ML20058E340

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Proposed Tech Specs Reflecting Elimination of RTD Bypass Sys
ML20058E340
Person / Time
Site: Farley 
Issue date: 10/26/1990
From:
ALABAMA POWER CO.
To:
Shared Package
ML19310C816 List:
References
NUDOCS 9011070137
Download: ML20058E340 (13)
v • d • e


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A1TACHMENT 1

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9011070137 901026 DR ADOCK 05000348 p

PDC

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' 670 Unacceptable Operation F-660 2400 psia 650 2250 psia.

640

-630 2000 psia 1875 psia C '620 u

.p

.m-

. r 610 Acceptable Operation

.600 590

.H*

580 t

570

+

T 560 N

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.2

.3

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.9 1.0

' 1.1 1.2 POWER (fraction of nominal).

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rigure 2.1-1 Reactor Core Safety Limit Three Loops in Operation r

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' " FARLEYi-UNIT 1 2-2 AMDiDfENT NO,

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TABLE 2.2-1;

$h REACTOR TRIP SYSTEM ' INSTRUMENTATION TRIP SETPOINIS

-; 5

~ FUNCTIONAL UNIT TRIP SETPOINT ALLOVABLE VALUES c:

i'i 1.

Manual. Reactor Trip

-Not Applicable.

Not Applicable

? C=)

~

2. - Power Rarge, Neutron Flux Low Setpoint -;f 25% of. RATED Low Setpoint - f 26% of RATED

' THERMAL POVER-THERMAL POWER-High Setpoint -'$ 109% of RATED High Setpoint - $ 110%.of RATED THERMAL POWER THERMAL POWER.

3.

Power Range, Neutron Flux, 5 5%.of RATED THERMAL POWER vith

. f 5.5% of RATED THERMAL POWER High Positive Rate a time constant 1 2 second with a time constant 1 2 second 4.

Power Range, Neutron Flux,

< 5% of RATED THERMAL POVER vith

< 5.5% of RATED THERMAL POVER

.High Negative Rate

'a time constant 1 2 second with a time constant 2 2 second 5.

Intermediate Range, Neutron f 25% of RATED THERMAL POWER

$ 30% of RATED THERMAL POVER Flux 6.

Source Range, Neutron Flux

.f 10' counts per second f 1.3 X 10' counts per second 7.

Overtemperature aT See Note 1 See Note 3 8.

Overpower AT See Note 2 See Note 6 l

9.

Pressurizer Pressure-Lov 1 1865 psig 1 1855 psig 10.

Pressurizer' Pressure-High

.f 2385 psig i 2395 psig 11.

Pressurizer Vater f 92% of instrument' span f 93% of instrument span Level--High 12.

Loss of Flov-

- 2 90% of design flow per loop

  • 1 88.5% of design flow per loop
  • l g
  • Design flow is 87,200 gpm per loop.

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. TABLE 2.2-1 (Continued)-

5;.

.-. REACTOR TRIP ~STSTEM INSTRUMENTATION TRIP SETPOIPrrs m~

-NOTATION 8

-Note'1:

Overtemperature af

~

.q AT (1l+ JT,s) i AT, [K - Kj (1-.'+T s) _

-(T (;

1 -.- )

.'T') + K.(P

.P') - f (AI)]

1 1

3 t

~

(1 + T s)'

(l '+T s).

- I +T,s 3

2 where:

AT = Measured aT by RTD' instrumentation; Et, = Indicated AT at' RATED THERMAL POVER;

'T

=~ Average temperature, "F; T' $ 577.~2'F (Maximum Reference T,,, at RATED THERMAL POVER);

Pressurizer-pressure, psig; P

=

P' = 2235 psig-(Nominal RCS-operating pressure);

Y 1+T8

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

-1 1+T8 2 T &T

= Time constants utilized in theilead-lag controller for T,,, T = 30 secs, T 4 secs;

=

1 2

t 2

I^Ts 4

= The function generated by the lead-lag controller for AT dynamic compensation; 1+Ts3 T, &T

= Time constants. utilized in the lead-lag controller for ar, T,

=T

= 0 seconds; 3

3

-1

= Lag compensator on measured T,,,;

l ! +T,s T,-

= Time constant. utilized in the' measured T,,, lag compensator, T,

= 0 see; Laplace transform operator, sec *;-

s =

Operation with 3 loops Operation with 2 loops E

(values blank pending l

E K

1.18; K

=

=

1 1

m K.=.0.0154;-

K NRC approval of

=

2 3

5 2-loop operation) 0.000635; K

K

=

=

3 3

. -..=_- - -.-

-[

'. TABLE '2.2-1 (Continue'd).

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_ 4

_ REACTOR' TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS "4

NOTATION (Continued)

U

'and:fg (AI) is a functioniofJthelindicated difference.'between top and bottom detectors of the power-range nuclear l ion chambers; with gainsLto.be selected based on measured instrument response during plant startup

~

~

tests such that:

~

^

(i)' lfor q - q :between -35 percent.and.+9 percent, f (AI) - O (where q and q are percent RATED THERM 1LPODERin-the.topandbottomhalvesofthecorerespectively,tand q,,+ g g

is total THERMAL -

~

g POWER-in percent of RATED THERMAL POWER);-

(ii) for each percent that the" magnitude of-(q, - q,) exceeds -35 percent, the aT trip setpoint shall be. automatically reduced by.1.37 percent of its value'at RATED. THERMAL POWER; and (iii) for each percent'that the magnitude of (q, - q,) exceeds +9 percent, the aT trip setpoint shall be automatically reduced by 1.75 percent.ofLits value at RATED THERMAL POWER.

l Note 2: Overpower aT AT (1 + T,s) f aT, [K,- K'3 j - T") - f,(AI)]

(

3

) (

)T-K, (T (

1 Ts 1

(1 + T s) 1 +T s 1 + T,s 1 + T,s 3

3 where:

aT = Measured AT by RTD instrumentation; AT,= Indicated ar at RATED THERMAL POWER; T~ = Average temperature, 'F; T" = Reference T,,,.at RATED THERMAL'POVER (Calibration temperature for AT instrumentation, I

f 577.2*F);

l 1.08; K

=

4 l

l K. = 0.02/*F for increasing average temperature and 0 for decreasing average temperature;

~

3 l

E K, =.0.00109/'F for T.> T" K, = 0 for T $ T";

l E

U Ts _ = The function generated by the rate lag controller for-T,,, dynamic compensation; i

3 i

g 1+T s 3

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TABLE 2.2-1 (Continued):

1 J

' REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPODrrs NOTATION (Continued)

H-

+4 w

10 secs; T ~= Time constant utilized.in the rate. lag controller for T,,,

T

=

3 3

I+Ts = The-function generated by the' lead-lag controller for'aT dynamic compensation; 4

1+Ts 3

= T = 0 seconds; T, & ' T = Time constants utilized in the lead-lag controller for aT, T,

3 3

1'

= Lag. compensator on measured T,,,;

1 + T, s -

T, = Time constant utilized in the measured T,,,

lag compensator, T, = 0 sec;

~*

s-=

Laplace transform operator,'sec f2(aI) = 0 for all aI.

Note 3: The channel's maximum ^td p point shall not exceed its computed trip point by more than 3.4 percent.

l Note 4: Pressure value to be determined during-initial startup testing. Pressure value of $ 55 psia to be used prior to" determination of revised value.

Note 5: Pressure value to be determined during initial startup testing.

Note 6: The channel's maximum ~ trip point shall not exceed its computed trip point by more than 2.9 percent.

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

- -... - - - - = - - - - -. - -

- - - - ~

M l

l LIMITING SAFETY SYSTEM SETTINGS I

f BASES

,mmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmm.

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Intermediate and Source Range, Nuclear Flux l

g The Intermediate and Source Range, Nuclear Flux trips provide reactor core protection during reactor startup.

These trips provide redundant protection to the lov setpoint trip of the Power Range, Neutron Flux chapnels.<.The Source Range Channels vill initiate a reactor trip at about 10*

counts per second unless manually blocked when P-6 becomes active.

The Intermediate' Range Channels vill initiate a reactor trip at a current level proportional to approximately 25 percent of RATED THERMAL POVER unless manually blocked when P-10 becomes active.

No credit was taken for operation of-the trips associated with either the Intermediate tr Source Range. Channels in the accident analyses however, their functional: capability at the

'specified trip settings is required by this specific 4 tion to enhance the 1

overall; reliability of the Reactor Protection Syster..

I Overtemperature oT The Overtemperature delta T_ trip provides core protection to prevent-DNB-m for all: combinations of-pressure, power, coolant temperature,.and axial power distribution,-provided that the transient is slov vith respect to piping

' transit, thermovell, and RTD response time delays from the core to the

'l

~

' temperature detectors (about 4 seconds), and pressure is withinLthe range between;the High.and Lov Pressure reactor trips. This setpoint includes

corrections for changes in density and heat capacity. of
vater with temperature and dynamic compensation for. transport, thermovell,'and RTD l

response time delays-(tom the core to.RTD output indication. ;Vith. 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,Jas! indicated by.the. difference between-top and bottom power range nuclear ~ detectors,-the reactor trip is automatically reduced according-to

.the1 notations in. Table 2.2-1.

'y Operation with a reactor coolant loop out of service'below the 3 loop iP-8'setpoint does not require reactor 1 protection system setpoint (modification because the P-8tsetpoint and associated' trip vill prevent'DNit=

~

during 2 loop operation exclusive of the Overtemperature delta T'setpoint.

sTwo? loop. operation above-the 3 loop'P-8 setpoint is permissible after Jresetting the K1, 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 u!{

lat the reduced power level.

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FARLEY - UNIT 1 B 2-4 AMENDMENT NO.'

M s4

F LIMITING SAFETY SYTEM SETTINGS 8^SES......................................................................

Overpover 6T

-The Overpower delta T reactor trip provides assurance of fuel integrity (e.g., no fuel pellet melting) under all'possible overpover conditions, l

limits,the required range for Overtemperature delta T protection, and provides a backup to the High Neutron Flux trip.

The setpoint includes corrections for axial power distributien, changes in density and heat capacity of water vith temperature,.and dynamic compensation for-transport, thermovell, and RTD response time delays from the core to RTD output indication. 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 Lov Pressure trips are provided to limit the pressure range in which reactor operation is permitted. The High Pressure trip is backed up by'the pressurizer code safety. valves for RCS overpressure protection,.and is therefore' set lover than the set pressure for these valves (2485 psig). The Lov

Pressure'. trip provides protection by tripping the reactor in the event of a loss-of reactor coolant pressure.,

Pressurizer Vater Level

'The Pressurizer High Vater Level trip ensures protection against Reactor Coolant System overpressurization by. limiting the water level to a volume Esufficient 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 specifieu trip

setting;is required:bysthis specification'to enhance the overall reliability.of thel Reactor Protection System.

a a

Loss of Flov-

-The! Loss of Flov ' trips provide core protection to prevent DNB in the event of a' loss:of one or more reactor coolant pumps.

i I

-Above 10 percent.of RATED THERMAL POWER, an' automatic reactor trip vill occur if the-flov in any two loops. drop belov.90% of nominal. full loop flov.

a,

,Above 36% (P-8)'of RATED THERMAL'POVER, automatic reactor trip vill-occur if 1

the flov in any single loop drops belov 90% of nominal full-loop flow. 'This i

FARLEY - UNIT 1 B 2-5 AMENDMENT N0.

-w 3.-

TABLE 3.2-1 E

.- Q DNB PARAFETERS.

I C.$

. LIMITS

.a PARAMETER ~

. 3' Loops in Operation 2 Loops in Operation

' Reactor Coalant. System T,,,

f 581.5'F-(**)

l Pressurizer Pressure 1 2220 psia *

(**)

Reactor Coolant System 1 267,400 gpm***

(**)

j.

Total Flow Rate U

Y G

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

Values blank pending NRC approval of 2. loop operation.

5

      • Value includes a 2.2% flow uncertainty.

l 5

5 5

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,, a e-

--m

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Y I TABLE 3.3-2

--f g-

~ REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES FUNCTIONAL UNIT RESPONSE TIME u

Not Applicable:

1.

Manual Reactor Trip 2.

Power Range,~~ Neutron Flux f 0.5 seconds

  • a.

High y

Not Applicable b.

Lov 3.-

Power Range,. Neutron Flux, High Positive Rate Not Applicable 4.

Power Range, Neutron Flux, High Negative Rate f 0.5 seconds

  • 5.

Intermediate' Range, Neutron Flux ^

Not Applicable 6.

Source Range, Neutron Flux Not Applicable 7.

Overtemperature AT f 6.0 seconds

  • l Y

8.

Overpower AT Not Applicable 9.

Pressurizer Pressure-Lov f 2.0 seconds 1

10.

Pressurizer Pressure-High f 2.0 seconds 11.

Pressurizer Vater Level-High -

Not Applicable

  • Neutron detectors are exempt from response time testing. Response time of the neutron flux signal portion-of the channel shall be measured from detector output or input of ' first electronic component in channel.

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7 TABLE 3o3-4 (Continued)-

E Q.

ENGINEERED SAFETT FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOIlfrS f

c:$

- FUNCTIONAL UNIT.

TRIP SETPOIPfr ALLOVABLE VALUES e

4.

STEAM'LINE ISOLATION a.

Manual Not Applicable.

Not Applicable i

b.

Automatic Actuation Not Applicable-Not Applicable-

. Logic c.

Containment Pressure-

.f 16.2 psig.

f 18.2 psig High-High d.

Steam Flov in Two Steam f'A function defined as follovs:

f A function defined as follows::

Lines--High,-Coincident' A op corresponding [to 40% of full A op corresponding to 44% of full-vith T

-Low-Lov steam flow between 0%~and 20% load steam flov between 0%-and'20%. load and then a op increasing linearly and then a op increasing linearly y

to a op corresponding to 110% of to a op corresponding to 111.5% of o

- full steam-flow at full load with full steam flow at full load with y

1,,, 2 543'F T,,, 2 540'F l

n" e.

Steam Line Pressure--Lov 2 585 psig 2 575 psig 5.

TURBINE TRIP AND FEED VATER ISOLATION a.

Steam Generator Water 2 5 75% of narrow range instrument f 76% of narrow range instrument Level---High-High span each steam generator span each steam generator E

tri 5

gml 7 u.(

w,

.n o

n

.nr,_.-,_

--l,.

w g

,. - ~,~,

w

+,'. - -...

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TABLE 3.3 4 (Continued)

ENGINEERED SAFETT FEATURE ACTUATItJ STSTEM INSTRUPNGETATION TRIP SETPOINTS 3

,yW TRIP SETPOINT ALLOUABLE VALUES FUNCTIONAL UNIT-E 6.

AUXILIARY FEEDUATER g

N.d.

3 Automatic Actuation N.A.

a.

'~

Logic b.

H eam Generator Vater 2 17% of narrow range instrument 2 16% of narrow cange instrument

.avel-Low-Lov span each steam generator span each steam generator c.

Undervoltage - RCP 2 2680 volts 2 2640 volts

?

See 1 above (all SI Setpoints) d.

S.I.

N.A.

e.

Trip of Main Feedvater N.A.

Pumps 7.

LOSS OF POWER M

a.

4.16 kv Emergency Bus 2 3255 volts bus voltage

  • 2 3222 volts bus voltage
  • 5 3418 volts bus voltage
  • Undervolta n (Loss of i'

Voltage) b.

4.16 kv Emergency Bus 2 3675 volts bus voltage

  • 2 3638 volts bus voltage
  • c$

5 3749 volts bus voltage

  • Gndervoltage (Degraded Voltage) 8.

ENGINEERED SAFETT FEATURE ACTUATION SYSTEM INTERLOCKS a.

Pressurizer Pressure, f 2000 psig

$ 2010 psig P-11 (Increasinig}, ?-12 Low-Lov T b.

544'F f 547'F l

(Decreasing) 543'F 2 540'F E

Steam Generator Level, (See 5. above)

Q c.

o P-14 E

N. A-.

E d.

Reactor Trip, P 4 N.A.

5 E efer to appropriate relay setting sheet calibration requirements.

R

ATTACRMWff 2 i

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VCAP-12613,=

"RTD Bypass Elimination Licensing Report for J. M. Farley

[

Revision 1 Nuclear Plant Units 1 and 2" (Vestinghouse Proprietary Class 2) l VCAP-12614, "RTD Bypass Elimination Licensing Report for J. M. Farley Revision 1 Nuclear Plant Unita 1 and 2" (Vestinghouse Proprietary l

Class 3) l

)

CAV-90-078

" Application for Vithholding Proprietary Information from Public Disclosure" (Vestinghouse letter with enclosures dated October 15, 1990) i l

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