Text

Proposed Tech Specs Re Core Operating Limits Rept
	ML20012E828
Person / Time
Site:	North Anna
Issue date:	03/29/1990
From:	; VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	;
Shared Package
ML20012E826	List: ML20012E825; ML20012E828; ML20012E831; ML20012E832;
References
	NUDOCS 9004060380
	Download: ML20012E828 (61)
	v • d • e

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c t Attachment 1 Proposed Technical Specification Change North Anna Unit 1 Virginia Electric and Power Company 90040Ao380 900320 PDR MDOCK 05000338 P P DC-l- u

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- DEFINITIONS s MicTm EE r

F 1.0 DEFINIT 10hlS l

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! Action...................................................................................................................... 11 l A t t al F lu x Diff er e noe . . . .. . ........... .... . .. . ... . . ...... ..... .. . .. . .. . .. .... .... ... . .. ... ... .. . .. . ... . . . .. .. . .. . . 11 C h a n nel C alibra lio n........ ... .. .. .. ... .. ..... ....... .... .... . .. .. ..... .... .. ... . . ... . . .. .. . . ... .. .. ... . . .. .. ... .. . 11 Channe!Chook....................................................................................................... 11 k C ha n n el F u nction at T e 5t...... .. .. .. . .. .. .... . .. ....... .. . .. .. .... ... .. ... .. .. .. ... .. .. .. . .. ... .... .. .. . .. . ...... 11 C o n t aln me n t i n i og rit y .. . . . .. . .. . .. . .. .. . . .. . .. . . .. .. . .. .. .. . .. . ... . . . ... ... .. . . . . .. . . . . ... .. ... . .. . .. . .. . . .. . .. . . 11 l- C on tr olle d Le aka ge .. ... .. ... .. ... .. ..... ... .. . .. ..... .. ... .. ... .. . .. .. . .. .. ... .. .. . .. ... . . ... .. ... . . .. ... .. . . . .. .... 12 C o r e Alt e r at lo n . . . .. . . . . . ,. . . . . . . . . . . .. . . . .. . . . . . . . . . . . .. . . . . . . . . .. . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 s

I Cor e Ope rating Limits R eport.............. .. .. .. ................ ......... .. .. ......... ........... ........... 12 l

Do se E quiv ale n t 1 131.. . .. ... . ... .. . ..... .. . .. . .. ... .. ....... .. . .. .. . .. . .. . .. . .. .. . .. ... . .. ... ..... . .. . .. . .... . 12 s

E- Average Disinteg ratio n E ne rg y......................................................................... 12 E ngineerod Saf ety Feaiure Response Timc............................................................ 13 F r eq u e ncy Nota tio n ..... .. .. .. ... .. . .. ... ... .. .. .... .... ... . .. .. .. .. ... .. .. . . . . .. .... . . .. .. .. .. . . ... .. . . .. . . .. . . . . 13 G aseou s Rad waste T reatment System..................................................... ................ 13 Ide n li fied Le ak a9 e... . .. .. .. ..... .. . .... . . .. ..... .. . . .. ... . ... .. .. . .. .. .... . .. .. ... . . . . ..... .. .. .. .. . .. .. .. 13 ... .. .. ..

M e m be r (s) o f th e P u blic...... . .. . .. . .. . . . . .. . . . . . .. . ... .. . .. . . . . . .. . .. . .. . .. . .. . . .. . .. . . . . .. . .. . . . . . . . . .. . .. . 13 i

i Off site Dose Calculation Manual (ODCM)............................................................. 14

Ope r a bl e Ope r abit it y. .. ... . .... . . .... .. . .. . . . .. .. . .. . .. . . .. . .. . .. . .. . . .. . .. ... ... . . .. ... ... . ... . . . ... . . . . . . ., 14 Ope r a t ion al Mode . Mooe .. ..... ..... . ...... . .. ... .. . .. .. .. . .. . .. .. .. . .. ... .. ... .. .. . ... . . . . ... .. . .. .. .. . . . ... .

14 PhysicsTests........................................................................................................ 14 Pr e ss ur e Bou nda ry L e ak a9e ... . ... .. ........ .. .... . .. .. .. ... .. .. . .. .. .. . .. . .. .. .. .. . .. . . ... .. . .. . . . . . . ... .. . 14 Proce ss Control Prog ram (PC P) .. ... ........ ..... ........ .......... . .. ... ..... .. .... ....... . .. .. .... .. .. 14 Purge. Purging..................................................................................................... 14 Qu ad r a n t Po we r Tilt R at io .. .. .. . ... .. .. .. .. . ... . .. . . .. .. .. .. .. .. .. .. . . .. . . . . . . . .. .. .. .. . . . . . . . . . . . . . . . . . . .. . 15 NORTH ANNA. UNIT 1 1 Amendment No.

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1.0 DEFINIT 60NS (Continuad) a- Capable of being closed by an OPERABLE containment automatic isolation valve system, or

b. Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Table 3.61 of Specification 3.6.3.1 1.6.2 All equipment hatches are closed and sealed, 1.6.3 Each air lock is OPERABLE pursuant to Specification 3.6.1.3, 1.6.4 The containment leakage rates are within the limits of Specification 3.6.1.2, and 1.6.5 The sealing mechanism associated with each penetration (e.g. welds, bellows or O rings) is OPERABLE.

OCNTROLLED LEAKAGE 1,7 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.

CORE ALTERATION 1.8 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuelin the vessel Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

CORE OPERATING LIMITS REPORT 1.9 The CORE OPERATING LIMITS REPORT is the unit specific document that provides core operating limits for the current operating reload cycle. These cycle specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.7. . Plant operation within these operating limits is addressed in individual specifications.

DDSE EQUIVALENT l 131 1,10 The DOSE EOUlVALENT l 131 shall be that concentration of l 131 (microcurle/ gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I 131, 1132,1133,1134 and 1135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table lli of TID 14844,

Calculation of Distance Factors for Power and Test Reactor Sites?

E AVERAGE DISINTEGRATION ENERGY 1,11 E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than lodines, with half lives greater than 15 minutes, making up at least 95% of the total non lodine activity in the coolant.

NORTH ANNA. UNIT 1 12 Amendment No.

e 1.0 DEFINITIONS (Continued) _

EhnNEERED SAFETY FEATURE RESPONSE TIME 1.12 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e.. the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.

FT:tEOLENCYNOTATICH 1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the Intervals defined in Table 1.2.

GASEOUS RADWASTE TREATMEh[T SYSTEM 1.14 A GASEOUS RADWASTE TREATMENT SYSTEM is the system designed and installed to l reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total i radioactivity prior to release to the environment. The system is composed of the waste gas decay tanks, regenerative heat exchanger, waste Oas charcoal filters, process vent blowers, waste gas surge tanks and waste gas diaphram compressor.

IDENTIFIED LEAKAGE 1.15 IDENTIFIED LEAKAGE shall be:

a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or

b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not I? be PRESSURE BOUNDARY LEAKAGE, or

c. Reactor coolant system leakage thrcugh a steam generator to the secondary system.

MEMBERIS) OF THE PUBLIC 1.16 MEMBER (S) OF THE PUBLIC shall include all individuals who by virtue of their occupational status have no formal association with the plant. This category shall include non-employees of the licensee who are permitted to use portions of the site for recreational, occupational, or other purposes not associated with plant functions. This category shall apj include non. employees such as vending machine servicemen or postman who, as part of their formal job function, occasionally enter an area that is controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, i

l NORTH ANNA UNIT 1 13 Amendment No.

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! 1.0 DEFINITIONS (Continued)

OFFSITE DOSE CALCULATION MANUAL (ODCM) 1,17 The OFFSITE DOSE CALCULATION MANUAL shall contain the current methodology and l parameters used in the calculation of offslie doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints and the specific monitoring locations of the environmental radiological monitoring program.

OPERABLE- QPERABILITY 1,18 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s), and when all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubriention or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function (s) are also capable of performing their related support function (s),

OPERATIONALMODE MODE 1,19 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination l of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.1.

PHYSICS TESTS 1.20 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related Instrumentation and 1) described in Chapter 14,0 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

PRESSURE BOLNDARYif.AKAGE 1.21 PRESSURE BCONDARY LEAKAGE shall be leakage (except steam generator tube leakage) l through a non Itolable fault in a Reactor Coolant System component body, pipe wall or vessel wall.

PROCESS CONTROL PROGRAM (PCP) 1,22 The PROCESS CONTROL PROGR' 9 shall contain the current formula, sampling, analyses, l tests and determinations to be made ' ensure that the processing and packaging of solid radioactive wastes based on demons"'ad processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Part 20,10 CFR Part 71 and Federal and State regulations and other requirements governing the disposal of the radioactive waste.

PURGE PURGING 1,23 PURGE or PURGING is the controlled process of discharging air or gas from a j confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

l NORTH ANNA - UNIT 1 14 Amendment No.

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n 1.0 DEFINITIONS (Continued) f OUADRANT POWERTILT RATIO !

1.24 OUADRANT POWER TILT RATIO shall be the ratno of the maximum upper exoore detector calibrated output to the average of the upper exoore detector calibrated outputs, or the ratio of the maximum lower exoore detector calibrated output to the average of the lower exoore detector l calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining )

I three detectors shall be used for computing the average, j l

RATEDTHERMAL POWER 1.25 RATED THERMAL POWER shall be a total reactor oore heat transfer rate to the reactor coolant of 2893 MWt. l l REACTORTRIP SYSTEM RESPONSE TME 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the tim.9 interval from when the l f monitored parameter exceeds its trip setpoint at the channel sensor until loss of stationary i gripper coil voltage.

l REPORTABLEEVENT I 1.27 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50. ;

SHUTDOWN td&BGlb[ f 1.28 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is suberitical or would be suberitical from its present condition assuming all full length rod ;

cluster assemblies (shutdown and control) are fully inserted sxcept for the single rod cluster i assembly of highest reactivity worth which is assumed to be fully withdrawn.

l SITE BOUNDARY [

1.29 The SITE BOUNDARY shall be that line beyond which the land is not owned, leased or l' otherwise controlled by the licensee. l SLAVE RELAYTEST I 1.30 A SLAVE RELAY TEST shall be the energiration of each slave relay and verification of l OPERABILITY of each relay. The SLAVE RELAY TEST shall include a continuity check, as a i minimum, of associated testable actuation devices. I SOUDIFICATION

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1.31 SOLIDIFICATION shall be the conversion of wet wastes into a solid form that meets

! shipping and burial ground requirements. i l

SOURCECECK !

1.32 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to radiation. This applies to installed radiation monitoring systems. ,

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! NORTH ANNA UNIT 1 15 Amendment No, t

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I l 1.0 DEFINITIONS (Continued) i t

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! STAGGEREDTESTRASIS ,

1.33 A STAGGERED TEST BASIS ahall oonaist of. 7 i

a. A test schedule for n systems, subsystems, trains or other designated components l obtained by dividing the specified test interval into n equal subintervals, i l :

b. The testing of one system, subsystem, train or other designated component at the {

baginning of each subinterval. ,

THERMAL PCMER i

1.34 THERMAL POWER shall be the total reactor oore heat transfer rate to the reactor coolant. !

UNIDENTIFIED LEAKAGE I i

1.35 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or l' COfRROU.ED LEMAGE.

UNRESTRICTEDAREA 1.36 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY where sooess is not controlled by the licensee for purposes of protection of individuals from exposure to i radiation and radioactive materials or any area within the SITE BOUNDARY used for residential !

quarters or for industrial, commercial, institutional, and/or recreational purposes. j VENTil.ATION EXHAUST TREATMENT SYSTEM ;

1.37 A VENTILATION EXHAUST TREATMENT SYSTEM is the system designed and installed to l ;

reduce gaseous radiolodine or radioactive material in particulate form in effluents by passing !

ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the l purpose of removing lodines or particulates from the gaseous exhaust stream prior to the l release to the environment (such a system is not considered to have any effect on noble gas l effluents). Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to i be VENTILATK)N EXHAUST TREATMENT SYSTEM components. !

VENTING 1.38 VENTING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in '

system names, does not imply a VENTING process.

i NORTH ANNA UNIT 1 16 Amendment No.

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T REACTMTYCONTROL SYSTEMS MODERATOR TEMPERATURE COEFF1CIEhri LIMmNG CONDmON FOR OPERATION 3.1.1.4 The moderator temperature coefficient (MTC) shall be within the limits specified in the CORE OPERATING LIMITS REPORT (COLR). The maximum upper limit shall be s 0.6 x 10 4 AVV'F below 70 percent RATED THERMAL POWER and s 0.0 x 10-4 AVi<rF at or above 70 percent RATED THERMAL POWER.

APPLICABfLITY: Beginning of Cycle (BOC) Limit . MODES 1 and 2' only #

End of Cycle (EOC) Limit MODES 1,2 and 3 only#

6CIIB

a. With the MTC more positive than the BOC limit specified in the CORE OPERATING LIMITS REPORT:

1. Establish and maintain control rod withdrawal limits sufficient to restore the MTC to within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

These withdrawal timits shall be in addition to the insertion limits of Specification 3.1.3.6,

2. Maintain the control rods within the withdrawal limits established above until subsequent measurement verifies that the MTC has been restored to within its limit for the all rods withdrawn condition.

3. Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 10 days, describing the value of the measured MTC, the interim control rod withdrawal limits and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition.

4. With the MTC more negative than the EOC limit specified in the COP.E OPERATING LIMITS REPORT, be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , i I

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'With Kef t 2: 1.0 [

See Special Test Exception 3.10.3 ;

I NORTH ANNA UNIT 1 3/4 16 Amendment No.

m PEACTMTY00NTRQLSYST1NIS i

NTORTSMPERAT1RE CQEFFICIENT I, SURVEILLANCE REQUIREMENTS l

L 4.1.1.4 The MTC shall be determined to be within its limits during each fuel cycle as follows:

a. The MTC shall be measured and compared to the BOC limit specified in the CORE OPERATING LIMITS REPORT. prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading. !

I b. The MTC shall be measured at any THERMAL POWER and compaled to the 300 ppm surveillance limit specified in the CORE OPERATING LIMITS REPORT (all rods ly ,

L withdrawn, RATED THERMAL POWER condition) within 7 EFPD after reaching an equilibrium boron concentration of 300 ppm. In the event this comparison indicated ,

the MTC is more negative than the 300 ppm surveillance limit, the MTC shall be !

remeasured, and compared to the EOC MTC limit specified in the CORE OPERATING l LIMITS REPORT, at least once per 14 EFPD during the remainder of the fuel 3 i cycle.(1) {

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i (1) Once the equilibr;um boron concentration (all rods withdrawn, RATED THERMAL POWER f condition) is 60 ppm or less, further ineasurement of the MTC in accordance with :

4.1.1.4.b may be suspended providing that the measured MTC at an equilibrium boron concentration of 5 60 ppm is less negative than the 60 ppm surveillance limit specified in

the CORE OPERATING UMITS REPORT.

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RSCTMTY00MTROL SYSTEMS i

! LIMITING CONDITION FOR OPERATION (Continued)

b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

I c) A power distribution map is obtained from the movable incore detectors and h

Fo(Z) and AF"H are verified to be within their limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or d) Either the THERMAL POWER levelis reduced to s 75% of RATED THERMAL POWER within one hour and within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the high neutron flux trip setpoint is reduced to s 85% of RATED THERMAL POWER, or e) The remainder of the fods in the group with the inoperable rod are aligned to within i 12 steps of the inoperable rod within the hour while maintaining the rod sequence and insertion limits of Specification 3.1.3.6. l The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation.

SURVEILLANCE REOUIREMENTS 4.1.3.1.1 The position of each fulllength rod shall be determined to be within the group demand limit by verifying the individual rod positions at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the Rod Position Deviation Monitor is inoperable, then verify the group positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4.1.3.1.2 Each full length 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 31 days.

NORTH ANNA UNIT 1 3/4 1 19 Amendment No.

mwTwvoouunsysTen SHUTDOWN ROD INSERTION LIMIT

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f LIMfTNG CONDfTION FOR OPERATK)N 3.1.3.5 All shutdown rods shall be limited in physical insertion as specified in the CORE OPERATING LIMITS REPORT. ;

APPLICABILITY: MODES 1' and 2'# ;

ACTICH:

With a maximum of one shutdown rod inserted beyond the insertion limit specified in the CORE OPERATING LIMITS REPORT, except for surveillance testing pursuant to Specification 4.1.3.1.2, within one hout either:

a. Restore the rod to within the Insertion limit specified in the CORE OPERATING LIMITS l REPORT, or r

b. Declare the rod to be inoperable and apply Specification 3.1.3.1 l i

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SURVEILLANCE REQUIREMENTS i

4.1.3.5 Each shutdown rod shall be determined to be within the insertion limit specified in the }

CORE OPERATING LIMITS REPORT i t

a. Within 15 minutes prior to withdrawal of any rods in control banks A, B, C or D !

during an approach to reactor criticality, and l t

b. At least onca per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter. !

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Te~e 5pecIal TEt 5Eceptions 5.10.2 and 3.10.3. :'

With K,gg 21.0 NORTH ANNA UNIT 1 3/4 1 24 Amendment No.

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(( c REACTMTY 00WTROL SYSTEMS QCWTROL RQDINSERTION LIMITS ,

LIMITING CONDITION FOR OPERATION 3.1.3.6 The control banks shall be limited in physical insertion as specified in the CORE ,

OPERATING LIMITS REPORT. , . t APPLICABILITY: MODES 1* AND 2'#, l ACTICH:

With the control banks inserted beyond the insertion limits, except for surveillance testing pursuant to Specification 4.1.3.1.2, either:

a. Restore the control banks to within the limits within two hours, or l t

b. Reduce THERMAL POWER within two hours to less than or equal to that fraction of l RATED THERMAL POWER which is allowed by the group position using the insertion I

limits specified in the CORE OPERATING LIMITS REPORT, or

c. Be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . ,

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SURVEILLANCE REO'JIREMENTS ,

t 4.1.3,6 The position of each control bank shall be determined to be within the insertion limits !

at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the Rod insertion Limit Monitor is i inoperable, then verify the individual rod positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . i i

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'See Special Test Exceptions 3.10.2 and 3.10.3.

With K egg h 1.0 l I

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NORTH ANNA UNIT 1 3/4 1 25 Amendment No.

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NORTH ANNA UNIT 1 3/4 1 27 Amendment No.

I 1/L2 POWER DlSTRIBUTlON LIMITS i

[ AX1AL FLUX DIFFERENCE fAFD)

UMITING OONDITION FOR OPERATION 3.2.1 The indicated AXtAL FLUX DIFFERENCE (AFD) Shall be maintained within the limits spoolfed in the CORE OPERATING UMITS REPORT.

APPUCAktuTY: MODE 1 ABOVE 50% RATED THERMAL POWER l o.

ACTON,

a. With the indicated AXIAL FLUX DIFFERENCE outside of the limits specified in the CORE OPERATING UMITS REPORT, 4

1. Elther restore the indicated AFD to within the limits within 15 minutes, or i 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux . High Trip setpoints to less than or equal to 55 percent of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ,

b. THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the Indicated AFD is within the limits specified in the CORE OPERATING LIMITS REPORT.

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l NORTH ANNA UNIT 1 3/4 21 Amendmont No.

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POWER Dl3TRIBUTION LIMITS l

LIMITING CONDITION FOR OPERATION (Continued)

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SURVEILLANCE REQUIREMENTS ;

4.2.1.1 The indicated AXlAL FLUX DIFFERENCE shall be determined to be within its limits during POWER OPERATION above 50% of RATED THERMAL POWER by:

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a. Monitoring the indicated AFD for each OPERABLE exoore channel:

1.. At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and i

2. At least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after restoring the AFD Monitor l Alarm to OPERABLE status. ,

b. Mon'ioring and logging the indicated AXIAL FLUX DIFFERENCE for each OPERABLE l exoore channel at least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least once per 30 ;

minutes thereafter, when the AXlAL FLUX DIFFERENCE Monitor Alarm is incperable, i The logged values of the indicated AXIAL FLUX DIFFERENCE shall be assumed to exist ,

during the interval preceding each logging, u .

4.2.1.2 The indicated AFD shall be considered outside of its limit when at least 2 OPERABLE '

exoore channels are indicating the AFD to be outside of the limits specified in the CORE l OPERATING LIMITS REPORT. l l l

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l NORTH ANNA UNIT 1 3/4 22 Amendment No. j

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POWER DISTRIBUTION LIMITS

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I HEAT FIIIX HOT CHANNFl FACTOR-Fp o

e UMITING CONDITION FOR OPERATION 3.2.2 Fo(Z) shall be limited by the following relationships; Fo(Z) s ( f) [K(2)] for P > 0.5 Fo(Z) s ( ) [K(Z)] for P s 0.5 where CFO = the Fo limit at RATED THERMAL POWER specified in the CORE OPERATING LIMITS REPORT, THERMAL POWER P = RATED THERMAL POWER , and K(Z) = the normalized Fo limit as a function of core height specified in the CORE OPERATING UMITS REPORT.

APPLICABILITY: MODE 1.

ACTION:

With Fo(Z) exceeding its limit:

a. Reduce THERMAL POWER at least 1% for each 1% Fo(Z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux.High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION may proceed provided the Overpower AT Trip Setpoint (value of K4) has been reduced at least 1% (in AT span) for each 1% Fo(Z) exceeds the limit,

b. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above the reduced limit required by a, above; THERMAL POWER L may then be increased provided Fo(Z) is demonstrated through incore mapping to be 1 within its limits.

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l NORTH ANNA- UNIT 1 3/4 2-5 Amendment No.

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( POWER DISTRIBUTION LIMITS

' SURVEILLANCE REQUIREMENTS , 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 Fo(Z) shall be evaluated to determine if Fo(Z) is within its limit by:

a. Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER Oreater than 5% of RATED THERMAL POWER.

b. Increasing the measured Fo(Z) component of the power distribution map by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties.

c. Satisfying the following relationship:

for P > 0.5 Fo(z) s PxN .

F (z) s z) 0.5 for P s 0.5 where Fo(z)is the measured Fo(Z) increased by the allowances for manufacturing tolerances and measurement uncertainty, and N(z) is the cycle dependent fanction that accounts for power distribution transients encountered during normal operation. This function is specified in the CORE OPERATING LIMITS REPORT as por :

Specification 6.9.1.7.

d. Measuring F (z) according to the following schedule:

1. Upon achieving equilibrium conditions after exceeding the THERMAL POWER at >

which Fo(Z) was last determined by 10% or more of RATED THERMAL POWER *, ,

.or (

2. At least once per 31 effective full power days, whichever occurs first.

e. With measurements indicating j maximum FO (z)'

over z ( K(z) j has increased since the previous determination of F (z) either of the following actions shall be taken:

During power escalation, the power level may be increased until a power level for extended operation has been achieved and a power distribution map obtained.

NORTH ANNA UNIT 1 3/4 2-6 Amendment No.

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POWER DISTRIBUTION LIMITS ,

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l SURVEILLANCE REQUIREMENTS (Continued) ;

i t I 1. F (z) o shall be increased by 2% over that specified in 4.2.2.2.c, or i

2. F (z)

O shall be measured at least once per 7 effective full power days until 2 [

successive inaps indicate that j maximum FO (z)' f is not increasing. ;

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over2 (K(z)s i :

f. With the relationships specified in 4.2.2.2.c above not being satisfied:

1. Calculate the percent Fo(Z) exceeds its limit by subtracting one from the ,

measurement /llmit ratio and multiplying by 100: !

maximum b# '

x 100 for P 2 0.5

$ , y p ;; CFO x K(z) >

( P x N(z) ) -

pOM(z) f T maximum *

x 100 for P < 0.5 over z CFO x K(z)

(0.5 x N(z)) .

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2. Elther of the following actions shall be taken: I

a. Power operation may continue provided the AFD limits of Specification 3.2.1 l are reduced 1% AFD for each percent Fo(Z) exceeded its limits, or i

b. Comply with the requirements of Specification 3.2.2 for Fo(Z) exceeding its :

limit by the percent calculated above. l t

g. The limits specified in 4.2.2.2.c 4.2.2.2.0, and 4.2.2.2.f above are not applicable in the following core plane regions:

4

1. Lower core region 0 to 15 percent inclusive.

2. Upper core region 85 to 100 percent inclusive.

4.2.2.3 When Fo(Z) is measured for reasons other than meeting the requirements of Specification 4.2.2.2, an overall measured FO(Z) shall be obtained from a power r distribution map and increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty.

NORTH ANNA - UNIT 1 3/4 27 Amendment No.

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i NORTH ANNA- UNIT 1 3/4 28 Amendment No.

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h~ POWER DISTRIBUTION LIMITS

< - ' Ntrl FAR ENTHALPY HOT CHANNEL FACTOR - ._,

i UMITING CONDITION FOR OPERATION i

3.2.3. (H shall be limited by the following relationship:

H s CFDH [1 + PFDH (1 P)}

where CFDH - The F AH limit at RATED THERMAL POWER specified in the CORE OPERATING LIMITS REPORT, THERMAL POWER P = RATED THERMAL POWER L

PFDH - The Power Factor Multiplier for F 3y specified in the CORE OPERATING LIMITS REPORT,and l

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F H - measured value of F H obtained by using the movable incore detectors to obtain a power distribution map.

! APPLICABILITY: MODE 1 ACTION: ,

. With FAH exceeding its limit

a. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip Setpoints to s 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ,-

b. Demonstrate through in core mapping that F H is within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and

c. Identify and correct the cause of the out of limit condition prior to increasing

!' THERMAL POWER above the reduced limit required by a or b, above; subsequent j

~

"-POWER OPERATION may proceed provided othat F"H ~ls demonstrated through in core !

mapping to be within its limit at a nominal 50% of RATED THERMAL POWER prior to l exceeding this THERMAL POWER, at a nominal 75% of RATED THERMAL POWER L prior to exceeding this THERMAL POWER and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after attalning 95%

or greater RATED THERMAL POWER.

NORTH ANNA UNIT 1 3/4 29 Amendment No.

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. i T4.9 REFUELING OPERATIONS j i

BORONCGICENTRATION I UMITING CONDITION FOR OPERATION :

3.9.1 With the reactor vessel head unbolted or removed, the boron concentration of all fliled l portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive of the following reactivity conditions is met: l

a. Elther a Keff of 0.95 or less, or ,

b. A boron concentration limit as specified in the Core Operating Limits Report. !

APPLICABILITY: MODE 6',

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all C operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue :

boration at 210 ppm of 212,950 ppm boric acid solution or its equivalent until Keft i s reduced to s 0.95 or the boron concentration is restored to 2 2300 ppm, whichever is the more restrictive. The provisions of Specification 3.0.3 are not applicable.

?

SURVEILLANCE REQUIREMENTS 4.9.1.1 The more restrictivo of the above two reactivity conditions shall be determined prior to:

a. Removing or unbotting the reactor vessel head, and 5

b. Withdrawal of any full length control rod located within the reactor pressure vessel, in excess of 3 feet from its fully inserted position. 1 4.9.1.2 The boron concentration of the reactor coolant system and the refueling canal shall be determined by chemical analysis at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . k

The reactor shall be maintained in MODE 6 when the reactor vessel head is unbolted or removed.

NORTH ANNA - UNIT 1 3/4 91 Amendment No,

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i ADMINISTRATIVE CONTROLS t

h' CORE OPERATINGIJMITS REPORT

! 6.9.1.7.a Core operating limits shall be established and documented in the CORE OPERATING f LIMITS REPORT before each reload cycle or any remaining part of a reload cycle i for the following:

1. Moderator Temperature Coefficient BOC and EOC limits, and 300 ppm and 60 ppm surveillance limits for Specification 3/4.1.1,4,

2. Shutdown Rod Insertion Limit for Specification 3/4.1.3.5,

3. Control Rod Insertion Limits for Specification 3/4.1.3.6,

4. Axlal Flux Difference limits for Speellication 3/4.2.1,

5. Heat Flux Hot Channel Factor, K(Z), N(Z) for Specification 3/4.2.2, and

6. Nuclear Enthalpy Rise Hot Channel Factor, and Power Factor Multiplier,

, for Specification 3/4.2.3.

7. Boron Concentration for Specification 3/4.9.1.

6.9.1.7.b The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC and identified in the Core -

Operating Limits Report.

The core operating limits shall be determined so that all applicable limits (e.g.,

fuel thermal mechanical limits, core thermal hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met. The CORE OPERATING LIMITS REPORT, ,

! including any mid cycle revisions or supplements thereto, shall be provided i upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

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4 NORTH ANNA UNIT 1 6-17 Amendment No. ;

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l o-3/4.1 REACTMTY CONTROL SYSTEMS r

BAcES .

3/4.1.1.4 MODERATOR TEMPERATURE COEFFICIENT (MTC) (Continued) conditions other than those explicitly stated will require extrapolation to those conditions in order to permit an accurate comparison.

The most negative MTC value was obtained by incrementally correcting the MTC used in the FSAR analyses to nominal operating conditions. These corrections involved adding the incremental change in the MTC associated with a core condition of Bank D inserted to an all rods withdrawn condition and an incremental change in MTC to account for measurement uncertainty at RATED THERMAL POWER conditions. These corrections result in the End of Cycle (EOC) MTC limit. The 300 ppm surveillance limit MTC value represents a conservative value (with corrections for bumup and soluble boron) at a core condition of 300 ppm equilibrium boron concentration and is obtained by making these corrections to the EOC MTC limit, a

Once the equilibrium boron concentration falls below about 60 ppm, diiution operations take an extended amount of time and reliable MTC measurements become more difficult to obtain due to the potential for fluctuating core conditions over the test interval. For this reason, MTC measurements may be suspended provided the measured MTC value at an equilibrium full power boron concentration s 60 ppm is less negative than the 60 ppm surveillance limit. The difference between this value and the EOC MTC limit conservatively bounds the maximum credible change in MTC between the 60 ppm equilibrium boron concentration (all rods withdrawn, RATED THERMAL POWER conditions) and the licensed end of cycle, including the effect of boron concentration, burnup, and end of cycle coastdown.

The surveillance requirements for measurement of the MTC at the beginning and near the end of each fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.

l 3/4.1.1.5 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Rea: tor i Coolant System average temperature less than 541*F. This limitation is required to ensure 1) l the moderator temperature coefficient is within its analyzed temperature range,2) the protective instrumentation is within its normal operating range, and 3) the P 12 Interlock is i above its setpoint, and 4) compliance with Appendix G to 10 CFR Part 50 (see Bases 3/4.4.9).

3/4.1.2 BORATION SYSTEMS i i The boron injection system ensures that negative reactivity control is available during ;

each mode of facility operation. The components required to perform this function include 1) borated water sources,2) charging pumps,3) separate flow paths,4) boric acid transfer l pumps,5) associated heat tracing systems, and 6) an emergency power supply from OPERABLE diesel generators. ,

NORTH ANNA - UNIT 1 B 3/4 12 Amendment No.

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POWER DISTRIBUTION LIMITS BASES 3!422 and 3/42.3 HEAT FLUX AND NUCLEAR ENTHALPY HOT CHANNEL FACTORS-g(M and F The limits on ' heat flux and nuclear enthalpy hot channel factors ensure that 1) the design limits on peak local power density and minimum DNBR are not exceeded and 2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200*F ECCS acceptance criterla limit.

Each of these hot chtnnel factors are measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance Is sufficient to insure that the hot channel factor limits are maintained provided:

a. Control rod in a single group move together with no individual rod insertion differing by more than i 12 steps from the group demand position.

b. Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6.

c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are ;

maintained. ;

d. The axial power distribution, expressed in terms of AXlAL FLUX DIFFERENCE, is maintained within the limits.

r The relaxation in F H as a function of THERMAL POWER allows changes in the radial N '

power shape for all permissible rod insertion limits. FAH will be maintained within its limits provided conditions a thru d above, are maintained.

When an Fo measurement is taken, both experimental error and manufacturing tolerance must be allowed for. 5% is the appropriate allowance for a full core map taken with r the incore detector flux mapping system and 3% is the appropriate allowance for manufacturing tolerance. ,

The specified limit for FAH contains a 4% error allowance.' Normal operation will l result in a measured F H less than or equal to the limit specified in the CORE OPERATING [

1 LIMITS REPORT. The 4% allowance is based on the following considerations:

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NORTH ANNA UNIT 1 B 3/4 2-4 Amendment No.

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POWER DISTRIBtJTION LIMITS !

BASES

a. abnormal perturbations in the radial power shape, such as from rod misalignment, :

L N

effect FAH more directly than Fo,

b. although rod movement has a direct influence upon limiting Fo to within its limit, such control is not readily available to limit F H,and !

c. errors in prediction for control power shape detected during startup physics tests can be compensated for in Fo by restricting axial flux distributions. This compensation for F H si less readily available.

Fuel rod bowing reduces the value of the DNB ratio. Credit is available to offset this i reduction in the margin avaltable between the safety analysis design DNBR values (1.46 for ;

Vieginia Electric and Power Company statistical methods) and the limiting design DNBR valuo (1.26 for Virginia Electric and Power Company statistical methods). A discussion of the rod bow penalty is presented in the FSAR.

The hot channel factor Fo(Z) is measured periodically and 'rncreased by a cycle and

' height dependent power factor, N(Z), to provide assurance that the limit on the hot channe, factor, Fo (Z), is met. N(Z) accounts for the non-equilibrium effects of normal operation transients and was determined from expected power control maneuvers over the full range of i burnup conditions in the core. The N(Z) function for normal operation is specified in the CORE OPERATING LIMITS REPORT per Specification 6.9.1.7. .

3/4.2.4 OUADRANT POWER TILT RATIO The quadrant power till ratio limit assures that the radial power distribution satisfies .

. the design values used in the power capability analysis. Radial power distribution measurements are mado during startup testing and periodically during power operation, ,

i The limit of 1.02 at which corrective action is required provides DNB and linear heat >

generation rate protection with x y plane power tilts.

The two hour time allowance for operation with a till condition greater than 1.02 but I less than 1.09 is provided to allow identification and correction of a dropped or misaligned rod.

In the event such action does not correct the tilt, the margin for uncertainty on Fo is reinstated by reducing the power by 3 percent for each percent of tilt in excess of 1.0.

For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the moveable Incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the OUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore flux map or two sets of 4 symmetric thimbios. The two sets of 4 symmetric thimbles is a unique set of 8 detector locations. These locations are C 8, E 5, E 11, H 3, H 13, L 5, L 11 and N 8 NORTH ANNA UNIT 1 B 3/4 2 5 Amendment No.

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Attachment 2 ]

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Proposed Technical Specification Change - !

North Anna Unit 2 f 3

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DEFINITIONS o

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! 1.0 DEFINITIONS

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Action...................................................................................................................... 11 Ax lal F lu x Ditfe re nee . . . .. . .. .. . .. . .. . .. ..... . .. ... . ..... ... .. . .. .. . .. . .. . . . . .. .. . .. . .. . .. .. . .. . .. . .. . . . . . . . . . .. . . 11 C h a n n el C alibr a tio n. . . . .. . . ... .. . . .. ....... .. . .. .. .. . . . .. .. .. . .. .. .. .. . . . .. . . . .. .. .. . . . . . .. .. . . . .. . . . . . . . .. .. .. . .. . 11

! ChannelCheck......................................................................................................... 11

' C h an n el Fu nction al T e st. .. ... .. .. ... .. .. . .. .. .... ... .. .. ... .. .. .. . .. .. . . . .. .. .. .. . .. . . ... .. . . ... .... . .. . . . . .. .. . 1-1 i' C o nt ain me nt I n t eg rit y. . .. . .. .. . . . . . .. . .. . .. . . . . .. . .. ... . . .. . . . . .. . .. . . . . . . . . . . . . . .. . .. . .. . .. . .. . . . . . . . . . . . .. . .. . . 11 C o n t r olle d Le ak a g e . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 l'

C o r e Al t e r a t io n . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 r Core Ope rating Limits Report.. ..... ................ ............... ..... .... .. ......... ... .. .... ............. 1-2 D o se E qu lv ale n t 1 131.. . .. ... . .. . .. . . . .. . .. . .. . .. . .. .. . . . . .. . .. . .. . . . .. . . . . .. ... . .. . . . .. . .. . . . . .. . .. .. . . . . .. . .. 1 2-E. Ave rage Disinteg ratlon En e rgy....... ...................................... ...... .... ...... .. . .. ... ... . 12 i

l Engineered Safety Feature Response Time............................... ............................ 1-3 s

l L Fr eq u e ncy No ta tlo n. . . .. . . .. . . .. .. .. .. ... . . . . . . .. . . . . . . .. . . .. . . .. .. .. ... . . . . . . .. . ... .. .. .. . . .. . . . . . . . .. .. . . .. .. . . . . 13 1'

Gaseou s Radwaste Treatment Syste m..................................................................... 13 o

! 1-3 I d e n t i f l ed Le a k ag e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

M e m be r(s) o f t h e P u blic .. . . . . . . . . .. . . . . . . .. . . . .. . . . . .. . . . . .. . . .. . . . . . . . . . . .. . .. . . . . . . . . . . . .. . . . . . . . . . . . . . . .. . 13 l

l Off site Dose C alculation Manual (ODCM)............................................. ................ 14 Ope rable Ope r ability. . . . .. . . . . . . . . . .. .. . .. . . . . . . . . .. . . . . . .. . .. . . . . .. . . . . . .. . . . . . . . . . . . .. . .. . . . . . . . . .. . . . . . . . . 1-4 Ope ra tio n al Mo de - Mode .. . . . . . . . . . . .. . .. .. . . . . . . . . ... . . .. . .. . .. . . . . . .. . .. . . ... . . . . . . . ... . . . . . . . . . . . . . .. . . . . . . 1-4 l 4

PhysIcsTests........................................................................................................ 1-4 P re s s u re Bo u nd ary Le a kage . . .. . . .. . . . . .. . . . . . . . . . .. . . . .. . .. . . . .. . . . . .. . .. .. . .. . . . . . . . . .. . . . .. . .. .. . . . . . . . . 1-4 Proce ss Control Prog ra m ( PC P) .. .. .. . .... . .. .. . .. . .. ..... . .. .. ... . .. .. .... .... . ... .. .. . .. . .. .. . .. ... .. 1-4 Purge-Purging.................................................................................................... 1-4 -7 O u ad r an t Po we r Tilt R atio . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 NORTH ANNA - UNIT 2 1 Amendment No.

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, 1.0 DEFINITIONS (Continued) l a. Capable of being closed by an OPERABLE containment automatic isolation valve

, system, or l 1

[ b. Closed by manual valves, blind flanges, or deactivated automallo valves ;

secured in their closed positions, except as provided in Table 3.61 - of :

Specification 3.6.3.1,

[

1.6.2 All equipment hatches are closed and sealed, 1.6,3 Each air lock is OPERABLE pursuant to Specification 3.6.1.3, 1.6.4 The containment leakage rates are within the limits of Specification 3.6.1.2, and l l

1.6.5 The sealing mechanism associated with each penetration (e.g. welds, bellows or ,

0 rings) is OPERABLE.

CONTROLED LEAKAGE +

1.7 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals. ,

COREALTERATION 1.8 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE l ALTERATION shall not preclude completion of movement of a component to a safe conservative ;

position. ,

fpRE OPERATING LIMITS REPORT 1.9 The CORE OPERATING LIMITS REPORT is the unit specific document that provides core '

operating limits for the current operating reload cycle. These cycle-specific core operating :

limits shall be determined for each reload cycle in accordance with Specification 6.9.1.7. Plant operation within these operating limits is addressed in Individual specifications, i .

(

DOSE EQUIVALENT l 131 l ,

1.10 The DOSE EQUIVALENT l 131 shall be that concentration of I-131 (microcurlo/ gram) l which alone would produce the same thyrold dose as the quantity and isotopic mixture of I 131, 1132,1133,1134 and 1135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table ll1 of TID 14844, " Calculation of Distance Factors for Power and Test Reactor Sites."

E. AVERAGE DISINTEGRATION ENERGY ;

1.11 E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than lodines, with half lives greater than 15 minutes, making up at least 95% of the total non lodine activity in the coolant.

NORTH ANNA UNIT 2 12 Amendment No.

1.0 DEFINITIONS (Continued) l ENGINEERED SAFETY FEATURE RESPONSE TIME 1.12 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from l when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the I ESF equipment is capable of performing its safety function (i.e., the valvec travel to their q required positions, pump discharge pressures reach their required values, etc.). Times shall :

Include diesel generator starting and sequence loading delays where applicable. l FREQUENCYNOTATION 1.13 The FREQUENCY NOTATION specified fN the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

t GASEOUS RADWASTETREATMENT SYSTEM 1.14 A GASEOUS RADWASTE TREATMENT SYSTEM is the system designed and installed to '

reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment. The system is composed of the waste gas decay >

tanks, regenerative heat exchanger, waste gas charcoal filters, process vent blowers, waste gas surge tanks and waste gas diaphram compressor.

IDENTIFIED LEAKAGE 1.15 IDENTIFIED LEAKAGE shall be:

a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or ;

valve packing leaks that are captured and conducted to a sump or collecting tank, or

b. Leakage into the containment atmosphere from sources that are both specifically

( located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or l

c. Reactor coolant system leakage through a steam generator to the secondary system.

MEMBER (S) OF THE PUBUC ;

1.16 MEMBER (S) OF THE PUBLIC shall include all individuals who by virtue of their l occupational status have no formal association with the plant. This category shall include non-employees of the licensee who are permitted to use portions of the site for recreational, L occupational, or other purposes not associated with plant functions. This category shall D.03 I

include non-employees such as vending machine servicemen or postman who, as part of their formal job function, occasionally enter an area that is controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materlats.

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NORTH ANNA - UNIT 2 1-3 Amendment No.

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i 1.0 DEFINITIONS (Continued) i OFFSITE DOSE CALCULATION MANUAL IODCM) 1.17 The OFFSITE DOSE CALCULATION MANUAL shall contain the current methodology and l parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints and the specific monitoring locations of the environmental radiological monitoring program.

OPERABLE - OPERABILITY 1.18 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s), and when all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxillary equipment that are required for the system, subsystem, train, component, or device to perform its function (s) are also capable of performing their related support function (s).

OPERATIONAL MODE - MODE 1.19 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination l of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.1.

PHYSICSTESTS 1.20 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear .

characteristics of the reactor core and related instrumentation and 1) described in Chapter 14.0 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise

approved by the Commission.

PRESSURE BOUNDARY LEAKAGE 1.21 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tubo leakage) l through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall.

l PROCESS CONTROL PROGRAM (PCP) l 1.22 The PROCESS CONTROL PROGRAM shall contain the current formula, samplir.g, analyses, tests and determinations to be made to ensure that the processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will j l

be accomplished in such a way as to assure compliance with 10 CFR Part 20,10 CFR Part 71 l l and Federal and State regulations and other requirements governing the disposal of the !

radioactive waste.

l PURGE PURGING ,

1.23 PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement, i

l.

NORTH ANNA - UNIT 2 1-4 Amendment No.

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,' 1.0 DEFINITIONS (Continued)

OUADRANT POWERTILT RATIO !

1.24 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper exoore detector i calibrated output to the avera0e of the upper excore detector calibrated outputs, or the ratio of .

the maximum lower exoore detector calibrated output to the average of the lower excore detector '

calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average. !

RATEDTHERMAL POWER !

1.25 RATED THERMAL POWER shall be & total reactor core heat transfer rate to the reactor coolant of 2893 MWl. l REACTOR TRIP SYSTEM RESPONSE TIME 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the l ,

monitored parameter exceeds its trip setpoint at the channel sensor until loss of stationary -

gripper coil voltage.

FIEPORTA1:lLE EVENT 1.27 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 l CFR Part 50. I SHUTDOWN MARGIN i 1.28 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be suberitical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

SITE BOUNDARY 1,29 The SITE BOUNDARY shall be that line beyond which the land is not owned, leased or

otherwise controlled by the licensee.

l SLAVE RELAYTEST 1.30 A SLAVE RELAY TEST shall be the energization of each slave relay and verification of OPERABILITY of each relay. The SLAVE RELAY TEST shall include a continuity check, as a minimum, of associated testable actuation devices.

SOUDIFICATION -

1.31 SOLIDIFICATION shall be the convers! .o of wet wastes into a solid form that meets shipping and burial ground requirements.

l SOURCECECK 1

L 1.32 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to radiation. This applies to installed radiation monitoring systems.

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L NORTH ANNA- UNIT 2 15 Amendment No.

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).

1.0 DEFINITIONS (Continued)

STAGGEREDTESTBASIS 1.33 A STAGGERED TEST BASIS shall consist of:

a. A test schedule for n systems, subsystems, trains or other designated components p obtained by dividing the specified test interval into n equal subintervals,

b. The testing of one system, subsystem, train or other designated component at the L,eginning of each subinterval.

THERMAL POWER 1.34 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

UNIDENTIFIED LEAKAGE u

K 1.35 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLEDLEAKAGE.

UNRESTRICTED AREA 1.36 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY where access is not controlled by the licensee for purposes of protection of Indivic'uals from exposure to radiation and radioactive materials or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.

VENTILATION EXHAUST TREATMENT SYSTEM 1.37 A VENTILATION EXHAUST TREATMENT SYSTEM is the system designed and Installed to reduce gaseous radiolodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing lodines or particulates from the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents). Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

VEhmNG 1.38 VENTING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such i a manner that replacement air or gas is not provided or required during VENTING. Vent, used in I system names, does not imply a VENTING process.

9 NORTH ANNA - UNIT 2 1-6 Amendment No.

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.. 'I REACTMTYQQNTROLSYSTEMS i MODERATORTEMPERATURE COEFFICIENT I

UMtTING CONDITION FOR OPERATION l

I -

3.1.1.4 The moderator temperature coefficient (MTC) shall be within the limits specified in the CORE OPERATING LIMITS REPORT (COLR). The maximum upper limit shall be 5 0.6 x 10 4 Ak/krF below 70 percent RATED THERMAL POWER and s 0.0 x 10-4 Ak/kPF at or above 70 percent RATED THERMAL POWER.

l-APPLICABILITY: Beginning of Cycle (BOC) Limit . MODES 1 and 2* only # ;

End of Cycle (EOC) Limit - MODES 1,2 and 3 only# :

I

a. With the MTC more positive than the BOC limit specified in the CORE OPERATING LIMITS REPORT, operations in MODES 1 and 2 may proceed provided: ,

1. Control rod withdrawal limits are established and maintained sufficient to restore ,

the MTC to within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY within the next ,

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . These withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6. L

2. The control rods are maintained within the withdrawal limits established above

until subsequent measurement verifies that the MTC has been restored to within its -

limit for the all rods withdrawn condition.

3. Prepare and submit a Special Report to the Commission pursuant to Specification i 6.9.2 within 10 days, describing the value of the measured MTC, the interim control rod withdrawal limits and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition.

4. With the MTC more negative than the EOC limit specified in the CORE OPERATING LIMITS REPORT, be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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With Keft 21.0

See Special Test Exception 3.10.3 NORTH ANNA UNIT 2 3/4 15 Amendment No.

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3 REACTIVITYOONTROL SYSTEMS

-]

MODERATORTEMPERATURE COEFFICIENT 3 i

SURVEILLANCE REQUIREMENTS __

i

'4.1.1.4

The MTC shall be determined to be within its limits during each fuel cycle as follows: ,

s. The MTC shall be measured and compared to the BOC limit specified in the CORE- ,

OPERATING LIMITS REPORT, prior to initial operation above 5% of RATED THERMAL .

POWER, after each fuel loading. .

b. The MTC shall be measured at any THERMAL POWER and compared to the 300 ppm surveillance limit specified in the CORE OPERATING LIMITS REPORT (all rods l '

{

withdrawn, RATED THERMAL POWER condition) within 7 EFPD after reaching an '

equilibrium boron concentration of 300 ppm. In the event this comparison indicated the MTC is more negative than the 300 ppm surveillance limit, the MTC shall be ,

remeasured, and compared to the EOC MTC limit specified in the CORE OPERATING LIMITS REPORT, at least once per 14 EFPD during the remainder of the fuel '

cycle.(1) ;

t t

F 1

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l.

(1 ) Once the equilibrium boron concentration (all rods withdrawn, RATED THERMAL POWER condition) is 60 ppm or less, further measurement of the MTC in accordance with 4.1.1.4.b may be suspended providing that the measured MTC at an equilibrium boron l concentration of s 60 ppm is less negative than the 60 ppm surveillance limit specified in the CORE OPERATING LIMITS REPORT.

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NORTH ANNA - UNIT 2 3/4 1-6 Amendment No.

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F FtEACTIVITYCONTROLSYSTEMS UMITING CONDITION FOR OPERATION (Continued) 1 b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , c) A power distribution map is obtained from the movable incore detectors and Fo(Z) and F H are verified to be within their limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , c d) Either:

1) The THERMAL POWER levelis reduced to s 75% of RATED THERMAL' POWER within one hour and within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the high neutron flux trip setpoint is reduced to is less than or equal to 85% of RATED THERMAL POWER,or

2) 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 rod sequence and insertion limits of Specification 3.1.3.6. The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation.

-l SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each full length rod shall be determined to be within the group demand limit by verifying the Individual rod positions at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the Rod Position Dev5 tion Monitor is inoperable, then verify the group positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

L 4.1.3.1.2 ' Each rod not fully inserted shall be determined to be OPERABLE by movement of at least 10 steps in any one direction at least once per 31 days.

1

NORTH ANNA UNIT 2 3/4 1-17 Amendment No.

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[ REACTMTYCONTROLSYSTEMS

< i SHUTDOWN ROD INSERTION LIMIT UMITING CONDITION FOR OPERATION l i

c r L. - 3.1.3.5 All shutdown rods shall;be limited in physical insertion.as specified in the CORE !

OPERATING UMITS REPORT.

APPLICABILITY: MODES 1* and 2'# j ACTION: ;

.l With a maximum of one shutdown rod inserted beyond the insertion limit specified in the CORE -t

- OPERATING LIMITS REPORT, except for surveillance testing purt,uant to Specification v 4.1.3.1.2, within one hour either: .

l

~l

a. Restore the rod to within the Insertion limit specified in the CORE OPERATING LIMITS REPORT, or :

b. Declare the rod to be Inoperable and apply Specification 3.1.3.1 ;

, SURVEILLANCE REQUIREMENTS 4.1.3.5 Each shutdown rod shall be determined to be within the insertion limit specified in the CORE OPERATING UMITS REPORT s

a. Within 15 minutes prior to withdrawal of any rods in control banks A, B, C or D during an approach to reactor criticality, and

b. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

1

~

'See Special Test Exceptions 3.10.2 and 3.10.3. )

With Koff graater than or equal to 1.0 .

)

NORTH ANNA UNIT 2 3/4 1 22 Amendment No.

i

l REACTMTYCONTROLSYSTEMS !

CONTROL RODINSERTION LIMITS I UMITING CONDITION FOR OPERATION ,

l 3.1.3.6' The control banks shall be limited in physical insertion as-specified in the CORE i OPERATING LIMITS REPORT.

APPLICABil.LTX: MODES 1* AND 2*#.

i ACTION:

L .With the control banks inserted beyond the Insertion limits, except for surveillance testing ~

pursuant to Specification 4.1.3.1.2, either: l

+

a. Restore the control banks to within the limits within two hours, or l I b. Reduce THERMAL POWER within two hours to less than or equal to that fraction of ~!

RATED THERMAL POWER which is allowed by the group position using the insertion limits specified in the CORE OPERATING LIMITS REPORT, or

c. Be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . _

r .

SURVEILLANCE REQUIREMENTS 1

i 4.1.3.6 The position of each control bank shall be determined to be within the insertion limits .l at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the Rod insertion Limit Monitor is inoperable, then verify the individual rod positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

i 4

See Special Test Exceptions 3.10.2 and 3.10.3. y

With Keff greater than or equal to 1.0 NORTH ANNA UNIT 2 3/4 1-23 Amendment No.

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NORTH ANNA UNIT 2 3/4 1 24 Amendment No. l l

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NORTH ANNA UNIT 2 3/4 1 25 Amendment No, t 9

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4 3/4.2 POWER DISTRIBUTION LIMITS

5 _

AXIAL FLUX DIFFERENCE (AFD)

UMITING CONDITION FOR OPERATION I

3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the limits ,

i,ie specified in the CORE OPERATING UMITS REPORT. :

APPLICABILITY: MODE 1 ABOVE 50% RATED THERMAL POWER .

ACTION:

a. With the indicated AXIAL FLUX DIFFERENCE outside of the limits specified in the CORE -

OPERATING UMITS REPORT, l

1. Either restore the Indicated AFD to within the limits within 15 minutes, or

2. Reduce THERMAL POWGR to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux High Trip setpoints to less )

than or equal to 55 percent of RATED THERMAL POWER with!n the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ,

b. THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD is within the limits specified in the CORE OPERATING LIMITS REPORT.

s SURVEILLANCE REQUIREMENTS r

>. 4.2.1.1 The indicated AX1AL FLUX DIFFERENCE shall be determined to be within its limits during POWER OPERATION above 50% of RATED THERMAL POWER by: ,

a. Monitoring the indicated AFD for each OPERABLE excore channel:

1. At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and

2. At least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after restoring the AFD Monitor Alarm to OPERABLE status. [

b. Monitoring and logging the indicated AXIAL FLUX DIFFERENCE for each OPERABLE excore channel at least once par hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least once per 30 minutes thereafter, when the AX!AL FLUX DIFFERENCE Monitor Alarm is inoperable. '

i

, The logged '/alues of the Indicated AXIAL FLUX DIFFERENCE shall be assumed to exist l

during the interval preceding each logging.

3 l-l 1

NORTH ANNA- UNIT 2 3/4 2-1 Amendment No.

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'3 POWER DISTRIBUTION LIMITS l

h ' I' &)DAL Ft11X DiFi'ERENCE (AFD) p, ]

l SURVEILLANCE REQUlF'IMENTS (Continued) j i

i

.i 4.2.1.2 The indicated AFD shall be considered outside of its limit when at least 2 OPERABLE 'i

. exoore channels are indicating the AFD to be outside of the limits specifh.* 'n the CORE .

f OPERATING LIMITS REPORT. .!

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NORTH ANNA - UNIT 2 3/4 22 Amendment No. ;

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NORTH ANNA UNIT 2 3/4 23 Amendment No.

.a .-

f."

POWER DlSTRIBUTION LildliS

{

HEAT FII fr HOT CHAhlNrl FACTOR-Fp i LIMITNG OONDITON FOR OPERATION

! 3.2.2> Fo(Z) $ hall be limbed by the following relationships:

l Fo(Z) s (CFO) p [K(Z)) for P > 0.5 Fo(Z) s (0[, ) [K(Z)) for P s 0.5 whore CFO = the Fo limit at RATED THERIAAL POWER specifed in the CORE OPERATING LI!AlTS REPORT.

THERMAL POWER

,and P = ' RATED THERMAL POWER K(2) = the normalized Fo limit as a function of core height speelfied in the CORE OPERATING LilAfTS REPORT.

APPLICABILITY: MODE 1.

ACTION-With Fo(Z) exceeding its limit:

a. Reduce THERMAL POWER at least 1% for each 1% Fo(Z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux High Trip Setpoints within the next 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s: POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION may proceed provided the Overpower AT Trip r Selpoints (value of K 4) have been reduced at least 1% (in AT span) for ear,:h 1%

Fo(Z) exceeds the limit,

b. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above the reduced limit required by a, above: THERMAL POWER may then be increased provided Fo(Z) is demonstrated through incore mapping to be within its limit.

l i

NORTH ANNA UNIT 2 3/4 25 Amendment No.

! POWER DISTR!RLrTION LIMITS l

SURVEILLANCE REQUIREMENTS l

9

,, 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 Fo(Z) shall be evaluated to determine if Fo(Z) is within its limit by:

a. Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RAT ED THERMAL POWER.

l b. Increasing the measured Fo(Z) component of the power distribution map by 3% to

! sooount for manufacturing 10lerances and further increasing the value by 5% to I account for measurement uncertaintles,

c. Satisfying the following relationt. hip:

O

F (z) s p,Nz for P > 0.5 F (2) s Nz)x .

I r P s 0.5 where F$(z)is the measured Fo(Z) increased by the allowances for manufacturing tolerances and measurement uncertainty, and N(z) is the cycle dependent function that accounts for power distribution transients encountered during normal operation. This function is specified in the CORE OPERATING LIMITS REPORT as per Specification 6.9.1.7.

d. Measuring F$(z) according to the foKowing schedule:

1. Upon achieving equilibrium conditions after exceeding the THERMAL POWER at which Fo(Z) was last determined by 10% or more of RATED THERMAL POWER',

or

2. At least once per 31 effective full power days, whichever occurs first,

e. With measurements indicating maximum 'F ( z )'

over2 ( K(z) s has increased since the previous determination of F$(z) either of the following actions shall be taken:

'During power escalation, the power level may be increased until a power level for extended operation has been achiev. J snd a power distribution map obtained.

NORTH ANNA UNIT 2 3/4 26 Amendment No.

1 ~0 POWER DISTRIBUTION LMTS SURVEILLANCE REQUIREMENTS (Corninued)

1. F o(z) shall be increased by 2% over that specified in 4.2.2.2.c, or f

2. F (r) o shall be measured at least once per 7 effective full power days until 2 successive maps indicate that maximum (*)

is not increasing,

[ over2 ( K(z) ;

f, With the relationships specified in 4.2.2.2.c above not being satisfied:

1. Calculate the percent Fo(Z) exceeds its limit by subtracting one from the measurement /ilmit ratio and multiplying by 100:

I D maximurn F (z)

1 > x 100 for P 2 0.5 over 2 CFO x K(11

( P x N(z) ) .

I D maximum F (z)

d CFO x X(rj

x 100 for P < 0.5 overz g (0.5 x N(2)) ,

s 2 E!!her of the following actions shall be taken:

a. Power operation may continue provided the AFD limits of Specification 3.2.1 are reduced 1% AFD for each percent Fo(Z) exceeded its limit, or l

b. Comply with the requirements of Specification 3.2.2 for Fo(Z) exceeding its limit by the percent calculated above.

g. The limits specified in 4.2.2.2.c, 4.2.2.2.e, and 4.2.2.2.f above are not applicable in the following core plane regions:

1

1. Lower core region 0 to 15 percent inclusive.

2. Upper core region 85 to 100 percent inclusive.

4.2.2.3 When Fo(Z) is measured for reasons other than meeting the requirements of Specification 4.2.2.2, an overall measu'ed Fo(Z) shall be obtained from a power distribution map and increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty.

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l NORTH ANNA UNIT 2 3/4 27 Amendment No.

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b Qi NORTH ANNA. UNIT 2 3/4 28 Amendment No.

_a

POWER DISTRlRUTION LIMITS tMJ FAR ENTHALPY HOT CHAtML FACTOR-

! UMITING CONDITION FOR OPERATION 3.2.3 H shall be limited by the following relationship:

H 5 CFDH [1 + PFDH (1.P))

where CFDH .The FAH limit at RATED THERMAL POWER specified in the CORE OPERATING UMITS REPORT, THERMAL POWER P = RATED THERMAL POWER PFDH = The Power Factor Multiplier for F"AH specified in the CORE OPERATING LIMITS REPORT, and F"H = measured value of F H obtained by using the movable incore detectors to obtain a power distribution map, APPLICABILITY: MODE 1 ACTION:

With F H exceeding its limit:

a. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux High Trip Setpoints to s 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ,

b. Demonstrate through in core mapping that FAH is within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and

c. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above the reduced limit required by a or b, above; subsequent

' ~ -- -

" POWER OPERATION may proceed provided that FAH si demonstrated through in core mapping to be within its limit at a nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THERMAL POWER prior to NORTH ANNA UNIT 2 3/4 29 Amendment No.

O

.Q ild.9 REFUELlhlG OPERATIONS BMcNOm0ENTRATI%l .

UMITNG OONDITION FOR OPERATION 3.9.1 Whh the reactor vessel head unbolted or removed, the boron concentration of all filled l portens of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive of the following reactivity conditions is met-t

a. Elther a Kett of 0.95 or less, or f

b. A boron concentration limit as specified in the Core Operating Limits Report.

APPLICABILITY: MODE 6*.

l MTICN: j I

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue ;

boration at 210 ppm of 212,950 ppm boric acid solution or its equivalent until Kett i s l

reduced to s 0.95 or the boron concentration is restored to 2 2300 ppm, whichever is the more i restrictive. The provisions of Specification 3.0.3 are not applicable, j i

SURVEILLANCE REQUIREMENTS 4.9.1.1 The more restrictive of the above two reactivity conditions shall be determined prior i to:

a. Removing or unbotting the reactor vessel head, and

b. Withdrawal of any full length control rod located within the reactor pressure vessel, in excess of 3 feet from its fully inserted position. ,

4.9.1.2 The boron concentration of the reactor coolant system and the refueling canal shall be I determined by chemical analysis at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . l f

The reactor shall be maintainod in MODE 6 when the reactor vessel head is unbolted or i removed.

4 l I l

l NORTH ANNA UNIT 2 3/4 91 Amendment No.

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- i ADMNSTRATIVE CONTROLS CORE OPERATNGl.MffS REPORT 6.9.1.7.a Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following: ,

1. Moderator Temperature Coefficient BOC and EOC ltnits, and 300 ppm and i 60 ppm surveillance limits for Specification 3/4.1.1.4,

2. Shutdown Rod insertion Limit for Specification 3/4.1.3.5,

! 3. Control Rod insertion Limits for Speelfication 3/4.1.3.6,

4. Axial Flw Difference limits for Specification 3/4.2.1,

5. Heat Flux Hot Channel Factor, K(Z), N(Z) for Specification 3/4.2.2, and

6. Nuclear Enthalpy Rise Hot Channel Factor, and Power Factor Multiplier, for Specification 3/4.2.3.

7. Boron Concentration for Specification 3/4.9.1.

6.9.1.7.b The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC and identified in the Core Operating Limits Report, The core operating limits shall be determined so that all applicable limits (e.g.,

fuel thermal mechanical limits, core thermal hydraulle limits ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met. The CORE OPERATING LIMITS REPORT, including any mid cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the red onal i Administrator and Resident it'spoctor.

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NORTH ANNA. UNIT 2 6 17 Amendment No.

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[ O-1/41 REACTMTY CONTROL SYSTEMS IMSES it4.1.1.4 MODERATOR TEMPERATURE COEFFICIENT rMTC) (Continued)

The limitations on MTC are provided to ensure that the value of this coefficient remains within the limiting conditions assumed for this parameter in the FSAR accident and transient analyses.

The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of MTC values at conditions other than those explicitly stated will require extrapolation to those conditions in order to permit an accurate comparison.

The most negative MTC value was obtained by incrementally correcting the MTC used in l

the FSAR analyses to nominat operating conditions. These corrections involved adding the I incremental change in the MTC associated with a core condition of Bank D inserted to an all rods withdrawn condition and an incremental change in MTC to account for measurement uncertainty at RATED THERMAL POWER conditions. These corrections result in the End of Cycle (EOC) MTC limit. The 300 ppm surveillance limit MTC value represents a conservative value (with corrections for burnup and soluble boron) at a core condition of 300 ppm equilibrium boron concentration and is obtained by making those corrections to the EOC MTC limit.

Once the equilibrium boron concentra' ion falls below about 60 ppm, dilution operations take an extended amount of time and reliable MTC measurements become more difficult to obtain l due to the potential for fluctuating core conditions over the test intervat. For this reason, MTC measurements may be suspended provided the measured MTC value at an equilibrium full power boron concentration s 60 ppm is less ragative than the 60 ppm surveillance limit. The difference between this value and the EOC MTC limit conservatively bounds the maximum credible change in MTC between the 60 ppm equilibrium boron concentration (all rods withdrawn, RATED THERMAL POWER conditions) and the licensed end of cycle, including the effect of boron oor centration, burnup, and end of cycle coastdown.

The survelliance requirements for measurement of the MTC at the beginning and near the end of each fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.

3'4.1.1.5 MINIMUM TEMPERATURE FOR CRITICAUTY This specification e7sures that the reactor will not be made critical with the Reactor Coolant System average tOmperature less than 541*F. This limitation is required to ensure 1) the moderatot temperatur4 coefficient is within its analyzed temperature range,2) the protective instrumentation is within its normal operating range, and 3) the P 12 Interlock is

! above its setpoint,4} tne pressurizer is capable of being in an OPERABLE status with a steam

bubble, and 5) the reactor pressure vessel is above its minimum RT NDT emperature, t

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NORTH ANNA UNIT 2 B 3/4 12 Amendment No.

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POWER DISTRISLfTION LIMITS EASES 1

When FAH si measured,4% is the appropriale experimental error allowance for a full core map taken with the incore detection system. The specified limit for F H contains a 4%

error allowance. Normal operation will result in a measured F H ess l than or eoual to the limit t.pecified in the CORE OPERATING LIMITS REPORT. The 4% allowance is based on the following i considerations: j l

a. abnormal perturbations in the radial power shape, such as from rod mitalignment, i effect F H more directly than Fo,

b. although rod movement has a direct influence upon limiting F O o t within its limit, j such control is not readily available to limit F"AH,and

c. errors in prediction for control power shape detected during startup physics tests t can be compensated for in Fo by restricting axial flux distributions. This compensation for F H s i less readily available, f

f Fuel rod bowing reduces the value of the DNB ratio. Credit is available to offset this l reduction in the margin available between the safety analysis design DNBR value (1.46 for l Virginia Electric and Power Company statistical methods) and the limiting design DNBR value !

(1.26 for Virginia Electric and Power Company statistical methods). A discussion of the rod i bow penalty is presented in the FSAR. !

The hot channel factor F$(Z) is measured periodically and increased by a cycle and height dependent power factor, N(Z), to provide assurance that the limit on the hot channel !

factor, Fo (Z), is met. N(2) accounts for the non equilibrium effects of normal operation l transients and was determined from expected power control maneuvers over the full range of burnup conditions in the core. The N(Z) function for normal operation is specified in the CORE l OPERATING LIMITS REPORT per Specification 6.9.1.7.

l 314.2.4 OUADRANT POWER TILT RATIO The quadrant power till ratio limit assures that the radial power distribution satisfies ,

the design values used in the power capability analysis. Radial power distribution ;

measurements are made during startup testing and periodically during power operation. '

i The limit of 1.02 at which corrective action is required provides DNB and linrear heat l generation rate protection with x.y plane power tilts.

The two hour time allowance for operation with a till condition greater than 1.02 but less than 1.09 is provided to allow identification and corroction of a dropped or misaligned rod. ;

In the event such action does not correct the tilt, the margin for uncertainty on Fo is reinstated by reducing the power by 3 percent for each percent of till in excess of 1.0. i NORTH ANNA UNIT 2 B 3/4 2 5 Amendment No.

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' Discussion of Proposed Change-j.

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l The proposed changes are consistent with the requirements of 10 CFR 50.36 and the staff's proposed policy for improving Technical Specifications, delineated in SECY 86-10,

Recommendations for improving TS.* The policy allows process variables such as core operational limits to be controlled by specifying them numerically in the Technical Specifications or by specifying the method of calculating their numerical values if the staff finds that the correct limits will be followed in operating the plant. The !

proposed revision references the NRC approved calculation methodologies. The I development of cycle specific core operating limits will continue to be performed by !

the referenced methodologies which have been accepted by the NRC.

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The proposed changes to the Technical Specifications are also considered to be i improvements and are consistent with the NRC stated policy for improving Technical i Specifications (52 FR 3788, February 6,1987).

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Safety Evaluation I J

The current Technical Specification method of controlling reactor physics parameters to assure conformance to 10 CFR 50.36 (which requires the lowest functional ;

l performance levels acceptable for continued safe operation) is to specify the values (

determined to be within the acceptance criteria using an NRC approved calculation I methodology. As previously discussed, the methodologies for calculating these ;

parameter limits have been reviewed and approved by the NRC and are consistent !

with the applicable limits in the Updated Final Safety Analysis Report (UFSAR).

The removal of cycle dependent variables from the Technical Specifications has no impact upon plant operation or safety. No safety related equipment, safety function, or plant operations will be altered as a result of this proposed change. Since the i applicable UFSAR limits will be maintained and the Technical Specifications will l continue to require operation within the core operational limits calculated by these i NRC approved methodologies, this proposed change is administrative in nature, i Appropriate actions to be taken if limits are violated will also remain in the Technical ;

Specifications, i This proposed change will control the cycle specific parameters within the acceptance criteria and assure conformance to 10 CFR 50.36 by using the approved methodology !

instead of specifying Technical Specification values. The COLR will document the .

specific parameter limits resulting from Virginia Electric and Power Company ,

calculations, including mid cycle or other revisions to parameter values. Therefore, '

the proposed change is in conformance with the requirements of 10 CFR 50.36. !

Any changes to the COLR will be made in accordance with the provisions of 10 CFR :

50.59. From cycle to cycle, the COLR will be revised such that the appropriate core i operating limits for the applicable cycle will apply. Technical Specifications will not be i changed.

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'iL T~ s 10 CFR 50.92 Evaluation :

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a Determination of Sionificant Hazards Pursuant to 10 CFR 50.91 Virginia Electric and Power Company has determined that operation of the facility in accordance with the proposed license amendment request i

does not involve any significant hazards considerations as defined by NRC regulations in 10 CFR 50.92. The following discussion describes how the proposed amendment satisfies each of the three standards of 10 CFR 50.92(c)..

1) The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The removal of cycle specific core operating limits from the North Anna Technical Specifications has no influence or impact on the probability or consequences of any accident previously evaluated. The cycle specific core operating limits, although not in Technical Specifications, will be followed in the operation of North Anna. The proposed amendment still requires exactly the same actions to be taken when or il limits are exceeded as is required by current Technical Specifications. Each accident analysis addressed in the Nor1h Anna UFSAR will be examined with respect to changes in cycle-dependent parameters, which are obtained from application of the NRC-approved reload design methodologies, to ensure that the transient evaluation of new reloads are bounded by previously accepted analyses.

This examination, which will be performed per requirements of 10 CFR 50.59, ensures that future reloads will not involve an increase in the probability or consequences of an accident previously evaluated.

2) The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

As stated earlier, the removal of the cycle specific variables has no influence l or impact, nor does it contribute in any way to the probability or l consequences of an accident. No safety related equipment, safety function, ;

or plant operation will be altered as a result of this proposed change. The cycle specific variables are calculated using the NRC approved methods and submitted to the NRC to allow tha Staff to continue to trend the values of l these limits. The Technical Specifications will continue to require operation t within the required core operating limits and appropriate actions will be i taken when or if limits are exceeded. Therefore, the proposed amendment does not in any way create the possibility of a new or different kind of !

accident from any accident previously evaluated, j

3) The proposed change does not result in a significant reduction in the margin of safety.

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l' The margin of safety is not affected by the removal of cycle specific core operating limits from the Technical Specifications. The margin of safety ,

presently provided by current Technical Specifications remains unchanged.

Appropriate measures exist to control the values of these cycle specific !

limits. The proposed amendment continues to require operation within the i

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core limits as obtained from the NRC approved reload design '

methodologies and appropriate actions to be taken when or if limits are violated remain unchanged.

l The development of the limits for future reloads will continue to conform to l those methods described in NRC approved documentation. In addition, each future reload willinvolve a 10 CFR 50.59 safety review to assure that operation of the unit within the cycle specific limits will not involve a ,

reduction in a margin of safety, i

Therefore, the proposed changes are administrative in nature and do not impact the operation of North Anna in a manner that involves a reduction in ;

the margin 01 safety.

The Commission has provided guidance concerning the application of the standards for determining whether a significant hazards consideration exists. This guidance (51 l FR 7750) includes examples of the type of amendments that are considered not likely to involve significant hazards considerations. The change proposed is similar to the examples of administrative changes identified in 51 FR 7750. Additionally, the i proposed change is consistent with the NRC policy for improving technical '

specifications (52 FR 3788) and the proposed change is consistent with 10 CFR 50.36 ,

and 10 CFR 50.59.

In view of the preceding, Virginia Electric and Power Company has determined that the proposed license amendment does not involve any significant hazards considerations, i

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North Anna Unit 1, Cycle 8 !

Core Operating Limits Report ;

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