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| Attachment 2 Technical Specifications Changes North Anna Power Station i
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| 9404290179 940419 PDR ADOCK 05000280 P PDR _
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| REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) b) The SHUTDOWN MARGIN requimment of Specification 3.1.1.1 is determined at least once per 12 hours, and c) A power distribution map is obtained from the movable incom detectors and FQ(Z) andF[,, are verified to be within their limits within 72 hours, or d) Either the THERMAL POWER level is educed to s 75% of RATED THERMAL POWER within one hour and within the next 4 hours the high neutron flux trip setpoint is reduced to s 85% of RATED THERMAL POWER, or c) The remainder of the rods in the group with the inoperable rod are aligned to within i 12 steps of the inoperable rod within the hour while maintaining the thermal power, rod sequence, and insertion limits of Specification 3.1.3.6 during subsequent operation.
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| SURVEILLANCE REQUIREMENTS 1
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| 4.1.3.1.1 The position of each rod shall be detennined to be within the group demand limit by l verifying the individual rod positions at least once per 12 hours except during time intervals when the Rcxl Position Deviation Monitoris inoperable, then verify the group positions at least once per 4 hours.
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| 4.1,3.1.2 Each rod not fully inserted in the core shall be determined to be OPERABLE by movement of at least 10 steps in any one direction at least once per 92 days. l l
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| NORTH ANNA - UNIT 1 3/4 1-19 Amendment No. 46r448, l
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| REACTIVITY CONTROL SYSTEMS
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| : LIMITING CONDirION FOR OPERATION (Continued) b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours, and c) A power distgbution map is obtained from the movable incore detectors and Fq(Z) and Fg are verified to be within their limits within 72 hours, or d) Either:
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| '1) The THERMAL POWER level is reduced to less than or equal to 75%
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| of RATED THERMAL POWER within one hour and within the next 4 hours the high neutron flux trip setpoint is reduced to less than or equal. ~l to 85% of RATED THERMAL POWER, or
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| : 2) The remainder of the rods in the group with the inoperable rud are aligned to within 12 steps of the inoperable rod within the hour while maintaining the thermal power, rod sequence, and insertion limits of
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| . Specification 3.1.3.6 during subsequent operation.
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| SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each rod shall be determined to be within the group demand limit by verifying the individual rod positions at least once per 12 hou'rs except during time intervals when the Rod Position Deviation Monitor is inoperable, then verify the group positions at least once per -
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| ' 4 hours.
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| .4.1.3.1.2 Each rod not fully inserted in the core shall be determined to be OPERABLE by movement of at least 10 steps in any one direction at least once per 92 days. _l j
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| NORTH ANNA - UNIT 2 3/4 1-17 ' Amendment No. 424,
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| Attachment 3 Technical Specifications Changes Surry Power Station
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| l TABLE 4.1-2A MIN! MUM FREQUENCY FOR EQUIPMENT TESTS FSAR SECTION -
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| TEST FREQUENCY REFERENCE DESCRIPTION Rod drop times of a!! full Prior to reactor criticality: 7
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| : 1. ControlRod Assemblies length rods at hot conditions a. For all rods following each removal of the reactor vessel head
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| : b. For specially affected individisat rods following any maintenance on or modi-fication to the control rod drive system which could affect the drop time of those specific rods, and
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| : c. Each refueling shutdown.
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| Partial movement of all rods Quarterly 7 l
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| : 2. ControlRo-d Assemblies Functional Each refteling shutdown 6
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| : 3. Refueling Water Chemical Addition Tank Setpoint Per TS 44.3 4
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| : 4. Pressurizer Safety Valves Setpoint Per TS 4.0.3 10
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| : 5. Main Steam Safety Valves Containment isolation Trip
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| * Functional Each refueling shutdown 5 6.
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| * Functional Prior to refueli.g 9.12
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| : 7. Refueling System Interbcks Service Water System
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| * Functional Each refueling shutdown 9.9 8.
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| Fire Protection Pump and Functional Monthly 9.10 9.
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| Power Supply
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| * Evaluate Daily 4
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| : 10. Primary System Leakage
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| : 11. Diesei FuelSupply
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| * Fuel Inventory 5 days / week 8.5 .g Monthly 9.1 .#
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| : 12. Boric Acid Piping Heat
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| * Operational Tracing Circuits {
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| cr Functional Before each startup (TS 4.7) 10
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| : 13. Main Steam Line Trip Valves .
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| (Fu!! Closure) The provisions of Specification 4.0.4.
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| are not appfcable
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| Attachment 4 Significant. Hazards Consideration Determination
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| ' Surry and North Anna Power Stations ,
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| , nt j Sionificant Hazards Considerations i
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| Ac documented in- NUREG-1366, " Improvements to Technical Specifications Surveillance Requirements," dated December 13,1992, the NRC has completed a l comprehensive examination of surveillance requirements in technical specifications that !
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| require testing at power. The NRC staff found that while the majority of testing at power is important, safety can be improved, equipment degradation decreased, and an unnecessary burden on personnel resources eliminated by reducing the amount of testing that is required by technical specifications at power. Generic Letter (GL) 93-05, "Line-Item Technical Specifications Improvements to Reduce Surveillance Requirements for Testing During Power Operation," dated September 27, 1993, provides guidance for preparing license amendments to implement the recommendations of NUREG-1366. Consistent with GL 93-05, we are requesting a l change to the control rod surveillance frequency from monthly to quarterly for Surry and from once per 31 days to once per 92 days for North Anna.
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| Virginia Electric and Power Company has reviewed the proposed changes against the !
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| criteria of 10 CFR 50.92 and has concluded that the changes as proposed do not pose a significant hazards consideration. The proposed changes to the surveillance requirements for the controi rods at Surry and North Anna, are consistent with the intent of Generic Letter 93-05, "Line-Item Technical Specifications improvement to Reduce j Surveillance Requirements for Testing During Power Operation," dated September 27, 1993, which is to improve safety, decrease equipment degradation, and reduce unnecessary burden on personnel resources by reducing testing requirements that are .!
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| marginal to safety. Specifically, operation of Surry and North Anna Power Stations in i accordance with the proposed Technical Specifications changes will not: i l
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| : 1. '9volve a significant increase in the probability of occurrence or consequences of an accident previously evaluated.
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| The proposed change to the surveillance frequency for controi rods does not increase the probability of an accident occurrence. Surveillance testing is a means of determining control rod operability and does not of itself contribute to control rod inoperability. Although reduced testing also implies a less frequent confirmation of mechanical operability, operational experience ' ' established that the reduced testing does not decrease plant safety. Furthermore, reduced l frequency testing reduces the probability of an inadvertent operational transient or misaligned control rod. There are other means available (e.g., Individual Rod Page 1 of 3
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| Po,sition ' Indicators, flux distributions anomalies) to detect a misaligned control rod.
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| Reducing the frequency of surveillance testing will decrease the possibility of finding an inoperable control rod. Industry experience has shown that most inoperable (stuck) control rods are identified during rod drop testing and unit startup after refueling outages. Therefore, the NRC has determined that a reduced frequenc'y surveillance test during power is acceptable to determine control rod operability (trippable).
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| The control rods will continue to be operated in the same manner during the surveillance testing and will be available to shutdown the reactor if a Reactor Protection System trip setpoint is reached. The operability requirements, alignment and insertion limits for the control rods remain unchanged. Since the control rods remain available (trippable) to perform their intended safety function, testing of the control rods at the proposed reduced frequency will not increase l the consequences of an accident previously evaluated.
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| : 2. Create the possibility of a new or different kind of accident from any accident previously evaluated. l l
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| l The proposed reduced frequency testing of the control rods does not change the l way the Control Rod Drive System or the control rods are operated. The reduced frequency of testing of the control rods does not alter the operation of the Control Rod Drive System or the control rods ability to perform their intended safety function. Therefore, the reduced frequency testing of the control rods does not generate any new accident precursors. In fact, industry experience has shown that this surveillance testing may result in inadvertent reactor trips, dropped control rods, or unnecessary challenges to safety systems. Therefore, the possibility of a new or different kind of accident than previously evaluated is not created by the proposed changes in surveillance frequency of the control rods.
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| 3.' involve a significant reduction in a margin of safety.
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| The proposed reduced frequency testing of the' control rods does not change the control rod operability requirement or the way the Control Rod Drive System is.
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| operated. NUREG-1366, concluded that most stuck control rods are discovered during plant startup after refueling or during control rod drop testing. Therefore, Page 2 of 3
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| routine surveillance testing of the control rods at the proposed reduced frequency
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| , is considered adequate to identify inoperable (stuck) control rods during ]
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| operation. The reduced surveillance requirements do not affect the margin of safety in that the operability requirements remained unchanged and the existing ;
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| safety analysis, which assumes the most reactive control rod sticks out of the I core during accident scenarios, remains bounding. Therefore, no margins of safety are adversely affected.
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