Text

Amends 84 & 71 to Licenses NPF-4 & NPF-7,respectively, Changing Tech Specs to Increase Max Power Level to 2,893 Mwt
	ML20206Q388
Person / Time
Site:	North Anna
Issue date:	08/25/1986
From:	Harold Denton; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20206Q394	List: ML20206Q388; ML20206Q419;
References
	NUDOCS 8609040384
	Download: ML20206Q388 (88)
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UNITED STATES NUCLEAR REGULATORY COMMISSION o

B rl WASHINGTON, D. C. 20566

s.,...../

VIRGINIA ELECTRIC AND POWER COMPANY OLD DOMINION ELECTRIC COOPERATIVE DOCKET NO. 50-338 NORTH ANNA POWER STATION, UNIT NO. 1 AMEN 0 MENT TO FACILITY OPERATING LICENSE Amendment No. 84 License No. NPF-4 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by Virginia Electric and Power Company, et al., (the Itcensee) dated May 2,1985, as supplemented February 6, April 30, June 4, July 3, and August 20, 1986, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I;

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B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonab1'e' assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

8609040384 860825 DR ADOCK 0500 8

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2.

Accordingly, the license is amended by changes to the Technical Speci-2 fications as indicated in the attachment to this license amendment, and paragraph 2.D.(2) of Facility Operating License No. NPF-4 is hereby j

amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 84, are hereby incorporated i

in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

l 3.

Paragraph 2.D.1 to Facility Operating license No. NPF-4 is hereby

]

revised to read:

2.D.1 Maximum Power Level l

(a) VEPCO is authorized to operate the North Anna Power Station, Unit No.1, at reactor core levels not in j

excess of 2893 megawatts (thermal).

4.

This license amendment is effective within 60 days from its date of issuance with the exception that the change to Table 3.3-2 requiring

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a response time test of the source range, neutron flux trip at least once per 18 months shall be effective prior to restart after the 1

forthcoming sixth (6th) refueling outage.

FOR THE NUCLEAR REGULATORY COMMISSION i

L Harold R. Denton, Direc r

j Office of Nuclear Reactor Regulation l

Attachment:

Changes to the Technical Specifications Date of !ssuance: August 25,1986

i ATTACHMENT TO LICENSE AMENDMENT NO. 84 TO FACILITY OPERATING LICENSE NO. NPF-4 DOCKET NO. 50-338 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages as indicated. The revised pages are identified by amendment number and contain vertical lines indicating the area of change. The corresponding overleaf pages are also provided to maintain document completeness.

P,ag 1-5 2-2 2-6 2-8 2-9 2-10 B 2-1 B 2-2 8 2-4 8 2-6 3/4 2-5 3/4 2-8 3/4 2-9 3/4 2-10 3/4 2-15 3/4 2-16 3/4 3-2 3/4 3-10 3/4 3-16 3/4 3-31 4

3/4 5-6

)

b 3/4 2-1 l

B 3/4 2-4

~

B 3/4 2-E B 3/4 2-6 B 3/4 7-1 i

1 I

l

s 1.0 DEFINITIONS (Continued)

QUADRANT POWER TILT RATIO 1.23 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper ex-core detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore 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.

RATED THERMAL POWER i

1.24 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2893 MWt.

I REACTOR TRIP SYSTEM RESPONSE TIME 1.25 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until loss of s,tationary gripper coil voltage.

REPORTABLE EVENT 1.26 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.

SHUTDOWN MARGIN 1.27 SHUTDOWN MARGIN sha11 be the instantaneous amount of reactivity by which the reactor is subcrttical or would be subcritical from its present condition assuming all full length rod cluster assemblies (shutdown ar.d control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

SITE BOUNDARY' 1.28 The SITE BOUNDARY shall be that line beyond which the land is not owned, j

leased or otherwise controlled by the licensee.

SOLIDIFICATION i

1.29 SOLIDIFICATION shall be the conversion of wet wastes into a solid form that meets shipping and burial ground requirements.

SOURCE CHECK i

1.30 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.

NORTH ANNA - UNIT 1 1-5 Amendment No. J$, AS,4/. 84

1.0 DEFINITIONS (Continued)

STAGGERED TEST BASIS 1.31 A STAGGERED TEST BASIS shall consist of:

A test schedule for n systems, subsystems, trains or other s.

desigrated components 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 beginning of each subinterval.

THERMAL POWER 1.32 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

UNIDENTIFIED LEAKAGE 1.33 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.

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UNRESTRICTED AREA 1.34 An UNRESTRICTED ARTi'shall be any area at or beyond the SITE BOUNDARY where access is not controlled by the licensee for purposes of protection of individuals from exposure tb radiation and radioactive materials or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.

i VENTILATION EIRAUST TREATMENT SYSTEM 1.35 A VENTILATION EIRAUST TREATMENT SYSTEM is the system designed and installed to reduce gaseous radiciodine or radioactive material in particulate form in effluents by p s.s sing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines 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 affluents).

Engineered Safety Feature (EST) atmospheric cleanup systems are not considered to be VENTILATION EIRAUST TREATMENT SYSTEM components.

VENTING 1.36 VEttrING is the controlled process of discharging air or gas f rom 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.

NORTH ANNA - UNIT 1 1-6 Amendnent No. I6, 4 g

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the limits shown in Figures 2.1-1 for 3 loop oper$lio)n and 2.1-2 and 2.1-3 highest operating loop coolant temperature (T shall not exceed the for 2 loop operation.

APPLICABILITY: MODES 1 and 2.

ACTION:

Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure lir,r, be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.

APPLICABILITY: MODES 1, 2, 3, 4 and 5.

ACTION:

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MODES 1 and 2 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

MODES 3, 4 and 5 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes.

NORTH ANNA-UNIT 1 2-1

r

.i

.4 Nominal T

= 586.8'T avs i

heinal RCS flov = 289200 GFM 640 l

l 655 -

2400 psia 558-l 645 '

2250 psia 840 655 -

7 650 2000 psia

, gag,

&- $28 -

[ $15 -

1860 psia

.i..

I

$45 H8 l

i 595 <

54-545 s

500 -

675 8.

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.2

.8 4

.5

.6 7

.0 9

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1.1 I.3 P0wCe treaction er noelnell-Figure 2.1-1 REACTOR CORE SAFETY LIMITE FOR TEREE LOOP OPERATION i

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I NORTH ANNA - UNIT 1 22 Amendment No. AS. 5A, 84

I 4

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The reactor trip system instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.

APPLICABILITY: As shown for each channel in Table 3.3-1.

ACTION:

With a reactor trip system instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION state-ment requirement of Specification 3.3.1.1 until the. channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

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NORTH ANNA - UNIT 1 2-5

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TA8tE 2.2-1 i

3 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 5

FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES E

R 1.

Manual Reactor Trip Not Applicable Not Applicable l

2.

Power Range, Neutron Flux Low Setpoint 1 25% of RATED Low Setpoint - 1 26% of RATED c-l i'i THERMAL POWER THERMAL POWER w

t I

High Setpoint - 1 109% of RATED High Setpoint - 1 110% of RATED THERMAL POWER THERMAL POWER 3.

Power Range, Neutron Flux,

< 5% of, RATED THERMAL POWER with

< 5.5% of RATED THERMAL POWER High Positive Rate a time,consftant > 2 seconds with a time constant > 2 seconds 4.

Power Range, Neutron Flux,

< 5% of RATED THERMAL POWER with

< 5.5% of RATED THERMAL POWER High Negative Rate a time constant > 2 seconds with a time constant > 2 seconds E

5.

Intermediate Range, Neutron 1 25% of RATED THERMAL POWER

< 30% of RATED THERMAL POWER Flux 5

5 6.

Source Range, Neutron Flux

< 10 counts per second

< 1.3 x 10 counts per second I

j 7.

Overtemperature AT See Note 1 See Note 3 8.

Overpower AT See Note 2 See Note 3 l

g 9.

Pressurizer Pressure--Low.

> 1870 psig

> 1860 psig 1

m l

3

10. Pressurizer Pressure--High

,1 2385 psig i 2395 psig l

11. Pressurizer Water Level--High 1 92% of instrument span 1 93% of instrument span D

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12. Loss of Flow

> 90% of design flow per loop *

> 39% of design flow per loop

I E

1 l

Design flow is 96,400 gpm per loop.

i 5

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TABLE 2.2-1 (Continued) l 5

2 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS g

R FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES Y

13. Steam Generator Water

> l8% of narrow range instrument c-

> 17% of narrow range instrument l 2

Level--Low-tow span-each steam generator span-each steam generator

14. Steam /Feedwater Flow

< 40% of full steam flow at

< 42.5% of full steam flow at Mismatch and Low Steam EATED THERMAL POWER coincident KATED THERMAL POWER coincident Generator Water Level with steam generator water level with steam generator water level

> 25% of narrow range instru-

> 24% of narrow range instru-

, men,t span--each steam generator ment span--each steam generator r

15..Undervoltage-Reactor

> 2905 volts-each bus

> 2870 volts-each bus l

Coolant Pumo Busses

16. Underfrequency-Reactor

> 56.1 Hz - each bus

> 56.0 Hz - each bus mL Coolant Pump Busses

17. Turbine Trip A.

Low Trip System

> 45 psig

> 40 psig Pressure B.

Turbine Stop Valve

> 1% open

> 0% open 3

Closure l

j

18. Safety Injection Input Not Applicable Not Applicable from ESF I

i 2

19. Reactor Coolant Pump

ot Applicable Not Applicable

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g Breaker Position Trip 5

i.

if N

1

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TABLE 2.2-l(Continued) l 8

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS l

g NOTATION l

3 E

'IN b l

b NOTE 1: Overtemperature AT 1 aT, [K -g 1

(T-T')+K(P-P')-f(AI)]

g 3

g lh 5 5

2

-A where:

AT,

= Indicated aT at RATED THERMAL POWER w

T

= Average temperature, OF r

0 T'

= Indicated T,,9 at RATED THERMAL POWER 1 586.8 F P

= Pressurizer pressure, psig

'?*

P'

= 2235 psig (indicated RCS nominal operating pressure) i IM S

= The function generated by the lead-lag controller for T dynamic compensation 3

14 $

2 rg & r2

= Time constants utilized in the lead-lag controller for T,yg r3 = 25 secs, 2 = 4 secs.

T F

3 S

= Laplace transform operator (sec-1)

Y an N

E em

l TABLE 2.2-1 (Continued)

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g REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS z5 NOTATION (Continued) g Operation with 3 Loops Operation with 2 Loops Operation with 2 Loops q

(no loops isolated)*

(1 loop isolated)*

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K 1.264 K

(

)

K

(

)

=

g i

g K

0.0220 K

(

)

E

=

(

)

2 2

2 I')

E K

0.001152 K

I

)

=

3 3

3 4

and fi (AI) is a function of the indicated difference between top and bottom detectors

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of the power-range nuclear ion chambers; with gains to be selected based on measured i

instrument response during plant startup tests such that:

4 (i) for q, - qu between - 44 percent and + 3 percent, f3 (a!) = 0 l

(wher6q Mnd q are percent RATED THERMAL POWER in the top and bottom halves o the c re respective

, and qt

9b is total THERNAL POWER in percent of RATED THERMAL POWE k

(ii) i for each percent that the magnitude of (qfy rqbu)ced by 1.67 percent of exceeds - 44 percent, 1

a the AT trip setpoint shall be automatical e

l g

its value at RATED THERMAL POWER.

(iii) for each percent that the magnitude of (qfy reau)ced by 2.00 percent of qh exceeds + 3 percent, P

the AT trip setpoint shall be automatical its value at RATED THERMAL POWER.

w i

1 w

=

l

Values dependent on NRC approval of ECCS evaluation for these operating conditions.

N.

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TABLE 2.2-1 (Continued) i S

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f REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 23 NOTATION (Continued) c

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3 E

T3 E

N p

Note 2: Overpower AT 1 AT, [K 45 1+1 S 6

2 4

3 Where:

AT, Indicated AT at RATED THERMAL POWER

=

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Average temperature, "F T

=

0 Indicated T,,9 at RATED THERMAL POWER 1 586.8 F l

T'

=

I 1.0M K

=

4

?

0 0.02/ F for increasing average temperature 5

K

=

5 K

0 for decreasing average temperatures

=

5 I

0.00164 for T > T'; K6 = 0 for T 1 T' K

=

6 3

The function generated by the rate lag controller for T T 3

=

avg g

1+r35 dynamic compensation i

E!

'3 Time constant utilized in the rate lag controller for T,yg

=

3 = 10 secs.

T e

Laplace transform. operator (sec~I)

.E S

=

w f

l

?

f(AI)= 0 for all AI 2

l E

Note 3: The channel's maximum trip point shall not exceed its computed trip point by more than j

.E.

2 percent span.

T

n 2.1 SAFETY LIMITS BASES J

2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel and 1

possible cladding perforation which would result in the release of fission

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products to the reactor coolant.

Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure

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from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient.

DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been i

related to DNB through the W-3 correlation.

The W-3 DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform d

and non-uniform heat flux distributions.

The local DNB heat flux ratio DNBR, defined as the ratio of the heat flux that would cause ONB at a particular core location to the local heat flux, is indicative of the margin to DNB.

j The DNB design basis is as follows:

there'must be at least a 95 percent i

probability that the minimum DNBR of the limiting rod 'during Condition I and II events is greater than or equal to the DNBR limit of the DNB correlation being used (the WRB-1 correlation in this application). The correlation DNBR limit is established based on the entire applicable experimental data set such that there is a 95 percent probability with 95 percent confidence that DNB i

will not occur when the minimum DNBR is at the DNBR limit.

In meeting this design bassis, uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered statistically such that there is at least a 95% probability that the minimum l

DNBR for the limiting rod is greater than or equal to the DNBR limit.

The uncertainties in the above plant parameters are used to determine the plant

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DNBR uncertainty.

This DNBR uncertainty, combined with the correlatin DNBR i

limit, establishes a design DNBR value which must be met in plant safety analyses using values of input parameters without uncertainties.

The curves of Figures 2.1-1, 2.1-2, and 2.1-3 show the loci of points l

of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than the design limit DNBR, or the average I

enthalpy at the vessel exit is equal to the enthalpy of saturated ifquid.

I

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NORTH ANNA - UNIT 1 8 2-1 Amendment No.84 l

=-

4 SAFETY LIMITS BASES N

The curves are based on an enthal

/

and a reference cosine with a peak of fy hot channel factor, F.55 for axial power s An N

allowance is included for an increase in F at reduced power based on aH the expression:

F3H = 1.49 [1+ 0.3 (1-P)]

l N

where P is the fraction of RATED THERMAL POWER These limiting heat flux conditions are higher than those calculated I

for the range of all control rods fully withdrawn to the maximum allowable i

control rod insertion assuming the axial power imbalance is within the l

limits of the f(AI) function of the Overtemperature trip. When the axial power imbalance is not within the tolerance, the axial power im-balance effect on the Overtemperature AT trips will reduce the setpoints I

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to provide protection consistent with core safety limits.

2.1.2 REACTOR COOLANT SYSTEM PRESSURE l

l The restriction of'this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the 4

release of radionuclides contained in the reactor coolant from reaching the containment atmosphere, The reactor pressure vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plant which permits a maximum transient pressure of 110%(2735 psig) of design pressure. The Reactor Coolant System piping, valves and fittings, were initially designed to ANSI i

B 31.1 1967 Edition and ANSI B 31.7 1969 Edition (Table 5.2.1-1 of FSAR) which permits a maximum transient pressure of 120% (2985 psig)ofcomponent design pressure. The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirements.

The entire Reactor Coolant System is hydrotested at 3107 psig 125%

of design pressure, to demonstrate integrity prior to initial operation.

i I

1 i

NORTH ANNA-UNIT 1 B 2-2 Amendment No. 4 g, 84-.

8 a

2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the values at which the Reactor Trips are set for each parameter. The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits. Opera-tion with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that each Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.

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Manual Reactor Trip The Manual Reactor Trip is a redundant channel to the automatic' protective instrumentation channels and provides manual reactor trip capability.

Power Range, Neutron plux l

The Power Range, Neutron Flux channel high setpoint provides reactor I

core protection against reactivity excursions which are too rapid to be protected by temperature and pressure protective circuitry. The low set point provides redundant protection in the power range for a power excursion beginning from low power. The trip associated with the low setpoint may be manually bypassed when P-10 is active (two of the four power range channels indicate a power level of above approximately 10 percent of RATED THERMAL POWER).

Power Range, Neutron Flux. High Rates The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of rod ejection events from any power level. Specifically, this trip complements the. Power Range Neutron Flux High and Low trips to ensure that the criteria are met for rod ejection from partial power.

NORTH ANNA-UNIT 1

,B 2-3

B i

LIMITING SAFETY SYSTEM SETTINGS BASES The Power Range Negative Rate Trip provides protection for control rod drop accidents. At high power, a rod drop accident could cause local flux peaking which could cause an unconservative local DNBR to exist. The Power Range Negative Rate Trip will prevent this from occurring by tripping the reactor. No credit is taken for operation of the Power Range Negative Rate Trip for those control rod drop accidents for which the DNBR's will be greater than the applicable design limit DNBR value for each fuel type.

Intermediate and Source Range, Nuclear Flux The Source and Intermediate Range, Nuclear Flux trips provide reactor core protection during shutdown (Modes 3, 4 and 5) when the reactor trip system breakers are in the closed position. The Source and Intermediate Range trips in addition to the Power Range trips provide core protection during reactor startup (Mode 2). Reactor 'startup is prohibited unless the Source, Intermediate and Power Range trips are cperable in accordance with Specification 3.3.1.1.

The Source Range Channels will initiate a reactor trip at about 10+5 counts per second unless manually blocked when P-6 becomes active. The Intermediate Range Channels will initiate a reactor trip at a current level proportional to approximately 25 percent of RATED THERMAL POWER unless manually blocked when P-10 becomes active.

In the accident analyses, bosnding transient results are based on reactivity excursions from an initially critical condition, where the source range trip is assumed to be blocked. Accidents initiated from a subcritical condition would produce less severe results since the source range trip would provide core protection at a lower power level. No credit was taken for operation of the trip associated with the Intermediate Range Channels in the accident analyses; however, their functional capability at the specified trip settings is required by-this specification to enhance the overall reliability of the Reactor Protection System.

Overtemperature AT The Overtemperature aT trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping transient delays from the core to the temperature detectors (about 4 seconds),

and pressure is within the range between the High and Low Pressure reactor trips. This setpoint includes corrections for changes in density and heat capacity of water with temperature and dynamic compensation for piping delays from the core to the loop temperature detectors. With normal axial power distribution, this reactor trip limit is always below the core safety limit as shown in Figure 2.1-1.

If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the reactor trip is automatically reduced according to the notations in Table 2.2-1.

k i

NORTH ANNA - UNIT 1 B 2-4 Amendment No. 84

t I

L LIMITING $AFETY SYSTEM SETTINGS BASES Operation with a reactor coolant loop out of service below the 3

-loop P-8 set point does not require reactor protection system set point modification because the P-8 set point and associated trip will prevent DNB during 2 loop operation exclusive of the Overtemperature AT set point. Two loop operation above the 3 loop P-8 set point is permis-sible after resetting the K1, K2 and K3 inputs to the Overtemperature AT channels and raising the P-8 set point to its 2 loop value.

In this mode of operation, the P-8 interlock and trip functions as a High Neutron Flux trip at the reduced power level.

Overpower AT o

The Overpower AT reactor trip provides assurance of fuel integrity, e.g., no melting, under all possible overpower conditions,' limits the required range for Overtemperature AT protection, and provides a backup to the High Neutron Flux trip. No credit was taken for operation of this trip in the accident analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System.

Pressurizer Pressure '

l' The Pressurizer sith and Low Pressure trips are provided to limit the pressure range in which reactor operation is pennitted. The High Pressure trip is backed up by the pressurizer code safety valves for RCS overpressure protection, and is therefore set lower than the set pressure for these valves (2485 psig). The Low Pressure trip provides protection by tripping the reactor in the event of a loss of reactor coolant pressure. The low pressure trip is blocked below P-7.

Pressurizer Water Level The Pressurizer High Water Level trip ensures protection against Reactor Coolant System overpressurization by limiting the water level to a volume sufficient to retain a steam bubble and prevent water relief NORTH ANNA-UNIT 1 B 2-5

'8 i

LIMITING SAFETY SYSTEM SETTINGS BASES through the pressurizer safety valves. No credit was taken for operation of this trip in the accident analyses; however, its functional capability at the l

specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System. The pressurizer high water level trip is blocked automatically below the P-7 setpoint.

Loss of Flow The Loss of Flow trips provide core protection to prevent DNB in the event of a loss of one or more reactor coolant pumps.

Above 11 percent of RATED THERMAL POWER, an automatic reactor trip will occur if the flow in any two loops drop below 90% of nominal full loop flow.

Above 31% (P-8) of RATED THERMAL POWER, automatic reactor trip will occur if the flow in any single loop drops below 90% of nominal full loop flow. This latter trip will prevent the minimum value of the DNBR from going below the design limit during normal operational transients and anticipated transients when 2 loops are in operation and the Overtemperature AT trip setpoint is

-ndjusted to the value specified for all loops in operation. With the Over-temperature AT trip setpoint adjusted to the value specified for 2 loop operation, the P-8 trip at 71% RATED THERMAL POWER with the loop stop valves closed in the nonoperating loop, will prevent the minimum value of the DNBR l

fcom going below the design limit during normal operational transients with 2 loops in operation.

Steam Generator Water Level '

The Steam Generator Water Level Low-Low trip provides core protection by

~

preventing operation with the steam generator water level below the minimum l

volume required fer adequate heat removal capacity. The specified setpoint provides allowance that there will be sufficient water inventory in the steam generators at the time of trip to allow for starting delays of the auxiliary feedwater system. The steam generator water level low-low trip is blocked when the loop stop valves are closed. A steam generator water level high-high l

signal trips the turbine which in turn trips the reactor if above the P-7 l

setpoint.

Steam /Feedwater Flow Mismatch and Low Steam Generator Water Level The Steam /Feedwater Flow Mismatch in coincidence with a Steam Generator Low Water Level trip is not used in the transient and accident analyses but is included in Table 2.2-1 to ensure the functional capability of the specified trip settings and thereby enhance the overall reliability i

l l

NORTH ANNA - UNIT 1 B 2-6 Amendment No. 84

POWER DISTRIBUTION LIMITS HEATFLUXHOTCHANNELFACTOR-Fg LIMITING CONDITION FOR OPERATION 3.2.2.F (Z) shall be limited by the following relationships:

q F (Z) 1 [2.151 [K(Z)]for P > 0.5 q

P F (Z) 1 [,4.30] [K(Z)]for P 10.5 l

q where P = THERMAL POWER RATED THERMAL POWER and K(Z) is the functica obtained from Figure 3.2-2 for a given core height location.

' APPLICABILITY: MODE 1.

ACTION:

With F (Z) exceeding its limit:

q Comply with either of the following ACTIONS:

a.

1.

Reduce THERMAL. POWER at least 1% for each 1% F (Z) exceeds the q

limit within 15 minutes and similarly reduce the Power Range Neutron Flux-Hfgh 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 l

Setpoints have been reduced at least 1% for each 1% F (Z) q

' exceeds the limit.

The Overpower AT Trip Setpoint reduction shall be performed with the reactor in at least HOT STANDBY.

2.

Reduce THERMAL POWER as necessary to meet the limits of Specification _ 3.2.6 using the APDMS with the latest incore map and updated R.

b.

Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above the reduced limit required by a, j

above; THERMAL POWER may then be increased provided F (Z) is q

demonstrated through incore mapping to be within its limit.

i l

l NORTH ANNA - UNIT 1 3/4 2-5

/gnendment No. 3, 5, 76,N9, W.

84 l

l

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 F*Y shall be evaluated to determine if F (Z) is within its q

limit by:

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

a.

Increasing the measured F component of the power distribution b.

map by 3% to account for IIdnufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties.

j c.

Comparing the F computed (F ) obtained in b, above to:

xy 1.

The F limits for RATED THERMAL POWER (FRTP) for the xy appropriate measured core planes given in e and f, below, and 2.

The relationship:

P F

=F Il + 0.2(1-P)]

x where'F is the limit for fractional THERMAL POWER operation L

P expressed as' a function of F and P is the fraction of RATED THERMAL POWER at which F was measured.

xy d.

Remeasuring F according to the following schedule:

xy P

1.

When F

.is greater than the F limit for the appropriate x.

l measured core plane but less than the F relationship, xy additional power distribution maps shall be taken and F compared to F andFh:

RTP x

x Either within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding by 20% of a)

RATED THERMAL POWER or greater, the THERMAL POWER dt which F was last determined, or f

x NORTH ANNA-UNIT 1 3/4 2-6 i

I 4

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)

(b) At least once per 31 EFPD, whichever occurs first.

C 2.

When the F,y is less than or equal to the F, limit for the appropriate measured core plane, additional power distribution C

maps shall be taken and F,y compared to and F at least once per 31 EFPD.

The F, limits for Rated Thermal Power (F RTP) shall be provided e.

g for all core planes containing Bank "D" control rods and all unrodded core planes, in a Core Surveillance Report per Technical Specification 6.9.1.7.

{

f.

The F limits of e, above, are not applicable in the following core plane regions as measured in percent of core he.ight from the bottom of the fuel:

1.

Lower core region from 0 to 15%, inclusive.

2.

Upper core region from 85 to 100%, inclusive.

3.

Grid plane regions at 17.8 12%, 32.1 12%, 46.422%,

60.612% and 74.9 2%, inclusive (17 x 17 fuel elements).

t 4.

Core plane regions within t2% of core height (22.88 inches) about the bank demand position of the bank "D" control rods.

C With F,y exceeding F the effects of F,y on F (Z) shall be g.

q evaluated to determine if F (Z) is within its limit.

q 4.2.2.3 When F (Z) is measured for other than F determination, an overall 9

measured F (Z) shall be obtained from a power distribution map and increased q

by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty.

l l

NORTH ANNA-UNIT 1 3/4 2-7 Amendment No. M. 22,//; 63 i

E 1.2

.t- :=

-:L~5L 1.0 e=_i. _

.=-

~=.-.

0.8

.=.

t.,,.

n

.==: -

wE

- :=-_.

m w

'1 Cf 0.6

+1

==

, ;,(

x=-

0.4

-- b :

wW

- F_ EEil=_ -..h.

E:....u:

~-

r' 0.2

+ t.:: -

l

.i.

="W 8

0.0

~

O 2

4 6

8

'10 12

)

CORE HEIGHT (FT) i i

Figure 3.2-2 NORMALIZEDF(z)ASAFUNCTIONOFCOREHEIGHT g

1 NORTH ANNA - UNIT 1 3/4 2-8 Amendment No. 3,5,75,3),#5, 84

^ - - - - - -

i POWER DISTRIBUTION LIMITS NUCLEAR ENTHALPY HOT CHANNEL FACTOR - FaH LIMITING CONDITION FOR OPERATION F"H shall be limited by the following relationship:

3.2.3 a

F"Ha1 1.49 [1 + 0.3 (1-P)]

l

~

where P = THERMAL POWER ItXTED THERMAL POWER N

F{g = measured value of Fdetectors to obtain $ obtained by using the movable incore power distribution map.

~

APPLICABILITY: MODE 1 ACTION:

N With F exceeding its limit:

3H a.

Reduce THERMAL PONER 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 155% 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" 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 N reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 j

hours, and I

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, aboge; subsequent POWER OPERATION may proceed provided that F is demonstrated through in-core mappingtobewithinit$glimit at a nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THERMAL POWER 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 attaining 95% or greater RATED THERMAL POWER.

HORTH ANNA - UNIT 1 3/4 2-9 Amendment No. fj/, fe, 84

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS N

4.2.3.1 F

shall be determined to be within its limit by using the movableineBredetectorstoobtainapowerdistributionmap:

a.

Prior to operation above 75% of RATED THERMAL POWER after each fuel loading, and b.

At least once per 31 Effective Full Power Days, c.

The provisions of Specification 4.0.4 are not applicable.

l

^

~

l l

i l

\\

NORTH ANNA - UNIT 1 3/4 2-10 Amendment No. 84

\\

l TABLE 3.2-1 E

l DNB PARAMETERS R5 LIMITS g

2 Loops in Operation **

q 3 Loops in

& Loop Stop

& Isolated Loop i

PARAMETER Operation Valves Open Stop Valves Closed 0

Reactor Coolant System T,,9

< 591 F Pressurizer Pressure 62205psig*

i Reactor Coolant System 2.

Total Flow Rate 289,200 gpm R

'?

M l

l i

y

.&5

Limit not applicable during either a THERMAL POWER ramp increase in excess of 5% RATED THERMAL g

POWER per minute or a THERMAL POWER step increase in excess of 10% RATED THERMAL POWER.

M

- Values dependent on NRC approval of ECCS evaluaticri for these conditions.

4 1

4 i

POWER DISTRIBUTION LIMITS AXIAL POWER DISTRIBUTION LIMITING CONDITION FOR OPERATION 3.2.6 The' axial power distribution shall,be limited by the following relationship:

[F (Z))S " (R )(P{)(1.03)(1 + a )(1.07) 3 3

3 Where:

F (Z) is the normalized sxial power distribution from thimble a.

3j at core elevation Z.

b.

P is the fraction of RATED THERMAL POWER.

K(Z) is the function obtained from Figure 3.2-2 for c.

a given core height location.

d.

R, for thimble j, is determined from at least n=6 incore flux maps covering the full configuration of permissible rod patterns above Py of RATED THERMAL POWER in accordance with:

I = 1 [n Rij j

n 1,1 Where:

Qi R

= Pij

) Max gg g3 and(Fg3(Z)]g,x is the maxh value of the normalized axial distribution at elevation Z from thimble j in map i which had a measured peaking factor without uncertainties or densification allowance of F NORTH ANNA - UNIT 1 3/4 2-16 Amendment No. 3, 5, 76, 22, 37./3'g(5 84

s 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1.1 As a minimum, the reactor trip system instrumentation channels and interlocks of Table 3.3-1 shall be OPERAELE with RESPONSE TIMES as shown in Table 3.3-2.

APPLICABILITY: As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1.

SURVEILLANCE REQUIREMENTS

[

j i

4.3.1.1.1 Each reactor trip system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL, FUNCTIONAL TEST operations for the MODES and at the frequencies shown in Table 4.3-1.

'~

l l

4.3.1.1.2 The logic for the interlocks shall be demonstrated OPERABLE prior to each reactor startup unless performed during the prece'eding 92 days. The total interlock function shall be demonstrated OPERABLE at l

least once per 18 months during CHANNEL CALIERATION testing of each channel affected by interlock operation.

4. 3.1.1. 3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function shall be demonstrated to be within its limit at least once per 18 months. Each test shall include at least one logic train such that both logic trains are tested at least once per 36 months and one channel' per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shown in the " Total No. of Channels" column of Table 3.3.1.

NORTH ANNA - UNIT 1 3/4 3-1

TABL'E'3.3-1 h

REACTOR TRIP SYSTEM INSTRUMENTATION Y

MINIMUM 2

E TOTAL NO.

CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION E

1.

Manual Reactor Trip 2

1 2

1, 2 and

12 4

2 3

1, 2.

2 g

2.

Power Range, Neutron Flux 3.

Power Range, Neutron Flux 4

2 3

1, 2 2

High Positive Rate 4.

Power Range, Neutron Flux, 4

2 3

1, 2 2

High Negative Rate 5.

Intermediate Range, Neutron Flux 2

1 2

1, 2 and

3 i

6.

Source Range, Neutron Flux g2, 4

Y A.

Startup 2

1 2

B.

Shutdown 2

1 2

3*,4* and 5*

15 C.

Shutdown 2

0 1

3, 4 and 5 5,

7.

Overtemperature AT Three Loop Operation 3

.2 2

1, 2 7#

Two Loop Operation 3

1**

2 1, 2 9

E E

M "i

5

TABLE 3.3-1 (Co tinued)

' REACTOR TRIP SYSTEM INTERLOCKS ALLOWABLE DESIGNATION CONDITION SETPOINT VALUES FUNCTION 2

P-7 (Cont'd) 3 of 4 Power range below 8%

>7%

Prevents reactor trip on:

g setpoint Low flow or reactor coolant y

and pump breakers open in more 2 of 2 Turbine Impulse 8%

>7%

than one loop, m

chamber pressure below Undervoltage (RCP busses),

4 Underfrequency (RCP biisses),

j setpoint (Power level decreasirtg) f Turbine Trip, Pressurizer low pressure, and Pressurizer high level.

P-8 2 of 4 Power range above 30%

<31%

Permit reactor trip on low setpoint flow or reactor coolant pump w4 breaker open in a single (Power level increasing) loop.

3 of 4 Power range below 28%

>27%

Blocks reactor trip on low setpoint flow or reactor coolant pump breaker open in a single (Power level decreasing) loop.

TABLE 3.3-2 g

5 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES x

=

e.

FUNCTIONAL UNIT RESPONSE TIME

=

]

1.

Manual Reactor Trip liOT APPLICABLE 2.

Power Range, Neutron Flux 1 0.5 seconds

3.

Power Range, fleutron Flux, NOT APPLICABLE High Positive Rate f

4.

Power Range, Neutron Flux, R

High Negative Rate 1 0.5 seconds

Y 5.

Intermediate Range, Neutron Flux HOT APPLICABLE 1

G 1

5 seconds

l 0

6.

Source Range, Neutron Flux 7.

Overtemperature AT 1 4.0 seconds

NOT APPLICABLE 8.

Overpower AT 9.

Pressurizer Pressure--Low 1 2.0 seconds

10. Pressurizer Pressure--High 1 2.0 seconds y

11. Pressurizer Water Level--High NOT APPLICABLE l

E e

,Neutron detectors are exempt from response time testing.

Response of the neutron flux signal z

portion of the channel time shall be measured from detector output or input of first electronic g

component in channel.

1 S

INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2.1 The Engineered Safety Feature Actuation System (ESFAS) instrumen-tation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the valuds shown in the Trip Setpoint column of Table 3.3-4 and with RESPONSE TIMES as shown in Table 3.3-5.

APPLICABILITY: As shown in Table 3.3-3.

ACTION:

~

a.

With an ESFAS instrumentation channel trip setpoint less conserva-tive than the value shown in the Allowable Values column of Table 1

3.3-4, declare the channel inoperable and apply the applicable ACTION requirement of Table 3.3-3 until the channel is restored to OPERABLE status with the trip setpoint adjusted consistent with the Trip Setpoint value.

b.

With an ESFAS instrumentation channel inoperable, take the ACTION shown in Table 3.3-3.

s SURVEILLANCE REQUIREMENTS 4.3.2.1.1 Each ESFAS instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shown in Table 4.3-2.

4.3.2.1.2 The logic for the interlocks shall be demonstrated OPERABLE during the automatic actuation logic test. The total interlock function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by interlock operation.

4.3.2.1.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be demonstrated to be within the limit at least once per 18 months. Each test shall include at least one icgic train such that both logic trains are tested at least once per 36 months and one channel per function such that all channels ere tested at least once per N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in the " Total ha. of Channels" Column of Table 3.3-3.

NORTH ANNA - UNIT 1 3/4 3-15

TABLE 3.3-3 h

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION 2

MINIMUM h

TOTAL NO.

CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT _

0F CHANNELS _

TO TRIP OPERABLE MODES ACTION E

1.

SAFETY INJECTION, TURBINE TRIP AND M

FEEDWATER ISOLATION a.

Manual Initiation 2

1 2

1,2,3,4 18 b.

Automatic Actuatio~n 2

1 2'

1, 2, 3, 4 13 c.

Containment 3

2 2

1,2,3,4 14 Pressure-High d.

Pressurizer 3

2 2

1,2,3 14 i

. Pressure-Low-Low y

1,2,3 Y

e.

Differential M

Pressure Between Steam Lines -High Three Loops 3/ steam line 2/ steam line 2/ steam line 14 Operating twice and 1/3 a-steam lines g

1. 1. steam 2/ operating 15 k

Two Loops 3/ operating 2

/

A Operating steam line line twice steam line in either l

z operating steam line M

1,2,3 k

f.

Steam Flow in Two Steam Lines-High S

]

TABLE'4~.3-2 O

]

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS

!!i!

CHANNEL MODES IN WHICH i

CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE g

FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED 2

1.

SAFETY INJECTION, TURBINE TRIP AND I

FEEDWATER ISOLATION i

a.

Manual Initiation N.A.

N.A.

M(1) 1,2,3,4

/

b.* Automatic Actuation Logic A.A.

N.A.

M(2) 1,2,3,4 c.

Containment Pressure-High S

R M

1,2,3,4 l

[

d.

Pressurizer Pressure--Low-Low S

R M

1,2,3 e.

Differential Pressure S

R M

1,2,3 Between Steam Lines--High f.

Steam Flow in Two Steam S

R M

1,2,3 j

Lines--High coincident with T

--Low-Low or Steam Line 3

3 PPIIsure--Low 2

2.

CONTAINMENT SPRAY a

2 a.

Manual Initiation N.A.

N.A.

M(1) 1, 2, 3, 4 b.

Automatic Actuation Logic N.A.

N.A.

M(2) 1, 2, 3, 4 g

c.

Containment Pressure--High-S R

M 1,2,3 High t

i l

g TABLE 4.8-2 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION E

SURVEILLANCE RE0UIREMENTS E

e CHANNEL MODES IN WHICH s

CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE 5

FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED l

3.

CONTAINHENT ISOLATION 4

a.

Phase "A" Isolation

1) Manual N A.

N.A.

M(1) 1, 2, 3, 4 f

l 2)

From Safety Injection N.A.

N.A.

M(2) 1,2,3,4 Automatic Actuation Logic b.

Phase "B" Isolation w1

1) Manual N.A.

N.A.

M(1) 1,2,3,4

[

2) Automatic Actuation N.A.

N.A.

M(2) 1, 2, 3, 4 Logic m

3) Containment Pressure--

S R

M(3) 1, 2, 3 High-High 4.

STEAM LINE ISOLATION a.

Manual N.A.

N.A.

R 1,2,3 g

g b.

Automatic Actuation Logic N.A.

N.A.

M(2) 1, 2, 3 k

c.

Containment Pressure--

S R

M 1,2,3 g

Intennediate High-High W

d.

Steam Flow in Two Steam S

R-M 1,2,3 Lines--High Coincident with

-- Lo -Low or Steam Line T

g Pfdsurew-Low e

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 2.

Verifying that each of the following pumps start auto-matica11y upon receipt of a safety injection test signal:

a)

Centrifugal charging pump, and b)

Low head safety injection pump.

f.

By verifying that each of the following pumps develop the indicated discharge pressure (after subtracting suction pres-sure) on recirculation flow when tested pursuant to Specifica-tion 4.0.5.

1.

Centrifugal charging pump >_2410 psig.

2.

Low head safety injection pump >_156 psig g.

By verifying that the following manual valves requiring adjustment to prevent pump " runout" and subsequent component

~ _

damage are locked and tagged in the proper position for injection:

1.

Within 4' hours following completion of any repositioning or maintenance on the valve when the ECCS subsystems are required'to be OPERABLE.

l 2.

At least once per 18 months.

1.

1-SI-188 Loop A Cold Leg 2.

1-SI-191 Loop B Cold Leg

-3.

1-SI-193 Loop C Cold Leg 4.

1-SI-203 Loop A Hot Leg 5.

1-SI-204 Loop B Hot Leg 6.

1-SI-205 Loop C Hot Leg h.

By performing a flow balance test, during shutdown, following completion of modifications to the ECCS subsystems that alter the subsystem flow characteristics and verifying that:

1.

For high head safety injection lines, with a single pump running:

a) The sum of the injection line flow rates, excluding the highest flow rate, is > 384 gpm, and b) The total pump flow rate is 1650 gpm.

j NORTH ANNA-UNIT 1 3/4 5-5 Amendment No. 6, 19

EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - T,yg < 350*F LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

One OPERABLE centrifugal charging pump #,

a.

b.

One OPERABLE low head safety injection pump #, and c.

An OPERABLE flow path capable of automatically transferring fluid to the reactor coolant syste'm'when taking suction from the refueling water storage tank or from the containment sump when the suction is transferred during the fecirculation phase of operation or from the discharge.of the outside recirculation spra9 pump.

APPLICABILITY: MODE 4.

~

ACTION:

With no ECCS s'ubsystem OPERABLE because of the inoperability a.

of either the gentrifugal charging pump or the flow path from the refueling water storage tank, restore at least one ECCS subsystem to OPERABLE status within.1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in COLD SHUTDOWN within the next 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months , b.

With no ECCS subsystem OPERABLE because of the inoperability of the low head safety injection pump, restore at least one ECCS subsystem to OPERABLE status or maintain the Reactor heatremovalmetMEs.essthan350*Fbyuseofalternate l

Coolant System T c.

In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Comission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.

A maximum of one centrifugal charging pump and one low head safety injection pump shall be OPERABLE whenever the temperature of one or more of the RCS cold legs is less than or equal to 320*F.

NORTH ANNA-UNIT 1 3/4 5-6 Amendment No. 3. M, 84

^5 3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by:

(a) maintaining the minimum DNBR in the core from going beyond the design limit DNBR during normal operation and in short term transients, and (b) limiting the fission gas release, fuel pellet temperature & cladding mechanical properties to within assumed design criteria.

In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200*F is not exceeded.

The definitions of certain hot channel and peaking factors as used in these specifications are as follows:

F (Z)

Heat Flux Hot Channel Factor, is defined as the maximum local 9

heat flux on the surface of a fuel rod at core. elevation Z divided by the average fuel rod heat flux, allowing for manufac.turing tolerances on fuel pellets and rods.

N F

Nuclear Enthalpy Rise Hot Channel' Factor, is defined as the AH ratio of the integral of linear power along the rod with the

~

highest integrated power to the average rod power.

F,Y(Z)

Radial Peaking Factor, is defined as the ratio of peak power density to the average power density in the horizontal plane at core elevation Z.

3/4 2.1 AXIAL FLUX DIFFERENCE (AFD)

The limits on AXIAL FLUX DIFFERENCE assure that the F (Z) upper bound 0

envelope, as given in Specification 3.2.2, is not exceeded during either normal operation or in the event of xenon redistribution following power changes.

Target flux difference is determined at equilibrium xenon conditions.

The full length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady state operation at high power levels. The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL' POWER level. The periodic updating of the target flux difference value is necessary to reflect core burnup considerations.

NORiH NNA - UNIT 1 B 3/4 2-1 Amendment No. 3,5,B,39,75,84

POWER DISTRIBUTION LIMITS BASES Although it is intended that the plant will be operated with the AXIAL FLUX DIFFERENCE within the + 5% target band about the target flux difference, during rapid plant THTRMAL POWER reductions, control rod motion will cause the AFD to deviate outside of the target band at reduced THERMAL POWER levels. This deviation will not affect the xenon redistribution sufficiently to change the envelope of peaking factors which may be reached on a subsequent return to RATED THERMAL POWER (with the AFD within the target band) provided the time duration of the deviation is limited. Accordingly, a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months penalty deviation limit cumulative during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is provided for operation outside of the target band but within the limits of Figure 3.2-1 while at THERMAL POWER levels between 50% and P % of RATED THERMAL POWER.

For THERMAL POWER levels between l 15% and 50% of RATED THERMAL POWER, deviations of the AFD outside of the target band are less significant. The penalty of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months actual time reflects this reduced significance.

Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The.

computer determines the one minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are outside the target band and the THERMAL POWER is greater than P % of RATED THERMAL POWER.

f During operation.at THERMAL POWER levels between 50% and P,% and 15% and 50% RATED THERMAL POWER, the computer outputs an alarm message when the penalty deviation accumulates beyond the limits of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , respectively.

Figure B 3/4 2-1 sh'oks a typical monthly target band.

NORTH ANNA - UNIT 1 B 3/4 2-2 Amendment No. 3, 5, 22, 37

1 b

Percent of Rated M

P'**'

5%

100%

90%

4 00%

70%

Target Flux Difference M

I 50%

l

~

1 30%

20%

4 l

10%

30%

20%

10%

0

+10%

+20%

+30%

i INDICATED AXIAL FLUX DIFFERENCE Figure 3 3/4 21 TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS THERMAL POWER NORTH ANNA - UNIT 1 B 3/4 2-3

~

POWER DISTRIBUTION LIMITS BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX AND NUCLEAR ENTHALPY HOT CHANNEL FACTORS-F(Z)andF n

H 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 criteria limit.

Each of these hot channel factors are measurable but will nomally only be detemined 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 + 12 steps from the group demand position.

b.

Control rod groups are sequenced with overlapping groups as describedinS(ecification3.1.3.6.

c.

The control rod insertion limits of Specifications 3.1.3.5 and

~

3.1.3.6 are maihtained.

d.

The axial power distribution, expressed in tems of AXIAL FLUX DIFFERENCE, is maintained within the limits.

N The relaxation in F as a function of THERMAL POWER allows changes intheradialpowershap4Hfor all pemissible rod insertion limits. F will be maintained within its limits provided conditions a thru d aboN, are maintained.

When an F measurement is taken, both experimental error and man-ufacturing tolkrance must be allowed for.

5% is the appropriate allowance for a full core map taken with the incore detector flux mapping system and 3% is the appropriate allowance for manufacturing tolerance.

The Specified limit for F contginsa.4%errorallowance. Normal The.4% allowance is basedon.thefollow1.ngconsiderationsgH..=_1.49.

operation will result in a mea Ured F NORTH ANNA - UNIT 1 B 3/4 2-4 Amendment No.J6,84

o C

POWER DISTRIBUTION LIMITS BASES a.

abnormal perturbati us in the radial power shape, such as from rod N

misalignment, effect F more directly than F,

3 g

- b.

although rod movement has a direct influence upon limiting Fn to wjthinitslimit,suchcontrolisnotreadilyavailabletolYmit FaH, and c.

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

AH Fuel rod bowing reduces the value of the DNB ratio. Credit is available to offset this reduction in the margin available between the safety analysis design DNBR values (1.57 and 1.59 for thimble and typical cells, e

respectively) and the limiting design DNBR values (1.39 for thimble cells.and 1.42'for typical cells). The applicable value of rod bow penalties can be obtained from the FSAR.

3/4.2.4 QUADRANT POWER TILT RATIO The quadrant power tilt 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.

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 tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correc-tion of a dropped or misaligned rod.

In the event such action does not correct the tilt, the margin for uncertainty on F 0 is reinstated by reducing the power by 3 percent for each percent of tilt id 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 synsnetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore flux map or two sets of 4 syninetric thimbles. The two sets of 4 symmetric thimbles is a unique set of 8 detector locations. These loca-tions are C-8, E-5, E-11, H-3, H-13, L-5, L-ll, and N-8.

NORTH ANNA - UNIT 1

.B 3/4 2-5 AmendmentNo.35,b,84

POWER DISTRIBUTION LIMITS t

BASES

[

3/4.2.5 DNB PARAMETERS l

The limits on the DNB related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR greater than the design limit throughout each analyzed transient. Measurement uncertainties must j

be accounted for during the periodic surveillance.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months periodic surveillance of these parameters thru instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

The 18 month periodic measurement of the RCS total flow rate is adequate to detect flow degradation and ensure correlation of the flow indication channels with measured flow such that the indicated percent flow will provide su,fficient verification of flow rate on a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months basis.

3/4.2.6 AXIAL POWER DISTRIBUTION The limit on axial power distribution ensures that Fn will be controlled and monitored on a more exact basis through use of the APDMS when operating above P,% of RATED THERMAL POWER. This additional limitation on F is n

necessary in order to provide assurance that peak clad temperatures will remain' below the ECCS acceptance criteria limit of 2200*F in the event of a LOCA. The value for P is based on the cycle dependent potential violation of the F xK(Z) limit, Ehere K(Z) is the graph shown in Figure 3.2-2.

The

-amount oh potential violation is determined by subtracting 1 from the maximum ratio of the predicted F (Z) ahalysis (flyspeck) results for a particular fuel cycle to the Fn K(Zi limit. This amount of potential violation, in x

percent, is subtracted from 100% to determine the value for P If P is equal to 100%, no axial power distribution surveillance is re$u. ired. "P,wiU i

not exceed 100%.

i i

NORTH ANNA - UNIT 1 B 3/4 2-6 Amendment No. 3,5,22,37,84 i

. -. -. _, -. _ _ -. _ _ _ _ - _ _ _. _ _ _ - _. - _ _=_-._- _ -. - -_. __

L 3/4.7 PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line code safety valves ensure that the secondary system pressure will be limited to within 110% of the system design pressure, during the most severe anticipated system opera-tional transient.

The maximum relieving capacity is associated with a turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).

The specified valve lift settings and relieving capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure Code,1971 Edition.

The total relieving capbcity for all safety valves on all of the. steam lines is 12.83 x 10 lbs/

the total secondary steam flow of 12.77 x 10gt which is greater than lbs/hr at 100% RATED THERMAL POWER. A minimum of 2 OPERABLE safety valves per steam generator ensures that sufficient relieving capacity is avail-able for the allowable THERMAL POWER restriction in Table 3.7-1.

j STARTUP and/or POWE.R OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction 4q secondary system steam flow and THERMAL POWER required by the reduced reactor trip settings of the Power Range Neutron Flux channels. The reactor trip setpoint reductions are derived on the following bases:

For 3 loop operation 3p, (X)

(Y)(V) x 109 For 2 loop operation with stop valves closed SP = (X) - (Y)(U) x 7) l X

l For 2 loop operations with stop valves open 3p, (X) - (Y) (U) x 66 X

NORTH ANNA - UNIT 1 B 3/4 7-1 Amendment No. 84

f PLANT SYSTEMS BASES Where:

SP = reduced reactor trip setpoint in percent of RATED THERMAL POWER V=

maximum number of inoperable safety valves per steam line maximum number of inoperable safety valves per operating U

=

steam line Power Range Neutron Flux-High Trip Setpoint for 3 loop 109

=

operation Maximum percent of RATED THERMAL POWER permissible by 71

=

P-8 Setpoint for 2 loop operation with stop valves closed.

Maximum percent of RATED THERMAL POWER permissible 66

=

by P-8 setpoint for 2 loop operation with stop valves open.

X=

Total relieving capacity of all safety valves per steam line in Ibs/ hour

4,275,420 Y=

Maximum relieving capacity of any one safety valve in 1bs/ hour = 855,084 3/4. 7.1. 2 AUXILIARY FEEDWATER SYSTEM The OPERABILITY of the auxiliary feedwater system ensures that the Reactor Coolant System can be cooled down to less than 350*F from normal operating conditions in the event of a total loss of off-site power.

i Each electric driven auxiliary feedwater pump is capable of de-livering a total feedwater flow of 340 gpm at a pressure of 1064 psig to the entrance of the steam generators. The steam driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 700 gpm at a pressure of 1064 psig to the entrance of the steam generators.

This capacity is sufficient to ensure that adequate feedwater flow is NORTH ANNA - UNIT 1 B 3/4 7-2

-.. - ~ ~

[

UNITED STATES y

g NUCLEAR REGULATORY COMMISSION n

ej WASHINGTON, D. C. 20655

...../

VIRGINIA ELECTRIC AND POWER COMPANY OLD DOMINION ELECTRIC COOPERATIVE DOCKET NO. 50-339 NORTH ANNA POWER STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 71 License No. NPF-7 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendmer.t by Virginia Electric and Power Company, et al., (the licensee) dated May 2, 1985, as supplemented February 6, April 30, June 4, July 3, and August 20, 1986, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; i

D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

. ~

.m.

. 2.

Accordingly, the license is amended by changes to the Technical Speci-fications as indicated in the attachment to this license amendment, L

and paragraph 2.C.(2) of Facility Operating License No. NPF-7 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 71, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

Paragraph 2.C.(1) to Facility Operating license No. NPF-7 is hereby revised to read:

2.C.(1)

Maximum Power Level VEPC0 is authorized to operate the facility at steady state reactor core power levels not in excess of 2893 megawatts (thermal).

4.

This license amendment is effective within 30 days from its date of issuance with the exception that the change to Table 3.3-2 requiring a response time test of the source range, neutron flux trip at least once per 18 months shall be effective prior to restart after the forthcoming fifth (5th) refueling outage.

FOR THE NUCLEAR REGULATORY COMMISSION sff, -

Harold R. Denton, Dire or Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: August 25, 1986

ATTACHMENT TO LICENSE AMENDMENT NO. 71 TO FACILITY OPERATING LICENSE NO. NPF-7 DOCKET NO. 50-339 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages as indicated. The revised pages are identified by amendment number and contain vertical lines indicating the area of change. The corresponding overleaf pages are also provided to maintain document completeness.

?*S*

1-5 2-2 2-6 2-8 2-9 2-10 B 2-1 B 2-2 B 2-3 B 2-4 B 2-6 3/4 2-5 3/4 2-6 3/4 2-7 3/4 2-8 s

3/4 2-9 3/4 2-10 3/4 2-16 i

3/4 3-2 1

3/4 3-10 3/4 3-16 3/4 3-33 3/4 5-6 B 3/4 2-1 B 3/4 2-5 B 3/4 2-6 8 3/4 7-1 t

t I

m

1.0 DEFINITIONS (Continued)

QUADRANT' POWER TILT RATIO 1.23 QUADRANT POWER TILT RATIO shall be the ratio of the maximum ~ upper ex-core detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector i

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.

RATED THERMAL POWER 1.24 RATED THERMAL POWER shall be a total reactor core heat transfer rate l

to the reactor coolant of 2893 MWt.

REACTOR TRIP SYSTEM RESPONSE TIME l

l 1.25 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from l

when the monitored parameter exceeds its trip setpoint at the channel sensor until lo,ss of stationary gripper coil voltage.

REPORTABLE EVENT 1.26 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.

~

SHUTDOWN MARGIN i

s i

1.27 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is suberitical or would be subcritical 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.

i S1TE BOUNDARY' 1.28 The SITE BOUNDARY shall be that line beyond which the land is not owned, leased or otherwise controlled by the licensee.

SOLIDIFICATION 1.29 SOLIDIFICATION shall be the conversion of wet wastes into a solid form l

that meets shipping and burial ground requirements.

SOURCE CHECK 1.30 A SOURCE CHECK shall be tha qualitative assessment of channel response.

when the channel sensor is exposed to radiation. This applies to installed radiation monitoring systems.

NORTH ANNA - UNIT 2 1-5 Amendment No. 31.47, 71

~

e 1.0 DEFINITIONS (Continued)

STAGGERED TEST BASIS 1.31 A STAGGERED TEST BASIS shall consist of:

A test schedule for n systems, subsystems, trains or other a.

designated components obtained by dividing the specified test interval into n equal subintervals, b.

Tlie testing of one system, subsystem, train or other designated component at the beginning of each subinterval.

i THERMAL POWER l

j 1.32 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

UNIDENTIFIED LEAKAGE 1.33 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.

UNRESTRICTED AREA 1.34 An UNRESTRICTED ARE$'shall be' any area at or beyond the SITE BOUNDARY where access is not controlled by the licensee for purposes of protection of individuals from exposure tb radiation and radioactive materials or any area within the SITE BOUNDARY used for residential quarters or for industrial, consercial, institutional, and/or recreational purposes.

VENTILATION EXRAUST TREATMENT SYSTEM 1.35 A VENTILATION EIRAUST TREATMENT SYSTEM is the system designed and installed to reduce gaseous radiciodine or radioactive material in particulate in effluents by passing ventilation or vent exhaust gases through form charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or from the gaseous exhaust stream prior to the release to the particulates (such a system is not considered to have any effect on noble gas environment affluents).

Engineered Safety Feature (ESP) atmospheric cleanup systems are not considered to be VENTILATION EIEAUST TREATMENT SYSTEM components.

VENT'ING 1.36 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.

fT.T6RRim - UNIT 2 1-6 Ame ndme nt N o. 3 1

6

0 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE,

2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest shall not exceed the limits shown in operating loop' coolant temperature (TFigures 2.1-1 for 3 loop operation an6 APPLICABILITY: MODES 1 and 2.

ACTION,:

Whenever the point defined by the combination of the highest operating loop, average temperature and THERMAL POWER has exceeded -the appropriate pressurizer pressure line, be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

REACTOR COOLANT SYSTEM PRESSURE 2.1. 2 The Reactor Coolant System pressure shall not exceed 2735 psig.

APPLICABILITY: MODES 1 E, 3, 4 and 5.

s ACTION:

MODES 1 and 2 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

MODES 3, 4 and 5 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes.

NORTH ANNA - UNIT 2 2-1

Nominal T,,,= 586.8'T Iteeinal RCS flov = 289200 GPM 663 '

855 2400 psia 558 -

8'S '

2250 psia 840 -

555 -

1 553 2000 psis 625

$ 528 -

( 815 1860 psia si. -

685 N.-

595 -

f 55 u..

675 O.

.8

.2

.5 4

.5

.6

.7

.0 9

8.

3.1 13 PtM e treaction er noelneti REACTOR CORE SAFETT LMT4 701 N WP WERATION Figure 2.1-1 i

NORTH ANNA - UNIT 2 2-2 Amendment No. 29,$7,.71 i

l I

i --

l SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The reactor trip system instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.

APPLICABILITY: As shown for each channel in Table 3.3-1.

ACTION:

With a reactor trip system instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1.1 until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

S NORTH ANNA - UNIT 2 2-5

{

TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRNMENTATION TRIP SETPOINTS 2

E i

y FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES i

l E

1.

Manual Reactor Trip Not Applicable Not applicable 2.

Power Range, Neutron Flux Low Setpoint - 1 25% of RATED Low Setpoint - 1 26% of RATED l

E THERMAL POWER THERMAL POWER i

G l

High Setpoint - 1 109% of RATED High Setpoint - 1 110% of RATED THERMAL POWER THERMAL POWER m

l 3.

Power Range, Neutron Flux,

<5%ofRATEDTHERMALPbWERwith

< 5.5% of RATED THERMAL POWER High Positive Rate i time cqnstant > 2 seconds with a time constant > 2 seconds.

4.

Power Range, Neutron Flux,

< 5% of RATED THERMAL POWER with

< 5.5% of RATED THERMAL POWER High Negative Rate i time constant > 2 seconds with a time constant > 2 seconds.

m l

5.

Intermediate Range, Neutron 1 25% of RATED THERMAL POWER 1 30% of RATED THERMAL POWER Flux 5

5 6.

Source Range, Neutron Flux

< 10 counts per second

< 1.3 x 10 counts per second j

7.

Overtemperature AT See Note 1 See Note 3 8.

Overpower AT See Note 2 See Note 3 a

9.

Pressurizer Pressure--Low

> 1870 psig

> 1860 psig R

R

10. Pressurizer Pressure--High 1 2385 psig i 2395 psig

11. Pressurizer Water Level--High 1 92% of instrument.. span 1 93% of instrument span E

D

12. Loss of Flow

> 90% of design flow per loop *

> 89% of design flow' per loop

l

Design flow is 96,400 gpm per loop.

4

TABLE 2.2-1 (Continued) g

=

jii!

REACTOR TRIP SYSTEM INSTRtMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES g

8

13. Steam Generator Water 1 18% of narrow range instrument 1 17% of narrow range instrument Level--Low-Low span each steam generator span-each steam generator 4

14. Steam /Feedwat.er Flow

< 40% of full' steam flow at

< 42.5% of full steam flow at m

Mismatch and Low Steam NATED THERMAL POWER coincident RATED THERMAL POWER coincident Generator Water Level with steam generator water level with steam generator water level 1 25% of narrow range instru-1 24% of narrow range instru-ment span--each steam generator ment span--each steam generator

15. Undervoltage-Reactor 12905Iolts-eachbus 1 2870 volts-each bus Coolant Pump Busses

16. Underfrequency-Reactor 1 56.1 Hz - each bus 1 56.0 Hz - each bus Coolant Pump Busses m

17. Turbine Trip A.

Low Trip System 1 45 psig 1 40 psig Fressure 8.

Turbine Stop Valve 1 1% open 1 0% open Closure

18. Safety Injection Input Not Applicable Not Applicable from ESF

19. Reactor Coolant Pump Not Applicable Not Applicable Breaker Position Trip e

I6 e

l g

TABLE 2.'2-1 (Continued)

= -

s I

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

- g j

g NOTATION f

h 1

(T-T')+K I IA NOTE 1:

Overtemperature AT 1 AT, [K -K2 3

I

~

j 1+1 5 2e m.

1 where:

AT, Indicated AT at RATED THERMAL POWER

=

Average temperatur.e. *F T

=

Indicated T,,, at RATED THERMAL POWER S 586.S*F l

T'

=

Pressurizer pressure, psig P

=

to4 P'

2235 psig (indicated RCS nominal cperating pressure)

=

1+t S i

j j

3 - The function generated by the lead-lag controller for T,,, dynamic compensation I

t2= Time constants utilized in the lead-lag controller for T-t = 25 secs, "8

2 " 4 *****

Laplace transform operator (sec~1) k 5

=

a o"

O I

wD D

i i

i I

_.-..-_._-i-4._m.

._m l

TABLE 2.2-1. (Continued) i REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS g

NOTATION (Continued)

I w

Operation with 3 loops Operation with 2 Loops Operation with 2 Loops 3E (no loops isolated)*

(1 loop isolated)*

5

(

)

l

(

)

K j

-K) 1.264 K

=

j c

}

K I

)

K I

)

l 0.0220 K

=

g 2

2

(

)

l 0.001152 K

I'

)

K K

=

3 3

3 and fj (a!) is a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

{

m w

]

(1) for qt -9b between - 44 percent and + 3 percent, fj (AI) = 0 l

l (where q and qb are percent RATED THERMAL POWER in the top and bottom t

j halves of the core respectively, and qt + 9b is total THERMAL POWER in j

percent of RATED THERMAL POWER).

I

{

(ii) for each percent that the magnitude of (qt - 9'b) exceeds - 44 percent, 1

I E

the AT trip setpoint shall be automatically reduced by 1.67 percent of l

l l

its value at RATED THERMAL POWER.

1 g:

(iii) for each percent that the magnitude of (qt - 9 ) exceeds + 3 percent, I

b i

the aT trip setpoint shall be automatically reduced by 2.00 percent of l

~

U its value at RATED THERMAL F0kER.

.M M

i j

Values dependent on NRC approval of ECCS evaluation for these operating conditions.

I

TABLE 2.2-1 (Continued)

\\

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION (Continued)

E5 S,T-K6(T-T")-fIAI)3 2

jf3 Note 2:

' Overpower AT 1AT,[K-K5 2

4 2-Indicated 6T at RATED THERMAL POWER where:

AT

=

o cx Average temperature

'F

.Q T

=

Indicated T,,9 at RATED THERMAL POWER 1586.8'F.

T"

=

1.079 l

K

=

4 0.02/*F for increasing average temperature K

=

5 to 0 for decreasing average temperatures i

K

=

5 0.00164 for T > T";

K6 = 0 for T 1 T" K

=

6 3

T3 The function generated by the rate lag controller for T,yg

=

9 jg 3 3

dynamic compensation I

Time constant utilized in the rate lag controller for T,,g l

g 3

T

=

3 = iO secs.

1 g

Laplace transform operator (sec-j) s 3

S

=

i l

,E f (aI) = 0 for all al 2

l The channel's maximum trip point shall not exceed its computed trip point by more than M

Note 3:

i l

2 percent span.

a-I U

l l

l i

O t

2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling. (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through the W-3 correlation. The W-3 DNB correlation has been developed to predict the DNB flux and the location of

. DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio. DNBR, defined as the ratio of the heat flux that would cause DNB at a particQlar core location to the local heat flux, is iridicative of the margin to DNB.

The DNB design basis is as follows: there must be at least a 95 percent probability that the minimum DNBR of the limiting rod during Condition I and IJ events is greater than or equal to the DNBR limit of the DNB correlation being used (the WRB-1 correlation in this application). The correlation DNBR limit is established based on the entire applicable experimental data set such that there is a 95 percent probability with 95 percent confidence that DNBwillnotoccurwhenthem(nimumDNBRisattheDNBRlimit.

In meeting this design basis, uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered statistically such that t'iere is at least a 95% probability that the minimum DNBR for the limiting rod is greater than or equal to the DNBR limit. The uncertainties in the above plant parameters are used to determine the plant 0NBR uncertainty. This DNBR uncertainty, combined with the correlation DNBR limit, establishes a design DNBR value which must be met in plant safety analyses using values of input parameters without uncertainties.

The curves of Figures 2.1-1, 2.1-2, and 2.1-3 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than the design limit DNBR, or the average l

enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.

NORTH ANNA - UNIT 2 B 2-1 Amendment No. 71

l 4

i SAFETY LIMITS BASES The curves are based on an enthalpy hot channel factor, dH, of.1.49 and a l

i reference cosine with a peak f 1.55 for axial power shape.

An allowance is included for an increase in at reduced power based on the expression:

(H = 1.,49 [1+0.3 (1-P)]

}

where P is the fraction of RATED THERMAL POWER 1

These limiting heat flux conditions are higher than those calculated for the l

range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f(delta 1) function of the Overtemperature trip.

When the axial power I

imbalance is not within the tolerance, the axial power imbalance effect on the Overtemperature 4T trips will reduce the setpoints to provide protection consistent with core safety limits.

l 2.1.2 REACTOR COOLANT SYSTEM PRESSURE l

i The restriction of this Safety Limit protects the integrity of 'the Reactor

{

Coolant System from overpressurization and thereby prevents the release of l

radionuclides contained in the reactor coolant from reaching the containment

]

atmosphere.

_The reactor pressure vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plant which permits a maximum transient pressure of 110% (2735 psig) of design pressure.

The Reactor Coolant System piping.

l valves and fittings, were initially designed to ANSI B 31.1 1967 Edition and ANSI B 31.7 1969 Edition - (Table 5.2.1-1 of FSAR) which permits a maximum tranatent pressure of 120% (2985 psis) of component design pressure.

The Safety Limit of 2735 psig is therefore consistent with the design criteria and l

associated code requirements.

The entire Reactor Coolant System is hydrotested at 3107 psig, 125% of design pressure, to demonstrate integrity prior to initial operation.

8 1

i i

l

~

NORTH ANNA - UNIT 2 3 2-2 Amendment No. 4 f,71

o 2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the values at which the Reactor Trips are set for each parameter. The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits. Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.

Manual Reactor Trip The Manual Reactor Trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.

Power hnge. Neutron Flux The Power Range, Neutron Flux channel high setpoint provides reactor core protection against reactivity excursions which are too rapid to be protected ~

by temperature vid pressure protective circuitry. The low setpoint provides redundant protection in the power range for a power excursion beginning from low power. The trip associated with the low setpoint may be manually bypassed when P-10 is active (to of the four power range channels indicate a power level of above approximately 10 percent of RATED THERMAL POWER).

Power Range. Neutron Flux. High Rates The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of rod ejection events from any power level. Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for rod ejection from partial power.

The Power Range Negative Rate Trip provides protection for control rod drop accidents. At high power, a rod drop accident could cause local flux peaking which could cause an unconservative local DNBk to exist. The Power Range Negative Rate Trip will prevent this from occurring by tripping the reactor. No credit is taken for operation of the Power Range Negative Rate Trip for those control rod drop accidents for which the DNBR's will be greater than the applicable design limit DNBR value for each fuel type.

NORTH ANNA - UNIT 2 B 2-3 Amendment No.71

~-

i

}

I LIMITING SAFETY SYSTEM SETTINGS l

BASES Intermediate and Source Range, Nuclear Flux The Source and Intermediate Range, Nuclear Flux trips provide reactor core protection during shutdown (Modes 3, 4, and 5) when the reactor trip system breakers are in the closed position. The Source and Intermediate Range' trips in addition to the Power Range trips provide core protection during reactor startup (Mode 2). Reactor startup is prohibited unless the Source, Intermediate and Power Range trips are operable in accordance with Specification 3.31.1.

The Source Range Channels will initiate a reactor trip at about 10+$ counts per second unless manually blocked when P-6 becomes active. The Intermediate Range Channels will initiate a reactor trip at a current level proportional to approximately 25 percent of RATED THERMAL POWER unless manually blocked when P-10 becomes active.

In the accident analyses, bounding transient results are based on reactivity excursions from an initially critical condition, where the source range trip is assumed to be blocked. Accidents initiated from a subcritical condition would produce o

less severe results since the source range trip would' provide core protection at a lower power level. No credit Was taken for operation of the trip associated with the Intermediate Range Channels in the accident analyses; however..their functional capability at the specified trip settings is required by this specification to enhance the overall reliability of the Reactor Protection System.

Overtemperature Delta T The Overtemperature Delta T trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that that transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds),

and pressure is within the range between the High and Low Pressure reactor trips. This setpoint includes corrections for changes in density and heat capacity of water with ~teniperature and dynamic compensation for piping delays from the core'to the loop temperature detectors. With normal axial power distribution, this reactor trip limit is always below the core safety limit as shown in Figure 2.1-1.

If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the reactor trip is automatically reduced according to the notations in Table 2.2-1.

Operation with a reactor coolant loop out of service below the 3 loop P-8 setpoint does not require reactor protection system setpoint modification because the P-8 setpoint and associated trip will prevent DNB during 2 loop operation exclusive of the Overtemperature Delta T setpoint. Two loop operation above the 3 loop P-8 setpoint is permissible after resetting the Kl. K2, and K3 inputs to the Overtemperature Delta T channels and raising the P-8 setpoint to its 2 loop value.

In this mode of operation, the P-8 interlock and trip functions as a High Neutron Flux trip at the reduced power level.

NORTH ANNA-UNIT 2 8 2-4 Amendment No. 71

LIMITING SAFETY SYSTEM SETTINGS BASES Overpower Delta T The Overpower Delta T reactor trip provides assurance of fuel integrity, e.g., no melting, under all possible overpowcr conditions, limits the required range for Overtemperature Delta T protection, and provides a backup to the High Neutron Flux trip.

No credit was taken for operation of this trip in the accident analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System.

Pressurizer Pressure

'The Pressurizer High and Low Pressure trips are provided to limit the pressure range in which reactor operation is permitted.

The High Pressure trip is backed up by the prassurizer code safety valves for RCS overpressure protection, and is therefore set lower than the set pressure for these valves (2485 psig). The Low Pressure trip provides protection by tripping the reactor in the event of a loss of reactor coolant pressure.

The low pressure trip is blocked below P-7.

- s Pressurizer Water Level The Pressurizer High Water Level trip ensures protection against Reactor Coolant System overpressurization by limiting the water level to a volume sufficient to retain a steam bubble and prevent water relief through the pressurizer safety valves.

No credit was taken for operation of this trip in the accident analyses; however,.its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection Syster.

The pressurizer high water level trip is blocked automatically below the P-7 setpoint.

Loss of Flow The Loss of Flow trips provide core protection to prevent DNB in the cvent of a loss of one or more reactor coolant pumps.

Above 11 percent (P-7) of RATED THERMAL POWER, an automatic reactor trip will occur if the flow in any two loops drops below 90% of nominal full loop flow.

Above 31% (P-8) of RATED THERMAL POWER, automatic reactor trip will occur if the flow in any single loop drops below 90% of nominal full loop flow..

NORTH ANNA - UNIT 2 8 2-5

I 1

l LIMITING SAFETY SYSTEM SETTINGS BASES h

This latter trip will prevent the minimum value of the DNBR from going below the design limit during normal operational transients and anticipated l

transients when 2 loops are in operation a6d the Overtemperature AT trip setpoint is adjusted to the value specified for all loops in operatiori. With the Overtemperature AT trip setpoint adjusted to the value specified for 2 loop operation, the P-8 trip at 71% RATED THERMAL POWER with the loop stop valves closed in the nonoperating loop, will prevent the minimum value of the DNBR from going below the design limit during norma ~1 operational transients l

with 2 loops in operation.

1 Steam Generator Water Level The Steam Generator Water Level low-low trip provides core protection by preventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity. The specified setpoint provides allowance that there will be sufficient water inventory in the steam generators at the time of trip to allow for starting delays of the auxiliary j

feedwater system. The steam generator water level low-low trip is blocked when the loop stop valves are closed. A steam generator water level high-high j

signal trips the turbine which in turn trips the reactor if above the P-7 setpoint.

I Steam /Feedwater Flow Mismatch and Low Steam Generator Water Level The Steam /Feedwater Flow Mismatch in coincidence with a Steam Generator l

Low Water Level trip is not vsed in the transient and accident analyses but is included in Table 2.2-1 to ensure the functional capability of the specified trip setting and thereby enhance the overall reliability of the Reactor Protection System. This trip is redundant to the Steam Ge'nerator Water Level i

Low-Low trip. The Steam /Feedwater Flow Mismatch portion of this trip is activated when the steam flow exceeds the feedwater flow by greater than 1.616 x 100 lbs/ hour of full steam flow at RATED THERMAL POWER. The Steam Generator Low Water level portion of the trip is activated when the water level drops below 25 percent, as indicated by the narrow range instrument.

These trip values include sufficient allowance in excess of normal operating values to preclude spurious trips but will initiate a reactor trip before the steam generators are dry. Therefore, the required capacity and starting time requirements of the auxiliary feedwater pumps are reduced and the resulting thermal transient on the Reactor Coolant System and steam generators is l

minimized.

i l

NORTH ANNA - UNIT 2 B 2-6 Amendment No. 71 i

i

POWER DISTRIBUTION LIMITS EATFLUXHOTCHANNELFACTOR-Fg H

LIMITING CONDITION FOR OPERATION 3.2.2 F (Z) shall be limited by the following relationships:

q F (Z) 1 d [K(Z)]for P > 0.5

.l q

P F (Z) 1 [4.30] [K(Z)]for P 10.5 q

where P = THERMAL POWER RATED THERMAL POWER and K(Z) is the function obtained from Figure 3.2-2 for a given o

core height location.

APPLICABILITY : MODE 1.

ACTION:

With F (Z)_ exceeding its limit:

q Reduce THERMAL P01(ER at least 1% for each 1% Fq(Z) exceeds the a.

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 Setpoints (value of K4) have been reduced at least 1% (in AT span) fdr each 1% F (Z) exceeds the limit.

Q 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 F (Z) is demonstrated through incore mapping to be Q

within its limit.

NORTH ANNA - UNIT 2 3/4 2-5 Amendment No. 29,28, N,71

POWER DISTRIBUTION LIMITS SERVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

F (Z) shall be evaluated to determine if Fn(Z) is within,its 4.2.2.2 Q

limit by:

Using the movable incore detectors to obtain a power distribu-a.

tion map at any THERM.AL POWER greater than 5% of RATED THERMAL POWER.

I_ crea::ing the measured FQ(Z) component of the power distribu-b.

n tion 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:

M Fq(z)1 2.15 x K(z) for P > 0.5 P x N(z)

M Fg (z) 1 2.15 x K(z) for P $ 0.5 N(z) x 0.5 where Fg (z) is the measured F (z) increased by the allowances M

Q for manufacturing tolerances and measurement uncertainty, 2.15 l

is the Fo limit, K(z) is given in Figure 3.2-2, P is the relative THERMAL POWER, and N(z) is the cycle dependent function that accounts for power distribution transients encountered during normal operation. This function is given in the Core Surveillance Report as per Specification 6.9.1.7.

Measuring Fg (z) according to the following schedule:

M d.

1.

Upon achieving equilibrium conditions after exceeding the THERMAL POWER at which Fg(OWER*, orz) was last determi or more of RATED THERMAL P 2.

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

e.

With measurements indicating M

maximum

[Fg(z)\\

over z

( Mz) j M

has increased since the previous determinati,on of FQ(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 2 3/4 2-6 Amendment No. 6W, 71

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)

M Fg (z) shall be increased by 2% over that specified in 1.

4.2.2.2.c, or M

Q (z) shall be measured at least once per 7 effective 2.

F full power days until 2 successive maps indicate that

[ve*r 0

is not increasing.

( K(z) j f.

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

l 1.

Calculate the percent Fo(z) exceeds its limit by substracting one from the measurement / limit ratio and multiplying by 100:

[Fg(z)

Y-1 M

. maximum x 100 for P > 0.5 l

< om z 2.15 kP x N(z) x g(g)/

L O

-1, x 100 for P < 0.5 rz z)

}/

k0.5x s

2.

Either of the following actions shall be taken:

a.

Power operation may continue provided the AFD limits of Figure 3.2-1 are reduced 1% AFD for each percent Fg(z) exceeded its limit, or b.

Comply with the requirements of Specification 3.2.2 for Fo(z) exceeding its limit by the percent calculated a6ove.

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.

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 require-ments of Specification 4.2.2.2, an overall measured FQ(z) shall be obtained from a power distribution map and increased by 3% to account for manufactur-ing tolerances and further increased by 5% to account for measurement uncertainty.

NORTH ANNA - UNIT 2 3/4 2-7 Amendment No. J7,27.U. M, 71

1.2 i-- :.._

~

l:2: r 2

=

1.0

.u-

.+

'=

0.8 u.

-=. -

y sa g.._

taJ 16 0

+;

T 0.4 C

'S ".'.

M y-r-

.H.

~ r t --.......

k==

0.2

- n- _.

-i

W l

8 0.0 0

2 4

6 8

10 12 COREHEIGHT(FT)

Figure 3.2-2 NORMAL.!ZEDF(z)ASAFUNCTIONOFCOREHEIGHT g

NORTH ANNA - 2 3/4 2-8 Amendment No. 20,28,71 I

i

}

POWER DISTRIBUTION LIMITS NUCLEAR ENTHALPY HOT CHANNEL FACTOR -

H LIMITING CONDITION FOR OPERATION N shall be limited by the following relationship:

3.2.3 FbH P;,11.4, c1 0.3 ci-P>i THERMAL POWER

, and 4

where: P= RATED THERMAL POWER measured value of F obtained by using the movable incore detectors F

=

0 toobtainapowerdk!tributionmap.

I APPLICABILITY: MODE 1.

i

~

ACTION:

With F H'*"'*di"8 I** 1I"i"*

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 i

to less than or equalto 55% of RATED THERMAL POWER within the next 2

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ,

b. Demonstrate throu'ih in-core mapping that FfHis 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 n6xt 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and i

l

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

b., above; subsequent POWER OPERATION may proceed provided that is demonstrated through in-core mapping to be within its limit Ha nominal 50% of RATED THERMAL POWER prior to exceeding this a

THERMAL POWER, at a nominal 75% of RATED THERMAL POWER prior to I

I 1

I 4

l I

l NORTH ANNA - UNIT 2 3/4 2-9 Amendment No. 4/g,58, 71

=

POWER DISTRIBUTION LIMITS ACTION Continued exceeding this THERMAL POWER and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after attaining 95%

or greater RATED THERMAL PWER.

l l

SURVEILLANCE REQUIREMENTS 1

4.2.3.1 F

shall be determined to be within its limit by using the H

movable incore detectors to obtain a power distribution map:

' Prior to operation above 75% of RATED THERMAL POWER after each I

a.

fuel loading, and b.

At least once per 31 Effective Full Power Days.

f The provisions of Spccification 4.0.4 -are not applicable.

c.

)

~.

l

- s i

l NORTil ANNA - UNIT 2 3/4 2-10 Amendment No.11 l

l

POWER DISTRIBUTION LIMITS DNB PARAMETERS LIMITING CONDITION FOR OPERATION The following DNB related parameters shall be maintained within the 3.2.5 limits shown on Table 3.2-1:

Reactor Coolant System T,yg a.

a b.

Pressurizer Pressure Reactor Coolant System Total Flow Rate c.

APPLICABILITY: MODE 1

~

ACTION:-

With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to Tess than 5% o

- RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

s

' ~

SURVEILLANCE REQUIREMENTS Each of the parameters of Table 3.2-1 shall be verified to be within 4.2.5.1 their limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The Reactor Coolant System total flow rate shall be determined to be i

4.2.5.2 within its limit by measurement at least once per 18 months.

l f

D' 3/4 2-15 NDRTH_ ANNA - U_ NIT 2

L__

1 i

.5 TABLE 3.2-1 4

i DNB PARAMETERS

~

l l

s LIMITS e

2 Loops in Operation **

c5 3 Loops in

& Loop Stop

& Isolated Loop PARAMETER Operation Valves Open Stop ifalves Closed ro

< 591*F l

Reactor Coolant System T,yg Pressurizer Pressure

>2205,psig*

Reactor Coolant System

>289,200 gpm l

Total Flow Rate R

s

'?

E

Limit not applicable during either a THERMAL POWER ramp increase in excess of 5% RATED THERMAL POWER per. minute or a THERMAL POWER step increase in excess of 10% RATED THERMAL POWER.

[

- Values dependent on NRC approval of ECCS evaluation for these conditions.

E a

E

.N M

M G

~

l 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMFNTATION LIMITING CONDITION FOR OPERATION 3.3.1.1 As a minimum, the reactor trip system instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE with RESPONSE TIMES as shown in Table 3.3-2.

i 1

APPLICABILITY:

As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1.

SURVEILLANCE REQUIREMENTS 4.3.1.1.1 Each reactor trip. system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCIJONAL TEST operations for the MODES and at the frequencies shown in Table 4.3-1.

4.3.1.1.2 The logic for the interlocks shall be demonstrated OPERABLE prior to each reactor startup unless performed during the preceeding 92 days. The total interlock function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIERATION testing of each channel affected by interlock operation.

4.3.1.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function shall be demonstrated to be within its limit at least once per 18 months.

Each test shall include at least one logic train such that both logic trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shown in the " Total No. of Channels" coltan of Table 3.3.1.

NORTH ANNA - UNIT 2 3/4 3-1

TABLE 3.3-1 REACTORTRIPSYSTEMINSTRUMENTATIdN g

i x

E MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION E

Q 1.

Manual Reactor Trip 2

1 2

1, 2 and

12 N

2.

Power Range, Neutron Flux 4

2 3

1, 2 2f 3.

Power Range, Neutron Flux 4

2 3

1, 2 2f High Positive Rate

'4 2

3 1, 2 2#

4 Power' Range, Neutron Flux High Negative Rate g

5.

Intermediate Range, Neutron Flux 2

1 2

1, 2 and

3 3

{

6.

Source Range, Neutron Flux A.

Startup 2

1 2

2ff 4

8 Shutdown 2

1 2

3*, 4* and 5*

15 C.

Shutdown 2

0 1

3, 4 and 5 5

7 Overtemperature AT g

Three Loop Operation 3

2 2

1, 2 7f g

Two Loop Operation 3

1**

2 1, 2 9

R

=>

rt h

i e

~

TABLE 3.3-1 (Centinued)

E

=

REACTOR TRIP SYSTEM INTERLOCKS Y

ALLOWABLE DESIGNATION CONDITION SETPOINT FUNCTION VALUES d

P-7 (Cont'd) 3 of 4 Power range below 8%

>7%

Prevents reactor trip on:

setpoint Low flow or reactor coolant 2

H and pump breakers open in more N

2 of 2 Turbine Impulse.

8%

>7%

than one loop, chamber pressure below Undervoltage (RCP busses),'

setpoint Underfrequency (RCP busses),

(Power level decreasing)

Turbine Trip, Pressurizer low pressure, and Pressurizer high level.

P-8 2 of 4 Power range above 30%

<31%

Permit reactor trip on low setpoint flow or reactor coolant pump breaker open in a single (Power level increasing) loop.

3 of 4. Power range below 28%

>27%

Blocks reactor trip on low y

setpoint flow or reactor coolant pump breaker open in a single to (Power level decreasing) loop.

a G

TABLE 3.3-2 55 REACTOR TRIP SYSTEM INSTRUMCNTATION RESPONSE TIMES R

R FUNCTIONAL UNIT RESPONSE TINE 1.

Manual Reactor Trip NOT APPLICABLE i'i

[

2.

Power Range, Neutrcn Flux 5 0.5 seconds

3.

Power Range, Neutron Flux, High Positive Rate NOT APPLICABLE 4.

Power Range, Neutron Flux,

$ 0.5 seconds *

~

High Negative Rate f

5.

Intermediate Range, Neutron Flux NOT APPLICABLE 6.

Source Range, Neutron Flux

< 0.5 seconds

l ta 7.

Overtemperature AT

$ 4.0 seconds

o 8.

Overpower AT NOT APPLICABLE 9.

Pressurizer Pressure--Low 1 2.0 seconds I

10.

Pressurizer Pressure--High 5 2.0 seconds 11.

Pressurizer Water Level--High NOT APPLICABLE h"Neutrondetectorsareexemptfromresponsetimetesting.

Response of the neutron flux signal

@ portion of the channel time shall be measured from detector output or input of first electronic

" component in channel.

.E

.s e

INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2.1 The Engineered Safety Feature Actuation System (ESFAS) instrumen-tation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4 and with RESPONSE TIMES as shown in Table 3.3-5.

1 APFLICABILITY: As shown in Table 3.3-3.

ACTION:

With an ESFAS instrumentation channel trip setpoint less conservative a.

i than the value shown in the Allowable Values column of Table 3.3-4, declare the channel inoperable and apply the applic'able ACTION require-ment of Table 3.3-3*until the channel is restored to OPERABLE status with the trip setpoint adjusted consistent with the Trip Setpoint value.

l b.

With an ESFAS instrumentation channel inoperable, take the ACTION shown in Table 3.3-3.

SURVEILLANCE REQUIREMENTS _

- s 4.3.2.1.1 Each ESFAS instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shown in Table 4.3-2.

4.3.2.1.2 The logic for the interlocks shall be demonstrated OPERABLE during i

the automatic actuation logic test.

The total interlock function shall be demonstrated OPERABLE at least once per 18' months during CHANNEL CALIBRATION testing of each channel affected by interlock operation.

4.3.2.1.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function I

shall be demonstrated to be within the limit at least once per 18 months.

Each test shall include at least one logic train such that both logic trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once per N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in the " Total No. of Channels" Column of Table 3.3-3.

NORTH ANNA - UNIT 2 3/4 3-15

' TABLE 3.3-3

.g h

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM 5

TOTAL NO.

CHANNELS CHANNELS APPLICABLE l

8 FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION Eq 1.

SAFETY INJECTION, TUR8INE TRIP AND FEEDWATER ISOLATION m

a.

Manual Initiation 2

1 2

1,2,3,4 18 b.

Automatic Actuation 2

1 2

1,2,3,4 13 c.

Containment 3

2 2

1,2,3,4 14 Pressure-High A

g d.

Pressurizer 3

2 2

1,2,3 14 Pressure --

Low-Low w

e.

Differential 1,2,3 J.,

Pressure Between Steam Lines - High m

A Three Loops 3/ steam line 2/ steam line 2/ steam line 14 Operating twice and 1/3 steam lines Two Loops 3/ operating 2 * / steam 2/ operating 15 Operating steam line line twice steam line in either g

operating 3

steam line g

D f+

0 I

e w,w

w u u,.., u

.n.......,

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHAN*iEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED 3.

CONTAli4MENTISOLATION a.

Phase "A" Isolation

1) Manual N.A.

N.A.

M(1) 1, 2, 3, 4

2) From Safety Injection N.A.

N.A.

M(2) l', 2, 3, 4 Automatic Actuation Logic e'

b.

Phase "B" Isolation

1) Manual N.A.

N.A.

M(1) 1, 2, 3, 4

2) Automatic Actuation N.A.

N.A.

M(2) 1, 2, 3, 4 Logic h

3) Containment Pressure--

S R

M(3) 1, 2, 3 High-High o

I

)

\\

~

TABLE 4.3-2 S

-4 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 2

E CHANNEL MDDES IN WHICH e

CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCT/0NALUNIT CHECK CALIBRATION TEST REQUIRED c;="

1.

SAFETY INJECTION, TUR8INE TRIP AND FEEDWATER ISOLATION 4

a.

Manual Initiation N.A.

N.A.

M(1) 1, 2, 3, 4 b.

Automatic Actuation Logic N.A.

N.A.

M(2) 1, 2, 3, 4 c.

Containment Pressure-High S

.R M(3) 1, 2, 3, 4 d.

Pressurizer Pressure--Low-Low S

R M

1,2,3 1

e.

Differential Pressure S

R M

1,2,3 l {

8etween Steam Lines--High Y

f.

Steam Flow in Two Steam S

R M

1,2,3 U

Lines--High Coincident with T

--Low-PPIIsure-Low or Steam Line Low 2.

CONTAIMENT SPRAY k

a.

Manual Initiation N.A.

N.A.

M(1) 1, 2, 3, 4 E

2 b.

Automatic Actuation Logic N.A.

N.A.

M(2) 1, 2, 3, 4 5

2 c.

Containment Pressure--High-S R

M(3) 1, 2, 3 P

High 2

D 0

~

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) i f.

By verifying that each of the following pumps develop the indicated discharge pressure (after subtracting suction pressure) on recircula-tion flow when tested pursuant to Specification 4.0.5.

1.

Centrifugal charging pump greater than or equal to 2410 psig.

2.

Low head safety injection pump greater than or equal to 156 psig g.

By verifying that the following manual valves requiring adjustment to prevent pump " runout" and subsequent component damage are locked o

and tagged in the proper position for injection:

1.

Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following completion of any repositioning or maintenance on the valve when the ECCS subsystems are required to be OPERABLE.

2.

At least once per 18 months.

1.

2-51-89 Loop A Cold Leg 2.

2-SI-97 Loop B Cold Leg 3.

2-5I-103 s Loop C Cold Leg 4.

2-SI-116 Loop A Hot Leg 5.

2-51-111.,

Loop 8 Hot Leg 6.

2-SI-123 Loop C Hot Leg

~

h.

By performing a flow balance test, during shutdown, following completion of modifications to the ECCS subsystems that alter the subsystem flow characteristics and verifying that:

1.

For high head safety injection lines, with a single pump running:

a)

The sum of the injection line flow rates, excluding the highest flow rate, is > 384 gpa, and b)

The total pump flow rate is < 650 gpm.

NORTH ANNA - UNIT 2 3/4 5-5

~

EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - T less than 350*F ava LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

a.

One OPERABLE centrifugal charging pump #,

b.

One OPERABLE low head safety injection pumpf, and An OPERABLE flow path capable of automatically transferring fluid c.

to the reactor coolant system when taking suction from the refueling water storage tank or from the containment sump when the suction is transferred during the recirculation phase of operation.

APPLICABILITY: MODE 4.

i l

ACTION:

a.

With no ECCS subsystem OPERABLE because of the inoperability of either the centrifugal charging pump or the flow path from the refueling water storage tank, restore at least one ECCS subsystem

~

to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in COLD SHUTDOWN within the next 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months , b.

With no ECCS subsystem OPERABLE because of the inoperability of the low head safety injection pump, restore at least one ECCS subsystem to OPERABhE status or maintain the Reactor Coolant System T,yg less than 350' by use of alternate heat removal methods.

c.

In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected safety injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

i A maximum of one centrifugal charging pump and one low head safety l

injection pump shall be OPERABLE whenever the temperature of one or more i

l of the RCS cold legs is less than or equdl to 340'F.

I NORTH ANNA - UNIT 2 3/4 5-6 Amendment No. 71 l

l

=

3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Modera.te Frequency) events by.:

(a) maintaining the minimum DNBR in the core from going beyond the design limit DNBR during normal operation and in short term transients, and (b) limiting the fission gas release, fuel pellet temperature & cladding mechanical properties to within assumed design criteria.

In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200'F is not exceeded.

The definitions of certain hot channel and peaking factors as used in

' these specifications are as follows:

F (Z)

Heat Flux Hot Channel Factor, is defined as the maximum local 0

heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods.

F Nuclear Enthalpy Rise Hot Channel Factor, is defined as the H

ratio of the integral of linear power along the rod with the highest, integrated power to the average rod power.

3/4.2.1 AXIAL FLUX DIFFERENCE (AFD)

~

The limits on AXIAL FLUX DIFFERENCE assure that the Fn(Z) upper bound envelope, as given in Specification 3.2.2, is not exceeded during either normal operation or in the ' event of renon redistribution following power changes.

i NORTH ANNA - UNIT 2 B 3/4 2-1 Amendment No. 29,28,H, 71 1

A O

POWER DISTRIBUTION LIMITS BASES Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer determines the one minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are outside the allowed AI-power operating space and the THERMAL POWER is greater than 50% of-RATED THERMAL POWER.

.o

~.

- ~

NORTH ANNA - UNIT 2 B 3/4 2-2 Amendment No. 77, 64

I 4

POWER DISTRIBUTION LIMITS BASES N

When F is measured, 4% is the appropriate experimental error allowance for a full fre map taken with the incore detection system. The specified limit for F egntainsa4%errorallowance. Normal operation will result inameasur$gdF less than or equal to 1.49. The 4% allowance is based on thefollowingceWsiderations:

abnormal perturbations in ghe radial power shape, such as from a.

more directly than F,

rod misalignment, effect FAH q

although rod movement has a direct influence upon limiting F b.

to withgn its limit, such control is not readily available th limit FAH, and c.

errors in prediction for control power shape detected during startup physics tests can be compensated for in Fggyrestricting axial flux distributions. This compensation for F is less AH readily available.

Fuel rod bowing reduces the value of DNB ratio. Credit is available to offset this reduction in the margin available between the safety analysis design DNBR values (1.57 and 1.59 for thimble and typical cells, respectively) and the limiting design DNBR values (1.39 for thimble cells and 1.42 for typical l

f cells). The applicable value of rod bow penalties can be obtained from the FSAR.

The hot channel factor F (Z) is measured periodically and increased by l

a cycle and height dependent ~

wer factor, N(Z), to provide assurance that the limit on the hot channel factor, Fn(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 burnup condi-tions in the core. Th'e N(Z) function for normal operation is provided in the Core Surveillance Report per Specification 6.9.1.7.

3/4.2.4 QUADRANT POWER TILT RATIO The quadrant power tilt ratio limit assures that the radial power distri-bution satisfies the design values used in the power capability analysis.

Radial power distribution measurements are made during startup testing and periodically during power operation.

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 tilt condition greater than 1.02 but 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 is reinstated by reducing the power tilt, the margin for uncertainty on Fg by 3 percent for each percent of tilt in excess of 1.0.

NORTH ANNA - UNIT 2 B 3/4 2-5 Amendment No. 75,55,54,71 l

POWER DISTRIBUTION LIMITS BASES For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the movable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full.incore flux map or two sets of 4 symmetric thimbles. 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.

3/4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR greater than the design limit throughout each analyzed transient.

Measurement uncertainties must be accounted for during the periodic surveillance.

~

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months periodic surveillance of these parameters thru instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation. The 18 month periodic measurement of the RCS total flow rate is adequate to detect flow degradation and ensure correlation of the flow indication channels with measured flow such that the indicated percent flow will provide sufficient veri.fication of flow rate on a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months basis, s

NORTH ANNA - UNIT 2 B 3/4 2-6 Amendment No. 77, 6A 71

3/4.7 PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line code safety valves ensure that the secondary system pressure will be limited to within 110% of the system design pressure, during the most severe anticipated system operational transient.

The maximum relieving capacity is associated with a turbine trip from 100%

RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).

The specified valve lift settings and relieving capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure Code,1971 Edition. The total relieving capacity for all safety 6 lbs/hr which is greater than c

valves on all of the steam lines is 12.83 x g0lbs/hr at 100% RATED THERMAL the total secondary steam flow of 12.77 x 10 POWER. A minimum of 2 OPERABLE safety valves per steam generator ensures that sufficient relieving capacity is available for the allowable THERMAL POWER restriction in Table 3.7-1.

STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis

-of the reduction in secondary system steam flow and THERMAL POWER required by the reduced reactor trip settings of the Power Range Neutron Flux The reactor trip-setpoint reductions are derived on the following channels.

bases:

For 3 loop operation SP = (X)

(Y)(V) x 109 X

For 2 loop operation with stop valves closed SP = (X)

(Y)(U) x 71 X

NORTH ANNA - UNIT 2 B 3/4 7-1 Amendment No. 71

/

l PLANT SYSTEMS t

BASES For 2 loop operations with stop valves open 3p,(X) - (Y) (U) x 66 h,

Where:

1 l

SP = reduced reactor trip setpoint in percent of RATED THERMAL POWER V=

maximum number of inoperable safety valves per steam line maximum number of inoperable safety valves per operating

~

U =

steam line Power Range Neutron Flux-High Trip Setpoint for 3 loop I

109

=

operation Maximum percent of RATED THERMAL POWER permissible by 71

=

P-8 Setpoint for.2 loop operation with stop valves closed.

Maximum percent of RATED THERMAL POWER permissible 66 =

by P-8 setpoint for 2 loop operation with stop valves open.

l X=

Total relieving capacity of all safety valves per steam l

line in 1bs/ hour = 4,275,420 Y=

Maximum relieving capacity of any one. safety valve in 1bs/ hour = 855,084 314.7.1.2 AUXILIARY FEEDWATER SYSTEM l

The OPERASILITY of the auxiliary feedwater system ensures that the Reactor Coolant System can be cooled down to less than 350*F from normal operating cenditions in the event of. a total loss of off-site power.

l s

i

.

ML20206Q388

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