainty analysis and the BVPS Unit No. 1 steam generators are currently scheduled to be replaced in the 2006 refueling outage, which will require a complete new uncertainty analysis be established in 2006 for the replacement SGs.

Thus, there was limited benefit to now revise station calculations/procedures to address a new LONF SAL for a non-bounding event with only one more cycle of operation left at BVPS Unit No.1 with these original steam generators. The current LONF SAL value of 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 setpoint analyses, the bounding design basis accident uncertainties applicable for the SG water level-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 to increase the nominal low-low level trip setpoint from 15.1% to 20.1% level with an Allowable Value of 19.6% level for BVPS Unit 1. Thus, TS Proposed Change No. 1 consists of revising the BVPS Unit I Technical Specification allowable value for the SG water level-low-low reactor trip system function to 19.6% level and TS Proposed Change No. 2 consists of revising the BVPS Unit No. 1 Technical Specification allowable value for the SG water level-low-low auxiliary feedwater actuation function to 19.6% level. The evaluation also indicated that the nominal low-low level trip setpoint should be increased from 16.5% to 20.5% level with an Allowable Value of 20% level for BVPS Unit 2. Thus, TS Proposed Change No. 3 consists of revising the BVPS Unit 2 Technical Specification allowable value for the SG water level-low-low reactor trip system function to 20% level and TS Proposed Change No. 4 consists of revising the BVPS Unit No. 2 Technical Specification allowable value for the SG water level-low-low auxiliary feedwater actuation function to 20% level.

The steam generator level control will continue to maintain steam generator water level constant at 44% at full power. Therefore, the steady-state full power operating margin to the low-low level reactor trip setpoint (difference between normal water level and the trip setpoint) will decrease from 28.9% to 23.9% with the new proposed BVPS Unit No. 1 Allowable Value and from 27.5% to 23.5% with the new proposed BVPS Unit No. 2 Allowable Value. A steady state margin of 23.9%/23.5% for BVPS Unit Page 11

Beaver Valley Power Station Unit Nos. I & 2 License Amendment Request No. 327 & 197 No. 1/2 is considered adequate based on field experience to avoid unnecessary reactor trips on SG water level-low-low during normal expected transients where a reactor trip is neither wanted or expected. This includes plant startups, shutdowns, steady-state operation and design basis load swings (+/- 10% load change). The other area of concern is the steam generator level during a 50% load rejection transient. The 50% load rejection transient was analyzed for the (future anticipated) BVPS Unit 1 extended power uprate (EPU). This EPU analysis showed that margin to SG 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 setpoint will still provide at least 5% margin. Considering that the EPU analysis is conservative with respect to the current licensed power conditions, a margin of 5% is considered adequate.

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

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

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

Page 12

Beaver Valley Power Station Unit Nos. I & 2 License Amendment Request No. 327 & 197 4.2 Review of BVPS Unit No. 1 Safety Analyses Relating to the Proposed Technical Specification Changes The new proposed SG level-low-low trip and Allowable Value setpoints were evaluated to 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
  • ATWS Mitigation System Actuation Circuitry (AMSAC)

Margin to Trip

  • Radiological Analyses The results of these evaluations are summarized below for BVPS-1. In this section describing 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 events are 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 uncertainty of +/-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. The faulted loop assumes a conservatively high initial SG mass in order to delay the time of reactor trip on SG water level-low-low and the intact loops assume a lower than nominal initial SG mass in order to minimize the SG mass available for post trip heat removal. Based upon sensitivity studies by Westinghouse, the uncertainties applied to the 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 the analysis results. The acceptance criterion applied to a FLB analysis is that the hot legs remain subcooled. The current analysis of record shows significant margin (>300 F) to Page 13

Beaver Valley Power Station Unit Nos. 1 & 2 License Amendment Request No. 327 & 197 boiling in the hot legs. Based on the amount of available margin, the fact that the safety analysis limit is unaffected, and that the initial faulted SG mass is conservative, it is concluded that the new issues addressed by these proposed changes in this LAR would not 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 of Offsite Power to the Station Auxiliaries -14.1.1 1)

The current licensing basis LONF/LOAC analysis for BVPS Unit No. 1 assumes a SG water 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 generator water level. The current licensing basis analysis assumes the correct trip setpoint and an overly conservative initial steam generator mass (nominal + 6% rather than nominal

+3.5%). This results in conservatively late protection via reactor trip and auxiliary feedwater initiation which allows the heat-up to progress longer prior to protective actions occurring. Thus, current licensing basis LONF/LOAC analysis for BVPS-1 remains acceptable.

Other Non-LOCA Events No other non-LOCA event analysis credits the SG water level-low-low trip functions for protection. Thus, this aspect of this issue impacts no other non-LOCA event. The initial SG mass assumed in the non-LOCA analyses is biased high or low depending on the direction of conservatism. If a particular event is not sensitive to the initial SG mass, then the initial mass is set to the nominal mass. Any event that is analyzed with the initial mass set to the nominal mass is not impacted by the new changes on the initial mass. The analysis for any event for which a high initial SG mass is conservative remains conservative because the positive uncertainty has decreased. There are no non-LOCA events for which a low initial SG mass -yields more limiting results than the FLB 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 outside containment typically use SG water level uncertainties for both the initial conditions and the low-low reactor trip setpoint.

Page 14

Beaver Valley Power Station Unit Nos. 1 & 2 License Amendment Request No. 327 & 197 The current licensing-basis analysis of record for the SLB M&Es inside containment is documented using the MARVEL code (UFSAR Section 14D.10.3) based on conservative generic values for the initial SG water mass. Design mass values for initial SG water mass are greater than current safety analysis values and were the standard input to the analysis performed with MARVEL in the 1980s. Since no allowance was made in the MARVEL analysis for water level control, and since conservatively large initial SG wvater mass values were modeled, the revised uncertainties do not alter the validity of the analysis values at any power. The trip setpoint is not assumed in this analysis.

The current licensing-basis analyses of record for the SLB M&Es outside containment are generic M&E releases documented using the LOFTRAN code (UFSAR Section 14D.10.4) as part of a WOG program. The initial conditions related to the SG water mass are generic for the plants in the category for which the M&E releases were calculated for BVPS Unit No. 1. Any increase in the level uncertainty used for those analyses can be absorbed by conservatisms associated with other assumptions included in the analyses. The reactor trip setpoint used in the generic analyses has been evaluated (1999) and determined to be less than 0% for BVPS Unit No. 1. Thus, the trip setpoint is conservative with respect to the increased uncertainties. These uncertainties 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 Analysis 14.3.4.2.1)

The potential effect of the NSAL on the LOCA M&E related analysis is the change to the steam generator fluid mass. The long term LOCA M&E calculations are typically initialized at 100% full power steady state conditions and the initial secondary side fluid mass is biased high. The current licensing basis long term LOCA M&E for Unit 1 was calculated using the LOCTIC code. This is the same code which is used to calculate the containment response. The maximum steam generator liquid mass assumed in this analysis accounts for the 1.4% uprate, maximum steam generator liquid mass corresponding to 100% full power steady state conditions @ 44% NR level, with 10%

uncertainty. The uncertainty used in the current safety analysis of record bounds the initial condition instrumentation uncertainties. Therefore, the current analysis remains bounding.

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

Page 15

Beaver Valley Power Station Unit Nos. I & 2 License Amendment Request No. 327 & 197 The steam generator tube rupture analyses are not affected by the proposed changes since SG level control (initial condition uncertainties) and SG level trips are not modeled 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 since SG level control (initial condition uncertainties) and SG level trips are not modeled in these 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 flow condition. This AMSAC logic is Logic 2 of the generic WOG AMSAC designs provided in WCAP-10858-P-A (Reference 10). Since the AMSAC at Beaver Valley Unit 1 is not actuated on a SG water level-low-low condition, there is no impact on the operation 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 from

15. 1%to 20.1% and the associated Allowable Value is being proposed to be raised from 14.6% to 19.6%. The steam generator level control will continue to maintain steam generator water level constant at 44% at full power. Therefore, the steady-state full power operating margin to the low-low level reactor trip setpoint (difference between normal water level and the trip setpoint) will decrease from 28.9% to 23.9% with the new proposed BVPS Unit No. 1 trip setpoint and Allowable Value. A steady state margin of 23.9% for BVPS Unit No. 1 is considered adequate based on field experience to avoid unnecessary reactor trips on SG water level-low-low during normal expected transients where a reactor trip is neither wanted or expected. This includes plant startups, shutdowns, steady-state operation and design basis load swings (+/- 10% load change).

Page 16

Beaver Valley Power Station Unit Nos. 1 & 2 License Amendment Request No. 327 & 197 The other area of concern is the steam generator level during a 50% load rejection transient. The 50% load rejection transient was analyzed for the (future anticipated)

BVPS Unit 1 extended power uprate (EPU). This EPU analysis showed that margin to SG 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 setpoint will still provide at least 5% margin. Considering that the EPU analysis is conservative with respect to the current licensed power conditions, a margin of 5% is considered adequate.

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 accidents will be within the acceptance criteria of 10 CFR Part 100 and 10 CFR 50.67, as applicable. There were no new radiological safety analyses performed since the previously assumed low-low and high-high actuation values used in the safety analyses for the steam generator water level setpoints were not altered.

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

In summary, the impact of the proposed changes has been evaluated for BVPS Unit No. 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 is sufficient margin in safety analyses to accommodate these impacts and continue to satisfy acceptance criteria.

4.3 Review of BVPS Unit No. 2 Safety Analyses Relating to the Proposed Technical Specification Changes The new proposed SG level-low-low trip and Allowable Value setpoints were evaluated to 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
  • Margin to Trip Page 17

Beaver Valley Power Station Unit Nos. 1 & 2 License Amendment Request No. 327 & 197 0 Radiological Analyses The results of these evaluations are summarized below for BVPS-2. In this section describing 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 water level-high-high trip setpoints. These are the Feedwater Malfunction (FWM), FLB and the LONF/LOAC analyses. These events are discussed in this section.

Feedwater Malfunction:

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

The current licensing basis FWM analyses for BVPS Unit No. 2 assumes a SG water level-high-high safety analysis limit of 100%. Thus, the current analysis remains applicable and includes sufficient conservatism to cover the additional uncertainties that must 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 uncertainty of +/-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. The faulted loop assumes a conservatively high initial SG mass in order to delay the time of reactor trip on SG water level-low-low and the intact loops assume a lower than nominal initial SG mass in order to minimize the SG mass available for post trip heat removal. Based upon sensitivity studies by Westinghouse, the uncertainties applied to the 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 the analysis results. The acceptance criterion applied to a FLB analysis is that the hot legs remain subcooled. The current analysis of record shows significant margin (>30'F) to boiling in the hot legs. Based on the amount of available margin, the fact that the safety analysis limit is unaffected, and that the initial faulted SG mass is conservative, it is concluded that the new issues addressed by these proposed changes in this LAR would not cause any safety limits to be violated.

Page 18

Beaver Valley Power Station Unit Nos. 1 & 2 License Amendment Request No. 327 & 197 Loss of Normal Feedwater/Loss of AC Power (BVPS Unit No. 2 UFSAR Section: Loss of Non-Emergency AC Power to the Plant Auxiliaries -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 SG water 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 generator water level. The current licensing basis analysis assumes the correct trip setpoint and an overly conservative initial steam generator mass (nominal + 6% rather than nominal

+3.5%). This results in conservatively late protection via reactor trip and auxiliary feedwater initiation which allows the heat-up to progress longer prior to protective actions occurring. Thus, current licensing basis LONF/LOAC analysis for BVPS-2 remains acceptable.

Other Non-LOCA Events No other non-LOCA event analysis credits the SG water level-low-low or SG water level-high-high trip functions for protection. Thus, this aspect of this issue impacts no other non-LOCA event. The initial SG mass assumed in the non-LOCA analyses is biased high or low depending on the direction of conservatism. If a particular event is not sensitive to the initial SG mass, then the initial mass is set to the nominal mass. Any event that is analyzed with the initial mass set to the nominal mass is not impacted by the new changes on the initial mass. The analysis for any event for which a high initial SG mass is conservative remains conservative because the positive uncertainty has decreased. There are no non-LOCA events for which a low initial SG mass yields more limiting 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 Postulated Secondary System Pipe Rupture Inside Containment - 6.2.1.4 The analyses for the SLB mass and energy (M&E) releases inside and outside containment typically use SG water level uncertainties for both the initial conditions and the low-low reactor trip setpoint.

The current licensing-basis analysis of record for the SLB M&Es inside containment is documented using the MARVEL code (UFSAR Section 6.2.1.4.4) based on conservative generic values for the initial SG water mass. No allowance was made in Page 19

I.

Beaver Valley Power Station Unit Nos. 1 & 2 License Amendment Request No. 327 & 197 the MARVEL analysis for water level control, and the values used are conservatively large. Thus these revised uncertainties do not alter the validity of the analysis values at any power. The trip setpoint is not assumed in this analysis.

The current licensing-basis analyses of record for the SLB M&Es outside containment are generic M&E releases documented using the LOFTRAN code (UFSAR Section 15.0.1) as part of a WOG program. The initial conditions related to the SG water mass are generic for the plants in the category for which the M&E releases were calculated for BVPS Unit No. 2. Any increase in the level uncertainty used for those analyses can be absorbed by conservatisms associated with other assumptions included in the analyses. 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 setpoint is conservative with respect to the increased uncertainties. These uncertainties do not alter 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 Postulated Loss of Coolant Accidents - 6.2.1.3)

The potential effect of the NSAL on the LOCA M&E related analysis is the change to the steam generator fluid mass. The long term LOCA M&E calculations are typically initialized at I 00% full power steady state conditions and the initial secondary side fluid mass is biased high. The current licensing basis long term LOCA M&E for Unit 1 was calculated using the LOCTIC code. This is the same code which is used to calculate the containment response. The maximum steam generator liquid mass assumed in this analysis accounts for the 1.4% uprate, maximum steam generator liquid mass corresponding to 100% full power steady state conditions ( 44% NR level, with 10%

uncertainty. The uncertainty used in the current safety analysis of record bounds the initial condition instrumentation uncertainties. Therefore, the current analysis remains bounding.

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 changes since SG level control (initial condition uncertainties) and SG level trips are not modeled in these analyses.

Page 20

Beaver Valley Power Station Unit Nos. 1 & 2 License Amendment Request No. 327 & 197 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 since SG level control (initial condition uncertainties) and SG level trips are not modeled in these analyses.

AMSAC (BVPS Unit No. 2 UFSAR Section: Anticipated Transients Without Trip - 4.3.1.7 and 15.8)

A review of Chapter 4.3.1.7 of the BVPS-2 UFSAR shows that the ATWS Mitigation System Actuation Circuitry (AMSAC) logic in place at BVPS-2 is actuated on a low feedwater flow condition. This AMSAC logic is Logic 2 of the generic WOG AMSAC designs provided in WCAP-10858-P-A (Reference 10). Since the AMSAC at BVPS-2 is not actuated on a low-low steam generator water level condition, there is no impact on 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 from 16.5% to 20.5% and the associated Allowable Value is being proposed to be raised from 16% to 20%. The steam generator level control will continue to maintain steam generator water level constant at 44% at full power. Therefore, the steady-state full power operating margin to the low-low level reactor trip setpoint (difference between normal water level and the trip setpoint) wvill decrease from 27.5% to 23.5% with the new proposed BVPS Unit No.2 trip setpoint and Allowable Value. A steady state margin of 23.5% for BVPS Unit No. 2 is considered adequate based on field experience to avoid unnecessary reactor trips on SG water level-low-low during normal expected transients where a reactor trip is neither wanted or expected. This includes plant startups, shutdowns, steady-state operation and design basis load swings (+/- 10% load change).

The other area of concern is the steam generator level during a 50% load rejection transient. The steam generator low low level margin to trip analysis for the 50% load rejection transient was performed for BVPS Unit 1. Because of the differences in the feedwater condensate pump design, BVPS Unit I was more limiting than Unit 2. As Page 21

Beaver Valley Power Station Unit Nos. 1 & 2 License Amendment Request No. 327 & 197 such, a detailed steam generator level response analysis for the 50% load rejection transient was performed for Unit 1. Because Unit 2 has a higher feedwater system flow capacity 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 from 80.6% to 92.2% and the associated Allowable Value is being proposed to be raised from 81.1% to 92.7%. The steam generator level control will continue to maintain steam generator water level constant at 44% at full power. Therefore, the steady-state full power operating margin to the high-high level trip setpoint (difference between normal water level and the trip setpoint) will increase from 36.6% to 48.2% with the new proposed BVPS Unit No. 2 trip setpoint and Allowable Value. This increases the margin to a high level trip, which will provide additional time for operator action to address any unnecessary high level condition before a high-high level condition is reached.

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 accidents will be within the acceptance criteria of 10 CFR Part 100 and 10 CFR 50.67, as applicable. There were no new radiological safety analyses performed since the previously assumed low-low and high-high actuation values used in the safety analyses for the steam generator water level setpoints were not adversely altered.

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

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

5.0 REGULATORY SAFETY ANALYSIS FirstEnergy Nuclear Operating Company (FENOC) requests to amend Operating License DPR-66 for Beaver Valley Power Station (BVPS) Unit No. 1 and License NPF-73 for BVPS Unit No. 2. The proposed amendment would revise Technical Specification 3/4.3.1, "Reactor Trip System Instrumentation" and Technical Specification 3/4.3.2, "Engineered Safety Feature Actuation System Instrumentation" to Page 22

Beaver Valley Power Station Unit Nos. I & 2 License Amendment Request No. 327 & 197 modify steam generator level allowable value setpoints. The proposed changes address recent generic issues involving new steam generator level uncertainty considerations and 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 involved with the proposed amendments by focusing on the three standards set forth in I OCFR50.92, "Issuance of amendment," as discussed below:

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

No. The SG water level-low-low setpoint and allowable value have been revised to address Westinghouse Nuclear Safety Advisory Letter NSAL-03-9 and other considerations on steam generator water level uncertainties. The revised setpoint and allowable value calculations continues to follow the setpoint methodology previously approved for BVPS Unit No.1 and No.2 while addressing newly identified level uncertainty considerations. The proposed changes to the SG water level-low-low Allowable Value for BVPS Unit No. 1 and No. 2 and to the SG water level-high-high Allowable Value for BVPS Unit No. 2 continue maintain the validity of the safety analysis limits used in the safety analyses that credit the actuations based on SG water level.

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

The proposed changes do not alter the accident analyses that credit the SG water level-low-low setpoint actuation or the associated accident acceptance criteria.

Therefore, they do not involve a significant increase in the consequences of an accident previously evaluated.

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

No. The SG water level-low-low setpoint and allowable value have been revised to address Westinghouse Nuclear Safety Advisory Letter NSAL-03-9 and other considerations on steam generator water level uncertainties. Implementation of the proposed setpoint changes have no significant effect on either the configuration of Page 23

Beaver Valley Power Station Unit Nos. 1 & 2 License Amendment Request No. 327 & 197 the plant, or the manner in which the plant is operated. The proposed changes to the SG water level-low-low allowable value for BVPS Unit No. 1 and No. 2 and to the SG water level-high-high allowable value for BVPS Unit No. 2 continue to maintain the validity of the safety analysis limits used in the safety analyses that credit the actuations based on SG water level.

Therefore, since the plant configuration is not adversely changed and the proposed changes do not alter the accident analyses that credit actuation based on SG water level, the proposed change does not create the possibility of a new or different accident 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 System setpoint analysis methodology and acceptance criteria provide the margin of safety.

The SG water level-low-low and SG water level-high-high actuation setpoint and allowable value have been calculated using the same methodology as previously approved for BVPS Unit No. I and No. 2 while addressing newly identified considerations needed to protect the limits used in the safety analyses. The applicable safety analyses have been performed and show acceptable results.

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

Based on the above, FENOC concludes that the proposed amendments present no significant hazards consideration under the standards set forth in 10CFR50.92(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 the proposed changes. General Design Criteria (GDC) 20 is potentially impacted and is assessed with respect to the need for a modification to the UFSAR description of BVPS Unit 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 of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational Page 24

Beaver Valley Power Station Unit Nos. I & 2 License Amendment Request No. 327 & 197 occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.

The proposed changes do not alter any current physical plant components. Only the setpoint and allowable value for steam generator water level are proposed to be modified. The proposed changes continue to maintain the underpinning safety analyses valid. 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 reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to 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 requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10CFR20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10CFR51.22(c)(9). Therefore, pursuant to 10CFR51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 REFERENCES

1. WCAP-1 1419, Revision 2, "Setpoint Methodology for Protection Systems for Beaver Valley Power Station - Unit 1," dated December, 2000.
2. WCAP-1 1366, Revision 4, "Setpoint Methodology for Protection Systems for Beaver 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 Beaver Valley Power Station Unit Nos. 1 & 2 License Amendment Request No. 327 & 197
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 Plate Pressure 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 Level Uncertainties," dated September 22, 2003.
8. Technical Bulletin, TB-04-12, "Steam Generator Level Process Pressure Evaluation,", dated June 23, 2004.
9. Nuclear Safety Advisory Letter, NSAL-02-4, "Maximum Reliable Indicated Steam Generator 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. 1 License 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 MINIMUM TOTAL NO. CHANNELS CHANNELS ALLOWABLE APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE VALUE MODES ACTION

7. Overtemperature AT 3 2 2 See Table 1, 2 7 Notation (A)
8. Overpower AT 3 2 2 See Table 1, 2 7 Notation (B)
9. Pressurizer Pressure-Low 3 2 2 2 1941 psig 1, 2 7 (Above P-7)
10. Pressurizer Pressure-High 3 2 2
  • 2389 psig 1, 2 7
11. Pressurizer Water Level- . 3 2 2
  • 92.5% of 1, 2 7 High (Above P-7) instrument span
12. Loss of Flow - Single Loop 3/loop 2/loop in 2/loop in Ž 89.8% of 1 7 (Above P-8) any each indicated loop operating operating flow loop loop
13. Loss of Flow - Two Loops 3/loop 2/loop in 2 89.8% of 1 7 (Above P-7 and below P-8) two each indicated loop operating loops operating loop flow e4Z
14. Steam Generator Water 3/loop 2/loop 2/loop 2 14.60 of narrow 1, 2 7 I Level-Low-Low range instrument (Loop Stop Valves Open) span-each steam generator BEAVER VALLEY - UNIT 1 3/4 3 -3 Amendment No. 213-4

TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS ALLOWABLE APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE VALUE MODES ACTION

7. AUXILIARY FEEDWATER
a. Steam Gen. Water Level-Low-Low (Loop Stop Valves Open)
i. Start Turbine Driven 3/stm. gen. 2/stm. 2/stm. 2 14.6% of narrow 1, 2, 3 14 I Pump gen. any gen. range ii ment stm. gen. span each stea generator 19.6%

ii. Start Motor Driven 3/Stm. gen. 2/stm. 2/stm. 2 narrow 1, 2, 3 14 I Pumps any 2 stm. gen. any gen. range instrument gen. 2 stm. span each steam gen. generator

b. Undervoltage-RCP (Start (3) -1/bus 2 2 2 71.2% rated RCP 1 14 Turbine Driven Pump) bus voltage
c. S.I. (Start All See 1 above (all S.I. initiating functions and requirements)

Auxiliary Feedwater Pumps)

d. (Deleted)
e. Trip of Main Feedwater 1/pump 1 1 Not Applicable 1, 2, 3 18 Pumps (Start Motor Driven Pumps)

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

Attachment A-2 Beaver Valley Power Station, Unit No. 2 License Amendment Request No. 197 Proposed Technical Specification Changes M - - ---- M -

The following are the affected pages:

3/4 3-3 3/4 4-19 3/4 4-20

TABLE 3.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS ALLOWABLE APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE VALUE MODES ACTION

7. Overtemperature AT 3 2 2 See Table 1, 2 7 Notation (A)
8. Overpower AT 3 2 2 See Table 1, 2 7 Notation (B)
9. Pressurizer Pressure-Low 3 2 2 2 1941 psig** 1, 2 7 (Above P-7)
10. Pressurizer Pressure-High 3 2 2
  • 2379 psig 1, 2 7
11. Pressurizer Water Level- 3 2 2
  • 92.5* of 1, 2 7 High (Above P-7) instrument span
12. Loss of Flow - Single Loop 3/loop 2/loop in 2/loop in 2 89.6* of 1 7 (Above P-8) any each indicated loop operating operating flow loop loop
13. Loss of Flow - Two Loops 3/loop 2/loop in 2/loop 2 89.6% of 1 7 (Above P-7 and below P-8) two each indicated loop operating operating flow loops loop
14. Steam Generator Water 3/loop 2/loop 2/loop 2 446 of narrow 1, 2 7 I Level-Low-Low range instrument span-each steam generator
    • 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 those values.

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

TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS ALLOWABLE APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE VALUE MODES ACTION

5. TURBINE TRIP & FEEDWATER ISOLATION
a. Automatic Actuation 2 1 2 N.A. 1, 2 42 Logic and Actuation Relays
b. Steam Generator Water 3/loop 2/loop 2/loop
  • 81.1- of narrow 1, 2, 3 14 1 Level--High-High, P-14 in any in each range instrument operating operating span loop loop
c. Safety Injection See Item 1 above for all Safety Injection initiating functions and requirements.
6. LOSS OF POWER
a. 4.16kv Emergency Bus
1. Undervoltage 2/4.16kv 2/4.16kv 2/4.16kv 2 71.2% of rated 1, 2, 3, 4 33 (Trip Feed) Bus Bus Bus Bus Voltage with a 1 i 0.1 second time delay
2. Undervoltage 1/4.16kv 1/4.16kv 1/4kv Bus 2 71.2% of rated 1, 2, 3, 4 33 (Start Diesel) Bus Bus Bus Voltage, 20 cycles

+/- 2 cycles

b. 4.16kv Emergency Bus 2/4.16kv 2/Bus 2/Bus 2 93.1% of rated 1, 2, 3, 4 34 (Degraded Voltage) Bus Bus Voltage with a 90 +/- 5 second time delay BEAVER VALLEY - UNIT 2 3 /4 3 -19 Amendment No. a2Go

%I TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS ALLOWABLE APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE VALUE MODES ACTION

6. LOSS OF POWER (Continued)
c. 480 Volt Emergency Bus 2/480v Bus 2/Bus 2/Bus > 93.1% of rated 1, 2, 3, 4 34 (Degraded Voltage) Bus Voltage with a 90 +/- 5 second time delay
7. AUXILIARY FEEDWATER(3)
a. Automatic Actuation 2 1 2 N.A. 1, 2, 3 42 Logic and Actuation Relays
b. Steam Gen. Water Level--

Low-Low

1. Start Turbine 3/stm. gen. 2/stm. 2/stm. 2 of narrow 1, 2, 3 14 Driven Pump gen. any gen. range in~ ment stm. gen. span
2. Start Motor 3/stm. gen. 2/stm. 2/stm. Ž 14-6of narrow 1, 2, 3 14 Driven Pumps gen. any gen. range instrument 2 stm. span gen.
c. Undervoltage-RCP (Start (3)-1/bus 2 2 2 71.2% of rated 1, 2 14 Turbine Driven Pump) bus voltage (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 Beaver Valley Power Station, Unit No. 1 License Amendment Request No. 327 Proposed Licensing Requirement Manual Changes Licensing Requirement Manual changes are provided for information only.

The following are the only affected pages:

3.9-2 3.9-6

BVPS-I LICENSING REQUIREMENTS MANUAL Providedfor Information Only.

TABLE 3.9-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT NOMINAL TRIP SETPOINT

1. Manual Reactor Trip Not Applicable
2. Power Range, Neutron Flux A. High Setpoint 109% of RATED THERMAL POWER B. Low Setpoint 25% of RATED THERMAL POWER
3. Power Range, Neutron Flux, High Positive 5% of RATED THERMAL POWER with a Rate time constant 2 2 seconds
4. Power Range, Neutron Flux, High Negative 5% of RATED THERMAL POWER with a Rate time constant 2 2 seconds
5. Intermediate Range, Neutron Flux 25% of RATED THERMAL POWER
6. Source Range, Neutron Flux A. With Rod Withdrawal Capability 105 counts per second B. With All Rods Fully Inserted and Not Applicable Without Rod Withdrawal Capability
7. Overtemperature AT See Technical Specification Table Notation (A) on Table 3.3-1
8. Overpower AT See Technical Specification Table Notation (B) on Table 3.3-1
9. Pressurizer Pressure-Low 1945 psig
10. Pressurizer Pressure-High 2385 psig I1. Pressurizer Water Level-High 92% of instrument span
12. Loss of Flow A. Single Loop 90.2% of indicated loop flow B. Two Loops 90.2% of indicated loop flow
13. Steam Generator Water Level-Low-Low 5.10%

°of narrow range instrument span-each I

/ steam generator

14. Deleted 3.9-2 Revision 24

BVPS-1_

LICENSING REQUIREMENTS MANUAL Providedfor lInformation Only.

TABLE 3.9-2 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT NOMINAL TRIP SETPOINT

6. LOSS OF POWER
a. 4.16 kv Emergency Bus Undervoltage I. Loss of Voltage (Trip Feed) 75% of rated bus voltage with a 1 +/- 0.1 second time delay
2. Loss of Voltage (Start Diesel) 75% of rated bus voltage with a < 0.9 second time delay (includes auxiliary relay times)
b. 4.16kv Emergency Bus Undervoltage 93.7% of rated bus voltage with a 90 +/- 5 (Degraded Voltage) second time delay
c. 480v Emergency Bus Undervoltage 93.7% of rated bus voltage with a 90 +/- 5 (Degraded Voltage) second time delay
7. AUXILIARY FEEDWATER 20.1%
a. Steam Generator Water Level-Low-Low
i. Start Turbine Driven Pump - of arrow range instrument span each steam nerator I ii. Start Motor Driven Pumps of narrow range instrument span each I steam generator
b. Undervoltage - RCP (Start Turbine Driven 75% rated RCP bus voltage Pump)
c. S.I. (Start All Auxiliary Feedwater Pumps) See 1 above (all SI Setpoints)
d. (Deleted)
e. Trip of Main Feedwater Pumps (Start Not Applicable Motor Driven Pumps)
8. ESF INTERLOCKS
a. Reactor Trip, P-4 Not Applicable
b. Pressurizer Pressure, P-1I 2000 psig
c. Low-Low Tavg, P-12 541 cF 3.9-6 Revision 49

Attachment B-2 Beaver Valley Power Station, Unit No. 2 License Amendment Request No. 197 Proposed Licensing Requirement Manual Changes Licensing Requirement Manual changes are provided for information only.

The following are the only affected pages:

3.10-3 3.10-6

BVPS-2_

LICENSING REQUIREMENTS MANUAL Providedfor InforTL3t0on Only.

TABLE 3.110-1 (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-each I steam generator
14. DELETED.
15. Undervoltage - Reactor Coolant Pumps 75% of rated bus voltage-each bus
16. Underfrequency-Reactor Coolant Pumps 57.5 Hz-each bus
17. Turbine Trip
a. Emergency Trip Header Low Pressure 1000 psig
b. Turbine Stop Valve Closure > 1% open
18. Safety Injection Input from ESF N.A.
19. Reactor Coolant Pump Breaker Position Trip N.A.
20. Reactor Trip Breakers N.A.
21. Automatic Trip Logic N.A.
22. Reactor Trip System Interlocks
a. Intermediate Range Neutron Flux, P-6 I x 10'10 amps
b. Power Range Neutron Flux, P-8 30% of RTP*
c. Power Range Neutron Flux, P-9 49% of RTP*
d. Power Range Neutron Flux, P-b 0 (Input to 10% of RTP*

P-7)

e. Turbine First Stage Pressure, P-13 (Input 10% of RTP* Turbine First Stage Pressure to P-7) Equivalent
  • = RATED THERMAL POWER
      • WVith the exception of Functional Unit number 17.b.

3.10-3 Revision l2

BVPS-2 LICENSING REQUIREMENTS MANUAL Providedfor Information Only.

TABLE 3.10-2 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT NOMINAL TRIP SETPOINT

5. TURBINE TRIP & FEEDWATER ISOLATION
a. Automatic Actuation Logic and Actuation N.A.

Relays

b. Steam Generator Water Level - High- 8A of narrow range instrument span High, P-14
c. Safety Injection See Functional Unit 1. above for all Safety Injection Trip Setpoints.
6. LOSS OF POWER
a. 4.16 kV Emergency Bus
1. Undervoltage 75% of rated Bus Voltage with a (Trip Feed) I +/- 0.1 second time delay
2. Undervoltage 75% of rated Bus Voltage, 20 cycles (Start Diesel) + 2 cycles
b. 4.16 kV Emergency Bus 93.4% of rated Bus Voltage with a (Degraded Voltage) 90 +/- 5 second time delay
c. 480 Volt Emergency Bus 93.4% of rated Bus Voltage with a (Degraded Voltage) 90 +/- 5 second time delay
7. AUXILIARY FEEDWATER*
a. Automatic Actuation Logic and Actuation N.A.

Relays 20.5%

w i

b. Steam Generator Water Level-Low-Low
1. Start Turbine Driven Pump 16.5%ofnarrow range instrument span I
2. Start Motor Driven Pumps 4  % of narrow range instrument span I

I it Attachment C Beaver Valley Power Station, Unit Nos. 1 & 2 License Amendment Request No. 327 & 197 Commitment Summary The following list identifies those actions committed to by FirstEnergy Nuclear Operating Company (FENOC) for Beaver Valley Power Station (BVPS), Unit Nos. I & 2 in this document. Any other actions discussed in the submittal represent intended or planned actions by Beaver Valley. These other actions are described only as information and arc not regulatory commitments. Please notify Mr. Larry R. Freeland, Manager, Regulatory Compliance, at Beaver Valley on (724) 682-5284 of any questions regarding this document or associated regulatory commitments.

Commitment Due Date None N/A

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