BVY 05-101, Proposed Technical Specification Change No. 263 - Supplement No. 41 Re Extended Power Uprate - Condensate and Feedwater Transient Response Engineering Analysis

Proposed Technical Specification Change No. 263 - Supplement No. 41 Re Extended Power Uprate - Condensate and Feedwater Transient Response Engineering Analysis

ML053120378
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	11/02/2005
From:	Thayer J Entergy Nuclear Northeast, Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
%dam200601, BVY 05-101, TAC MC0761
Download: ML053120378 (16)

Text

Entergy Nuclear Northeast Entergy Nuclear Operations, Inc.

Vermont Yankee P.O. Box 0500 185 Old Ferry Road v-~ lagsergv Brattleboro, VT 05302-0500 Tel 802 257 5271 November 2, 2005 Docket No. 50-271 BVY 05-101 TAC No. MC0761 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Subject:

Vermont Yankee Nuclear Power Station Technical Specification Proposed Change No. 263 - Supplement No.41 Extended Power Uprate - Condensate and Feedwater Transient Response

References:

1) Entergy letter to U.S. Nuclear Regulatory Commission, uVermont Yankee Nuclear Power Station, License No. DPR-28 (Docket No. 50-271), Technical Specification Proposed Change No. 263, Extended Power Uprate," BVY 03-80, September 10, 2003

2) Entergy letter to U.S. Nuclear Regulatory Commission, uVermont Yankee Nuclear Power Station, Technical Specification Proposed Change No. 263 - Supplement No. 36, Extended Power Uprate -

Response to NRC's Letter re:

License Conditions," BVY 05-096, October 17, 2005

3) Entergy letter to U.S. Nuclear Regulatory Commission, "Vermont Yankee Nuclear Power Station, Technical Specification Proposed Change No. 263 - Supplement No. 30, Extended Power Uprate -

Response to Request for Additional Information," BVY 05-072, August 1,2005 This letter provides additional information regarding the application by Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc. (Entergy) for a license amendment (Reference 1) to increase the maximum authorized power level of the Vermont Yankee Nuclear Power Station (VYNPS) from 1593 megawatts thermal (MWt) to 1912 MWt.

In Reference 2, Entergy committed to a license condition requiring transient testing of the condensate and feedwater system (CFS) at nominal full extended power uprate (EPU) reactor power as part of EPU power ascension testing. The testing will be designed to demonstrate that adequate margin exists for events involving the loss of a condensate pump (CP) to preclude transients that unnecessarily challenge plant safety systems. Following the transient testing involving the loss of one CP, Entergy will confirm, through the performance of additional transient testing and/or analysis, that the loss of one reactor feedwater pump (RFP) will not result in a reactor trip. Data gathered during the test will be used to validate and update (as necessary) analytical models and to verify the accuracy and conservatism of analytical AIuo

BVY 05-1 01 Docket No. 50-271 Page 2 of 3 assumptions used for predicting plant response to the loss of CFS flow events.

The CFS is designed to provide sufficient feedwater to maintain the reactor vessel level within a predetermined range during all modes of power operation. As described in previous submittals in support of the EPU license amendment, all three motor-driven RFPs and all three CPs will be required to satisfy feedwater demands under EPU conditions. Because of this change in plant operation for EPU, Entergy installed a reactor recirculation (RR) system runback modification to avoid a reactor trip upon the loss of a CP or RFP. In addition, to provide additional margin to tripping all RFPs upon loss of a CP, the RFP suction pressure trip setpoint was lowered (while still maintaining adequate NPSH for the RFPs), and a RFP sequential trip logic was installed to preclude simultaneous tripping of RFPs upon low suction pressure.

These measures are designed to preclude challenges to reactor safety systems resulting from the loss of a CFS pump, which occurs infrequently.

In preparation for the CFS power ascension test, an engineering review identified an input assumption in the calculation of RFP suction pressure that is non-conservative.

The input assumption in question resulted in an additional reduction in the calculated RFP suction pressure following a CP trip (previously calculated to be -124 psig as stated in Reference 3). In light of the reduction of margin and potential for loss of feedwater, Entergy is in the process of implementing a modification which will automatically trip a RFP upon loss of a CP.

This modification, which will be completed during the current refueling outage, will result-in a calculated RFP low suction pressure of -162 psig upon loss of a CP, thus increasing the margin to loss of all RFPs due to low suction pressure. Details of the proposed modification are included in Attachment 1 to this letter.

Because of revised analyses and the additional modification to the CFS trip circuitry, some of the results presented in Attachment 1 update information previously provided to the NRC staff. The net result of these changes is an increase in the margin to the low suction pressure RFP trip setting, thus preserving continued operation of the CFS and avoiding challenges to reactor safety systems.

Additional actions to evaluate the extent of condition relating the non-conservative input assumption in the referenced calculation are discussed in Attachment 2. No other instances of non-conservative assumptions in EPU calculations were identified.

There is one new regulatory commitment contained in this submittal. Attachment 3 provides a summary of the commitment.

This supplement to the license amendment request provides additional information to clarify Entergy's application for a license amendment and does not change the scope or conclusions in the original application, nor does it change Entergy's determination of no significant hazards consideration.

Entergy stands ready to support the NRC staff's review of this submittal and suggests telecons or meetings, as necessary, to resolve any remaining issues. If you have any questions or require additional information, please contact Mr. James DeVincentis at (802) 258-4236.

BVY 05-101 Docket No. 50-271 Page 3 of 3 I declare under penalty of perjury that the foregoing is true and correct.

Executed on November 2, 2005.

Sincerely, V~y8

)ayer e Vice President Vermont Yankee Nuclear Power Station Attachments (3) cc:

Mr. Samuel J. Collins (w/o attachments)

Regional Administrator, Region 1 U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406-1415 Mr. Richard B. Ennis, Project Manager Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 0 8 B1 Washington, DC 20555 USNRC Resident Inspector (w/o attachments)

Entergy Nuclear Vermont Yankee, LLC P.O. Box 157 Vernon, Vermont 05354 Mr. David O'Brien, Commissioner VT Department of Public Service 112 State Street - Drawer 20 Montpelier, Vermont 05620-2601

BVY 05-1 01 Docket No. 50-271 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 263 - Supplement No. 41 Extended Power Uprate - Condensate and Feedwater Transient Response Engineering Analysis Total number of pages in Attachment 1 (excluding this cover sheet) is 5.

- to BVY 05-101 Docket No. 50-271 Page 1 of 5 ENGINEERING ANALYSIS CONDENSATE AND FEEDWATER TRANSIENT RESPONSE BACKGROUND The capability of the condensate and feedwater system (CFS) at extended power uprate (EPU) conditions was evaluated and described in References 1 through 6. This document updates those submittals based on revised analyses and a new plant modification designed to increase reactor feedwater pump (RFP) low suction pressure margin following the trip of a condensate pump (CP).

A steady state CFS hydraulic analysis is used to predict minimum RFP suction pressure following a CP trip. The original EPU analysis of this event held reactor pressure constant at 1020 psig. A revised analysis has been performed to account for the reactor pressure drop that results from a reactor recirculation (RR) runback upon a CP trip. A modification is being implemented to increase (suction pressure) margin and assure RFP availability following a CP trip. This modification (being implemented during the current refueling outage-RFO-25) will automatically trip the "B" RFP upon the trip of any CP when operating at EPU conditions. In support of this modification the dynamic analysis of a CP pump trip was updated to confirm that no appreciable impact on margins to the reactor low level scram setpoint would result.

The following discussion describes the updated analyses performed and the modification being implemented during RFO-25.

STEADY STATE CFS HYDRAULIC ANALYSIS The steady state analysis was updated to revise the reactor pressure assumptions for CP trip cases. Additional tuning of the model to plant data was also performed with the objective of achieving a conservative prediction of RFP suction pressure for CP trip events. The CP trip model change is summarized as follows:

• The CP characteristic curve model was adjusted for better agreement with recent plant data.

The low pressure portion of the condensate system model from the CPs to the RFPs was benchmarked and found to be in good agreement with plant data.

• The feedwater sparger model was tuned to achieve improved agreement with plant data.

Initial conditions assumed in the analysis prior to the CP trip reflect normal steady state 120%

EPU conditions as follows:

3 CPs / 3 RFPs operating Condensate demineralizers at maximum differential pressure Reactor pressure at 1005 psig

... to BVY 05-101 Docket No. 50-271 Page 2 of 5 Reactor at 1912 MWt power Feedwater flow at 7.874 Mlb/hr Feedwater regulating valves (FRVs) at approximately 68% open RFP suction pressure at approximately 210 psig CP TRIP LOSS OF FEEDWATER AVOIDANCE Three different steady state conditions were evaluated to simulate the progression of a CP trip scenario:

Beginning of Transient - Reactor pressure at nominal 1005 psig, a CP trip and RFP trip result in 2 CPs and 2 RFPs operating. FRVs are initially in their pre-trip position. The RR runback is initiated.

End of Runback -

Reactor pressure at approximately 960 psig.

FRV position is assumed to be full open, which provides a bounding condition relative to low pressure.

Reactor Scram - Although the design intent is to preclude a reactor scram following a trip of CP, this case was evaluated because a reactor scram would result in -lower reactor pressure. It may also lower RFP suction pressure margin should the FRVs not close fast enough.

The results of the updated analysis which utilizes more realistic reactor pressure and CP trip model conditions, as described above, and incorporates the 'B" RFP trip modification are summarized in Table 1.

Table 1 CP Trip Analysis Conditions FRV Position RFP Suction Pressure Feedwater Flow 2 CP/2 RFP Condition

% Open psig / (margin to trip-psig)

Mlb/hr Time=0 CP Trip 68%

198 (98) 6.34- -

Reactor Pressure = 1005 psig RR Runback Complete, 100%

Reactor Pressure = 960 (bounding 162 (62) 6.69 psig.

assumption)

Post Reactor Scram 100%

@ 850 psig minimum (bounding 111 (11) 7.16 Reactor Pressure assumption)

The original analytical results were provided to the NRC in the response to request for additional information (RAI) SPLB-A-17 (Ref. 2) and concluded that the minimum calculated RFP suction pressure following the trip of one CP at EPU would be approximately 124 psig. Based on the to BVY 05-1 01 Docket No. 50-271 Page 3 of 5 updated analysis described above, and crediting the UB" RFP trip modification described below, the minimum calculated RFP suction pressure following the trip of a CP at EPU will be approximately 162 psig. As in the original analysis, the revised analysis assumes that a reactor scram does not occur. Therefore, with the new modification, the RFP suction pressure margin increases compared to the results reported earlier. Also, the new modification has sufficient margin, such that even in the event of a reactor scram (with a full open FRV) following a CP trip event, adequate margin to the RFP suction pressure trip setpoints remains and the HPCI and RCIC systems will not be unnecessarily challenged.

"B" RFP TRIP MODIFICATION As demonstrated by the updated steady state hydraulic analysis described above, implementing a trip of the "B" RFP on a CP trip provides a positive method to provide additional margin to preclude inadequate RFP suction pressure and to preserve feedwater flow.

The dynamic pump trip analysis, originally described in response to RAI SPLB-A-18 (Ref. 3),

has been updated to confirm that the new modification will adequately protect against the risk of a low water level scram following a CP trip that will now involve an automatic RFP trip in addition to a RR runback. This analysis is further described below.

Trip logic is being added to directly trip the 'B' RFP upon the trip of any CP, provided all RFP pump motor breakers are closed (i.e., all RFPs are operating). A keylock bypass switch will be installed on the 'B' RFP breaker cubicle door to bypass this feature. The bypass switch is required for infrequent conditions, such as when operators need to swap RFPs at low power and only two CPs and two RFPs are operating. During the RFP swap evolution, three RFPs will be in-service momentarily, thus satisfying the trip permissive, and without the bypass, the 'B' RFP would trip since only two CPs are running.

The 'A' and 'B' RFPs are powered by electrical Bus 1, and the 'C' RFP is powered by Bus 2. A trip of the 'B' RFP results in one RFP remaining on each Bus. Advantage was taken of this design feature by designating the 'B' RFP as the RFP to trip on any CP trip.

PUMP TRIP DYNAMIC ANALYSIS-LOW LEVEL SCRAM AVOIDANCE The CP / FWP trip dynamic analysis has been updated to demonstrate that the new 'B' RFP trip logic remains acceptable from a low reactor vessel (RV) water level scram avoidance perspective. The original analysis used a maximum FW flow (after FW or CP pump trip) of 7.2 Mlb/hr. Based on Table 1 above, a feedwater flow from approximately 6.3 to 7.2 Mlbm/hr is calculated when accounting for the trip of the "B" RFP following a CP trip.

The original analysis, as described in response to RAI SPLB-A-18, concluded that a minimum level margin of 16.2 inches to the low RV water level scram setpoint was calculated. In the updated analysis, a bounding case was evaluated to determine the effect of tripping a RFP upon a CP trip event. This evaluation takes no credit for the response of the feedwater level control system by assuming the flow drops to 5.53 Mlb/hr in 0.5 seconds and remains there throughout the transient. The assumed 5.53 Mlbm/hr feedwater flow is significantly less than the minimum calculated flow with 2 CPs and 2 RFPs operating (6.34 Mlb/hr), as described in Table 1. Even in this bounding case, the low RV water level setpoint is avoided.

to BVY 05-1 01 Docket No. 50-271 Page 4 of 5 The minimum calculated flow (6.34 Mlb/hr) was determined assuming the CP and RFP had tripped, the feedwater regulating valve had not yet responded by opening, and the reactor pressure was still at 1005 psig due to the fact that the RR runback had not yet begun. This minimum calculated flow is assumed to be conservative because the feedwater regulating valve will begin to open after the pumps trip due to the sensed level error, as well as the feedwater flow, steam flow mismatch, if operating in three element level control. Additionally, conservatism exists in the fact that the RR runback will begin to reduce reactor power, and therefore pressure, allowing the remaining pumps to provide more flow to the vessel.

The results of the low RV water level scram avoidance analysis provides confidence that the new modification, to institute an automatic trip of the UB3 RFP on a CP trip event, while increasing the RFP suction pressure margin, adequately protects against the risk of a low water level scram.

PUMP TRIP DYNAMIC ANALYSIS-HIGH LEVEL TURBINE TRIP/IFEEDWATER PUMP TRIP AVOIDANCE A CP trip event results in a RV level transient due to the initial drop in feedwater flow followed by the rapid power reduction (caused by the automatic reactor recirculation runback) and associated drop in steam flow. If reactor water level increases to the high water level trip, a main turbine trip occurs followed by tripping of the feedwater pumps.

A reactor scram is initiated as a consequence of the turbine trip. A high level trip signal is to be avoided because of the resulting loss (at least temporarily) of all feedwater flow. The Vermont Yankee Nuclear Power Station (VYNPS) has three motor-driven RFPs. To supply subsequent reactor coolant makeup following a high level trip, plant operators would normally (and rather easily) restart one RFP, rather than using the high pressure coolant injection system or the reactor core isolation cooling system for makeup.

As described above, the updated design and analysis provide confidence that adequate RFP suction pressure will be available on a CP trip to preclude loss of all RFPs and that the feedwater flow available in conjunction with the automatic RR runback will preclude a reactor scram on low reactor water level. To provide assurance of high level trip avoidance, the dynamic model used for reactor water level transients will be further benchmarked to plant data in order to improve the dynamic model's level response compared to the as-built plant.

Optimization of the runback and level controller parameters, as well as associated operator actions, may be made to further enhance system performance and to ensure with high confidence that a high water level trip can be avoided following a CP and/or RFP trip at EPU conditions.

ADDITIONAL ACTIONS As discussed above, additional high level scram avoidance analysis is planned with improved benchmarking to actual plant data.

Based on further analysis, improvements in the scram avoidance strategy may be implemented. Changes to the reactor recirculation runback scheme and level controller settings may also be implemented to further reduce the risk of a reactor scram following a CP and/or RFP trip.

Future changes will be implemented under the provisions of 10 CFR 50.59.

to BVY 05-101 Docket No. 50-271 Page 5 of 5 Operator actions in the event of a CP or RFP trip may also be enhanced based on the additional high level scram avoidance analysis.

REFERENCES

1. VYNPS Power Uprate Safety Analysis Report, Section 7.4, Attachments 4 (proprietary information) and 6 (non-proprietary information) to Entergy letter to U.S. Nuclear Regulatory Commission, uVermont Yankee Nuclear Power Station, License No. DPR-28 (Docket No. 50-271), Technical Specification Proposed Change No. 263, Extended Power Uprate," BVY 03-80, September 10, 2003

2. Response to RAI SPLB-A-17 (license amendment request PC-263, Supplement 28), to Entergy letter to NRC BVY 05-046, April 22, 2005

3. Response to RAI SPLB-A-18 (license amendment request PC-263, Supplement 28), to Entergy letter to NRC BVY 05-046, April 22, 2005

4. Response to RAI SPLB-A-27 (license amendment request PC-263, Supplement 30), to Entergy letter to NRC BVY 05-072, August 1, 2005

5. Response to RAI SPLB-A-28 (license amendment request PC-263, Supplement 30), to Entergy letter to NRC BVY 05-072, August 1, 2005

6. Response to RAI SPLB-A-30 (license amendment request PC-263, Supplement 32), to Entergy letter to NRC BVY 05-083, September 10, 2005

7. Entergy letter to U.S. Nuclear Regulatory Commission, Vermont Yankee Nuclear Power Station, Technical Specification Proposed Change No. 263, Supplement No. 36, Extended Power Uprate - Response to NRC's Letter re: License Conditions," BVY 05-096, October 17, 2005

BVY 05-1 01 Docket No. 50-271 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 263 - Supplement No. 41 Extended Power Uprate - Condensate and Feedwater Transient Response Extent of Condition l Total number of pages in Attachment 2 1 (excludina this cover sheet) is 4.

l to BVY 05-101 Docket No. 50-271 Page 1 of 4 EXTENT OF CONDITION BACKGROUND Vermont Yankee Nuclear Power Station (VYNPS) calculation VYC-2269 RO was prepared by Stone & Webster and reviewed and accepted by Entergy as part of the VYNPS extended power uprate (EPU) balance of plant (BOP) evaluation in August 2003. The objectives of the calculation were to evaluate the capability of the feedwater system to maintain adequate reactor makeup in the event of a condensate pump (CP) trip at EPU conditions and to provide an estimate of reactor feedwater pump (RFP) suction pressure in the event of a CP trip. A reactor pressure of 1020 psig was used as design input in VYC-2269 RO for all cases evaluated. The 1020 psig value was considered a maximum normal operating reactor pressure and appropriate for most of the condensate and feedwater performance cases contained in VYC-2269. This reactor pressure value was obtained from the VYNPS transient analysis input document (OPL-3). This value was considered as an upper bound, and therefore conservative, for the calculation cases that evaluated flow capacity of the system. Two of the analysis cases (G1 and G2), however, were performed to determine the minimum reactor feedwater pump (RFP) suction pressure and maximum CP runout following a CP trip. These cases were intended for use as input to a new low suction pressure RFP trip necessary for operation at EPU conditions and installed during RFO-24. This modification was designed to protect the feedpumps from damage due to loss of available NPSH in the event of a CP trip at EPU conditions. EPU operation will necessitate running three CPs and 3 RFPs to meet rated feed flow requirements. Entergy subsequently performed a Technical Evaluation (TE) that determined the RFP low suction pressure trip setpoint.

VYC-2269 was used as input to the TE for determining the minimum expected RFP suction pressure of 124 psig.

VYC-2269 cases G1 and G2 should have used a minimum, rather than maximum, expected reactor pressure for the CP trip pressure analysis. The lower reactor pressure would be more appropriate and conservative for cases G1 and G2.

Another EPU-related modification consisting of a reactor recirculation runback was installed to ensure that the reactor would not scram on low vessel level in the event of a RFP trip at EPU conditions. The runback feature is armed above a rated power level where it is expected that VYNPS will operate with three RFPs and CPs. GE performed the analysis of this transient event to confirm that the runback would function satisfactorily. This dynamic analysis was completed after approval of VYC-2269 and concluded that a reactor pressure drop of approximately-50 psi would occur as a result of the runback. These results were not discussed among cognizant Entergy, Stone & Webster (S&W) and General Electric (GE) engineers relative to VYC-2269 inputs nor were the results factored back into VYC-2269 or into the RFP suction trip setpoint determination.

The cognizant Entergy and S&W engineers should have applied results of the GE dynamic RFP trip analysis to validate the VYC-2269 reactor pressure input assumption prior to development of the RFP low suction pressure trip design change finalization.

VYC-2269 had an open item to perform further evaluation to determine the appropriate RFP suction pressure trip setpoint.

The conclusion section of the calculation states that a detailed setpoint evaluation based upon additional testing of the condensate pump and/or margin analysis to BVY 05-1 01 Docket No. 50-271 Page 2 of 4 that accounts for uncertainties and operating conditions including reactor dome pressure variations should be performed separately to determine the actual revised setpoint value. The TE that determined the low pressure suction trip setpoint served as the detailed evaluation. The Entergy engineers involved in preparation and review of the TE did not re-assess the calculation input assumptions thereby missing an opportunity to find the inappropriate reactor pressure value.

EXTENT OF CONDITION ASSESSMENT Entergy conducted an extent of condition assessment to determine whether there were other analyses or evaluations conducted in support of the VYNPS EPU project that used non-conservative input assumptions. The apparent cause of this condition was lack of self checking to confirm that use of an input assumption for a new or different application was appropriate.

The actions taken to assess extent of condition included:

1. Review of all input assumptions in the subject calculation to ensure conservatism.

2. Review corrective action and quality assurance documents related to power uprate audits, technical reviews and self assessments to identify other instances of non-conservative assumptions used in EPU analyses.

3. Review approximately 12% of EPU calculations that applied new input assumptions or used an existing input assumption for the first time in a different application.

4. Review Entergy Vermont Yankee corrective action program for similar conditions previously identified in EPU calculations. No previous conditions were identified.

GE EVALUATION REVIEWS GE's constant pressure power uprate (CPPU) project process was approved by the NRC staff in July 2003. The CPPU process consists of evaluations and analyses of plant systems, structures and components (SSCs) and safety evaluations. Many of the CPPU evaluations have been applied to other BWR EPUs tinder previous processes. GE has a well-founded understanding of the appropriate input assumptions required for CPPU evaluations. Every GE evaluation included supplying a Design Input Request (DIR) to Entergy requesting specific evaluation input information. Entergy provided the requested information and discussed its application with GE evaluation performers. This ensured that the information provided was appropriate for the intended evaluation application.

During the course of the VYNPS EPU project, Entergy conducted three technical reviews of GE Nuclear Energy power uprate-related design record files. These reviews occurred in May 2003, October 2003 and June 2004. Entergy also conducted a vendor surveillance through the Entergy Quality Assurance program. The reviews included assessment of input assumptions, methodologies and results in the GE design record files for the following evaluation tasks:

• Thermal Hydraulic Stability a RPV Structural Analysis to BVY 05-1 01 Docket No. 50-271 Page 3 of 4 RPV Internals Structural Analysis RPV Internals Differential Pressures and Fuel Lift RPV Flow Induced Vibration Steam Dryer Performance Containment Analysis ECCS-LOCA SAFERIGESTR

• Technical Specifications Instrument Setpoints Transients

• Anticipated Transients Without Scram These reviews did not reveal instances where non-conservative input assumptions had been applied.

ENTERGY CALCULATION REVIEWS Entergy performed multiple evaluations in support of the power uprate license amendment request and/or power uprate plant modifications. These evaluations were conducted in accordance with the calculation or engineering evaluation process in effect at the time. These calculations were verified in accordance with the VYNPS Appendix B program.

As a result of the RFP trip non-conservative input assumption condition, Entergy conducted an additional review of seven EPU calculations to determine the potential for similar conditions where a new input assumption was applied or an input assumption was used for the first time in a different application. The results of this review did not identify any instances where non-conservative input assumptions had been applied.

Entergy and its vendors performed numerous calculations and evaluations in support of the VYNPS EPU project. These calculations and evaluations are included in the VYNPS EPU design change package generated to facilitate implementation of EPU following issuance of the license amendment. The design change package is currently under review and includes engineering design verification of information contained in each of the EPU evaluation tasks.

This constitutes a comprehensive review of EPU evaluations and analyses and has not resulted in identification of non-conservative input assumptions similar to that associated with the RFP trip calculation.

STONE & WEBSTER CALCULATION REVIEWS S&W has initiated Corrective Action Report (CAR) 05-VY-001 to investigate and report on this condition. The results of the CAR will be entered into Entergy's corrective action program, as appropriate.

to BVY 05-101 Docket No. 50-271 Page 4 of 4 A review was performed to determine the extent of condition associated with this non-conservative input and to evaluate whether this was an isolated case or a programmatic problem. Four S&W calculations prepared in support of the VYNPS power uprate and one Condensate and Feedwater calculation associated with another plant were evaluated for extent of condition. In addition, recent S&W CARs, NUPIC Audit Reports, and Internal Audit Reports were reviewed for similar non-conservative design input findings. Based on the documents reviewed no inappropriate design inputs that would impact the conclusions of VYNPS calculations were identified and there was no evidence that a programmatic problem exists.

However S&W did identify areas for improvement and they have been entered into S&W's corrective action system.

BVY 05-1 01 Docket No. 50-271 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 263 - Supplement No. 41 Extended Power Uprate - Condensate and Feedwater Transient Response Regulatory Commitment Total number of pages in Attachment 3 (excludina this cover sheet) is 1.

I

. to BVY 05-101 Docket No. 50-271 Page 1 of 1 Licensee Identified Commitment Form This form identifies actions discussed in this letter for which Entergy Nuclear Operations, Inc.

(Entergy) commits to perform for the Vermont Yankee Nuclear Power Station (VYNPS). Any other actions discussed in this submittal are described for the NRC's information and are not commitments.

TYPE (Check one)

C M SCHEDULED CMIETE COMPLETION a:

t9FE DATE (if Required) 0 Implement a plant modification which will automatically trip a reactor feedwater pump upon X

RFO-25 loss of a condensate pump at EPU conditions.

(fall 2005)