ML18152A447

Application for Amend to License DPR-37,modifying TS to Support Operation of Unit 2 w/100 Psi Reduction in RCS Nominal Operating Pressure Through End of Operating Cycle 12,per Enforcement Discretion Granted by NRC on 930504

ML18152A447
Person / Time
Site:	Surry
Issue date:	05/06/1993
From:	Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML18152A448	List: ML18152A447 ML18153D333
References
93-283A, NUDOCS 9305170175
Download: ML18152A447 (17)

Text

I I VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 e

May 6, 1993 U.S. Nuclear Regulatory Commission Serial No. 93-283A Attention: Document Control Desk NL&P/: RO Washington, D.C. 20555 Docket No. 50-281 License No. DPR-37 Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNIT 2 PROPOSED TECHNICAL SPECIFICATIONS CHANGES REDUCED REACTOR COOLANT SYSTEM PRESSURE Pursuant to 10 CFR 50.90, the Virginia Electric and Power Company requests an amendment, in the form of changes to the Technical Specification to Facility Operating License No. DPR-37 for Surry Power Station Unit 2. Surry Unit 2 is currently beginning Cycle 12 operation at reduced Reactor Coolant System pressure in accordance with enforcement discretion granted by NRC on May 4, 1993. The proposed changes will incorporate the provisions of the enforcement discretion as Technical Specification requirements to allow continued operation of Unit 2 at reduced pressure for the duration of Cycle 12.

This Technical Specification change is requested on an emergency basis in accordance with 10 CFR 50.91 (a)(5). The basis for an emergency change request is included as Attachment 1. A discussion of the proposed Technical Specification changes is provided in Attachment 2. The proposed Technical Specification changes are provided in Attachment 3.

It has been determined that the proposed Technical Specification 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 4. The proposed Technical Specification 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.

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Very truly yours,

~JI ?~\P I *{z-- us:-0 L

W. L. Stewart Senior Vice President - Nuclear Attachments

,----~--------

140106 93051701.75 930506 .~. ~\ I I PDR ADOCK 05000281 I P PDR

cc: U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, N.W.

Suite 2900 Atlanta, Georgia 30323 Mr. M. W. Branch NRC Senior Resident Inspector Surry Power Station Commissioner Department of Health Room 400 109 Governor Street Richmond, Virginia 23219

e COMMONWEALTH OF VIRGINIA )

)

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.

Acknowledged before me this (p1J day of c!ll;i~ ,19 <f3.

My Commission Expires: LJlcug 31 , 19"9cf:-.

(;;;_,_. ~. &u Notary Public (SEAL)

Attachment 1 Basis For Emergency Change Request

e Basis for Emergency Change Request NRC regulations (1 O CFR 50.91 (a)(5)) require that, whenever an emergency situation exists, a licensee must explain why this emergency situation occurred and why it could not avoid this situation. The NRC will assess the licensee's reasons for failing to file an application sufficiently in advance of the event. An emergency situation exists when the NRC's failure to act in a timely way would result in derating or shutdown of a nuclear power plant, or in prevention of either resumption of operation or of increase in power output up to the plant's licensed power level. In such cases, the NRC may issue a license amendment involving no significant hazards consideration without prior notice and opportunity for a hearing or for public comment. Also, in such cases, the regulations require that the NRC be particularly sensitive to environmental considerations. Our discussion of why this proposed change meets the conditions necessary for emergency consideration is provided below.

Why Emergency Situation Occurred and Could Not Be Avoided On April 29, 1993, hydrostatic testing of the Unit 2 Reactor Coolant System (RCS) was being performed in accordance with ASME Section XI after modifications to the resistance temperature detector bypass lines and replacement of a safety injection check valve. During the hydrostatic test with RCS pressure of 2278 psig, indication of discharge from the "A" and "C" Pressurizer safety valves was observed. The RCS pressure was reduced to approximately 2050 psig. Further evaluation indicated that minor leakage from the "A" and "C" pressurizer safety valves occurred at the hydrostatic test pressure and was continuing to exhibit minor leakage at the reduced pressure.

Technical Specification 3.1 .A.3.a requires that all three pressurizer safety valves be operable when the head is installed on the reactor and RCS temperature is above 350°F. Mechanically securing the leaking pressurizer safety valves was evaluated as an acceptable method to aid in the reseating of the valves and completing the required hydrostatic test without potentially increasing leakage from the safety valves. In order to mechanically secure the safety valves, enforcement discretion from the requirement of Technical Specification 3.1.A.3.a was necessary. The enforcement discretion was requested and received on April 30, 1993. This enforcement discretion allowed two safety valves to be mechanically secured for 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

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On May 1, 1993, the "A" and "C" safety valves were mechanically secured and the hydrostatic test was completed. However, when the mechanical securing devices were

, removed following the test and RCS pressure was returned to normal operating pressure (2235 psig), indications from the pressurizer relief tank, safety valve acoustic monitors, and tail pipe temperature detectors continued to indicate minor leakage of the "A" safety valve. Reducing RCS pressure to approximately 1800 psig and holding at that pressure resulted in the leakage stopping and the associated safety valve temperature profile stabilizing, indicating reformation of the loop seal for the 'A' safety valve.

On May 3, 1993, RCS pressure was slowly increased to and held at 2135 psig without any further indication of the leakage phenomenon experienced previously. It was determined then, based on the most recent leakage experience to request a lowering of RCS operating pressure for Cycle 12 operation as a conservatively prudent action.

Operation with a reduced RCS pressure will provide additional margin between operating pressure and the onset of valve leakage. This reduces the potential for safety valve leakage at power, which can result in valve seat damage and any consequential plant transient that may result from increased RCS leakage.

Operation of Unit 2 at a reduced RCS pressure of 2135 psig was evaluated and found acceptable. In order to allow reduced pressure operation, changes to the Technical Specifications are required. Technical Specification 3.12.F.1 requires that RCS pressure be maintained ~2205 psig during power operation and Technical Specification 2.3.A.2(b) requires that the high pressurizer pressure reactor trip setting be~ 2385 psig.

To ensure continued overpressure protection and departure from nucleate boiling ratio (DNBR) margins at the proposed reduced operating pressure, these limits require revision. Discretionary enforcement from the requirements of Technical Specifications 3.12.F.1 and 2.3.A.2.(b) was requested and received on May 4, 1993 to allow resumption of Unit 2 operation. We are now proposing an emergency Technical Specification change within the provisions of that discretionary enforcement action to allow continued operation of Unit 2 at reduced RCS pressure through the end of Cycle 12.

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e Basis for Emergency Change Request Continued operation of Unit 2 at reduced RCS pressure is proposed as a conservatively prudent action to increase the margin between operating pressure and the onset of valve leakage which has been experienced during the recent hydrostatic testing.

Furthermore, the pressurizer safety valves are operable in their present condition and the unit can be safely operated at the proposed reduced RCS pressure. Operation at reduced pressure creates no safety consequences, as discussed in Attachments 2 through 6. The safety valves were removed, tested, and refurbished during the recently completed refueling outage. The as-left lift setpoints were within Technical Specification limits.

The observed minor safety valve leakage was believed to be the result of differential thermal expansion and the resultant slight misalignment of valve internals. Minor valve discharges can occur until internal stresses are relieved. This minor leakage is not indicative of a setpoint shift or other operability concern. The minor leakage which occurred on Unit 2 has no effect on the safety valve lift pressure or relieving capability.

We have concluded that the Unit 2 pressurizer safety valves are operable and that there is no safety or operational reason for placing the unit in cold shutdown for additional valve maintenance. Furthermore, based on the phenomenon as we understand it, valve maintenance is not likely to preclude a reoccurrence of valve leakage upon subsequent restart. The valve design in concert with the loop seal oven installation appears to be susceptible to the leakage phenomenon due to thermal changes at operating pressure.

While minor safety valve leakage in itself is not harmful, such leakage can lead to loss of the loop seal and steam cutting of valve seats if allowed to persist. Modifications which will reduce the potential for leakage are scheduled to be implemented during each unit's next refueling outage. The modifications will drain the loop seals, replace the safety valves with valves suitable for steam service, and eliminate the loop seal ovens. The loop seal ovens increase the propensity for leakage by maintaining high safety valve temperatures.

Failure to receive approval of the proposed Technical Specification changes on an emergency basis will increase the risk of RCS leakage and may necessitate a plant shutdown. Based on the recent experiences, operation of Unit 2 at the normal RCS pressure of 2235 psig could result in the resumption of minor safety valve leakage.

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e Continuing leakage would lead to steam cutting of the safety valve seats and increasing RCS leakage. With this concern in mind, we have concluded that operation at a reduced RCS pressure, which provides increased margin between operating pressure and the onset of valve leakage, is both prudent and conservatively safe. Since reduced RCS pressure operation is considered necessary to continue operation of Unit 2, this request satisfies the criteria of 10 CFR 50.91 (a)(5) for an emergency situation.

As discussed in Attachment 6, we have determined that the proposed Technical Specification changes do not involve significant hazards considerations. Therefore, we conclude that the condition in 10 CFR 50.91 (a)(5) regarding issuance only of a license amendment involving no significant hazards considerations is met.

Environmental Considerations Approval of the proposed Technical Specifications changes will not change the types of any effluents that may be released offsite, nor create a significant increase in individual or cumulative occupational radiation exposure. Operation at a reduced RCS pressure of greater than 2105 psig with the high pressurizer pressure reactor trip setpoint reduced to less than or equal to 231 O psig maintains the existing accident analysis margins and ensures RCS pressure will be maintained less than 110% of design for accident conditions. Therefore, the environmental consequences of any previously analyzed accident will remain unaffected.

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e e Attachment 2 Discussion of Changes

Discussion of Changes Introduction Surry Unit 2 is currently returning to operation following the end of Cycle 11 refueling outage. During Reactor Coolant System (RCS) hydrostatic testing and the subsequent return of the RCS to normal operating pressure, minor pressurizer safety valve leakage occurred. Reduction of RCS pressure from the normal operating pressure of 2235 psig to approximately 1800 psig terminated the leakage. Operation of Unit 2 through the end of Cycle 12 at the reduced RCS pressure of 2135 psig has been evaluated to be acceptable but requires changes to the Technical Specifications. Virginia Electric and Power Company therefore proposes changes to the Surry Technical Specifications to support operation of Unit 2 with a 100 psi reduction in RCS nominal operating pressure through the end of Operating Cycle 12.

Background

The proposed changes modify the Technical Specifications to allow operation of Unit 2 with a 100 psi reduction in RCS nominal operating pressure through the end of Cycle

12. A discussion of the proposed changes is provided below.

Technical Specification 2.3,A.2.(b) - Instrument Setting Limit for High Pressurizer Pressure Reactor Trip The purpose of the high pressurizer pressure reactor trip is to protect the RCS and its various components from overpressurization. Reduction of the high pressurizer pressure reactor trip setting limit from 2385 psig to 231 O psig is required to assure that a RCS overpressurization event does not cause RCS pressure to exceed the ASME Section Ill transient pressure limit of 110% design pressure. Reduction of this setpoint is necessary due to greater pressure overshoot during an overpressure transient. With initial RCS pressure 100 psi below the normal value, peak pressure could exceed the transient limit (2750 psia) unless the high pressurizer pressure trip setpoint were reduced. The proposed reduction of the setting limit is adequate to provide the required overpressure protection.

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e Technical Specification 3. 12.F. 1 - PNB Parameters Limits on RCS average temperature, pressurizer pressure, and RCS flow are specified to assure that each of the parameters is maintained within the normal, steady state envelope of operation assumed in the transient and accident analysis. These limits have been analytically demonstrated to be adequate to maintain a minimum DNBR which is greater than the design limit throughout each analyzed transient. The proposed changes decrease the specified minimum pressurizer pressure from 2205 to 2105 psig. The effect of this proposed decrease on DNBR margin has been evaluated and adequate margin is maintained.

Basis Sections of Technical Specifications 2.1, 2.2, and 3.3 The Basis Sections of these Technical Specifications are revised to reflect the change in RCS nominal operating pressure from 2235 to 2135 psig and the reduction of the high pressurizer pressure reactor trip setting limit from 2385 to 231 O psig. In addition, the Basis Section of Technical Specification 2.3 is revised to reflect a reduction in the power-operated relief valve (PORV) nominal setting from 2335 to 2235 psig. The setpoint change assures that the PORVs will provide a diverse source of RCS overpressure protection prior to reaching the high pressurizer pressure reactor trip setpoint.

Specific changes Technical Specification 2.3.A,2.(b) - Instrument Setting Limit for High Pressurizer Pressure Reactor Trip This specification provides the setting limit for the high pressurizer pressure reactor trip as follows:

(b) High pressurizer pressure - ~ 2385 psig page 2 of 8

e The proposed changes would modify this item as follows:

(b) High pressurizer pressure - ~ 2385* psig

• ~ 231 O psig for Unit 2 Cycle 12 operation at Reactor Coolant System nominal operating pressure of 2135 psig.

Technical Specification 3.12.F.1- DNB Parameters This specification provides a minimum pressurizer pressure to be maintained during power operation as follows:

Pressurizer Pressure~ 2205 psig The proposed changes would modify this item as follows:

Pressurizer Pressure ~ 2205* psig

• ~ 2105 psig for Unit 2 Cycle 12 operation at Reactor Coolant System nominal operating pressure of 2135 psig.

Basis Section for Technical Specification 2.1 - Safety Limit, Reactor Core This basis section states the following:

" ... based on steady state nominal operating power levels less than or equal to 100% steady state nominal operating Reactor Coolant System average temperatures less than or equal to 574.4°F and a steady state nominal operating pressure of 2235 psig."

The proposed changes would modify this discussion as follows:

" ... based on steady state nominal operating power levels less than or equal to 100% steady state nominal operating Reactor Coolant System average temperatures less than or equal to 574.4°F and a steady state nominal operating pressure of 2235* psig."

• For Unit 2 Cycle 12, Reactor Coolant System nominal operating pressure may be reduced to 2135 psig.

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Basis Section for Technical Specification 2.2 - Safety Limit. Reactor Coolant System Pressure This basis section states the following:

"The nominal settings of the power-operated relief valves at 2335 psig, the reactor high pressure trip at 2385 psig, and the safety valves at 2485 psig are established to assure never reaching the Reactor Coolant System pressure safety limit."

The proposed changes would modify this discussion as follows:

"The nominal settings of the power-operated relief valves at 2335* psig, the reactor high pressure trip at 2385** psig, and the safety valves at 2485 psig are established to assure never reaching the Reactor Coolant System pressure safety limit."

• 2235 psig for Unit 2 Cycle 12 operation at Reactor Coolant System nominal operating pressure of 2135 psig.

• • :::;; 231 O psig for Unit 2 Cycle 12 operation at Reactor Coolant System nominal operating pressure of 2135 psig.

Basis Section for Technical Specifjcatjon 3,3 - Safety Injection System This basis section states the following:

"The operating pressure of the Reactor Coolant System is 2235 psig ... "

The proposed changes would modify this discussion as follows:

"The operating pressure of the Reactor Coolant System is 2235* psig ... "

• For Unit 2 Cycle 12, Reactor Coolant System nominal operating pressure may be reduced to 2135 psig.

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Safety Significance The major issues considered in the evaluation of the proposed changes were (1) departure from nucleate boiling ratio (DNBR) performance margins, (2) the impact on transient performance parameters other than DNBR, and (3) fuel performance margin.

DNBR Performance Margins A reduction in nominal operating pressure, with all other operating parameters unchanged, results in a reduction in DNBR. Accidents which are DNB-limited and are protected by the overtemperature AT (OTAT) reactor trip will continue to have acceptable results due to the protection setpoint being automatically adjusted for variations in Reactor Coolant System pressure. Review of the pressure-dependent term (K3) of the overtemperature AT setpoint equation shows that a 100 psi reduction in operating pressure results in an approximate 5.7% full power AT reduction in nominal operating margin to the trip. With a current nominal K1 of 113.5%, a 100 psi pressure reduction would result in a moderate reduction in nominal operating margin to the reactor trip (from 13.5% to 7.8%). The corresponding margin to turbine runback would be 7.8% minus 2% equals 5.8% at hot full power conditions to preclude spurious run backs.

The Statistical DNBR Evaluation Methodology as implemented for Surry Unit 2 is not impacted by the reduced pressure operation. The statistics for the methodology implementation were developed over a range of pressures for power operation (from the high pressure limit to the low pressure limit). The previous package indicated a pressure sensitivity and the maximum sensitivity was used to determine the Statistical DNBR Limit (SOL). Therefore, there is no impact of the reduced pressure operation on the use of the Statistical DNBR Evaluation Methodology implementation for Surry Unit 2.

For DNBR-limited events that are not protected by OT~T. the effects of a 100 psi pressure reduction were determined by running explicit DNB analyses at the limiting statepoint conditions. The DNBR pressure sensitivity of the two most limiting events in this category, loss of flow and locked rotor, were specifically evaluated. These sensitivities of DNBR to pressure (a maximum 4.4% per 100 psi) were used to determine penalties to be applied against available retained DNBR margins. Retained DNBR margin consists of quantified conservatisms in the modeling and the difference page 5 of 8

between the DNBR limit applied in the accident analysis and the design DNBR limit.

Sufficient retained margins were available in the analyses supporting Surry Unit 2, Cycle 12 operation to accommodate the DNBR penalty from the 100 psi pressure reduction.

Thus, the results for all DNB-limited events will continue to be bounded by the currently applicable licensing basis analysis for reduced pressure operation of Unit 2.

Non-DNB Accident Performance In addition to the assessment described above for impact upon events limited by DNBR considerations, all events were evaluated to assess potential indirect impacts from operation at reduced RCS pressure. Such indirect effects are more likely for events with an RCS pressure acceptance criterion or for which RCS pressure has an influence upon key transient phenomena. The assessment of impact from reduced RCS pressure operation resulted in events being place into one of three classification categories: (1)

No Impact, (2) Potentially Impacted, (3) Impact Requires Quantification. Events in each category are summarized below:

Events Not Impacted These events were concluded to be insensitive to initial RCS pressure conditions, since the acceptance criterion is either not related to RCS pressure or initial pressure does not directly influence transient behavior. For example, the large and small break LOCA events are insensitive to initial pressure since significant RCS blowdown occurs well before the limiting transient conditions are reached. Therefore, these limiting conditions are not related to initial RCS pressure. The steam line break events are limited by conditions present following return to criticality. After reaching critical conditions for steam line break events, the RCS pressure is a function of cooldown rate and initial upper head temperature. The initial RCS pressure will have an insignificant impact on these conditions.

Uncontrolled Control Rod Assembly Withdrawal from Subcritical Condition Chemical and Volume Control System Malfunction (Boron Dilution)

Startup of an Inactive Reactor Coolant Loop Excessive Heat Removal Due to Feedwater System Malfunction Excessive Load Increase Incident Fuel Handling Accidents page 6 of 8

Volume Control Tank Rupture Waste Gas Decay Tank Rupture Loss of Coolant Accident (Small and Large Break)

Rupture of a Main Steam Line Events Potentially Impacted Events in this category either have RCS pressure as the key safety analysis acceptance criterion or have the potential for initial RCS pressure to influence transient behavior during the event. Each was evaluated and it was concluded that limiting conditions were either insignificantly impacted (or were beneficially impacted) by reduced initial RCS pressure. For example, peak RCS pressure for the loss of heat sink events (Loss of Feedwater, Loss of AC) is governed by characteristics of the pressurizer and main steam safety valves, rather than initial RCS pressure. For Locked Rotor (the other limiting overpressure event), reduced initial RCS pressure will result in reduced peak pressure, since reactor trip for this event occurs on low RCS coolant flow. The time for reaching this trip setpoint is not affected by reduced initial RCS pressure.

Loss of Normal Feedwater Loss of All Alternating Current Power to the Station Auxiliaries Steam Generator Tube Rupture Rupture of a Control Rod Drive Mechanism Housing (Control Rod Assembly Ejection)

Main Feedline Break Locked Reactor Coolant Pump Rotor (Overpressure Evaluation) page 7 of 8

Events For Which Impact Was Quantified There is one event in this category - the Loss of External Electrical Load. This event establishes the limiting conditions for RCS overpressurization, which is influenced by initial RCS pressure. Previous sensitivities have demonstrated that operation at reduced RCS pressure causes the predicted peak RCS pressure for this event to increase, assuming all other assumptions and protection system actions remain unchanged. The key reactor protection system function which terminates the event is the reactor trip on high pressurizer pressure. Peak RCS pressure is predicted to be greater than in the design analysis for operation at reduced initial pressure since the primary-secondary energy imbalance exists for a longer time interval. This causes more energy deposition in the RCS and greater pressurization.

The specific overpressurization event of interest was a complete loss of load transient from an initial pressure of 2105 psig (2235 psi - 100 psi reduction - 30 psi errors) combined with a 75 psi reduction in the assumed high pressurizer pressure reactor trip setpoint. The pressurizer safety valve relief behavior was modeled using the approach described in WCAP-12910, which was recently approved by NRC. (1) Consistent with the NRC's SER on this methodology, a 1% valve setpoint shift was assumed, since the Surry valves are tested in a steam environment and operate with water loop seals in their installed configuration. An isothermal temperature coefficient of reactivity which bounds the beginning of core life Technical Specification limit (+3 pcm/°F) was assumed. Acceptable results were obtained for this case, with a peak RCS pressure of 2741 psia (i.e., <110% of design pressure, or 2750 psia).

Fuel Performance Assessment The fuel rod design criteria were reviewed to assess the impact of a reduction in RCS pressure from 2235 psig to 2135 psig (100 psi reduction) for Surry Unit 2 Cycle 12 operation. A reduction in RCS pressure can impact the rod internal pressure evaluation since the differential between the fuel rod internal pressure and the RCS pressure could result in an increase in the fuel-to-clad gap. The rod internal pressure evaluation has been reanalyzed at the reduced system pressure of 2150 psia and it was determined that all design criteria will continue to be met.

(1) Letter from James E. Richardson (NRC-NRR) to T. E. Herrmann (PSV Working Group-W.OG), "Acceptance for Referencing of Licensing Topical Report WCAP-12910, 'Pressurizer Safety Valve Set Pressure Shift.' February 19, 1993."

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