ML20044G216

Application for Amends to Licenses NPF-4 & NPF-7, Respectively,Consisting of Proposed Change to Ts,Increasing as-found MSSV Setpoint Tolerance from 1% to 3% of Nominal Setpoint Pressure

ML20044G216
Person / Time
Site:	North Anna
Issue date:	05/24/1993
From:	Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20044G217	List: ML20044G216 ML20044G222
References
93-317, NUDOCS 9306020191
Download: ML20044G216 (21)

Text

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VIRGINIA ELECTRIC AND power COMPANY RICHMOND, VIRGINIA 23261 May 24, 1993 i U.S. Nuclear Regulatory Commission Serial No.93-317 Attention: Document Control Desk NL&P/MAE: R0 Washington, DC. 20555 Docket Nos. 50-338 50-339 Ucense Nos. NPF-4 NPF-7 Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNITS 1 and 2 PROPOSED TECHNICAL SPECIFICATIONS CHANGES MAIN STEAM SAFETY VALVE SETPOINT TOLFRANCE INCREASE Pursuant to 10 CFR 50.90, the Virginia Electric and Power Company requests amendments, in the form of changes to the Technical Specifications, to Facility Operating License Nos. NPF-4 and NPF-7 for North Anna Power Station Units 1 and 2, respectively. The proposed changes will increase the as-found main steam safety valve setpoint tolerance from 1% to 3% of nominal setpoint pressure.

A discussion of the proposed Technical Specifications changes is provided in Attachment 1. The proposed Technical Specifications changes. are provided in Attachment 2. It has been determined that the proposed Technical Specifications changes do not involve an unreviewed safety question as defined in 10 CFR 50.59 or a significant hazards consideration as defined in 10 CFR 50.92. The basis for our determination that these changes do not involve a significant hazards consideration is provided in Attachment 3. The proposed Technical Specifications changes have been reviewad and approved by the Station Nuclear Safety and Operating Committee and-the Management Safety Review Committee.

Should you have any questions or require additionalinformation, p! ease contact us.

Very truly yours,

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W. L. Stewart Senior Vice President - Nuclear Attachments P PDR 5 I

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oc: U.S. Nuclear Regulatory Commission Region ll 101 Marietta Street, N.W.  :

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Mr. D. R. Taylor-NRC Resident inspector North Anna Power Station Commissioner Department of Health Room 400 i 109 Governor Street '

Richmond, Virginia 23219 t

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COMMONWEALTH OF VIRGINIA )

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COUNTY OF HENRICO )

The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by W. L. Stewart who is Senior Vice President -

Nuclear, of Virginia Electric and Power Company. He is duly authorized to execute .

and file the foregoing document in behalf of that Company, and the statements in the document are true to the best of his knowledge and belief.

Aci:nowledged before me this 7//Yday of ( /b .19$.

My Commission Expires: ( -

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4 Attachment 1 Discussion of Changes 9

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DISCUSSION OF CHANGES INTRODUCTION The Technical Specification requirement for the main steam safety valve (MSSV) lift setpoint tolerance is currently 1%. However, MSSV setpoints have been found to drift during an operating cycle beyond the ,

allowable 1% tolerance, resulting in the relatively frequent production of Licensee Event Reports and evaluations to justify prior operation with equipment beyond the requirements of the Technical Specifications. As is demonstrated by the analyses and evaluations described herein, there is ample margin in the safety analyses and evaluations supporting North Anna operation to justify a relaxation of the allowable setpoint tolerance in the "as-found" condition to 3%. The main steam safety valves will continue to be reset to 1% following testing.

The increased MSSV lift setpoint tolerance could result in increased auxiliary feedwater (AFW) system backpressure, thus causing a reduction in the AFW delivered flow. AFW flow margin in existing safety analyses has been used to support a reduction in the ~ assumed AFW delivered flow from 340 gpm to 300 gpm.

BACKGROUND .

The MSSV lift setpoints and tolerance are specified in North Anna Unit  ;

I and 2 Technical Specifications Tables 3.7-2. The proposed. setpoint 3

tolerance increase does not change the nominal setpoints of the main steam 1

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safety valves. Only the as-found allowable tolerances about the existing setpoints are to be changed.

l In transient analysis, a "setpoint tolerance plus accumulation" model  :

is utilized in which it is assumed that each MSSV begins to open at its nominal setpoint plus the setpoint tolerance. At the setpoint tolerance pressure plus 3% " accumulation pressure", it is assumed that the safety valve achieves its rated flow. The flow rate is assumed to vary linearly  ;

between these two points.

An increased MSSV lift setpoint tolerance primarily affects the maximum pressure that will be attained in Nuclear Steam Supply System ,

(NSSS) transients. The limiting UFSAR Chapter 15 overpressurization transients include the Locked Reactor Coolant Pump Rotor and the loss of Load. The Loss of Normal Feedwater, Main Feedline Break, and Loss of Of f site Power transients are also considered in this evaluation. ,

l Because the increased MSSV lift setpoint tolerance could result in increased AFW system backpressure and therefore, in a reduced AFW delivered flow rate, credit has been taken for AFW flov margin in existing safety analyses to support a reduction of the _ assumed delivered AFW system flow rate from 340 gpm to 300 gpm. J I

The current Technical Specification Section 3/4.7.1.2, Auxiliary Feedwater System, was issued as part of the original Technical Specifications for Units 1 and 2. The conditions described were for the original system design with the AFW pumps aligned to common headers. The

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system alignment was modified so that each steam generator is fed by a dedicated AFW pump (1). This bases section is being modified to reflect the revised AFW delivered flow rate, expand the discussion on pump surveillance testing, and clarify the basis statement with respect to system configuration.

TECHNICAL SPECIFICATION CHANGES General -

The Technical Specification changes -described herein apply to North Anna Units I and 2. t Technical Specification Table 3.7-2 This table will specify a MSSV setpoint. tolerance of 3% for as-found i

conditions and 1% for as-lef t conditions following valve' testing.

i Technical Specification Bases 3/4.7.1.2 This basis statement is being adified to address the reduction in the i minimum AFW delivered flow rate, expand the discussion on pump-surveillance testing, and clarify the basis statement.

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SAFETY SIGNIFICANCE The following safety analysis design inputs or safety analysis '

criteria were identified to be affected by the proposed MSSV setpoint ,

tolerance increase:

1. The MSSV operating characteristic assumed in transient analysis (Locked Rotor and Loss of External Electrical Load analyses)

2. The AFW flow rate assumed in transient analysis (Loss of Normal Feedwater, Main Feedline Break, and Loss of Offsite Power analyses)  ;

3. Overtemperature AT (OTAT) reactor ' trip setpoint (the " steam generator safety line" assumed in the development of OTAT setpoints)

4. Operational margin among setpoints based on system pressure  !

A revised design auxiliary feedwater (AFW) minimum delivered flow rate of 300 gpm per motor-driven pump was developed to account for the impact of, among other things, a higher steam generator back pressure resulting from a 3% tolerance and 3% accumulation MSSV setpoint tolerance. The f revised AFW minimum flow rate assumption has been incorporated into NSSS  ?

transient analyses currently supporting North Anna operation. As such, '

the existing Main Feedline Break, Loss of Normal Feedwater, and Loss of Offsite power transient analyses of record surport the proposed increased MSSV setpoint tolerance. Similarly, the Locked Rotor and Loss of External '

Electrical Load transient analyses of record conservatively assume a MSSV  :

i setpoint tolerance of 3%. These analyses are discussed later in this section.

Because of the relatively small impact of the increased MSSV setpoint tolerance on delivered turbine-driven AFW pump flow rates, the AFW flow rates assumed in the Loss of Normal Feedwater " Appendix R" cases and

Station Blackout transient analyses continue to be provided by the turbine-driven AFW pump. There continues to be significant margin to the AFW flow rate assumption in these analyses.

Westinghouse has recently performed an evaluation of the operating characteristics of pressurizer safety valves (PSV) installed in a loop seal configuration (2). This model has been reviewed and approved by the NRC (3). The model assumes that the loop seal causes the PSV setpoint to be effectively increased, and the full opening of the safety valve is delayed while water in the loop seal is purged. After this evolution, the valve is assumed to " pop" to its full open position.

The analyses supporting the proposed increased MSSV lift 'setpoint tolerance utilized a conservative pressurizer safety valve (PSV) model.

In a limiting Reactor Coolant System (RCS) overpressurization analysis.

(Loss of External Electrical Load), this PSV modelling was determined to produce similar RCS overpressurization results with respect to the result obtained with the Westinghouse PSV model assuming the currently applicable 1% Technical Specification PSV setpoint tolerance. Further, the Loss of Load analysis demonstrated that the PSV setpoint modelling negligibly af fects the secondary side overpressurization results.

The sections below discuss the analyses and evaluations which support the proposed MSSV setpoint~ tolerance increase. The transient analyses have already been incorporated into the North Anna design basis via the provisions of 10 CFR 50.59, and conservatively include the application

of a 3% MSSV setpoint tolerance. Presentation of the transient analyses is provided here to support the proposed Technical Specifications change.

LOSS OF LOAD EVENT ANALYSIS The Loss of Load event is characterized by a rapid reduction in steam flow from the steam generator and a resultant rapid rise in secondary temperatures and pressures. Consequently, primary side _ temperatures and ,

pressures increase. The transient is terminated either by a ' direct reactor trip or in the limiting case by the high pressurizer ~ pressure trip. The transient has been shown to be not limiting with respect to core thermal margins.

The currently applicable analysis of the Loss of Load event included a transient sensitivity analysis to demonstrate that a 3% MSSV setpoint tolerance would not cause the results of the Loss of Load . analysis to violate safety criteria. The following assumptions were made in the Loss of Load analysis: ,

1. The loss of load is a 100% loss of load with no condenser dumps or power operated relief valves available.

2. The transient is initiated from 102% power.

3. Main feedwater is ramped to zero at the loss of load. i

4. A most negative Doppler power coefficient is assumed.

5. A most positive moderator temperature coefficient is  ;

assumed.

6. No credit is taken for a direct reactor trip on turbine trip.

7. No credit is taken for automatic rod control.

8. The increased MSSV setpoint tolerance is modelled by increasing the assumed nominal safety valve lift setpoints in the RETRAN model.

t For the caso of a 3% MSSV lift setpoint tolerance, the Loss _of Load transient sensitivity analysis demonstrated that the maximum primary and secondary pressures remained well below the prima ry and secondary pressure safety limits of 2750 psia and 1210 psia, respectively.

Graphs of the maximum RCS and steam generator transient pressure during the Loss of Load event are presented in Figures 1 and 2.

t LOCKED ROTOR EVENT ANALYSIS A Locked Rotor transient sensitivity study was performed to demonstrate that an increased MSSV lif t setpoint tolerance would not cause- '

the results of the Locked Rotor analysis to exceed analysis criteria.

The following assumptions were used in this analysis:

1. Initial reactor power is 102% of nominal.

2. Initial average core temperature is nominal-T(avg) + 4'F. .

3. Initial Pressurizer Pressure is 2280 psia, I nominal pressure +-30 psi.  !

4. Pressurizer sprays do not function. >

5. Pressurizer power operated relief valves do not $

function.  :

6. Condenser steam dump valves never open.

7. Atmospheric steam dump valves never open. l l

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8. Locked rotor is the initiating event at t=0.001 sec.

9. Reactor trip occurs on low RCS flow.

10. Coolant flow is divided into 50% through the core and 50% bypass.

11. Most Negative Doppler Power Coefficient

12. Most Positive Moderator Temperature Coefficient

13. Maximum value of delayed neutron fraction

14. Minimum trip reactivity The Locked Rotor transient sensitivity analysis with a 3% MSSV setpoint tolerance demonstrated that the maximum primary and secondary pressure remain below the primary and secondary pressure safety limits of 2750 psia and 1210 psia, respectively.

Graphs of the maximum RCS and steam generatur transient pressure during ,,

the Locked Rotor event are presented in Figures 3 and 4.

LOSS OF NORMAL FEEDWATER, MAIN FEEDLINE BREAK, AND LOSS OF 0FFSITE POWER A recent Auxiliary Feedwater System evaluation considered, among other AFW system configuration modifications, the impact of an increased MSSV

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setpoint tolerance on the motor-driven AFW pump deliverable flow rate.  ;

The analyses of record which currently support North Anna Units 1 and 2 operation assume motor-driven AFW flow rates that are equivalent to, or l conservative with respect to, the revised design AFW system deliverable ,

flow rate of 300 gpm.  :

The Loss of Normal Feedwater event analysis demonstrates that the AFW system delivers sufficient feedwater to prevent the relief of reactor ,

coolant water through the pressurizer relief or safety valves, and to prevent system overpressurization. For the cases with and without offsite power available, the feedwater flow rates required to provide adequate cooling were demonstrated to be well below the revised minimum AFW flow rate assumption.

The recent AFW system evaluation also considered the impact of the proposed MSSV lift setpoint tolerance increase on delivered turbine-driven AFW pump flow rates. The evaluation demonstrated that the flow rates assumed in the currently docketed Loss of Offsite Power analysis, and in the " Station Blackout" and " Appendix R" Loss of Normal i Feedwater analysis cases, continue to be met with a 3% MSSV lift setpoint t

tolerance.

The key safety considerations for the Main Feedline Break Transient l

are dependent on the presumed location of the break. If the break is presumed to occur upstream from the check valves, no discharge of steam  !

generator inventory through the break would occur. However, feedwater i supply to one or more steam generators would be degraded or interrupted.

The results of this scenario are bounded by the analysis of the Loss of Normal Feedwater event.

t' The Main Feedline Break analysis of record demonstrates that the revised minimum AFW flow rate is sufficient to ensure that a main feedline break will not result in overpressurization of the RCS, and that the

1 liquid level in the RCS will be sufficient to cover the core at all times.

Further, hot leg temperatures remain well below the saturation temperature.

3 EFFECT OF INCREASED MSSV LIFT SETPOINT TOLERANCE ON THE OVERTEMPERATURE DELTA-T TRIP FUNCTION The MSSV setpoint is an input into the analysis which produces overtemperature AT (OTAT) reactor trip setpoints (4). On a plot of allowable core inlet temperature versus core power level (i.e., the core thermal limits as defined in the Technical Specifications), the MSSV setpoint defines a line representing the set of RCS thermal / hydraulic conditions above which operation is precluded' by automatic MSSV actuation. The intersection of this " steam generator safety line" and the core thermal limit lines is a reference point utilized in the development of OTAT setpoints. '

The evaluation of the impact of an increased MSSV setpoint tolerance on the current OTAT trip setpoints concluded that inherent' margin in the ,

current OTAT setpoints easily compensates for the adverse effect;, of ,-

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increased MSSV lift setpoint tolerance. The proposed changes present no safety concerns with respect to OTAT protection.

OPERAfl0NAL MARGIN Because an increased allowable MSSV setpoint tolerance may result in '

actual MSSV setpoints which reduce operational margin between the MSSV  ;

setpoint and other NSSS pressure setpoints, an evaluation of the impact  ;

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I of the proposed increased MSSV setpoint tolerance on operational margins was performed.

The proposed setpoint tolerance has been chosen such that an inadvertent opening of the safety valves during normal operation will not i occur. The maximum secondary side operating pressure is approximately 1020 psia, which is also the lowest nominal setpoint for the condenser steam dumps. The nominal setpoint for the atmospheric steam dumps is 1050 -

psia. The 30 psi difference between the nominal setpoints of the condenser and atmospheric steam dumps adequately compensates for control channel uncertainties, thereby insuring that steam is not inadvertently ,

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dumped to the atmosphere.

i An evaluation of control channel uncertainties for the atmospheric steam dumps confirmed that there is no overlap of the atmospheric steam dump PORV and the MSSV tolerance bands. Specifically, the atmospheric steam dump nominal setpoint plus tolerance is 1067 psia. The nominal lift setpoint of the MSSV with the lowest lift pressure (1100 psia) minus 3% ,

tolerance is 1067 psia. Therefore, there is no overlap of these tolerance l bands, so the proposed setpoint tolerance increase presents no ,

1 operatioral considerations with respect to secondary side overpressure - 1 protection equipment. It may be concluded that there are no operational  ;

considerations resulting from the proposed setpoint tolerance increase.  !

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SUMMARY

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The foregoing analyses and evaluations demonstrate that the proposed l t

MSSV setpoint tolerance increase does not cause any UFSAR Chapter 15 l i

transient to exceed analysis criteria. Further, the impact of an

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increased MSSV setpoint tolerance on the current OTAT protection  !

setpoints and operational margin was evaluated and determined to be insignificant. I l

1 In summary, each pertinent safety criteria was evaluated for an i

.t increased MSSV setpoint tolerance and all were found to be acceptable. i i

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REFERENCES (1) Virginia Electric and Power Company Resprinse to NRC Staff Comment S15.18, FSAR Section S15.18, " Analysis of the Most Limiting Feedwater Break with Offsite Power Available and All Loops in .;

Operation," September 10, 1976.

(2) G. O. Barrett: " Pressurizer Safety Valve Set Pressure Shift,"

WOG Project MUHP2351, WCAP-12910, March, 1991.

(3) Letter from J. E. Richardson (USNRC) to T. E. Herrmann (WOG),

" Acceptance for Referencing of Licensing Topical Report WCAP-12910,

' Pressurizer Safety Valve Set Pressure Shift'," dated February 19, -

1993.

i (4) S. L. Ellenberger, et al.: " Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions," WCAP-8746, dated March,1977.

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