ML20087A112

Requests Amend to Licenses NPF-4 & NPF-7,increasing Pressurizer Safety Valve Lift Setpoint Tolerance as Well as Reduce Pressurizer High Pressure RT Setpoint & Allowable Value

ML20087A112
Person / Time
Site:	North Anna
Issue date:	07/26/1995
From:	Ohanlon J VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20087A114	List: ML20087A112 ML20087A117
References
95-366, NUDOCS 9508040177
Download: ML20087A112 (15)

Text

j VIHOINTA 1$LECTRIC AND POWER COMPANY Ricunown,VinoINEA 20261 July 26, 1995 U.S. Nuclear Regulatory Commission Serial No.95-366 i

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 3PECIFICATIONS CHANGES INCREASED PRESSURIZER SAFETY VALVE LIFT SETPOINT TOLERANCE Pursuant to 10 CFR 50.90, the Virginia Electric and Power Company requests amendments, in the form of changes to the Techt!! cal Specifications, to Facility Operating Ucense Nos. NPF-4 and NPF-7 for North Anna Power Station Units 1 and 2, respectively. The proposed changes will increase the pressurizer safety valve lift setpoint tolerance as well as reduce the pressurizer high pressure reactor trip setpoint and allowable value.

A discussion of the proposed Technical Specifications changes is provided in.

The proposed Technical Specifications changes are provided in.

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 reviewed and approved by the Station Nuclear Safety and Operating Committee and the Management Safety Review Committee.

Should you have any questions or require additional information, please contact us.

Very truly yours, j

w James P. O' Hanlon Senior Vice President - Nuclear Attachments 01005n

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U.S. Nuclear Regulatory Commission Region il 101 Marietta Street, N.W.

Suite 2900 Atlanta, Georgia 30323 Mr. R. D. McWhorter NRC Senior Resident inspector North Anna Power Station Commissioner Department of Radiological Health Room 104A 1500 East Main Street Richmond, Virginia 23219

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COMMONWEALTH OF VIRGINIA )

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The foregoing document was acknowledged before me, in'and for the County and Commonwealth aforesaid, today by J. P. O'Hanlon, who is Senior Vice President -

Nuclear, of Virginia Electric and Power Company. He is duly authorized to execute 1

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.

Acknowledged before me this QL day of Tuks

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My Commission Expires:

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Discussion of Changes

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I DISCUSSION OF CHANGES INTRODUCTION Virginia Electric and Power Company requests changes to the following Technical Specifications for North Anna Power Station Units 1 and 2:

Table 2.2-1, Reactor Trip System Instrumentation Trip Setpoints, item 10, Pressurizer Pressure -- High 3.4.2, Reactor Coolant System Safety Valves - Shutdown 3.4.3.1, Reactor Coolant System Safety and Relief Valves - Operating 3/4.4.2 AND 3/4.4.3 Bases, Reactor Coolant System Safety and Relief Valves A safety evaluation has been performed which justifies increasing the current Technical Specification pressurizer safety valve (PSV) at-power (Modes 1-3) lift setpoint tolerance from 1 % as-found and i1% as-left to + 2%/-3% averaoe as-found with no single valve outside 3% as-found and i1% per valve as-left. The as-found value is based on testing, the results of which are expressed as an error (i.e., positive or negative percentage deviation from the nominal lift setpoint). The " average" means that the errors of the tested valves are summed and the result divided by the number of valves tested. This result is compared to the acceptable range of +2% to -3%. No single valve is allowed be outside of the 13% tolerance.

l The safety evaluation also supports an increase to the Hot Shutdown (Mode 4) required PSV lift setpoint tolerance from 11% as-found and 1% as-left to i3% per valve as-found and 1% per valve as-left. These proposed changes will provide greater operational flexibility in meeting periodic test requirements establishad by the safety analyses.

A concurrent reduction in the pressurizer high pressure reactor trip setpoint and allowable value of TS Table 2.2-1 are also proposed. These changes ensure that the analysis results for the loss of external load accident continue to meet the acceptance criteria with the higher PSV tolerance.

The Loss of Load, Locked Rotor, and Rod Withdrawal event analyses demonstrate that increasing the at-power PSV lift setpoint tolerance to

+2%/-3% averaoe as-found with no single valve outside i3% as-found and 11% per valve as-left does not result in a transient pressure in excess of the overpressure safety limit. Further, the increased setpoint tolerance does not adversely impact the DNBR results of any North Anna UFSAR Chapter 15

transient analysis. Mode 4 overpressure protection is adequate with one PSV with a tolerance of 3%

Finally, the increased PSV setpoint tolerances and reduction of the high pressurizer pressure reactor trip setpoint do not present any operational considerations which would significantly impact the performance of the plant during normal operation or during postulated accident conditions, in summary, each pertinent safety criterion was evaluated for the proposed Technical Specification changes, and all were found to be acceptable. The proposed Technical Specification changes do not create an unreviewed safety question or a significant hazards consideration.

BACKGROUND l

Three code safety valves are installed on each unit's pressurizer. The valves have a nominal lift setpoint of 2485 psig and function to protect the reactor coolant system from overpressure, j

The PSVs have a history of drifting outside the currently allowed tolerance of 11%, resulting in Technical Specification violations. Because up to a i3%

tolerance is the permitted by ASME Code Section ill, Division 1, Subsection NB, Part 7513, for code safety valves, a project was initiated to justify an increase in the PSV tolerance to reduce the number of TS violations. The analyses and evaluations described herein support the proposed PSV setpoint tolerance increase.

The proposed Technical Specification changes do not affect the nominal lift setpoint of the pressurizer safety valves, nor the as-left tolerance requirement.

Only the allowable as-found tolerance about the existing lift setpoint is to be changed.

To ensure acceptable analysis results with the increased as-found PSV tolerance, a concurrent reduction in the pressurizer high pressure reactor trip TS setpoint is also proposed. This reduction provides a faster response of the reactor protection system to overpressure events without significantly impacting existing operating margin.

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SPECIFIC CHANGES The following specific Technical Specification (TS) changes apply to both Units 1 and 2:

Table 2.2-1 Item 10 Revise the existing trip setpoint from "s 2385 psig" to "s 2360 psig".

e Revise the allowable value from "s 2395 psig" to "s 2370 psig".

TS 3.4.2 Revise the safety valve lift setpoint tolerance from "i1%'" to "i3% as-found and 11% as-left'."

TS 3.4.3.1 Revise the safety valve lift setpoint tolerance from "i1%'" to "+2%/-3%

average as-found with no single valve outside i3%, and i1% per valve as-left '."

Bases for TS 3/4.4.2 AND 3/4.4.3 Add the following paragraphs to the bases section:

"The safety valve tolerance requirement for Modes 1-3 is expressed as an average value. That is, the as-found error (expressed as a pocitive or negative percentage) of each tested safety valve is summed and divided by the number of valves tested. This average as-found value is compared to the acceptable range of +2% to -3%. In addition, no single valve is allowed to be outside of i3%.

An average tolerance of +2%/-3% was confirmed to be adequate for Modes 1-3 accident analyses. For the overpressure events, the analyses considered several combinations of valve tolerance with the arithmetic average of the three valves' tolerance equal to +2% (with no valve outside of 3%). The case of a +2% tolerance on each of the three valves provided the most limiting results.

The -3% tolerance is limiting for the DNB acceptance criterion."

SAFETY SIGNIFICANCE An increase in the pressurizer safety valve lift setpoint tolerance affects the maximum pressure that will be attained in a system transient. Evaluation of the overall effect of changing the PSV setpoint tolerance was accomplished by examining the effect of the changes on those transients which experience the most limiting pressure increases. These transients are the Complete Loss of External Electrical Load, the Locked Reactor Coolant Pump Rotor, and Rod Withdrawal events initiated from subcritical and at power.

In the analyses described herein, the PSVs were assumed to open in accordance with the Reference 3 pressurizer safety valve model, hereafter termed the Westinghouse model. To support the proposed PSV lift setpoint tolerance increase, the PSVs were assumed to begin opening at a pressure 2%

above the nominal lift setpoint.

For PSVs, such as North Anna's, that are installed on a loop seal, the Westinghouse model requires application of an additional 1 % " medium shift" to account for the effects of setting the valves on steam while installing them on a water-filled loop seal. Also, an additional delay was assumed for the opening of the valve to simulate purging of the loop seal.

Lastly, the time required for the PSV to " pop" completely open was simulated by application of an additional 0.1% tolerance to the assumed 2% lift setpoint tolerance. For the overpressure analyses, the PSVs ara assumed to close at a pressure 3% below the setpoint pressure (3% blowdov.4 In past analyses, the three PSVs have been modeled as one valve with pressure relief characteristics equivalent to the three valves opening in tandem. In order to confirm the appropriateness of an averaae PSV tolerance requirement, Loss of Load sensitivity studies were performed modeling three separate valves and using several combinations of valve tolerances, each averaging +2%.

The results of these sensitivities showed that a tolerance of +2%/+2%/+2% is more conservative than any other combination averaging +2%. Therefore, the past practice of modeling the three valves as one valve will remain appropriate under a requirement of a +2% averace tolerance. The limit for any one valve will be 13%.

Because an increased low end tolerance potentially reduces the system pressure experienced at the point of minimum departure from nucleate boiling ratio (DNBR), the effect of an increased PSV setpoint tolerance on the DNBR results of affected transients was evaluated. The proposed changes were also evaluated in light of their impact on operational margins.

Transient analyses were porfornie.1 with the RETRAN system transient analysis code (References 3 and 4).

LOSS OF LOAD 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 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 not to be limiting with respect to core thermal margins.

The Loss of Load analysis was performed to establish that a Loss of Load event would not result in primary side pressures beyond the limit of 2750 psia nor secondary side pressures beyond the limit of 1210 psia when the pressurizer safety valve lift setpoint tolerance is increased to 2%.

The following assumptions were made in this analysis:

1.

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

2.

The transient is initiated from 102% of an uprated core power level (1.02 x 2893 MWt).

3.

Main feedwater isolation signal at the time of reactor trip with a five second main feedwater regulating valve ramp time.

4.

A least negative Doppler temperature coefficient is assumed.

5.

Pressurizer sprays are disabled.

6.

A zero moderator temperature coefficient at full power (+0.0 pcm/F) is assumed. Overpressurization results for cases initiated from 70% power with the MTC at the 70% power analysis limit

(+6.0 pcm/ F) are bounded by those for cases initiated from 100% power with the MTC at the 100% power analysis limit (0.0 pcm/ F), since the reduction in core power more than offsets the increased allowable temperature-driven reactivity feedback.

7.

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

8.

No credit is taken for automatic rod control.

9.

An average PSV lift setpoint tolerance of 2% is simulated in accordance with the Reference (1) Westinghouse model.

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10.

One second response time requirement for the pressurizer high pressure reactor trip. The response time requirement for this trip was formerly contained in the Technical Specifications and has since been relocated to the Technical Requirements Manual-(TRM).

The current requirement is 2.0 seconds.

The TRM.

requirement for this response time will be changed to s1.0 second prior to implementation of the revised PSV tolerance Technical Specifications.

The response time for this trip -is periodically tested and typical results are less than 0.5 second.

The maximum primary side (cold leg) pressure was determined to be 2740 psia which is below the overpressure safety limit (110%

of design pressure) of 2750 psia. The peak secondary pressure was 1181 psia which is below the acceptance criterion of 1210 psia.

LOCKED ROTOR ANALYSIS This analysis was performed in order to determine if an increased PSV lift setpoint tolerance would result in an overpressurization of the primary side during a postulated Locked Rotor transient.

The following assumptions were used in this analysis:

1.

Initial reactor power is 102% of an uprated core power level (1.02 x 2893 MWt).

2.

Initial average core temperature is nominal T(avg) + 4 F.

3.

Initial pressurizer pressure is 2280 psia (nominal pressure + 30 psi).

4.

Pressurizer sprays do not function.

5.

Pressurizer power operated relief valves do not function.

6.

Condenser steam dump PORVs do not open.

7.

Atmospheric steam dump PORVs do not open.

8.

A locked rotor is the initiating event, i-9.

Reactor trip occurs on low RCS flow (87% of full flow).

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-10. Coolant flow is divided into 50% through the core and 50% bypass to rnaximize coolant expansion in the core region,'and to simulate extreme local voiding during a locked rotor event.

11. A least negative Doppler temperature coefficient.

12. A zero moderator temperature coefficient at full power (0.0 pcm/*F).

Locked Rotor overpressurization results for cases initiated from 70%

power with the MTC at the 70% power analysis limit (+6.0 pcm/*F) are bounded by those for cases initiated from 100% power with the MTC at the 100% power analysis limit (0.0 pcm/ F), since the.

reduction in core power more than offsets the increased allowable temperature-driven reactivity feedback.

13. Minimum trip reactivity.

14. Main feedwater isolation signal at the time of reactor trip with a conservatively short five second main feedwater regulating valve ramp time.

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The RETRAN transient analysis of the Locked Rotor event with a 2% average

-1 PSV setpoint tolerance rendered a peak primary (cold leg) pressure of 2739 psia. This value is below the primary safety limit of 2750 psia. The maximum secondary side pressure was determined to be 1186 psia, which is below the l

overpressure limit of 1210 psia.

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ROD WITHDRAWAL EVENTS The Loss of Load and Locked Rotor events have historically been considered.

the limiting RCS overpressurization events. However, recent reanalyses of the Rod Withdrawal at Power (RWAP) and Rod Withdrawal from Subcritical (RWSC) events revealed that these events' may result in significant pressurization of the RCS, particularly those cases initiated from low power.

Therefore, the results of these accident analyses are reviewed here for completeness.

The impact of a 3% PSV lift setpoint tolerance (bounding the 2% average tolerance) on RWAP results was quantified. The limiting case was initiated from 8% power, and assumed a 30 pcm/sec reactivity insertion rate, a 3% PSV lift setpoint tolerance, a water loop seal (additional opening delay), and a -1.4 pcm/ F full power Doppler temperature coefficient. This case resulted in a maximum RCS pressure of 2725 psia.

Similarly, the impact of a 3% PSV lift setpoint tolerance on RWSC results was quantified. A case which assumed a 100 pcm/sec reactivity insertion rate, a 3% PSV lift setpoint tolerance, and a water loop seal was run. The peak RCS pressure in the analysis is 2587 psia.

Therefore, the results for both the RWSC and RWAP are below 110% of the RCS design pressure, and are therefore acceptable.

DNB CONSIDERATIONS Because an increased negative PSV lift setpoint tolerance potentially reduces the system pressure experienced at the point of minimum Departure from Nucleate Boiling Ratio (DNBR), the effect of -3% PSV setpoint tolerances on the DNBR results of affected transients was evaluated by examining the North Anna UFSAR Chapter 15 safety analysis results.

Of the affected transients, only the DNBR results of the Locked Rotor event are potentially adversely affected by the increased negative tolerance.

A conservative maximum impact on the Locked Rotor analysis was quantified and the DNBR acceptance criteria continue to be met.

OPERATIONAL MARGIN CONSIDERATIONS The proposed setpoint tolerances have been chosen such that an inadvertent opening of the safety valves during normal operation will not occur.

The proposed high primary pressure trip setpoint is 2370 psig with an uncertainty of 18.72 psi. The nominal setpoint plus uncertainty is, therefore, 2389 psig.

Because the nominal PSV lift setpoint minus 3% tolerance is 2425 psia, a reactor trip will occur before the PSVs open.

It may be concluded that the proposed setpoint tolerance change does not present any operational considerations.

MODE 4 CONSIDERATIONS The shutdown overpressure protection requirements were calculated.

The analysis used a tolerance of +3% on the pressurizer safety valve. Tolerance in the negative direction will provide additional margin. The analysis showed that for two charging pumps injecting at double the flow of a single pump, two PSVs provide adequate overpressure protection. Therefore, for the case of one operable charging pump, as required in mode 4 and below, one PSV will provide adequate overpressure protection with a tolerance of up to +3%.

Therefore, the Mode 4 requirement (i.e., TS 3.4.2) will be specified as *3%.

SUMMARY

AND CONCLUSIONS The Loss of Load, Locked Rotor, and Rod Withdrawal event analyses demonstrate that increasing the at-power PSV lift setpoint tolerance to

+2%/-3% averaoe as-found with no single valve outside i3% as-found and 1% per valve as-left does not result in a transient pressure in excess of the overpressure safety limit. Further, the increased setpoint tolerance does not adversely impact the DNBR results of any North Anna UFSAR Chapter 15 transient analyses. Mode 4 overpressure protection is adequate with one PSV with a tolerance of 3%.

The reduction in the pressurizer high pressure reactor trip setpoint ensures that the analysis results for the loss of external load accident continue to meet the acceptance criteria with the higher PSV tolerance. Finally, the increased PSV setpoint tolerances and reduction of the high pressurizer pressure reactor trip setpoint do not present any operational considerations which would significantly impact the performance of the plant during normal operation or during postulated accident conditions. In summary, each pertinent safety criteria was evaluated for the proposed Technical Specification changes, and all were found to be acceptable.

The proposed changes have been reviewed against the criteria of 10 CFR 50.59. This review concluded that these changes raise no unreviewed safety questions. The basis for this determination is as follows:

1.

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated are not increased. Affected safety related parameters were analyzed for a change to North Anna 1 and 2 Technical Specifications 3.4.2 and 3.4.3 and Table 2.2-1 item 10. It was determined that the overpressure safety limits would not be exceeded in the most limiting overpressure transients (Loss of Load, Locked Rotor, and Rod Withdrawal events) with the pressurizer safety valve lift setpoint positive side tolerance increased to an average of +2% and the high pressurizer pressure reactor trip reduced by 25 psi. The DNBR results of transients impacted by the proposed setpoint tolerance increase (-3%) meet the acceptance criterion after accounting for the impact of the proposed changes. The increased setpoint tolerance will not result in an inadvertent opening of the pressurizer safety valves. Mode 4 overpressure protection is adequate with one PSV with a tolerance of *3%.

Since the affected accidents have been evaluated and found to meet their acceptance criteria with the revised PSV tolerance, the probability of occurrence of an accident or malfunction of equipment important to safety previously evaluated is not increased.

2.

The possibility of an accident or malfunction of a different type than any

evaluated pmviously in the safety analysis report is not created. 'The proposed change to North Anna -'1 and 2 Technical Specifications 3.4.2 and 3.4.3 and Table 2.2-1 item 10 does not involve any alterations to the physical plant which would introduce any new or unique operational modes or accident precursors. Only the allowable as-found tolerance about the existing PSV lift setpoint will be changed, along with a reduction in the pressurizer high pressure reactor trip setpoint.

3.

The margin of safety as ' defined in the basis of the technical specifications is not reduced. It was determined that the most limiting overpressure transients do not result in maximum pressures in excess of

- the overpressure safety limits. The DNBR results of affected transients are not made more limiting by the proposed setpoint tolerance increase and high pressurizer pressure reactor trip setpoint reduction. Therefore, the margin of safety is unchanged by the proposed increase in the safety valve setpoint tolerance.

REFERENCES

1) " Pressurizer Safety Valve Set Pressure Shift," Westinghouse Owners Group Project MUHP2351, WCAP-12910, dated March,1991.

2) Letter from J. E. Richardson (USNRC) to T. E. -Herrman, Chairman, Pressurizer Safety Valve Working Committee, Westinghouse Owners Group,

" Acceptance For Referencing Of Licensing Topical Report WCAP-12910,

' Pressurizer Safety Valve Set Pressure Shift'", February 19,1993.

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3) Topical Report VEP-FRD-41 A, "Vepco Reactor System Transient Analysis l

using the RETRAN Computer Code," dated May,1985.

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4) Letter from W. L. Stewart (Virginia Power) to H. R. Denton (NRC), "Surry and North Anna Power Stations Reactor System Transient Analyses," Serial No.85-753, dated November 19,1985 (RETRANO2 MOD 003).

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Technical Specifications Changes i

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