ML20211H413

Informs That Util Revised Encl Bases for TS 2.2.1, Reactor Trip Sys Instrumentation Setpoints, Discussing Steam Flow/ Feed Flow Mismatch Portion of Steam Flow/Feed Flow Mismatch & Low SG Water Level Reactor Trip Setpoint

ML20211H413
Person / Time
Site:	North Anna
Issue date:	08/27/1999
From:	Hartz L VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20211H417	List: ML20211H413 ML20211H431
References
99-447, NUDOCS 9909020084
Download: ML20211H413 (8)

Text

• VIRGINIA EimcTRIC ann POWER COMPANY Riemioso, You;iNir 2.us!

August 27, 1999 U.S. Nuclear Regulatory Commission Serial No.99-447 Attention: Document Control Desk NL&OS/GSS/ETS R0 Washington, D.C. 20555 Docket Nos. 50-338/339 License Nos. NPF-4/7 Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNITS 1 AND 2 TECHNICAL SPECIFICATION BASES CHANGE 1 STEAM FLOW / FEED FLOW MISMATCH SETPOINT i Virginia Electric and Power Company has revised the Bases for Technical Specifications 2.2.1," Reactor Trip System instrumentation Setpoints." Changes update l the Bases section to discuss the Steam Flow / Feed Flow Mismatch portion of the i

" Steam Flow / Feed Flow Mismatch and Low Steam Generator Water Level" reactor trip setpoint. The Bases discussion previously included a steam flow / feed flow mismatch value in pounds-mass per hour. The value has been revised to include steam flow as a ,

percentage of the rated thermal power which is consistent with the limiting safety system setting required by Technical Specification 2.2.1. We are providing these Technical Specification Bases changes for your information.

The Technical Specifications Bases changes have been reviewed and approved by the Station Nuclear Safety and Operating Committee and the Management Safety Review Committee. It has been determined that these changes do not involve an unreviewed safety question as defined in 10 CFR 50.59. A discussion and the Technical Specifications Bases changes are provided in Attachments 1 and 2 respectively.

There are no commitments made in this letter. If you have any further questions, please contact us.

Very truly yours,

\

020021 l Leslie N. Hartz Vice President - Nuclear Engineering and Services Attachments O 3

9909020004 990827 PDR ADOCK 05000338 P PDR

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1 cc: , U.S. Nuclear Regulatory Commission L Region ll l Atlanta Federal Center 61 Forsyth Street, SW Suite 23T85 Atlanta, Georgia 30303 Mr. M. J. Morgan NRC Senior Resident inspector North Anna Power Station Commissioner Department of Radiological Health Room 104A 1500 East Main Street Richmond, Virginia 23219 -

Mr. J. E. Reasor Old Dominion Electric Cooperative Innsbrook Corporate Center 4201 Dominion Blvd.

Glen Allen, Virginia 23060 l

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4 Attachment 1 Discussion of Bases Change I

North Anna Power Station Units 1 and 2 Virginia Electric and Power Company

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DISCUSSION OF CHANGE 1

l Introduction The Steam Flow - Feed Flow Mismatch Reactor Trip Setpoint specified in the I North Anna Unit 1 and Unit 2 Technical Specifications Bases for Section 2.2-1 is incorrect. Specifically, the flow value of 1.616

• 10 6lbs/ hour represents 40% of l the pre-core uprate design flow for RATED THERMAL POWER conditions (i.e., l 4.04

• 10e Ibs/ hour).

For consistency between units and to avoid future bases changes, the Technical l Specification Bases Section for the Unit 1 and Unit 2 Steam - Feedwater Flow l Mismatch Reactor Trip are being changed so that the Trip Setpoint is expressed in terms of a percentage of nominal steam flow (i.e., Flownom) at RATED l THERMAL POWER instead of a specific flow value given in terms of Ibs/ hour.

Background

i in 1998, two precision feedwater flow tests (one per unit) were performed at North Anna in order to quantify the accuracy of the secondary calorimetric using both feedwater flow and steam flow. The results of the flow tests determined that measured feedwater flow was very accurate when compared to the reference flow instrumentation. In addition, it was also determined that the measured steam flow was reading approximately 2.0% higher than feedwater flow and thus, the units were not operating at their RATED THERMAL POWER. Subsequent to the flow tests, the North Anna secondary calorimetric was revised so that it would be based on feedwater flow. As part of the design change package (DCP) used to implement this change to the secondary calorimetric program, the impact of increasi g power by approximately 2.0% on the Westinghouse 7300 Protection and Control System was assessed.

This assessment determined that the increased power level would not adversely affect the 7300 Protection and Control System and that no immediate instrument scaling changes were necessary. It was demonstrated that the Reactor Protection System would satisfy the applicable Safety Analysis assumptions and Technical Specifications requirements. However, during the review it was determined that it would be advantageous to normalize the steam flow portion of the 7300 Protection System to " Reference Feedwater Flow" as determined by the FLOWCALC portion of the secondary calorimetric computer program. This normalization process enhances the accuracy of the Reactor Trips, ESFAS Actuations, and indications developed from this parameter.

In addition to normalizing steam flow to feedwater flow, it was also decided that i the Steam Flow - Feed Flow Mismatch Reactor Trip Setpoint would be changed Page 1 of 5

y from 34% of Flownom to 40% of Flownom consistent with the limit stated in Technical Specifications, Table 2.2-1. In the mid 1980's, the Steam Flow - Feed Flow Mismatch Reactor Trip Setpoint was changed from the nominal value of 40% of Flownom to 34% of Flownom to account for feedwater flow fouling. Since the performance of the recent flow tests, feedwater flow has been proven to be accurate and thus it has been decided to change the Steam Flow - Feed Flow l Mismatch Reactor Trip Setpoint back to the nominal value of 40% of Flownom.

During the preparation of the DCP, to implement the scaling changes discussed above, it was discovered that the setpoint value in the Technical Specification Bases Section for the Steam Flow - Feed Flow Mismatch Reactor Trip was i incorrect. I Discussion Since the mid-1980s, the nominal 100% power flowrate for North Anna Units 1 and 2 has changed two times. The nominal flowrate for pre-core uprate was 4.04

• 10e !bs/ hour. After the core uprate in 1986, the nominal flowrate for 100%

power conditions increased'to 4.26

• 108 lbs/ hour, in 1993 after the Steam Generator Replacement Project (SGRP), the nominal flowrate changed again and decreased to 4.247

• 108lbs/ hour. During this time, the Steam Flow - Feed j Flow Mismatch Reactor Trip Setpoint value given in the Bases Section remained i 8 8 l at 31.616 + 10 lbs/ hour for Unit 1 and > 1.616

• 10 lbs/ hour for Unit 2.

Further, as stated above, the actual setpoint installed in the plant for Unit 2 is 34% of Flownom which equates to 1.444

• 108 lbs/ hour (Post -SGRP). For Unit 1, the Steam Flow - Feed Flow Mismatch Reactor Trip Setpoint is set at 34% of the Pre-SGRP Flownom value which equates to 1.448

• 108 lbs/ hour. As can be seen, keeping the Bases up to date ucing an actual flow value given in terms of Ibs/ hour would have required multiple changes to the Bases Section over time.

The Trip Setpoint value given in Technical Specifications, Table 2.2-1, item 14,

" Steam Flow - Feed Flow Mismatch Reactor Trip" is 5 40% of full steam flow at RATED THERMAL POWER. This setpoint permits changes in the nominal flowrate and thus changes to the actual trip setpoint. Further, the setpoint in the table permits (he actual trip setpoint to be set more conservatively if required.

The Steam Flow - Feed Flow Mismatch Reactor Trip coincident with Steam Generator low Level is considered to be a backup reactor trip and is not credited in the Chapter 15 Safety Analysis. However, according to the North Anna UFSAR, Section 7.2.2.3.5 this reactor trip function is credited as a backup to the Steam Generator Low-Low Level Reactor Trip in the IEEE 279-1971 scenario for the interaction between protection and control for power levels 3 50%.

Therefore, the implied Design Basis Limit for this Reactor Trip function is 50%

power or 50% of nominal flow at RATED THERMAL POWER. Specifying the trip setpoint for this function as 5 40% of nominal steam flow at RATED THERMAL Page 2 of 5

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P,0WER provides a TOTAL' ALLOWANCE of 10% 'of. nominal steam flow 'at l

RATED THERMAL- POWER. The Channel Statistical Allowance (CSA) for the Steam Flow - Feed Flow Mismatch Reactor Trip is + 7.31% of Flownom.

- Subtracting the CSA value of 7.31% of Flownom from the TOTAL ALLOWANCE VALUE of 10% of Flownom yields a margin to the Design Basis Limit of 2.69% of Flownom for power levels t 50E Based on this analysis, the Steam Flow - Feed Flow Mismatch Reactor Trip Setpoint given in Table 2.2-1, item 14, is acceptable and provides positive margin ~to the Design Basis Limit. Changing the Technical Specifications Bases Section for the Steam Flow - Feed Flow Mismatch Reactor Trip to match the nomenclature given in Table 2.2-1 will more accurately reflect j actual plant conditions and will have no impact on the Safety Analysis, Technical '

Specifications or on plant operations.

Specific Changes i i

Unit 1 Technical Specifications The third sentence in the Unit 1 Technical Specifications Bases Section 2.2.1 for

- the Steam /Feedwater Flow Mismatch and Low Steam Generator Water Level

. Reactor Trip currently states:

The Steam /Feedwater Flow Mismatch portion of this trip is activated when the steam fiow exceeds the feedwater flow by 31.616

• 10e Ibs/ hour of full steam flow at RATED THERMAL POWER."

The third sentence in the Unit 1 Technical Specifications Bases Section 2.2.1 for L the Steam / Fee.dwater Flow Mismatch and Low Steam Generator Water Level is modified as follows:

"The Steam /Feedwater Flow Mismatch portion of this trip is activated before steam flow exceeds feedwater flow by 40% of nominal steam flow at RATED THERMAL POWER."

Unit 2 Technical Specifications The' third sentence in the Unit 2 Technical Specifications Bases Section 2.2.1 for the Steam /Feedwater Flow Mismatch and Low Steam Generator Water Level Reactor Trip currently states:-

i "The Steam /Feedwater Flow Mismatch portion of this trip is activated l when the steam flow exceeds the feedwater flow by > 1.616

• 108lbs/ hour of full steam flow at RATED THERMAL POWER."

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The third sentence in the Unit 2 Technical Specifications Bases Section 2.2.1 for the Steam /Feedwater Flow Mismatch and Low Steam Generator Water Level is modified as follows:

"The Steam /Feedwater Flow Mismatch portion of this trip is activated before steam flow exceeds feedwater flow by 40% of nominal steam flow at RATED THERMAL POWER."

Safety Significance This correction to the Unit 1 and Unit 2 Technical Specifications Bases for the Steam /Feedwater Flow Mismatch coincident with Low Steam Generator Water Level Reactor Trip does not create an unreviewed safety question as described below:

The Bases change does not increase probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report.

Changing the setpoint value given in the Bases for the Steam Flow - Feed Flow Mismatch podion of the reactor trip from an incorrect flow value given in terms of lbs/ hour to a flow value given in percent of nominal flow that is consistent with the trip setpoint specified in Table 2.2-1 has no ,

impact on the probability of occurrence or the consequences of an '

accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report. Note that the Steam Flow - Feed Flow Mismatch coincident with Steam Generator Low Level Reactor Trip is not credited in the Chapter 15 Safety Analysis.

This Bases change will now express the Trip Setpoint for the Steam Flow -

Feed Flow Mismatch portion of this reactor trip in a similar manner as it is presented in Technical Specification Table 2.2-1. This change will have no impact on the probability of occurrence or the consequence of any accident previously evaluated.

The Bases change does not create the possibility of an accident or malfunction of a different type than any evaluated previously in the safety analysis report.

This Bases change corrects a setpoint value that does not reflect current plant conditions. The revised basis will express the Trip Setpoint for the Steam Flow - Feed Flow Mismatch portion of this reactor trip in a similar manner as it is presented in Technical Specification Table 2.2-1. This change will make the bases consistent with the Technical Specification Page 4 of 5

i

' limit. Therefore it is concluded that no new or different kind of an accident or malfunction from any previously evaluated has been created. 3 1

This Bases change does not result in a reduction of margin of safety as defined in the basis for any Technical Specifications.

The margin of safety is not reduced as a result of this Bases change since the change has no effect on any safety analysis assumptions. As previously stated, the Steam Flow - Feed Flow Mismatch coincident with l Steam Generator Low Level Reactor Trip is not credited in the Chapter 15 Safety Analysis. This trip function is intended to serve as a backup for the Steam Generator (SG) Low-Low Level Reactor Trip during an IEEE 279-1971 Protection and Control System interaction when the SG Low-Low Level Reactor Trip is not functional. Further, it has been demonstrated j that this trip function still retains a positive margin of 2.69% of Flownom with '

respect to the implied Design Basis Limit of < 50% of Flow nom. Therefore, the proposed Bases changes do not reduce the margin of safety as it i applies to this reactor trip function.

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