ML20078D691

Proposed Tech Specs Changing Fractional Thermal Power Multiplier from 0.2 to 0.3 for 587.8 F Operation

ML20078D691
Person / Time
Site:	North Anna
Issue date:	09/29/1983
From:	VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML20078D685	List: ML20078D678 ML20078D691
References
NUDOCS 8310040590
Download: ML20078D691 (12)

Text

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North Anna Unit No.1 Technical Specifications Changes f

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NORTH ANNA - UNIT 1 2-2 1

- ~ . ,. .. ~ .

5

= '

$ TABLE 2.2-1 (Continued)

'f REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

NOTATION'(Continued)

E .

Q Operation wfth 3 Loops Operation with 2 Loops Operation with 2 Loops (no loops isolated)* (1 loop isolated)*

Kg = 1.085 Ky =

( ) Ky =

( .

)

K = 0.0150 K "

( ) K 2

( )

2 2 K = 0.000670 K "

I ) 3

( )

3 3 i

and f g ( AI) 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:

(i) for q between - 32 percent and + 9 percent, f (AI) =0 l t

(where q

- 9b and qb arepercentRATEDTHERMALPOWERintketopandbottom Y halvesoIthecorerespectively,andq + q is b

total THERMAL POWER in

• percent of RATED THERMAL POWER).

! (ii) for each percent that the magnitude of (q - qb) ex eeds - 32 percent, the AT trip setpoint shall be automaticalIy reduced by 1.92 percent of its value at RATED THERMAL POWER.

(iii) for each percent that the magnitude of (q - q ) exceeds + 9 percent, b

l theATtripsetpointshallbeautomaticalIyreducedby1.77percentof its value at RATED THERMAL POWER.

We Y

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

i

5 5

TABLE 2.2-1 (Continued)

'E REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS e

g ' NOTATION (Continued) w 8 i n f .T3 NOTE 2: Overpower ATsAT .[K4-K  !

6 (T-T )-f2(AI)]

5 'YiT3 8 /

Where: AT- = Indicated AT at RATED THERMAL POWER o

T = Average temperature, *F T" = Indicated T at RATED THERMAL POWER 5587.8'F avg K = 1.091 4

K = 0.02/*F for increasing average temperature "1 5 O

K = 0 for decreasing average temperatures 5

K " *

~ # '"

6 6 l l T S = The function generated by the rate lag controller for T ""8 dynamic compensation 1+T 8 3

T = Time constant utilized in the rate lag controller for T 3

T = 10 secs.

3 S = Laplace transform operator (sec~ )

l f2 (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.

i

SAFETY LIMITS BASES The curves are based on an enthalpy hot channel factor, F H' ti l.55 and a reference cosine with a peak of 1.55 for axial power shape. An allowance is included for an increase in F at reduced power based on the expression:

H F

H = 1.55 [1+0.3 (1-P)] l 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(delta I) 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 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) 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.

I NORTH ANNA - UNIT 1 B 2-2

I'~

POWER DISTRIBUTION LIMITS NUCLEAR ENTHALPY HOT CHANNEL FACTOR -

H LIMITING CONDITION FOR OPERATION

?

3.2.3 F H

shall be limited by the following relationship:

F * + * (- ~

(

H THERMAL POWER , and where: P= RATED THERMAL POWER RBP (BU) = Rod Bow Penalty as a function of region average burnup as shown in Figure 3.2-3, where a region is defined as those assemblies with he same loading date (reloads) or enrichment (first cores).

APPLICABILITY: MODE 1.

ACTION:

With F H ex eeding 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. Demonstrate through in-core mapping that F H is *.dthin 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 demonstrated through in-core mapping to be within its limit AH 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 1.

I North Anna Unit No. 2 Technical Specifications Changes Tavg = 587.8 F i

s e

E 4

h

+ .

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, NORTH ANMA - UNIT 2 2-2 f

m h

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'Ew .

$ TABLE 2.2-1 (Continued)_

f' REACTOR TRIP' SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION (Continued)

E s..

Ci Operation with 3 L' oops . Operation with 2 Loops Operation with 2 Loops n T (no loops isolated)* (1 loop isolated)*

i

, ~E,ai 1.085 s Kg =-( ) Kg =

( )

K 2

= 0.0150 K 2

=

'( 3

) K 2'" ( )

K = 0.000670 K 3

( ) '

3

( )

3 i and f g (61) is a function of the indicated difference between top and bottom detectors i of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that: .,

(i) for q 9 between - 32 percent.and +- 9 percent, f (51) =0

) t b l (where q and qb arePercentRATED~THERMALF0WERintketopandbottom j 7 halvesofthecorerespectively,andq t + 9b is total THERMAL POWER in  ;

percent of RATED THERMAL POWER).

l (ii) for each percent that the magnitude of (q -

qb) exceeds - 32 percent, theATtripsetpointshallbeautomatica1Iyreducedby1.92percentof 4 its value at RATED THERMAL POWER.

1 (iii) for each percent that the magnitude of (q -

qb) exceeds + 9 percent, 4

theATtripsetpointshallbeautomatica1Iyreducedby1.77percentof its value et RATED THERMAL POWER.

a we T 4

+

~

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

G

5-m '

'* TABLE 2.2-1 (Continued)

N REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS i i g NOTATION (Continued) y .

~

T8 r3 1  ::

NOTE 2: Overpower ATsAT .[K4-K5 'l

1. 6 (T-T )-f2(AI)] .

YU3 S }

Where: AT = Indicated AT at RATED THERMAL-POWER o

T = Average temperature, *F i

' T" = . Indicated Tavg at RATED THERMAL POWER s587.8'F K

4

= 1.0M w = 0.02/*F for increasing average temperature K

1 o

5 K = 0 for decreasing average temperatures 5

K ~ * "I " # ' "

6 6 TS = The function generated by-the rate lag controller for T ,

j 3

dynamic compensation <

1+T S 3

T 3

= Time constant utilized in the rate lag controller for T,y t = 10 secs.

3 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.

a

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

AH (H = 1.55 [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(delta I) 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 conteined 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 Coclant System piping, 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 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.

NORTH ANNA - UNIT.2 B 2-2

.m __

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

F 3.2.3 F shall be limited by the following relationship:

H H

sl .55 [1 + 0.3 (1-P)] [1-RBP ('BU)]

THERMAL POWER , and where: P= RATED THERMAL POWER RBP (BU) = Rod Bow Penalty as a function of region average burnup as shown in Figure 3.2-3, where a region is defined as those assemblies with he same loading date (reloads) or enrichment (first cores).

APPLICABILITY: MODE 1.

ACTION: >

With' N ex ee ing i s limit:

AH

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 t:7 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. Demonstrate through in-core mapping that 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 demonstrated through in-core mapping to be within its limit AH at a nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THERMAL POWER prior to NORTH ANNA - UNIT 2 3/4 2-9