ML20116N875

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Proposed Tech Spec Changes for 2,905 Mwt NSSS Rating
ML20116N875
Person / Time
Site: North Anna  Dominion icon.png
Issue date: 01/31/1985
From:
VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML20116N845 List:
References
NUDOCS 8505070384
Download: ML20116N875 (49)
v • d • e


Text

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ENCLOSURE 3 NORTH ANNA UNITS 1 AND 2

{ TECHNICAL SPECIFICATION CHANGES FOR 2905 MWt NSSS RATING I

s 4

i VIRGINIA POWER JANUARY 1985 a

8505070394 850502 PDR ADOCK 05000338

P PDR _.

.I 45-WBR-2010N-5

~

n North Anna Core Uprating-(2893 N t RATED THERMAL-POWER)

Facility.0perating License Affected Pages Unit 1 4

Unit;2 3

Technical Specifications Affected Pages Unit 1 1-5, 2-2, 2-6, 2-8, 2-9, 2-10, B2-1, B2-2, B2-4, B2-6, 3/4 2-5, 3/4 2-8, 3/4 2-9, 3/4 2-11,'3/4 2-15 B3/4 2-1, B3/4 2-4, B3/4 2-5, B3/4 2-6, B3/4 7-1, 3/4 2-10 Unit 2

1-5, 2-2, 2-6, 2-8, 2-9, 2-10, B2-1, B2-2,'B2-3, B2-6, 3/4 2-5, 3/4 2-8, 3/4 2-9, 3/4 2-10,-3/4 2-11,'3/4 2-16, B3/4 2-1, B3/4 2-5,-B3/4 2-6, B3/4.7-1 e

_ _ .. -. ._. ._ _ _ _ .. _.. _._ _ . __ _ ~ _ _ _ _ . . _ _

.(.

1 a (1) Maximum Power Level

$ VEPCO is authorized to operate the North Anna Power Station,-Unit No. 1, at reactor core power levels not in excess of.2893 megawatts (thermal).

(2) Technical Specifications E

'The' Technical Specifications - contained in Appendices A and B, as revised through Amendment i

No. 63, are hereby-incorporated in the license. .VEPCO shall operate the facility in accordance with the Technical Specifications.

(3) AdditionalCondi5 ions The matters specified in the following conditions shall- be completed ~to. the. satisfaction of the

Commission within the stated time periods following the issuance of this! amendment or' within the -

f operational restrictions indicated. The removal of

these conditions shall be made by an amendment to the license supported by a favorable evaluation by-i the-Commission:

i

c. Virginia Electric and Power Company shall not operate the reactor in operational modes 1 and 2 4 with.less than three reactor coolant pumps in operation.
e. If Virginia Electric and Power Company plans to remove or to make significant changes in the normal operation of equipment that controls the
amount of radioactivity in effluents from the North Anna Station, the Commission shall. be l

notified . in writing regardless of whether the ~

change affects the amount of radioactivity in t the effluents. .

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,A' e :1.0 DEFINITIONS-(Continued) 1-

-QUADRANT POWER TILT RATIO I'23..

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

1 calibrated (output , to the . average of the lower excore detector - calibrated outputs, ' whichever is greater. . With one excore detector inoperable, the

[ remaining three de'tectors 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.

REACTOR TRIP SYSTEM RESPONSE TIME

, 1.25 'The REACTOR' TRIP' SYSTEM RESPONSE TIME shall be the time interval from i- when the monitored parameter exceeds its trip setpoint at the channel sensor 1 .until loss of stationary gripper coil voltage.

REPORTABLE EVENT i

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

k SHUTDOWN MARGIN-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.

~

-SITE BOUNDARY I 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 that meets shipping and burial ground requirements.

SOURCE CHECK 1

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.

1 .

~

NORTH ANNA - UNIT'1- 1-5 7

l Nominal T,y,= 566.8'F

, Wominal RCS flow = 289200 GPM , .

663 -

' 555 - 2400 psia 658 .

445 -

2250 psia 1

840 655 -

650 -

2000 psia

_ $25

- s2s .

.I sis - 1860 psia

'sie -

I ses

-688 -

595 - .

598 -

585..

588 675

8. .I .2 .5 4 .5 .s .7 .8 9 8. l.I l.J P0vCR treaction or nomin.1

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Figure 2.1-1 REACTOR CORE SAFETY LIMITS FOR THREE LOOP OPERATION

-)

NORTH ANNA - UNIT 1 2-2

'I

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.8. TABLE 2.2-1 g

c g -REACTOR TRIP SYSTEM-INSTRUMENTATION TRIP SETPOINTS W' TRIP SETPOINT- ALLOWABLE VALUES

- FUNCTIONAL UNIT .

~

$ 'l. Manual Reactor Trip Not Applicable Not Applicable, i a

W 2. Power Range.-Neutron Flux' Low Setpoint - 5 25% of RATED Low Setpoint - s 26% of RATED THERMAL POWER THERMAL POWER ~

High Setpoint- s 109% of RATED High Setpoint - s 110% of RATED-THERMAL POWER THERMAL POWER-3.- Power Range,' Neutron Flux, .s 5% of RATED THERMAL POWER with 5 5.5% of RATED THERMAL POWER High Positive Rate; a' time constant 2 2 seconds- with a time constant 2 2 seconds i 4. Power Range, Neutron Flux, .s 5% of RATED THERMAL POWER with 5 5.5%'of RATED THERMAL POWER High Negative Rate- a time constant 2 2 seconds with a time constant 2 2 seconds n 5. Intermediate Range, Neutron s 25% of RATED THERMAL POWER s 30% of RATED THERMAL POWER-5 Flux 5 5

6. Source Range,' Neutron Flux s 10 counts per second s 1.3 x 10 counts per.second
7. Overtemperature AT See Note 1 See Note 3 8.. Overpower AT .See Note 2 See Note 3
9. -Pressurizer Pressure-Low 2 1870 psig 2 1860 psig j
10. Pressurizer Pressure-High s 2385 psig s 2395 psig i

11.. Pressurizer Water Level-High 5 92% of instrument span s 93% of instrument span

12. Loss of Flow 2 90% of design flow per loop
  • 2 89% of design flow per loop *
  • Design flow is 96,400 gpm per loop. .

i

s TABLE 2.2-1 (Continued) -

55 .

Bl REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS m ,

E; NOTATION E ,1+ T S, g

c: NOTE 1: Overtemperature ATs AT, [K g-K2 (T-T')+K3 ( ~ ~}~ k (OI) 3

$ ,1 + T S,

. g 2

. n.

Where: AT, = Indicated AT at RATED. THERMAL POWER .

T = Average temperature. *F T' = Indicated Tavg at. RATED THERMAL POWER s 586.8'F P = Pressurizer pressure, psig P' = 2235 psig (indicated RCS nominal operating pressure) 1+Ta S = The function generated by the lead-lag controller for T,y dynamic compensation y 1+T28 I

  • = Time constants utilized in the lead-lag controller for T T =.25 secs, T &T avg 1 2 T = 4 secs.

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a I a g 2 3 iht et eE pnvW abP v R ro K K K dt s abP e T en ie eiO m m p( d wt 4 ctP lL lL e 4 rc elA elA S R O t ;p eel h aM ham C O - ppa thR C T asu thR E C crt sM - sE sE A ier neeR t H t H f E db a e r r.E atT atT R nmt ea H h n h n o i as w eT tid ti D l h t br oE oE ect eqoD t pT t pT a h n b cE nt A ntA v t na d T eeR eeR o ol bn e A cs cs r r t r t p fi p 9ahR epa epa p o rg t f pi pi a nan ~ I o re re C oei qo ht u ht u il r t c l c l R s t cu te s n aTa aTa N p cud gree eAv eAv o nn evc n o u e rhl r res res .

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  • D g = a A S N D E g a d d r t T Y E T n r e e e o S T . A i e 6 t n z p n

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2.1 SAFETY LIMITS BASES J.

- 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 i temperature is slightly above the coolant saturation temperature.

$. Operation' above the upper. boundary of the nucleate boiling regime could l result -in excessive -cladding temperatures because of the onset. of departure from nucleate boiling (DNB) and the resultant-sharp reduction in heat transfer i 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 f and non-uniform heat! flux distributions. The local DNB heat flux ratio, DNBR, 7 defined as the ratio of the heat flux that would'cause DNB at a-particular
core location to the local heat flux, is indicative of the margin to DNB.

l . lThe 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

. II events is . greater than or equal to the DNBR -limit of the DNB - correlation j being used (the WRB-1 correlation -in this application). The correlation DNBR j limit.is established based on the entire applicable experimental data set such 1

that there is 'a 95 percent' probability with: 95 percent confidence that DNB will.not occur when the minimum DNBR is at the DNBR limit.

t In meeting this. design basis, uncertainties >in plant- operating

parameters, nuclear and thermal parameters -and fuel- fabrication parameters l are considered statistically 'such that 'there- 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 DNBR 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 POWElt, Reactor. Coolant System ' pressure and average ~ temperature for F

- which the. minimum DNBR is no.less than the design limit'DNBR, or the average

. enthalpy-at the vessel exit is equal to the enthalpy of saturated 111guid. ]

I

+

NORTH ANNA - UNIT 1 B 2-1 6 .

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' SAFETY LIMITS

- BASES.

'The curves are based on an-enthalpy hot channel factor, F of 1.49 and l

- areferencecosinewithapeagof1.55foraxialpowershape. kn,allowanceis included for an increase in FAH, at reduced power based on the expression:

H.

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

~ where P is the fraction of RATED THERMAL POWER

.These limiting heat flux conditions are higher than those calculated for

{ the 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(AI) function of the Overtemperature trip.- When the axial power imbalance is

- not. within- the;-tolerance, the axial power imbalance offact on the-Overtemperature - AT trips will reduce . the setpoints to provide protection consistent with core safety limits.

2.1.2 REACTOR COOLANT SYSTEM PRESSURE The - restriction of ' this Safety Limit protects the integrity of the Reactor: Coolant- System from overpressurization and. thereby prevents the

release of radionuclides . contained in the reactor coolant from reaching the containment atmosphere.-

~Th'e' 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%1(2735 psig) of design pressure'. The Reactor Coolant System

~

l piping, valves and' fittings , were initially designed Lto ANSI 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% J (2985 ~psig) of component design pressure.

[ The Safety Limit of 2735 psig is therefore consistent with the design criteria-j- - 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.

1 L tf e

l-NORTH' ANNA - UNIT 1 ' s 2-2 o

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LIMITING SAFETY SYSTEM SETTINGS E BASES i

The 'Powe'r Range . Negative Rate Trip provides protection for control rod I

drop c accidents. - At high power, a rod drop .4ccident 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 i than the applicable design limit DNBR value for each fuel type.

Intermediate and Source Range, Nuclear Flux The . Intermediate and Source Range, .' Nuclear Flux trips provide reactor

core protection- during reactor startup. These trips provide redundant
protectiontothelowsetpointtripofthePowerRange.NeutronFluxgannels.

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. No credit was taken for operation of the trips j . associated with either the Intermediate . or Source 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 i of the Reactor Protection System.

li Overtemperature AT i

i.'

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

transit delays from the core to the temperature detectors (about 4 seconds),
and pressure is within the range between the High and Lov.' Pressure reactor trips. This setpoint includes corrections ' for changes in density and heat capacity of water with temperature and dynamic-compensation for piping delays l 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 e ' 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. ,

c i NORTH ANNA - UNIT-1 B 2-4

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-l 1 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 ,

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 i

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 adjusted 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 t

from going below the design limit during normal operational transients with 2 l 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

-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 feedwater ' system. The . steam generator-1 water leve171ow-low trip ' is blocked when the loop.stop valves are closed. A steam' generator water level high-high signal. trips the turbine which in turn trips the reactor if above the P-7 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 ,

t i

e NORTH ANNA - UNIT'1 .B 2-6 m

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$3 _ j POWER DISTRIBUTION LIMITS HEAT FLUX HOT CHANNEL FACTOR-F (Z)

LIMITING CONDITION FOR OPERATION i

3.2.2 'F (Z) shall be limited by the following relationships:

q .

-Fq (Z) s [. 15] [ K(Z)] for P>~0.5 Fq (Z) s [4.30]_ [K(2)] .for Ps 0.5 where P = THERMAL POWER y RATED THERMAL POWER and K(Z) is the function obtained from Figure 3.2-2 for a given

~

1

' core height location.

' APPLICABILITY: MODE 1.

ACTION:= ,

L With F (Z) exceeding'its limit:

9

a. Comply with either of the following ACTIONS:

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

. 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,-

-above; THERMAL POWER may then be' increased providedgF (Z) is demonstrated through incore mapping to be within its limit.

6

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COREHEIGHT(FT)

Figure 3.2-2 NORMALIZEDF(z)ASAFUNCTIONOFCOREHEIGHT g

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! NORTH ANNA - UNIT 1 3/4 2-8 l .

I

POWER DISTRIBUTION LIMITS

' NUCLEAR'ENTHALPY HOT CHANNEL FACTOR'- F H

-LIMITING CONDITION FOR OPERATION 3.2.3 Fg shall.be limited by the following relationship:

H THERMAL' POWER where: P= RATED THERMAL POWER obtained by using the movable incore detectors rL-measur.dva1ueof7"!tributionmap.

toobtainapowerdk APPLICABILITY: MODE.1.

ACTION:

With Fg exceeding its limit:

a.. Reduce THERMAL POWER to-less than 50% of RATED THERMAL POWER within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and reduce the Power Range Neutron Flux-High Trip Setpoints to s 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, b'. Demonstrate through in-core mapping that Fg is within its limit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding the limit or reduce THERMAL POWER.to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and

c. Identify and correct the cause of the out of limit condition prior to' increasing THERMAL POWER above the reduced limit required by a o b, above; subsequent POWER OPERATION may proceed provided that AH is demonstrated through'in-core mapping to be within its limit at a nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THERMAL POWER prior to exceeding this THERMAL POWER and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after attaining 95%

or greater RATED THERMAL POWER.

NORTH ANNA UNIT 1 3/4 2-9 l.

i.

L

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS-4.2.3.1- F .shall-be determined to be within its limit by using the. movable incoredethktorstoobtainapowerdistributionmap:-

a. Prior to operation above 75% of RATED THERMAL POWER af ter 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.

].

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i NORTH ANNA - UNIT 1 3/4 2-10

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. NORTN ANNA . UNIT-1 3/4 2-11 U.----_-._--_____-_____J____-____---______-___-________

l 5 TABLE 3.2-1 l 4

  • ~

DNB PARAMETERS 5 LIMITS l

i .

2 Loops in Operation ** 2 Loops in Operation **

g 3 Loops in & Loop Stop & Isolated Loop p Operation Valves Open Stop Valves Closed PARAMETER l _

Reactor Coolant System T,y s591*F Pressurizer Pressure 22205 psig*

Reactor Coolant System 2289,200 gpa

. Total Flow Rate n

Y G

  • Limit not applicable during either a THERMAL POWER ramp increase in excess of 5% RATED THERMAL POWER per minute or a 111ERMAL POWER step, increase in excess of 10% RATED THERMAL POWER.
    • Values dependent on NRC approval of ECCS evaluation for these conditions.

9

~ . _

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 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:

Fq(Z) Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for man-i ufacturing tolerances on fuel pellets and rods.-

I 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.

F Radial Peaking Factor, is defined as the ratio of peak power gy(Z) density to 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 Fq (Z) upper bound 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 dif ference 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 nscessary to reflect core burnup considerations. ,

NORTd ANNA - UNIT 1 B 3/4 2-1 y

2 i

' POWER DISTRIBUTION LIMITS

,_ BASES 3/4.2.2 and 3/4.223 HEAT FLUX AND NUCLEAR ENTHALPY HOT CHANNEL FACTORS -

Fq(Z) and F H The limits fon heat flux and nuclear anchalpy 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 normally only be determined periodically as specified in Specifications 4.2.2. and 4.2.3.

This periodic surveillance is sufficient to insure that.the hot channel factor limits are maintained provided:

a. Control rod in a single group move together with no individual rod -

insertion differing by more than 212 steps from the group . demand position.

'b. Control rod groups are sequenced with overlapping groups as described

--in Specification 3.1.3.6.

c. The: control rod insertion limits of Specifications 3.1.3.5 -and 3.1.3.6 are maintained.
d. The axial power - distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is' maintained within the limits.

The relaxation in F as a' function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits. .F will be H

maintained within its' limits provided , conditions a thru d above, are maintained.

-When an F measurement is taken, both -experimental error and manufacturingto13rancemustbeallowedfor. 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 ontains a 4% error ! allowance. Normal operationwillresultinameasuNd s 1.49. The 4% allowance is based on. .

the following considerations: H NORTH ANNA - UNIT 1 B 3/4 2-4

_ ;_.._ .. _ _ _ , _ _ , _ . - - ~ _ _ _- _ .

y POWER = DISTRIBUTION LIMITS BASES

a. abnormalperturbationsingheradialpowershape,suchasfrom rod misalignment, effect FAH * #* *" 7 ""

Q'

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

towithgn-itslimit,suchcontrol-isnotreadilyavailableto limit FAH ,fand

c. errors in prediction for control power shape deteeted during startup. physics' tests.can be compensated for in F O I "~

ing' axial flux distributions. Thiscompensationfor[k"8 H i'

less readily available.

Fuel rod bowing' reduces the'value of DNB ratio. Credit is available

, 'to offset this -reduction in the margin available betwaan 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 cells). .The applicable value of rod bow penalties can be obtained from the FSAR.

i .

! -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 Ifmit cf 1.02 at which corrective action'is required provides DNB and j linear heat generation rate protection with x-y. plane power' tilts.

i The two hour time allowance for operation with a tile condition greater i 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 tilt, the margin for uncertainty on Fn is reinstated by reducing the power by 3 percent for each' percent of tilt in 1xcess of 1.0.

For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the moveable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the 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.

h f

i i

NORTH ANNA - UNIT 1 B 3/4 2-5 1.

m , # - -

-a,rwu-,,,- -w- w -r .,eoe --,--,---m-----nr-- w e ,-%mm +

3- m 7- ' ~ ~ ~ ~- -- - - - - ~ -- '-'- - - - - --

(

y 1

. - +,  : ,

, 1 POWER DISTRIBUTION LIMITS 1- ,

i(

f(7 _ BASES"

+ ,

3/S.2.5 DNB' PARAMETERS-t

~

-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 ~surve111ance.

, . ~ The' 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> - periodic surveillance of . these parameters i 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 j verification of flow rats on a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> basis.

i 3/4.2.6 AXIAL POWER DISTRIBUTION-l The limit on axial power distributilon ensures that' qF 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 q .is necessary t in order toi. 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, wheIe,K(Z) is the graph shown in. Figure 3.2-2. The' amount

< ofpokentialviolationisdeterminedbyaubtractingIfromthemaximum. ratio of.the predicted Ff(Z) analysis ~(flyspeck) results for a particular fuel' cycle tolthe Fg xK(Z) limit. This amount of potential. violation, in percent, is-5- subtracted from 100% to determine the value for P .' If P is equal to 100%,

Jno .' axial power distribution surveillance . is re"uired. " P" will not exceed i- - 100%.

L f' .

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- ' NORTH ANNA - UNIT 1- B 3/4 2 -

1:

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- . . , - . . . - ., . . - ,. , - ,. .-..._,.a-. .. ,_ - .

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 reli ving capacity for all safety valves on all.of the steam lines is 12.83 x 1 lbs/hr which is greater than the total secondary steam flow of 12.-77 x 10 lbs/hr at 100% RATED THERMAL 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 *he Power Range Neutron Flux channels. The reactor trip setpoint reductions are derived on the following 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 77 X

For ,2 loop operations with stop valves open SP = (X) - (Y) (U) x 66 X

I NORTH ANNA - UNIT 1 B 3/4 7-1 w - - - -

. ~ . - - . . . - . - . . . .. - . - -- - - - - _ , . - --

A i .

(4) Pursuant to the Act and 10 CFR Parts 30, 40 and 70. VEPC0 to receive, possess = and -- use - in amounts as required any ' byproduct, source or special nuclear material without restriction to chemical or physical i.. form. for . sample analysis or' instrument calibration or associated '

with radioactive apparatus or components; and (5) Pursuant to the Act and 10 CFR Parts 30, 40 and 70 VEPCO to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

- C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I i 1.

and is subject to all applicable provisions of the Act and to the rules, t regulations, and orders of the Commission now or hereafter in effect; and

  • ;is subject to the additional conditions specified or incorporated below:  !

(1) Maximum Power Level  !

I Vepco 's i authorized to operate the' facility at steady state 1 reactor core power levels not in excess of 2893 megawatts (thermal). l (2)lTechnicalSpecifications

) The Technical Specifications contained in ' Appendices ~ A and B, 4 as revised through Amendment No. 47, 'are hereby incorporated in the .

license. VEPCO shall operate the facility in accordance with the Technical Specifications.

(b) The current surveillance period for Surveillance Requirement l'

4.7.10.c may be extended beyond the time limit specified by j Technical Specification' 4.0.2.a. The ' required surveillance 4 shall be' completed prior to startup af ter the first refueling l outage.. The plant shal1~not be operated in Modes 1, 2, 3 or 4 until Surveillance Requirement 4.7.10.c has. been completed.

I' Upon -accomplishment of the surveillance, the provisions of

{ 4.0.2.s shall apply.

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- + - - , , + . pr, - - , ---+,e,- - se rn.er ---

, , , , - ~ , ,--e,,.-g,--, p- -- e nv---e ,,m-,~

- 1.'O DEFINITIONS (Continued) i.

QUADRANT POWER TILT RATIO l'.23 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore 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 3 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 to the reactor coolant of 2893 MWt. l 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 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 1.27 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be suberitical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

SITE BOUNDARY 1.28 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 that meets shipping and burial ground requirements.-

SOURCE CHECK 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 2 1-5 I

l Nominal T,yg= M6.8'F

! , m=inal RCS flov = 289200 CFM , ,

663 -

655 2400 psia 1 650 ,

445 - 2250 psia 640 l ,

655

! 650 625 2000 psia

'l W ses

[ 615 1860 psia

" GIS -

Set 480 6% ,

54 l sol -

t les i 675

8. .I .2 .5 4 .5 .6 7 .e .9 4. 1.8 1.1 P0vte treaction of neeln.Il

! Figure 2.1-1 REACT 0m COM SAFETY LMIT4 POR TIME LOOP OPERATION i

I i

i NORTH ANNA - UNIT 2 2-2

l l

z 3 TARLE 2.2-1 i

Y

> REACTOR TRIP SYSTDI INSTRtBIENTATION TRIP SETPOINTS z

ALLOWABLE VALUES

{ FL E ONAL UNIT TRIP SETPOINT Not Applicable

.E 1. Manual Reactor Trip Not Applicable U Low Setpoint - s 26% of RATED l u 2. Power Range, Neutron Flux Low Setpoint - s 25% of RATED THERMAL POWER THERMAL POWER l

High Setpoint- s 109% of RATED High Setpoint - s 110% of RATED t THERMAL POWER THERMAL POWER l

3. Power Range, Neutron Flux, s 5% of RATED THERMAL POWER with 5 5.5% of RATED THERMAL POWER High Positive Rate a time constant 2 2 seconds with a time constant 2 2 seconds i
4. Power Range, Neutron Flux, s 5% of RATED THERMAL POWER with s 5.5% of RATED THERMAL POWER High Negative Rate a time constant 2 2 seconds with a time constant 2 2 seconds l
5. Intermediate Range Neutron s 25% of RATED THERMAL POWER 5 30% of RATED THERMAL POWER l w i

& Flux -

5

6. Source Range, Neutron Flux s 105 counts per second s 1.3 x 10 counts per second
7. Overtemperature AT See Note 1 See Note 3
8. Overpower AT See Note 2 See Note 3
9. Pressurizer Pressure-Low 2 1870 psig 2 1860 p'sig
10. Pressurizer Pressure-High 5 2385 psig s 2395 psig
11. Pressurizer Water Level-High s 92% of instrument span s 93% of instrument span
12. Loss of Flow 2 90% of design flow per loop
  • 2 89% of design flow per loop *
  • Design flow is 96,400 gym per loop.

l l

TABLE 2.2-1 (Continued) 8 REACTOR TRIP STSTEM INSTRUMENTATION TRIP SETPOINTS h

$ NOTATION 5

'l+ r gS '

NOTE 1: Overtemperature AT 5 AT, [Kg - (T-T')+K3 (F-P')-f g(AO ]

l+T 8 y s 2s n

Were: AT o = Indicated AT at RATED THERMAL POWER T = Average temperature. *F T' = Indicated Tavg at RATED THERMAL POW 9 s 586.8*F P = Pressurizer pressure, psig P' = 2235 psig (indicated RCS nominal operating pressure) 1+T S = The function generated by the lead-lag controller for T dynamic compensation

" 8 1+T2S cm T

3+ty = Time constants utilized in the lead-lag controller for T, T g

= 25 secs, 2 = 4 secs.

S = Laplace transform cperator (sec~ )

t t

l l

I I

i TABLE 2.2-1 (Continued)

~

E REACTOR TRIP SYSTEM INSTRIBEENTATION TRIP SETPOINTS

!  := -

!  :=

NOTATION (Continued) operation with 3 Loops Operation with 2 Loops Operation with 2 Loops -

8 (no loops isolated)* (1 loop isolated)*

E Kg =

( ) Kg = ( )

y Kg = 1.264 Ky = 0.0220 K =

( ) K 2

( )

2

( )

~

K = 0.001152 K = ( ) K =

3 3 3 l

and f g ( I) 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: 1

  • (1) for q .- between - 44 percent and + 3 percent, f (AI) =0 l (where q qb g are percent RATED THERMAL POWER in tb top and bottom a I l b halvesofthecorerespectively,andq I" '

l t

  • 9 b I" * **

I percent of RATED THERMAL POWER).

(ii) for each percent that the magnitude of (q - q ) exceeds - 44 rercent, b

i the AT trip setpoint shall be automaticalIy reduced by 1.67 percent of l its value at RATED THERMAL POWER.

(iii) for each percent that the magnitude of (q - q ) exceeds + 3 percent, b

theATtripsetpointshallbeautomaticalIyreducedby2.00percentof its value at RATED THERMAL POWER.

l t

I l~

t

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

I

\

ij TABLE 2.2-1 (Continued)  ;

l z

t

$' REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS' e NOTATION (Continued)

E

'8 / .T S

- E- . NOTE 2: Overpower -AT s AT,[K4-K5 1+ T S ,

- - ' 6 (T-f)-f (AI)] 2 '

Q 3 M

Where: AT = Indicated AT at RATED THERMAL POWER o

T = Average temperature. *F T" = Indicated Tavg at RATED THERMAL POWER 5586.8*F 1.079

~

K = l 4

K = 0.02/*F for increasing average temperature 5

K = 0 for decreasing average temperatures u 3

-,8 O .K6 .=. 0.00164 for T > T";.K6 = 0 for T s T" l

rS- = The function generated by the rate lag controller for T 3 ~

"YE dynamic compensation 1+T 3

= Time constant utilized in the rate lag controller for T 73 "'8

= 10 secs.

. 13 S.= Laplace transform operator (sec~ )

  • f 2(AI) = 0'for all AI Note 3: The channel's maximum trip point.shall not exceed its computed trip point by more than 2 percent span.

ww w y r y 9 .- ~ r ---r -- --- + aig.

7 ,

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2.1 'I SAFETY' LIMITS BASES j 2 .1.' 1 REACTOR CORE h '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 . coef ficient is large and the' cladding surface atemperature.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' . building (DNB) and the resultant sharp reduction .in heat J transfer coefficient.. - DNB -is ' not ' a directly measurable parameter during l-. operation- and .- therefore THERMAL POWER and Reactor Coolant Temperature and j ' 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 i# 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 particular core location to the local heat flux, is indicative of the margin:to DNB.:

ThelDNBl 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  ;

p 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 establishe'd based.on the entire-applicable experimental data set such l

i that1 there is a 95 percent probability with 95 percent confidence that DNB l will not occur when the. minimum DNBR is at the DNBR limit.

]

7 In meeting this design basis, uncertainties in plant operating

j. parameters, nuclear and thermal parameters, and fuel fabrication parameters i are considered statist.ically such that there is at' least a 95% probability that the minimum DNBR for the limiting rod is greater than' or equal to the t .DNBR limit. The uncertainties in the above plaat parameters are used to

! determine the plant DNBR uncertainty. This. DNBR ' uncertainty, combined with the correlation DNBR limit, establishes a design DNBR value which must be met

~

I i in plant ' safety'- analyses ~ using values of input parameters without uncertainties.

4 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 l .which:the minimum DNBR is no less than~the design limit DNBR, or the average l

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

i p

L l

NORTH ANNA . UNIT.2 B.2-1 e- , r .- b w - e te+--t%s- ., e nw we eve -,-s-m m e#m 5, - w- v3..wev --*-+ve.e,m-. h - y. *,-me --evy.7-,w v, e---

)

SAFETY LIMITS BASES The curves are based on an enthalpy hot channel factor, F'A . and a l reference cosine with a peak f 1.55 for cxial power shape. H,An allowance is included for an increase in at reduced power based on the expression:

AH H

~ * * -}

where P is the fraction of RATED THERMAL POWER These limiting heat flux conditions are higher than those calculated for the 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(AI) function of the Overtemperature trip. When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Overtemperature AT trips will reduce the setpoints to provide protection consistent with core safety limits.

2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant ' System from overpressurization and thereby prevents the 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 B 31.1 1967 Edi. tion and ANSI B 31.7 1969 Edition (Table 5.2.1-1 of FSAR) which permits a maximum transient pressure of 120% (2985 psig) of component 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.

e NORTH ANNA - UNIT 2 B 2-2

. b

. ~ . - - . -. - . - . -- - - - . . - - .. . _ _ _

m I

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 t

are prevented _ from exceeding their safety limits. Operation with a trip set lless 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 drif t 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 Range, Neutron Flux b . .

The Power Range, Neutron Flux channel high setpoint provides reactor core i

protection against . reactivity excursions which are too rapid to be protected 2

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 j level.- Specifically, this trip complements the Power Range Neutron Flux High l and Low trips to ensure that the criteria are i met for rod ejection from partial power.

The : Power Range . Negative ' Rate Trip provides protection for control rod j ' drop accidents.. At 'high' power, a rod drop accident could cause local flux peaking which could' cause an unconservative n local DNBR to exist. ~The Power

. Range Negative Rate l' Trip will prevent c 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 L for .which s the . DNBR's will be

' greater than the applicable design limit DNBR value for each-fuel type.

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LIMITING SAFETY SYSTEM SETTINGS 1 BASES the minimum value of the DNBR from going below This latter trip.will prevent the design , limit during normal operational transients and anticipated transients when ~ 2. loops are in operation and the Overtemperature AT trip l

setpoint is adjusted to the_value specified for all loops in operation. With thes0vertemperature 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 normal operational transients with 2 loops in operation.

l 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 l 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 signal trips the turbine which in turn trips the reactor if above the P-7 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 to ensure the functional capability of the specified trip settings and thereby enhance 'the overall reliability of the Reactor Protection-System. This trip is redundant to the Steam Generator Water Level Low-Low trip. The - Steam /Feedwater Flow Mismatch portion of this trip is actigatedwhenthesteamflowexceedsthefeedwaterflowbygreaterthan1.616 x 10 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 f auxiliary , feedwater pumps are reduced and . the resulting o thermal ; transient on the Reactor Coolant - System and steam- generators is

minimized.L 5 .

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f ' NORTHIANNAI-UNIT 2 B 2 .

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- POWER DISTRIBUTION LIMITS

- HEAT FLUX HOT CHANNEL FACTOR-F (Z)

. LIMITING CONDITION FOR OPERATION

3.2.2 Fy(Z)shallbelimitedbythefollowingrelationships

Fq (Z) s [ 2.15] [ K(Z)] for P > 0.5 >

P Fq (Z) s [4.30] [K(Z)} for P s 0.5 where P = THERMAL POWER RATED THERMAL POWER and K(Z) is the function obtained from Figure 3.2-2 for a given core height location.

. APPLICABILITY
MODE 1.

4 ACTION:

With F (Z). exceeding its limit:

q 4

a.- Comply with.either of'the following ACTIONS:

1. Reduce THERMAL POWER at lear.t 1% for each 1%qF (Z) exceeds the

. limit within 15 minutes and similarly reduce the Power Range 2 Neutron Flux-High Trip Setpoints within the'next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; POWER

( OPERATION may proceed for up to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; subsequent POWER OPERATION may proceed provided the Overpower AT Trip Setpoints have been reduced at least 1% for each 1% Fq(Z)

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, above; THERMAL POWER may.then be increased provided nF (Z) is demonstrated:through incore mapping ~to be within its Ilmit.

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CORE HEIGHT (FT) 1 l

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l Figure 3.2-2 NORMALIZED F g (z) AS A FUNCTION OF CORE HEIGHT t

NORTH ANNA . UNIT 2 .3/4 2-8

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POWER DISTRIBUTION LIMITS NUCLEAR ENTHALPY HOT CHANNEL FACTOR - F H LIMITING CONDITION FOR OPERATION t

- 3.2.3 F H shall be limited by the following relationship:

~ '

, H l THERMAL POWER where: P= RATED THERMAL POWER

!r", measured va1ue of 7 obtained by using the movaMe incore detectors

, 3 toobtainapowerdk!cributionmap.

i 4 APPLICABILITY: MODE 1.

ACTION:

With(H exceeding its limit:

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

-b.Demonstratethroughin-coremappingthat(H is within its limit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding the limit or reduce THERMAL POWER to less than 5% of RATED THERMAL. POWER within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and

c. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above the reduced limit required by a og b, above; subsequent POWER OPERATION may proceed provided that F is dem natrated through in-cor,e mapping to be within its limit H

4 at a nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THERMAL POWER prior to f

1 6

NORTH ANNA - UNIT ~2 .3/4 2-9 t

t

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

or greater RATED THERMAL POWER.

-SURVEILLANCE REQUIREMENTS 4.2.3.1 'F . shall be determined to be within its limit by using the movable incore-detektorstoobtainapowerdistributionmap'

a. Prior -to operation above 75% of RATED THERMAL POWER af ter 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.

0 NORTH ANNA - UNIT 2 3/4-2-10 I

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NORTH ANNA - UNIT 2 3/4 2-11

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5 G TABLE 3.2-1

.g DNB PARAMETERS

. LIMITS e

.5-- '2 Loops in Operation ** 2 Loops in Operation **

N 3 Loops in &. Loop Stop _

& Isolated Loop Operation Valves Open. Stop Valves Closed PARAMETER

Reactor Coolant System Total Flow Rate 2289,200 gpm R

s~

5:

  • 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 r

l 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 Moderate Frequency) l 4

events by: (a) maintaining the minimum DNBR in the core from going beyond 1 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 f 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 accep'tance 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

[ Fq(Z) Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation Z i.

divided by the average fuel rod heat flux, allowing for man-2 ufacturing tolerances on fuel pellets and rods.

H Nuclear Enthalpy Rise. Hot Channel Factor, is defined as the ratio of'the integral of linear power along the rod with the highest integrated power to the average rod power.

F (Z) Radial Peaking Factor, is defined as the ratio of peak power density to average power density in the horizontal plane at

~

I core elevation Z.

I 3/4.2.1 AXIAL FLUX DIFFERENCE (AFD) ,

The limits on AXIAL FLUX DIFFERENCE assure that the'F (Z) upper bound Q p 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.

f . 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 p9aition 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 l

l .

NORTH ANNA - UNIT 2 .B.3/4'2-1 l-m

POWER" DISTRIBUTION LIMITS BASES The specified limit for F tains a 4% error allowance. Normal operationwillresultinameasuNd H less than or equal to 1.49. The 4%

allowance.is based on the following considerations:

a.:abnormalperturbationsintgeradialpowershape,suchasfrom rod misalignment, effect F AH m re directly than F q, n' b. although rod movement has a direct influence upon limiting Fq to with n its limit, such control is not readily available to limit AH ' ""

c. errors in prediction for control power. shape detected during

. startup physics tests can be compensated for ingFyby restrict-ing axial flux distribtions. This compensation fdr rAH is less readily available.

JFuel 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 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 neasurements 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 operaticn 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 tilt,.the margin for uncertainty-on Fn is. reinstated by 7

reducing the power by 3 percent for each percent of tilt in'fexcess of 1.0.

l For . purposes ' of monitoring QUADRANT POWER TILT RATIO when- one excore

! detector.is inoperable, the moveable incore detectors are used to confirm that

! . the normalized symmetric power distribution is consistent with : the 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, j .

i E-5, ; E-11,- H-3, H-13, - L-5, L--11, and N-8. .

NORTH ANNA - UNIT 2- :B 3/4'2-5 e + w y w - -

POWER DISTRIBUTION LIMITS 1

!~

BASES 1

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.-

4 The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 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 verification of flow rate on a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> basis.

3/4.2.6 AXIAL' POWER DISTRIBUTION The limit.on axial power. distribution ensures that F will be controlled q

and monitored on a more exact basis through use of the APDMS when operating above P,% of RATED THERMAL POWER. This additional limitation on Fq is-4 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,.wheIe K(Z) is the graph shown in Figure 3.2-2. The amount ofpokential-violationisdeterminedbysubtracting1fromthemaximumratio

of the predicted Fn (Z) analysis (flyspeck) results for a particular fuel cycle xK(Z)' limit._ This amount of potential violation, in percent,~is to' the I'hd from 100% to determine the value for subtract If P . equal
to 100%,

P -is no axial power distribution surveillance is re$uired.

  • P w ill - not exceed

- 100%.

4

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NORTH ANNA - UNIT 21 B 3/4'2-6'

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- 3/4.7 PLANT SYSTEMS

~ BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAPETY VALVES i

The OPERABILITY of the main steam.line code safety valves ensure that the secondary system pre'ssure.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%

l ' RATED THERMAL POWER coincident with an assumed loss of condenser heat sink

1. (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, 1_971 Edition. The total . relieging capacity for all safety valves on all of the steam lines is 12.83 x Ig lbs/hr which is-greater than l the total secondary steam flow- of 12.77 x 10 lbs/hr at 100% RATED ~ THERMAL l 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.

l- 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 channels. The reactor-trip setpoint reductions are derived on.the following bases:

For 3. loop operation SP = (X) (Y)(V) x 109 l X:

. For 2 loop opers:: ion with l stop valves closed

'SP = (X) (Y)'(U) x 71

! X

'A

(

I j

4; NORTH' ANNA - UNIT 2: B'3/4 7-1 L .. ,

.y a f

, ~ - , - . +, ..a , -  :, - -- -

ENCLOSURE 4 .

I NORTH ANNA UNITS 1 AND 2 1 2905 MWt NSSS RATING RESPONSE TO NRC COMENTS RAISED AT THE OCTOBER 11, 1984 MEETING 1

1 VIRGINIA POWER JANUARY 1985 f

l 45-WBR-2010N-6 l

l Page 1 of 3 3-18-85 l RESPONSE TO NRC STAFF COMMENTS  !

RAISED AT THE OCTOBER 11, 1984 MEETING l CORE UPRATING TO 2905 MWt NORTH ANNA POWER STATION UNITS 1 AND 2 I. Comment: Will Virginia Power update the Environmental Report as part of the request for an amendment to the Operating Licenses?

Response: -The Environmental Report for North Anna Power Station (Appendix K) and the NRC's Environmental Impact Statement (Reference 1) have already addressed operation up to the stretch rating of,2910 MWt. Therefore, Virginia Power does not deem it necessary to revise the Environmental Report. It is recognized, however, that the temperatures in Lake Anna have exceeded predictions originally made-in the Environmental Report and that the lake, at times, has exceeded State thermal limits. Virginia Power is currently conducting a 316(a) demonstration to resolve whether or not these temperatures adversely affect aquatic life. This study is being conducted in accordance to Section 316(a) of the Clean Water Act. The proposed core uprating is projected to increase the rejected heat by approximately 4.5%. The Virginia State Water Control Board and their Technical Advisory Committee are aware of the proposed core uprating and will consider it as they consider the results of the 316(a) demonstration.

l II. Consnent: Will the licensing submittal address the NRC's generic concerns regarding consistency between the Station Technical Specifications and accident' analyses?

i Response: Virginia Power has reviewed NRC concerns regarding Technical l

Specifications for: 1) Mode 3, 4 and 5 operation; 2) safety related equipment; 3) comparison of indicated parameter values with limits and 4) use of Non-Docketed documents to l

establish assumed initial conditions for analyses. These concerns have been reviewed for potential impact upon.the-proposed Technical Specification changes for uprated operation. It has been concluded that these concerns do not l require any revision in the proposed changes.for uprated l

operation. However, an additional submittal may be required to address the outstanding item discussed below.- The-Westinghouse Owners' Group (WOG) is considering the appropriate generic approach which should be taken to address these concerns for existing plant Technical Sp6cifications.

Virginia Power has therefore only addressed the generic issues as they potentially impact this licensing submittal, so as not to propose any resolution of these items which may be inconsistent with that determined through the WOG. efforts.

l t '45-WBR-2011N-l'

l 1'

Page 2 of 3 3-18-85 l l

The uprating analysis of the uncontrolled rod bank withdrawal from a subcritical condition was performed assuming two reactor coolant pumps are in operation. Virginia Power has i . separately prepared a submittal-(Reference 4) to support

,- operation with a positive moderator temperature coefficient.

The'subcritical rod withdrawal results.in this separate

. submittal bound operation at current plant conditions with a i

single coolant pump operating. The final Westinghouse.

analysis methodology for this case was not available for the

, uprating submittal. Furthermore, Westinghouse informed the NRC via _ Reference 2 that they do not plan to pursue- -

. development of a generic methodology which assumes only one i reactor _ coolant pump in operation.' A future submittal is

- proposed to provide results of a Virginia Power analysis of ,

. this event at uprated conditions.

III. Comment: The NRC has stated that the core uprating request is likely to be treated as involving a significant hazards consideration.

Response: .The evaluations performed to support uprated operation have addressed all items which were identified to be impacted by the proposed: power increase. ;In addition, the proposed power level was chosen to remain within existing margins for plant

analyses and equipment operation. The results of the
evaluations performed confinn that all accidents, systems and
equipment remain within the appropriate acceptance criteria.

On this basis, the small. increase in~ power is not considered' to represent a significant hazard.

~

IV. Coment: Relevant to the core uprating, how is Virginia Power planning.

! to address the NRC staff concern on the environmental L qualification of equipment in areas that.are potentially exposed to superheated steam? - Note that if the steam

, generator tubes are uncovered during-a steam line break,

superheated steam.could be released and this may not have

! been considered in past analyses.

Response: In. response to.NRC questions during review of Reference 3, Westinghouse has addressed the effects of steam generator

tube uncovery on containment temperature .for' steam line -

breaks inside containment. It was_ concluded that existing

. analyses remain bounding for steam -line breaks inside dry

~

. containments. A Virginia Power review of confined areas Loutside of containment was conducted to determine those areas potentially subject to steam releases from a steam line

break. It was concluded that the Main Steam Valve House is t the only.such area containing equipment requiring i - environmental qualification. The existing MSLB calculation for this zone _does not specifically address the existence of

- superheated steam,- but the Virginia Power-evaluation ~

p I

t. .

L  : 45-WBR-2011N-2--

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Page 3 of 3

~

3-18-85 conclude'd that the existing temperature envelope bounds the expected results in the presence of superheated steam. These results are being incorporated into the Environmental Zone Description updates currently being performed by Virginia Powar. Note that, as stated in the 80P Safety Evaluation, the core uprate will not impact steam _line break environments inside or outside the containment since those breaks are based on the limiting no-load power conditions which remain unaffected by the uprate.

V. Coment: The licensing submittal should contain enough information so i

. the NRC staff can determine that the Station Design Basis

criteria will continue to be met for the core uprating.

Response: The NSSS and B0P Licensing summaries document the evaluations

, performed in sufficient detail to demonstrate that the -

Station Design Basis Criteria will not be violated as a result of the uprating. Note, however, that the use of the Westinghouse Improved Thermal Design Procedure (ITDP) for DNB

-analysis represents a change in the design basis _for DNB.  ;

The reanalysis of DNB-limited events has confirmed that the revised design basis.is met. Appendix R evaluations are i being performed for North Anna Power Station Units 1 and 2.

i These evaluations are not addressed as part of this core  !

uprating submittal; however, the Appendix R evaluations will support operation up to a NSSS Rating of 2905 MWt. The Appendix R Report (Volume I), submitted to the NRC in

Reference 5, lists the Core Uprating as one of the j assumptions for the Appe.ndix R Analyses.

4

.VI.

References:

1. Final-Environmental Impact Statement, North Anna Power Station, U.S. Atomic Energy Commission, Directorate of Licensing, April 1973
2. Letter from E. P. Rahe (Westinghouse) to H. Thompson (NRC),NS-NRC-85-2997, dated January 22, 1985,

Subject:

Number of Operating Reactor Coolant Pumps in Mode 3

3. WCAP-8822, " Mass and Energy Releases Following Steam Line Rupture"

- 4. Letter from W. L. Stewart (Virginia Power) to H. R.-

Denton (NRC), dated February 7,- 1985 (Serial No. 666),

Subject:

Positive Moderator Temperature Coefficient,-

i Amendment to Operating Licenses NPF-4.and NPF-7, North Anna Power Station, Units 1 and 2 i

5. Letter from W. L. Stewart (Virginia Power) to H. R. '

! Denton (NRC), dated March 8, 1985 (Serial No.85-114),

Subject:

-10 CFR 50 Appendix R Reanalysis, North Anna L Power Station Unit Nos. I and 2 l

WBR-2011N-3

_ - __ ._.~._

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