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| | issue date = 11/10/1989 | | | issue date = 11/10/1989 |
| | title = Application for Amends to Licenses DPR-32 & DPR-37,modifying Pressurizer Safety Valve Setpoint Tolerance in Tech Spec 3.1.A.3.c for Remainder of Cycle 10 | | | title = Application for Amends to Licenses DPR-32 & DPR-37,modifying Pressurizer Safety Valve Setpoint Tolerance in Tech Spec 3.1.A.3.c for Remainder of Cycle 10 |
| | author name = STEWART W L | | | author name = Stewart W |
| | author affiliation = VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) | | | author affiliation = VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| | addressee name = | | | addressee name = |
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| | document type = OPERATING LICENSES-APPLIATION TO AMEND-RENEW EXISTING, TEXT-LICENSE APPLICATIONS & PERMITS | | | document type = OPERATING LICENSES-APPLIATION TO AMEND-RENEW EXISTING, TEXT-LICENSE APPLICATIONS & PERMITS |
| | page count = 4 | | | page count = 4 |
| | | project = |
| | | stage = Request |
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| {{#Wiki_filter:-e VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 November 10, 1989 United States Nuclear Regulatory Commission Attention: | | {{#Wiki_filter:- VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 e |
| Document Control Desk Washington, D. C. 20555 Gentlemen: | | November 10, 1989 United States Nuclear Regulatory Commission Serial No. 89-750A Attention: Document Control Desk NO/ETS R1 Washington, D. C. 20555 Docket Nos. 50-280 50-281 License Nos. DPR-32 DPR-37 Gentlemen: |
| VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 PRESSURIZER SAFETY VALVE SETPOINT Serial No. NO/ETS Docket Nos. 89-750A R1 50-280 50-281 License Nos. DPR-32 DPR-37 EMERGENCY TECHNICAL SPECIFICATION CHANGE REQUEST Pursuant to 10 CFR 50.90, the Virginia Electric and Power Company requests an emergency amendment, in the form of a change to the Technical Specifications, to Operating Licenses No. DPR-32 and DPR-37 for Surry Power Station Units 1 and 2, respectively. | | VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 PRESSURIZER SAFETY VALVE SETPOINT EMERGENCY TECHNICAL SPECIFICATION CHANGE REQUEST Pursuant to 10 CFR 50.90, the Virginia Electric and Power Company requests an emergency amendment, in the form of a change to the Technical Specifications, to Operating Licenses No. DPR-32 and DPR-37 for Surry Power Station Units 1 and 2, respectively. Based on the potential generic safety valve setpoint testing issue recently identified, our recently completed Unit 2 safety valve setpoint testing results and the subsequent recent lift of a Unit 2 pressurizer safety valve during pressure testing, we are requesting that the +/- 1% setpoint tolerance of Technical Specification 3.1.A.3.c be modified for the remainder of Cycle 1O for both units. This change is requested due to the potential that the Unit 1 and 2 safety valve setpoints are outside the current + 1% tolerance required by the existing Specifications due to setpoint testing methodology. The safety valve testing methodology issue and the results of the Unit 2 safety valve testing were reported to the NRC in our letter 89-750, dated October 30, 1989. The requested Technical Specification change, discussion, and significant hazards consideration are provided in Attachments 1 and 2. |
| Based on the potential generic safety valve setpoint testing issue recently identified, our recently completed Unit 2 safety valve setpoint testing results and the subsequent recent lift of a Unit 2 pressurizer safety valve during pressure testing, we are requesting that the +/- 1 % setpoint tolerance of Technical Specification 3.1.A.3.c be modified for the remainder of Cycle 1 O for both units. This change is requested due to the potential that the Unit 1 and 2 safety valve setpoints are outside the current + 1 % tolerance required by the existing Specifications due to setpoint testing methodology. | | At the time this issue was identified, Unit 2 was shutdown and the safety valves could be readily tested. We chose to test the Unit 2 valves on a water loop seal and on steam and have them reset on a water loop seal to eliminate the setpoint shift due to testing methodology. We applied the Unit 2 test findings to the Unit 1 valves and notified the NRC of the potential Technical Specification noncompliance (>+ 1 % |
| The safety valve testing methodology issue and the results of the Unit 2 safety valve testing were reported to the NRC in our letter 89-750, dated October 30, 1989. The requested Technical Specification change, discussion, and significant hazards consideration are provided in Attachments 1 and 2. At the time this issue was identified, Unit 2 was shutdown and the safety valves could be readily tested. We chose to test the Unit 2 valves on a water loop seal and on steam and have them reset on a water loop seal to eliminate the setpoint shift due to testing methodology. | | setpoint tolerance). As documented in a NRC letter of October 27, 1989, we received 6 weeks of discretionary enforcement to work with the N RC and the industry to resolve the issue. On November 6, 1989, during RCS pressure testing prior to Unit 2 return to service, the recently reset 'C' pressurizer safety valve lifted due to an apparent loss of loop seal. Based on this recent event, we have decided to reset the Unit 2 valves using the previous methodology (steam) to minimize the potential for challenges of the pressurizer safety valves and seek similar relief from Technical Specification 3.1.A.3.c for Unit 2 pending generic resolution of this issue. To date, generic resolution of this issue has not been reached by industry or the NRC; and therefore, we request this Technical Specification change ~ce*ssed as an emergency change per 10 CFR 50.91 for continued operation .of Unit 1 and the restart of Unit 2 after resetting the 8911160247 891110 \ |
| We applied the Unit 2 test findings to the Unit 1 valves and notified the NRC of the potential Technical Specification noncompliance | | PDR ADOCK 05000280 P PNlJ |
| (>+ 1 % setpoint tolerance). | | |
| As documented in a NRC letter of October 27, 1989, we received 6 weeks of discretionary enforcement to work with the N RC and the industry to resolve the issue. On November 6, 1989, during RCS pressure testing prior to Unit 2 return to service, the recently reset 'C' pressurizer safety valve lifted due to an apparent loss of loop seal. Based on this recent event, we have decided to reset the Unit 2 valves using the previous methodology (steam) to minimize the potential for challenges of the pressurizer safety valves and seek similar relief from Technical Specification 3.1.A.3.c for Unit 2 pending generic resolution of this issue. To date, generic resolution of this issue has not been reached by industry or the NRC; and therefore, we request this Technical Specification change ~ce*ssed as an emergency change per 1 O CFR 50.91 for continued operation .of Unit 1 and the restart of Unit 2 after resetting the 8911160247 891110 PDR ADOCK 05000280 P PNlJ \ | | pressurizer safety valves on steam. The explanation for the emergency processing of this change is provided in Attachment 3. |
| pressurizer safety valves on steam. The explanation for the emergency processing of this change is provided in Attachment | | The proposed Technical Specification change modifies pressurizer safety valve lift setpoint tolerances to -1 % and +5%, which encompasses any observed increase in setpoint shift identified during testing of the Unit 2 safety valves. This modified tolerance remains within the safety analysis bounds. Although no additional measures are necessary based on the setpoint tolerance proposed and the safety analyses, due to the uncertainties and the unquantified variables associated with safety valve testing, we will perform appropriate measures to provide added assurance that primary pressure can not exceed 2750 psig (110% of system design). |
| : 3. The proposed Technical Specification change modifies pressurizer safety valve lift setpoint tolerances to -1 % and +5%, which encompasses any observed increase in setpoint shift identified during testing of the Unit 2 safety valves. This modified tolerance remains within the safety analysis bounds. Although no additional measures are necessary based on the setpoint tolerance proposed and the safety analyses, due to the uncertainties and the unquantified variables associated with safety valve testing, we will perform appropriate measures to provide added assurance that primary pressure can not exceed 2750 psig (110% of system design). These measures, which will apply to each unit, are the same measures as we are currently applying as part of discretionary enforcement for Unit 1. They include the continued operability of at least one of the two Power Operated Relief Valves (PORV) and the anticipatory reactor trip on turbine trip circuitry.
| | These measures, which will apply to each unit, are the same measures as we are currently applying as part of discretionary enforcement for Unit 1. They include the continued operability of at least one of the two Power Operated Relief Valves (PORV) and the anticipatory reactor trip on turbine trip circuitry. With these measures in place, any analyzed UFSAR transient would result in peak pressure remaining below 2750 psig, even if the setpoints were to increase to a value higher than the +5% proposed limit. |
| With these measures in place, any analyzed UFSAR transient would result in peak pressure remaining below 2750 psig, even if the setpoints were to increase to a value higher than the +5% proposed limit. In addition, we wm continue to work with the NRC, industry and Owners groups to determine and expedite a satisfactory resolution to this potential generic issue in order to support the end of Cycle 10 application of this proposed Technical Specification change. These requests have been reviewed and approved by the Station Nuclear Safety and Operating Committee and the Safety Evaluation and Control Staff. It has been determined that the proposed Technical Specification change does not involve an unreviewed safety question as defined in 1 O CFR 50.59 or a significant hazards consideration as defined in 10 CFR 50.92. The basis for our no significant hazards consideration determination is included in Attachment | | In addition, we wm continue to work with the NRC, industry and Owners groups to determine and expedite a satisfactory resolution to this potential generic issue in order to support the end of Cycle 10 application of this proposed Technical Specification change. |
| : 2. Should you have any additional questions, please call. Very truly yours, J.s~ . L. Stewart e ior Vice President
| | These requests have been reviewed and approved by the Station Nuclear Safety and Operating Committee and the Safety Evaluation and Control Staff. It has been determined that the proposed Technical Specification change does 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 no significant hazards consideration determination is included in Attachment 2. |
| -Nuclear Attachments | | Should you have any additional questions, please call. |
| : 1. Proposed Technical Specification Change 2. Discussion of Proposed Change and Significant Hazards Consideration | | Very truly yours, J.s~ |
| : 3. Justification for Emergency Technical Specification Change Request | | . L. Stewart e ior Vice President - Nuclear Attachments |
| . '--cc: U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, N. W. Suite 2900 Atlanta, Georgia 30323 Mr. W. E. Holland NRC Senior Resident Inspector Surry Power Station Commissioner Department of Health Room 400 109 Governor Street Richmond, Virginia 23219 e COMMONWEAL TH OF VIRGINIA ) ) COUNTY OF HENRICO ) The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by J. L. Wilson who is Assistant Vice President
| | : 1. Proposed Technical Specification Change |
| -Nuclear Operations, for W. L. Stewart who is Senior Vice President | | : 2. Discussion of Proposed Change and Significant Hazards Consideration |
| -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 ./Jzff day oi<p,k.,; , 1931. My Commission Expires: ~.b~ ZS. 19~. (SEAL)-<: | | : 3. Justification for Emergency Technical Specification Change Request |
| .. -. .:::: -./ <" . ' '\' _ _&~ Notary--#i:iubl ic I' ! I}} | | |
| | cc: U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, N. W. |
| | Suite 2900 Atlanta, Georgia 30323 Mr. W. E. Holland NRC Senior Resident Inspector Surry Power Station Commissioner Department of Health Room 400 109 Governor Street Richmond, Virginia 23219 |
| | |
| | I' |
| | !I e |
| | COMMONWEALTH OF VIRGINIA ) |
| | ) |
| | COUNTY OF HENRICO ) |
| | The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by J. L. Wilson who is Assistant Vice President - Nuclear Operations, for 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 ./Jzffday oi<p,k.,; , 1931. |
| | My Commission Expires: ~.b~ ZS. 19~. |
| | __ & ~ |
| | Notary--#i:iubl ic (SEAL)-<: |
| | ..~ -. .:::: |
| | ~ ~ |
| | -./ <" |
| | '\'}} |
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Amend Consistent W/Info Provided in Rev 1 of NUREG-1431 ML18151A9791997-11-0505 November 1997 Application for Amends to Licenses DPR-32 & DPR-37, Respectively.Amends Will Revise TS to Increase Maximum Allowable Fuel Enrichment from 4.1 to 4.3 Weight Percent U-235.Description of Change Encl ML18152A0761997-11-0505 November 1997 Application for Amends to Licenses DPR-32 & DPR-37, Establishing Requirements for Use of Temporary Supply Line (Jumper) to Provide Svc Water to Component Cooling Heat Exchangers ML18153A6331996-11-26026 November 1996 Application for Amends to Licenses DPR-32,DPR-37,NPF-4 & NPF-7,respectively,eliminating Records Retention Requirements Section of Ts,Per GL 95-06 & Adminstrative Ltr 95-06 ML18153A6881996-04-15015 April 1996 Application for Amends to Licenses DPR-32 & DPR-37, Clarifying Applicability of Quadrant Power Tilt Ratio Requirements ML18153A5371996-03-21021 March 1996 Application for Amends to Licenses DPR-32 & DPR-37, Clarifying Requirements for Testing Charcoal Adsorbent in Auxiliary Ventilation & Control Room Air Filtration Sys as Outlined in Tech Specs 4.12 & 4.20,respectively ML18153A5781996-01-30030 January 1996 Application for Amends to Licenses DPR-32 & DPR-37, Eliminating Surveillance Requirement for Certain Reactor Coolant Liquid Samples ML18153A6741995-11-20020 November 1995 Application for Amends to Licenses DPR-32 & DPR-37,permiting Use of 10CFR50 App J,Option B Performance-Based Containment Leakage Rate Testing. ML18153A7121995-07-20020 July 1995 Application for Amends to Licenses DPR-32 & DPR-37,proposing Changes to Ts,By Establishing New Setpoint Limit for Steam Generator high-high Level & Provides More Restrictive Setting Limits for Certain Rps/Esfas Setpoints ML18153A6971995-07-14014 July 1995 Application for Amends to Licenses DPR-32 & DPR-37,providing Two H Allowed Outage Time for One RHR Pump to Accommodate Plant Safety,Emergency Power Sys Surveillance Testing & Permit Depressurizing Su Accumulators in Lieu of Isolation ML18152A1841995-02-14014 February 1995 Application for Amends to Licenses DPR-32 & DPR-37,revising TS 4.4.D to Permit Approved Exemptions to ILRT Frequency Requirements,Including one-time Exemption Request from Section III.D.1(a) of 10CFR50,App J for Surry Unit 2 ML18153B2111995-01-24024 January 1995 Application for Amends to Licenses DPR-32 & DPR-37,modifying as-found Test Acceptance Criterion for Pressurizer Safety Valves ML18153B1601994-11-29029 November 1994 Application for Amends to Licenses DPR-32 & DPR-37,revising TSs to Implement Zirlo Fuel Cladding ML18153B1561994-11-22022 November 1994 Application for Amends to Licenses DPR-32 & DPR-37,deleting Unnecessary Descriptive Phrases Re Number of Cells in Station & EDG Batteries ML18153B1481994-11-10010 November 1994 Application for Amends to Licenses DPR-32 & DPR-37 to TS Re Changes to TS to Clarify SR for Reactor Protection & Engineered Safeguard Sys Instrumentation & Actuation Logic ML18153B0921994-10-11011 October 1994 Application for Amends to Licenses DPR-32 & DPR-37,modifying Surveillance Frequencies of Hydrogen Analyzers in Accordance W/Generic Ltr 93-05 ML18153B0491994-08-30030 August 1994 Application for Amends to Licenses DPR-32 & DPR-37,revising TS to Accomodate Core Uprating,In Accordance w/WCAP-10263, Review Plan for Uprating Licensed Power of PWR Power Plant. ML18153B0041994-07-14014 July 1994 Application for Amends to Licenses DPR-32 & DPR-37, Requesting Amendments in Form of Changes to TS to Eliminate Remaining References to cycle-specific Parameters in Surry TS ML18153A9651994-06-0909 June 1994 Application for Amends to Licenses DPR-32 & DPR-37,modifying in Part,Chemical & Vol Control Sys Specs & Safety Injection Sys Specs in Accordance w/NUREG-0452,Rev 4,NUREG-1431 & Generic Ltr 93-05 ML18153A8801994-02-25025 February 1994 Application for Amends to Licenses DPR-32,DPR-37,NPF-4 & NPF-7,modifying Surveillance Frequency of Nozzles in Containment & Recirculation Spray Sys in Accordance W/Gl 93-05, Line-Item TS Improvements to Reduced Srs.... ML18153B4301993-12-27027 December 1993 Application for Amends to Licenses DPR-32,DPR-37,NPF-4 & NPF-7,changing TS to Revise Review Responsibilities of Station Nuclear Safety & Operating Committee & Mgt Safety Review Committee ML18153B3481993-10-19019 October 1993 Application for Amends to Licenses DPR-32 & DPR-37,adding Operability & Action Statements for RSHX Service Water Outlet Radiation Monitors to TS Table 3.7-6 & SRs for Monitors to TS Table 4.1-2 ML18153B3301993-09-29029 September 1993 Application for Amends to Licenses DPR-32 & DPR-37,modifying TS to Require Insp Frequency of Low Pressure Turbine Blades to Permit Blade Insp to Perform Concurrent W/Disk & Hub Insp ML18153D3951993-07-20020 July 1993 Application for Amends to Licenses DPR-32,DPR-37,NPF-4 & NPF-7 Proposing Changes of Deletion of Requirement for Station Nuclear Safety & Operating Committee & Audit Frequencies ML18153D3901993-07-16016 July 1993 Application for Amends to Licenses DPR-32 & DPR-37,proposing Changes to TS to Permit Operation W/Three Degree Increase in Svc Water Temp Limit for Containment Air Partial Pressures of 9.1,9.2 & 9.35 Psia ML18153D3931993-07-16016 July 1993 Application for Amends to Licenses DPR-32,DPR-37,NPF-4 & NPF-7 to Implement Revised 10CFR20,revise Frequency of Radiological Effluent Release Repts from Semiannual to Annual & Clarify Site Maps ML18153D3791993-07-0202 July 1993 Application for Amends to Licenses DPR-32 & DPR-37,updating Augmented Insp Program for Sensitized Stainless Steel to Incorporate Newer Code Requirements While Maintaining Increased Insp Philosophy of ASLB ML18153D3781993-07-0202 July 1993 Application for Amends to Licenses DPR-32 & DPR-37 Modifying TS to Include COLR Which Presents reload-specific Limits for Key Core Operating Parameters ML18152A4471993-05-0606 May 1993 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 ML18153D3141993-04-21021 April 1993 Application for Amends to Licenses DPR-32 & DPR-37,providing Clarification of Design Response Time of Containment Hydrogen Analyzers & Deleting Unnecessary Channel Check for Analyzers ML18153D3111993-04-21021 April 1993 Suppl to 930315 Application for Amends to Licenses DPR-32 & DPR-37,modifying LCO & Action Statements for Main Control Room & Emergency Switchgear Room Air Conditioning Sys ML18153D2771993-03-19019 March 1993 Application for Amends to Licenses DPR-32 & DPR-37,revising TS to Address Operation W/Control Rod Urgent Failure Condition ML18153D2691993-03-15015 March 1993 Application for Amends to Licenses DPR-32 & DPR-37,changing TS to Permit Use of Two New Main Control Room & Emergency Switchgear Room Air Conditioning Sys Chillers,When Fully Operational,To Meet LCO ML18153D2281993-01-26026 January 1993 Application for Amends to Licenses DPR-32 & DPR-37,adding NRC Std Fire Protection License Condition to Each OL & Relocating Fire Protection Requirements from TS to Updated Fsar,Per GL 86-10 & GL 88-12 ML20128A5221993-01-22022 January 1993 Amends 173 & 172 to Licenses DPR-32 & DPR-37,respectively, Revising TS to Modify Acceptance Criteria for Functional Testing of Anchor Darling Mechanical Snubbers ML18153D1971992-12-11011 December 1992 Application for Amends to Licenses DPR-32 & DPR-37, Eliminating Operability Requirements for RCS Loop Stop Interlocks & Establishing Requirements for Operation of Loop Stop Valves ML18153D1681992-11-10010 November 1992 Application for Amends to Licenses DPR-32 & DPR-37, Specifying Action Statement for Operable Svc Water Flow Paths to Main Control & Emergency Switchgear Room Air Conditioning Condensers Governed by TS 3.14.C ML18153D1611992-10-26026 October 1992 Application for Amends to Licenses DPR-32 & DPR-37 to Increase Limit for Intermediate Range High Flux Reactor Trip Setpoint & to Eliminate Redundant Stipulation That Predicted Critical Rod Position Be Above Zero Power Insertion Limit ML18153D1171992-09-0404 September 1992 Suppl to 911230 Amend Applications 165 & 164 to Licenses DPR-32 & DPR-37,respectively,reinstating Allowed Outage Time for Surveillance Testing of Nonessential Svc Water Isolation Actuation Logic & Reformatting Logic ML18153D1141992-09-0404 September 1992 Application for Amends to Licenses DPR-32 & DPR-37,changing TS to Modify Acceptance Criteria for Functional Testing of Anchor/Darling Mechanical Snubbers to Reflect Replacement of Pacific Scientific Snubbers W/Anchor/Darling Snubbers ML18153D0871992-08-0707 August 1992 Application for Amends to Licenses DPR-32 & DPR-37, Establishing Operating Requirements for Recirculation Mode Transfer (Rmt) Function,Including LCO Action Statement & SR for Rmt & Deleting RWST Max Vol Requirements ML18153D0861992-08-0404 August 1992 Suppl to 911008 Application for Amends to Licenses DPR-32 & DPR-37,upgrading Portions of Section 3.0 & 4.0,per Generic Ltr 87-09 to Restrict Changes of Operational Conditions While in LCO & Delaying Entry Into Action Statement ML18153D0701992-07-28028 July 1992 Application for Amends to Licenses DPR-32 & DPR-37,deleting Operability & SRs of Hydrogen Monitor from Explosive Gas Monitoring Instrumentation Requirements for Waste Gas Holdup Sys & Including Requirements to Submit Special Rept to NRC 1999-04-28
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Text
- VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 e
November 10, 1989 United States Nuclear Regulatory Commission Serial No. 89-750A Attention: Document Control Desk NO/ETS R1 Washington, D. C. 20555 Docket Nos. 50-280 50-281 License Nos. DPR-32 DPR-37 Gentlemen:
VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 PRESSURIZER SAFETY VALVE SETPOINT EMERGENCY TECHNICAL SPECIFICATION CHANGE REQUEST Pursuant to 10 CFR 50.90, the Virginia Electric and Power Company requests an emergency amendment, in the form of a change to the Technical Specifications, to Operating Licenses No. DPR-32 and DPR-37 for Surry Power Station Units 1 and 2, respectively. Based on the potential generic safety valve setpoint testing issue recently identified, our recently completed Unit 2 safety valve setpoint testing results and the subsequent recent lift of a Unit 2 pressurizer safety valve during pressure testing, we are requesting that the +/- 1% setpoint tolerance of Technical Specification 3.1.A.3.c be modified for the remainder of Cycle 1O for both units. This change is requested due to the potential that the Unit 1 and 2 safety valve setpoints are outside the current + 1% tolerance required by the existing Specifications due to setpoint testing methodology. The safety valve testing methodology issue and the results of the Unit 2 safety valve testing were reported to the NRC in our letter 89-750, dated October 30, 1989. The requested Technical Specification change, discussion, and significant hazards consideration are provided in Attachments 1 and 2.
At the time this issue was identified, Unit 2 was shutdown and the safety valves could be readily tested. We chose to test the Unit 2 valves on a water loop seal and on steam and have them reset on a water loop seal to eliminate the setpoint shift due to testing methodology. We applied the Unit 2 test findings to the Unit 1 valves and notified the NRC of the potential Technical Specification noncompliance (>+ 1 %
setpoint tolerance). As documented in a NRC letter of October 27, 1989, we received 6 weeks of discretionary enforcement to work with the N RC and the industry to resolve the issue. On November 6, 1989, during RCS pressure testing prior to Unit 2 return to service, the recently reset 'C' pressurizer safety valve lifted due to an apparent loss of loop seal. Based on this recent event, we have decided to reset the Unit 2 valves using the previous methodology (steam) to minimize the potential for challenges of the pressurizer safety valves and seek similar relief from Technical Specification 3.1.A.3.c for Unit 2 pending generic resolution of this issue. To date, generic resolution of this issue has not been reached by industry or the NRC; and therefore, we request this Technical Specification change ~ce*ssed as an emergency change per 10 CFR 50.91 for continued operation .of Unit 1 and the restart of Unit 2 after resetting the 8911160247 891110 \
PDR ADOCK 05000280 P PNlJ
pressurizer safety valves on steam. The explanation for the emergency processing of this change is provided in Attachment 3.
The proposed Technical Specification change modifies pressurizer safety valve lift setpoint tolerances to -1 % and +5%, which encompasses any observed increase in setpoint shift identified during testing of the Unit 2 safety valves. This modified tolerance remains within the safety analysis bounds. Although no additional measures are necessary based on the setpoint tolerance proposed and the safety analyses, due to the uncertainties and the unquantified variables associated with safety valve testing, we will perform appropriate measures to provide added assurance that primary pressure can not exceed 2750 psig (110% of system design).
These measures, which will apply to each unit, are the same measures as we are currently applying as part of discretionary enforcement for Unit 1. They include the continued operability of at least one of the two Power Operated Relief Valves (PORV) and the anticipatory reactor trip on turbine trip circuitry. With these measures in place, any analyzed UFSAR transient would result in peak pressure remaining below 2750 psig, even if the setpoints were to increase to a value higher than the +5% proposed limit.
In addition, we wm continue to work with the NRC, industry and Owners groups to determine and expedite a satisfactory resolution to this potential generic issue in order to support the end of Cycle 10 application of this proposed Technical Specification change.
These requests have been reviewed and approved by the Station Nuclear Safety and Operating Committee and the Safety Evaluation and Control Staff. It has been determined that the proposed Technical Specification change does 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 no significant hazards consideration determination is included in Attachment 2.
Should you have any additional questions, please call.
Very truly yours, J.s~
. L. Stewart e ior Vice President - Nuclear Attachments
- 1. Proposed Technical Specification Change
- 2. Discussion of Proposed Change and Significant Hazards Consideration
- 3. Justification for Emergency Technical Specification Change Request
cc: U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, N. W.
Suite 2900 Atlanta, Georgia 30323 Mr. W. E. Holland NRC Senior Resident Inspector Surry Power Station Commissioner Department of Health Room 400 109 Governor Street Richmond, Virginia 23219
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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 J. L. Wilson who is Assistant Vice President - Nuclear Operations, for 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 ./Jzffday oi<p,k.,; , 1931.
My Commission Expires: ~.b~ ZS. 19~.
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Notary--#i:iubl ic (SEAL)-<:
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