ML20112G973

Proposed Tech Specs,Providing Improved Operating Margins as Well as Increased Flexibility W/Respect to Core Designs & Plant Operating Strategy

ML20112G973
Person / Time
Site:	Farley
Issue date:	06/12/1996
From:	SOUTHERN NUCLEAR OPERATING CO.
To:
Shared Package
ML20013A548	List: L-96-158, Rev 2 to Jm Farley Nuclear Plant Units 1 & 2 Licensing Rept for TS Changes Associated W/Revised Core Limits,Revised Trip Setpoints & Inclusion of RAOC Control Strategy ML20112G971 ML20112G973 ML20112G976
References
NUDOCS 9606140004
Download: ML20112G973 (100)

Text

.

I l

FNP Unit 1 Technical Specifications Channed Panes Unit 1 Revision 1

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2 1

9606140004 960612 PDR ADOCK 05000348 P

PDR

i i

i 1

.i M1 ADNINISTRATIVE CONTROLS a.

j M

t E

i Review.....................................................

6-10 t

1 Audits....................................................

5-11 Authority..................................................

4-12 s

Records....................................................

5-12 6.5.3 TECHNICAL REVIEW Ale CONTROL Activities.................................................

5-12 Records....................................................

6-13 6.6 REPORTABLE EVENT ACTION......................................

4-14 i

6.7 SAFETY LIMIT VIOLATION.......................................

4-14 6.R PROCEDURES Alm PROGRAMS......................................

6-14 6.9 REPORTINA REQUIREMENTS 6.9.1 ROLITINE REPORTS Startup Report.............................................

6-15a Annual Report..............................................

6-16 Annual Radiological Environmental Operating Report.........

6-17 Annual Radioacti ve Effluent Release Report.................

5-17 l

Monthly Operating Report...................................

5-19

/ 6 Peaking Factor Limit Report.........................

6-19 Annual Diesel Generator Reliability Data Report............

4-19 Annual Reactor Coolant system specific Activity Report.....

6-20 Annual Sealed Source Leakage Report........................

6-20 4.9.2 SPECIAL REPORT 5............................................

4-20 4.10 RECDRD RETNTIM............................................ 4-20 6.11 RADIATIM PRDTECTIM PROGRAN................................6-21a 4.11 MItiq RADIATIM AREA.........................................

4-22 FARLEY-WIT 1 III AfE O WIT NO.57,79,102

b ggpxg e weTH N M PN"#

y 670" 660" UNACCEPTABLE OPERATION 2440 psia 650 "

250 psia 640 i

630 '

(

2000 pela p 820 1875 pela

$10" 1840 pela 600 ' >

$00" ACC ABLE TION

$40" i

6

/.

.1

.2

.3

.4 J

.4

.7

.8

.9 1.

1.1 1.2 POWER (PR/ ACTION OF RATED THERMAL POWER!

Figure 2.1-1 Reactor Core safety Limits Three Loops in Operation FARI.ET - UNIT 1 2-2 AMENDMDff NO. 37, 73.

87, 92

W) M F/S u R E 2.)-l g

680 UNACCEPTABLE 660 2440 psia OPERATION l

2250 psia 640 2000 psia

_U

~

5620 1840 psia h

Sa

~

ACCEPTABLE OPERATION 20 -

560 l

0.0 0.2 0.4 0.6 0.8 1.0 1.2 l

POWER (FRACTION OF RATED THERMAL POWER)

Figure 2.1-1 Reactor Core Safety Limits

[

Three Loops in Operation i

i

TARLE 2.2-1 (Continued) w E

REACTOR TRIP SYSTEN INSTRiRGENTATION TRIP SETrollfTS e-q NOTATION Note 1: Overtemperature M M (1 + T,s) f W, lK - E II *T s)

(T (

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

g a

2 a

(1 + T,s)

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

where:

M - Neasured R by RTD instrumentation; and f* beat 8, N* - Indicated N at RATED THERNAL FOUER; A

T - Average temperature. *F; T'

5??.**" ("r-I z Reference T,,, at RATED TMERNAL POWER 6: 677 2 *E P - Pressuriser pressure, psig; P' - 2235 psig (

inal RCS operating pressure);

Y 1 + T, s

- The function generated by the lead-lag controller for T,,, dynamic compensatlon; I+Ts3 T, &T

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

T, 30 sec, T,

- 4 sec; 3

I * ** *

= The function generated by the lead-lag controller for af dynaale compensation; 1 + T,s O sec b '

T, & T,

- Time constants utilized in the lead-lag controller for af.

T, T,

1

- Lag compensator on measured T,,,;

E I

• T s r~'" 3 4 a

T,

- Time constant utilized in the measured T,,, lag compensator, T,

ec; l

- Laplace transform operator, sec ';

s 8

operation with 3 loops Operat. ion with 2 loops i

/./7 i

K,-

-t-t+;

K, - (values blank pending l

l N.N 0.017 l

y K,=

0. 02',0; K, - NRC approval of l

i D.000825 I

K,-

0.00:215; K,

2 loop ope ation) l

TABLE 2.2

' Continued)

REACTOR TRIP SYSTEM INSidtNIENTATION TRIP SETPOtitFS g

w NOTATION (Continued) 4 e

g and f nuclea,r ton chambers; with gains to be selected based on measured instrument ( AI) is a function y

tests such that response during plant startup

-13

1. 15 (1) fot q -

between - W percent and 944 percent, f ( AI) = 0 (where q and q are percent RATED l

in the top and bottom halves of the c, ore respectively, a,nd q,,+ q, is total THERNAL TB L

POWER in percent of RATED TEERNAL POWER);

-23 (11) for each percent that the magnitude of (q, - q,) exceeds --M percent, the a trip setpoint shall be automatically reduced by-4see-percent of its value at RATED TRERNAL POWER; and 1.40

,95 (iii) for each percent that the magnitude of (q,f its value at RATED TBERNAL POWER

- q,) exceeds,44-percent, the M trip setpoint shall be automatically reduced by 4rif-percent o 2.o 5 Note 2: Overpouer at ar (1 + T,s) f M, (K,- K, (

T, s ) (

1

) T - K, (T (

1

) - 7") - f,(AI))

(1 + T,s) 1 + T, s I + T, s I + r,s I

where:

Er - Measured Sr by RfD instrumentation; and t h re.nSt. % g Er - Indicated af at RATED TEERNAL POWER; A

Average temperature, 'F;

[

T T* - Reference T at RATED THERNAL POWER (. lit;;;i; :- "::;;;;; f:: r e r : ---- -_ :: !--

5 577.2*F);*"

f.10

?

t K.

-heiPs I

K, = 0.02/*F for increasing average temperature and 0 for decreasing average temperature; y

0. 0010s A

K, = 4:49tMW'F for T > T", K, - O for T f T*;

l Ts

- The function generated by the rate lag controller for 7,y dynamic compensation; a

9

• *28 0

1

f, TABLE 2.2-1 (Continusd)

E REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOIlfrS g

NOTATION (Continued)

_U

- Time constant utilized in the rate lag controller for T,,,,

T T

10 sec; 3

3 I + T, s - The function generated by the lead-lag controller for AT dynamic compensation; I + T's Osec.,

6b k sec; T, &T - Time constants utilized in the lead-lag controller for aT, T, T

3 3

I Lag compensator on measured T,,,;

I +

T, s g4 7 lsec; T, - Time constant utilized in the measured T,,, lag compensator, T, A

~*

s - Laplace transform operator, sec f 2( AI) = 0 for all AI.

AT yn Note 3: The channel's maximum trip point shall not exceed its computed trip point by more than percent l

g Note 4:

Pressure value to be determined during initial startup testing. Pressure value of f 55 psia to be used prior to determination of revised value.

Note 5:

Pressure value to be determined during initial startup testing.

o.+

ATspn Note 6: The channel's maximum trip point shall not exceed its computed trip point by more than Asl3 percen.

l

^

5 Aii 8

m

1 Sefaty Limits Bases l

The curves of rigures 2.1-k and 2.1-2 are based on the most

\\

4 limiting result using an enthal 'y hot channel factor, FN H, of 1.10 for VANTAGE $ fuel and an F gg of..

l 2

N peak of 1.55 for axial power shape. for LOPAA fuel and a reference cosine with a AH at reduced power based on the expressionsAn allowance it included for an increase in FN rNgg. 1,70 (1 + o,3 (1. Pl] for VANTAGE $ fuel and 4

f. 3o l

NF ag = h is (1 + 0.3 (1 - P)] for LOPAR fuel l

}

where P is the fraction of RATED THERMAL POWER.

These Itaiting heat the range of all control rods fully withdrawn to the maximum allowable c rod insertion assuming the axial power imbalance is within the limits of the f (delta I) function of the overtemperature trip.

t i

When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Overtemperature delta T trips will reduce the setpoints to provide protection consistent with core safety 1Laits.

2.1.2 REACTOR C00LANT SYSTDI PRESSURE

'I "The restriction of this safety Limit protects the integrity of the Re Coolant System from overpressuriration and thereby prevents the release of actor radionuclides contained in the reactor coolant from reaching the containment j

acnosphere.

i piping and fittings are designed to Section III of the A3NE Code for N s

Power Plant which permits a maximum transient pressure of 1100 design pressure.

(2735 psig) of design criteria and associated code requirements.The Safety LLait of 2735 psig is the tyPfrt design pres ure t demon aei og in t al oper i n.

)

j s

i j

4 i

a rARLEY - UNIT 1 8 2-2 AMENDMENT NO. 37,73,87,92,109 1

LIMITING SAFETT SYTEM SETTINGS HlH.'............................................

Overpower ar (e.g., no fuel pellet melting) under all possible overpowe a

a backup to the Bigh Neutron Flux trip. limits the required range l

The aetpoint includes corrections with temperature, and dynamic coepensation for transport,

_ response ti_me delays from the core to RTD output Ladication..No credit wast _be N aken Er_ ope tion of thi trip in accident yees hi ver, its DELETC %

s%tionalcapa ity at the pecified rip setting a requir by this f

pect(1 cation to ce the o rall rol bility of Reactor rotection

_Systes.N x

Pressuriser Pressure range in which reactor operation is permitted.The Pressuriser Eigh up by the pressuriser code safety valves for BCE everpressure protec therefore set loser than the set pressure for these valves (2445 peig).

Pressure trip provides protection by tripping the remeter la the event of a loss The Low of reactor coolant pressure.

Pressuriser Water invol The Pressuriser Eigh Water 14 vel trip ensures protection assinst Reactor Coolant System overpressurisation by limiting the water level to a volume sufficient to rotata a steam bubble and prevent water relief through the pressuriser safety valves.

9 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 Systaa.

Loss of Flow a loss of one or more reactor coolant pumps.The less of Flow tri Above 10 percent of RATED TEERNAL POWER, an automatic reactor trip will occur if the flow ta any two loops drop belov 90% of aestaal full loop flow.

Above 36X (P-4) of RATED TIERNAL POVER, autoestic reactor trip will occur if the flow in any single loop drops below 90% of moniaal full loop flow.

This e

G FARLBT - WIT 1 5 2-5 AlWERIErr 18. 29. 87 9

e l

i 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE (AFO) r/E/.drs75 5N=-WS w F4M 3.2 ~ /.

LIMITING CONDITION FOR OPERATION l

T

'l

[

3.2.1 The indica,ted AXI,AL FLUX OIFFERENCE (AFD) shall be maintained within.e-m

.._,._s

_L_...

m_

A

..s....,

APPLICABILITY: MODE 1 above 505 of RATED THERMAL POWER"'

ACTION:

Lwurrs s n e m w w M e<53.2-/.*

With the indicated AXIA a.

... _ _.......,__..,L FLUX DIFFERENCE outside of the '.".*.'.t- ;;t e.

j.

g,.. -._,......

,o, -..-.........-.._-,2u.,_

j wow

-eb-Either restore the indicated AFD to within the t..

Mr." r I

)

i s rmW(5j init1 or

{

m,,,

__,~,...._,___u.._._.

._m 40WEA 2.

. ". - ^ _ _ _.... __... -.-

" ^^"

.'. " _ ' '.. ' " "... ' ~... ^. '

l

^

~ " -

j s

.aue......,, _., __......,....,-

7,.-.__

_ _..,2_2 1

1. %

TL J

JA.

A.J

,A.Om.

_ h.

....JJ_

L__

L___

..A_

_A AL _

mm A____.

L__2 as._ _ ____

AL_.

1

____ta.,.. _..

a.s__

L.._

2..J i

...-.1.....

2.. _2 m..,_

.k.

9.

L...__

__a 1

T. L _

A_2J._

_2 a

J.

.j i t,_

iu_

-.......... _ _ _ g s _ j i__ _ _ _.. _..

j

_t

.~..... _ _ _

"'r; 0'

^^2 = '

.IReduce THERMAL POWER to 1

lless than 505 of RATED THERMAL POWER within 30 minutes.

m._.. _...

3 3

e........

.u _

tur - -_> anues me.64.-.,...

__.._1

_,-_i

.m.

.6.

t.....

o

.._ _.i..

4..._, m.

u.n,,,, %

.L____1._

L_

___m_

-a g,.e___ggg,,g(,,

..__....a

__..,2_2,....

_2,...

2 m.

._2- _ _....

u.J.kk.

.k_

1. A J A. -. A.,A..___

..A _1..

,a A.

A_1

_A is L_

____.m,m.

.k m..i.._..2_-_. J. A..a. u.a.w_pr irm..*..J.2

...1 A

_e

• L -

g i

Osi.

(

u__,

2._>_-

-un._

.s L_..

v.,.. v u.

4

-_k..

,Mf ".'** *_A

.k.11

.A k_. A..___._.__J

.J P_

N.

_L-._

A

.ATP.

Tt.lf w At

n. _

u..,

..a rv 1,..

e, m L__

u___

• '"See Special Test Exception 3.10.2 I A)$iAT HM mn Mw U HT PhS E

~

FARLEY-UNIT 1 3/4 2-1 AMENOMENT NO. 26

POWER DISTRIBUTION LINITS ACTION (Continued) c.

R shall not be increased above 50% of RATED TH POWER unless icated AFD has not been ou band for more than 1 ho e 25% target cumulative during the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Powe ses 0% of RATED THERMAL POWER do not --"'..

ing within the target ded the ace ty deviation is not violated.

4 SURVEILLANCE REQUIREMENTS i

4. 2.1 The indicated AXIAL FLUX OIFFERENCE shall be deterstned to be within j

a, its limits tr'r.;.T 07:07:0" he ;"" et OT ", 7,;--^^'., ;;.' ;; by:

l Monitoring the indicated AFD for each OPERA 8LE excore channel:

a.

j 1.

At least once per 7 days when the AFD Monitor Alare is OPERA 8LE, and a

l 2.

At least once per hour for ^J f'r;^. 20 7.;.r; ;fu c re;t..;rg Wgthe AFD Monitor Alarm t. ^.7:71".'

j

t;t.;. inepe/Mp i

b.

1 and logging the indicated AXIAL FLUX OIFFERE

{

0PERA8LE e nel at least once per first 24 houri ch 4

and at least once per i

DIFFERENCE Monitor A r, when the AXIAL FLUX noper logged values of the indicat DIFFERENCE shall be assume during the

{

a preceding ear.h 1ogging.

1 m

' 2.1.0,The indicated AFD shall be considered outside of its ;~~ ter;;^. =

i Lun n$

i p

when at least 2 0PERA8LE excore channels are indicating the AFD to be outside 115 Aan 8

tr.: t;r;;; ist.

P =:'.t> t r hti = : t:i d ;f 'J.; '** t;;;;; 1 24 ;t..i' t:

1

>....a_

a.

u..>.

i uN a*t-e.,

minute penalty deviation for each one minute of POWER OPERA i

We

• outs of the target band at THERMAL POWER levels equal t above

{

d* y g 50% of THE24AL POWER, and PM8 One-half minute 1ty deviation for each one > ute of POWER i

CPERATION outside o e target band at L POWER levels between l

155 and 505 of RATED TH POWER.

4.2.1.3 The target flux difference o e

ERA 8LE excore channel shall be determined by measurement at les nce per 92 octive Full Power Days.

The provisions of Specification

.4 are not applicab 4.2.1.4 The targe ux difference shall be updated at lea nce per 31 Effective F Power Days by either determining the target f difference pursuant

.2.1.3 above or by linear interpolation between the nos ecently value and 0 percent at the end of the cycle life.

The provisio of meas cification 4.0.4 are not applicable.

FARLEY-UNIT 1 3/4 2-2 AMENDMENT NO. 26

REAAca wrH us.4 %H

...__._.___._.__________..........._............._.u...-.

........_........,......._......_.._.._....2,_

..__..._...r......................___

................__._.._.......r.

..._.....r_.....

a g

.......... _ = ~.. =.......

W....__...

__._.._._...._...,..........................r....................._

e

......._.................... = -.

N..........__.....................................,.............,.,................

ye.,

j.. T e

y 7

e 1.......

............ 31.

................................................ ~.......................

............................ _ i 40............................................................................................................................

3........................................................

........u............... ~........

20.................................................................

)

........ _ _.. _..... _. ~ _....

...................r 50 30 10 0

10 20 30 40 FLUX DIFFERENCE 41) %

FIGURE 3.21 AXIAL PLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL FOWER FARLEY-UNIT 1 3/4 2-3 AMEdOMDIT N0. 26

<e 3.j,.

120 i

( 12.100)

(+9.100) 100 l

l l

1

~

UNACCEPTABLE

(

t OPERATION

\\

UNACCEPTABLE g

/

\\

OPERATION I

80 I

I ACCEPTABLE k

/

OPERATION

{

i e

/

\\

I L

60 2

I

\\

. (-30, 50)

I

\\

g

(+24,50) 40 20 0

-60

-50

-40

-30

-20

-10 0

10 20 30 40 50 60 AXIAL FLUX DIFFERENCE (DELTA I)%

Figure # 3 d'/

Axial Flux Difference Limits as a Function of Rated Thermal Power for RAOC

4 i

POVER DISTRIBLITION LIMITS 3/4.2.2 BEAT PLUI BOT CRANNEL PACTOR - F (Zj 7

LIMITING CONDITION POR OPERATION

,/

l 3.2.2 F,(2) shall be limited by the fcllowing relationships:

i i

F (Z) - [ p--] [K(Z)] for P > 0.'5 ict VAFTAGE 5 fuel l

< 2.45 2

o l

)

P,(Z) 3 [4.9) [K(Z)) for P f 0.5 for VANTAGE 5 fuel and l

3 I

4 P (2) < [2.32] [K(Z)] for P > 0.5 for LOPAR fuel e

y--

l i

)

}

F,(Z) f [4.64) [K(Z)] for P f 0.5 for LOPAR fuel l

1 i

i where P = THERMAL POVER l

j M

and K(Z) is the function obtained from Figure 2-or a given core height location.

4 l

APPLICA8ILITT:

MODE 1 ACTION:

Vith P,(Z) exceeding its limits Reduce THIRMAL POWER at least 1% for each 1% P (Z) exceeds the limit a.

vithin15minutesandsimilarlyreducethePovErRangeNeutron Plux-Eigh Trip 5etpoints 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 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />s: subsequent POWER OPP. RATION may proceed provided the Overpower delta T Trip Setpoints have been reduced i

at least 1% for each 1% P (Z) exceeds the limit. The ? ::;:r : delt: T l

Tri; 5::--12: ::d ::ier g-'1 h ---fe r f -19 "

r-eter i-

-t 1---t

-405-8541583, b.

Ifr i# --f

th: ::::: :f th: ::: :f lied: :: ditier ;:i:: ::

' c r--- f -- """"?? "'** 2 5: :: 9: ::d ::d 1!rit :: tired 5; 2.

15: ::

TEERNAL POVER any 4 hen be increased provided P,(Z) is demonstrated through incore mapping to be within its limit FARLEY - UNIT 1 3/4 2-4 AMENDMENT NO.

26. 73 92

l i

POWER O!STRIBUTION LIMITS

$URVEILLANCE REQUIREMENTS I

1 4.2.2.1 The provisions of Specification 4.0.*4 are not applicable.

i g(t) 17 4.2.2.2 shall be evaluated to determine if p is within its limit by:

Using the movable incore utectors to obtain a power distribution a.

map at any THERMAL POWER greater than 55 of RATED THERMAL POWER.

b.

Increasing the measured F component of the power distribution map by 35 to account for manuNeturing tolerances and further increasin i

j the value by $5 to account for esasurement uncertainties.

M/M j

u.nT4 aring the F,Y e g utM (F, ) obtai W in b, above to:

c.

C

1. 055Af 1.

F limits for RATED THERMAL POWER (Fxy ) for th appropriate U

l xy meas core planes given in e and f below, and i

l 2.

The relat nship:

1 F'

= F " (1+0.2(1-P))

My where F ' is the it for fractio THERMAL POWER operation 1

W of F,RT9 l

expressed as a funct P is the fraction of RATED l

THERMAL POWER at which measured.

I d.

Ressasuring F according to lowing schedule:

4 4

1 C

1.

When F is gree than the F 1

it for the appropriate W

W L

esasured core ane but less than the relationship, additional C

j power distr ution maps shall be taken a compared to F,RTP and F ':

y i

l a)

Either within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding by 2 f RATED j

THEMAL POWER or greater, the THERMAL POWER at ch F,y was last determined, or l

b)

At least once per 31 EFPO, whichever occurs first.

4 3

I i

i i

FARLEY-UNIT 1 3/4 2-5 AMEN 0 MENT NO. 26 i

1 l

I AM24 7 - Ph4E 3/4 '3 - 5 Determining the computed heat ' lux hot channel factor F b.

C follows:

Q (7),,,

Increase the measured F (Z) obtained from the power distribution g

map by 3% to account for manufacturing tolerances and further increase the value oy 5% to account for measurement uncertainties.

C c.

Verifying that Fg (Z), obtained in Specification 4.2.2.2b

above, satisfies the relationship in Specification 3.2.2.

d.

Satisfying the following relationship:

Cg(Z)5(N x K(Z) for F > 0.5 F

P x w(Z)

C RTP Fg (7) g x K(Z) for P < 0.5 0.5 x W(Z)

~

C Where Fg (Z) is obtained in Specification 4.2.2.2b above, F RU q

is the F limit, K(Z) is the normalized F (Z) as a function of core g

q height, P is the fraction of RATED THERMAL POWER, and W(Z) is tne cycle dependent function that accounts for power distribution transients encountered during normal operation.

)

Measuring F (Z) according to the following schedule:

e.

g 1.

Upon achieving equilibrius conditions after exceeding by 20% or more of RATED THERMAL POWER, the THERMAL POWER at which F (Z) was last determined", or 9

2.

At least once per 31 Effective Full Power Days, whichever occurs first.

"During power escalation after each fuel loading, power level say be increased until equilibrius conditions at any power level greater than or equal to 50%

of RATED THERMAL POWER have been achieved and a power distribution map obtained.

f o"'

F

= 2.45 (Vantage 5 fuel)

= 2.32 (LOPAR fuel)

K(Z) provided in Figure 3.2 2 W(Z) provided in the Radial Peaking Factor Limit Report

j I

i I

POVER DISTRI8ttfION LIMITS SURVIILLANCE REQUIREMDrfs

...........................(Continued) j 2.

When the F

• is less than or equal to the F TP appropriate'measuredcoreplane,addignalp5verdistributio R

]

limit for the shall be taken and F,y* compared to F,y and F,y* at least 5

ETPD.

ps RWN ce per l

$m4 e.

The F 11 core Nanes e for RATED THERMAL POWER (FRTp) shall be 647 j

planes in a Radiaining bank "O" control r5ds and al vided for all 4

i 6.9.1.11.

Peaking Factor Limit Report rodded core pecification i

f.

The F, legions as sensured i limits of e, abov i

plane are not appl ble in the following core n

reent the fuel:

core height from the bottom of j

i 1.

Lower core region free 0 15%,

lusive.

i 2.

Upper core region f 85 to 100%, inel I

ve.

1 3.

Grid plane re l

of the gri ns within. 2% of core height ound the midpoint j

j 4.

Core j

ane regions within. 22 of core height (. 2.88 th demand position of the bank "D" control rods.

es) about g.

V F

exceeding F 'if Fthe effects of F *it evaluatIMi on F (2) shall be to determine

, (Z) is within sllaits.

onsens<,7we GQ*nes*Enn *FS*QQ,"y py 4.2.2.3 When F'shall be obtained fissensuredforothe (Z) measured F, t for annufacturing tol (Z) t;;..'.r.;.;hr...

an overall rom a power 3% to accoun ution map and in:reased by account for measurement uncertainty.erances and further increased by 5X to i

i i

i 1

1 i

i i

j 4

FARLEY-UNIT 1 3/4 2-6 i

AMENDMDrf 90 51.r 1

)

i, __.

/ 43 5 4. T - p n<a E, 3/4 2 -Q f.

With measurements indicating I, *(2) h maximum F

over 2

( K(Z) )

has increased since the previous detemination of F, *(Z) either of the following actions shall be taken:

1)

Increase F, *(2) by the N*ca) Nar ^<= specified in Tom #

FEA KnWe FMTCA k'M*T AE N Yah$ verify that this Vaiue satisfies the r'elationship in Specification 4.2.2.2d, or F, *(Z) shall be esasured at least once per 7 Effective Full 2)

Power Days until two successive maps indicate that F, "(Z)\\

is not increasing.

maximum over 2

\\ K(Z)

With the relationships specified in Specification 4.2.2.2d above not g.

being satisfied:

1)

Calculate the percent F,(Z) exc9eds its limits by the following expression:

(

[ maximum F *(Z) x WII) h 3

hx100forp>0.5 over Z F,

• x K(Z)

-1 p

I

[ maximum F 'II) x WIf1 F

- j h

over Z F,

• xK(Z)

-1 x 100 for p s 0.5, and

_W 2)

The following action shall be taken*

Within 15 minutes, control the AFD to within new AFD limits which are detemined by reducing the AFD limits specified in Aw12',w==ffw %

exceeds its limits as delem by 15 AFD for each percent F,(Z) ined in Specification 4.2.2.2g.1.

Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, reset the AFD alam setpoints to these modified limits.

h.

The limit specified in Specffication 4.2.2.2c are applicable in all core pla regions, i.e., 0 - 2005, inclusive.

1.

The limits specified in Specifications 4.2.2.2d, 4.2.2.2f, and 4.2.2.2g above are not applicable in the following core plane regions:

1)

Lower core region from 0 to 155, inclusive.

2)

Upper core region free 45 to 1005, inclusive.

Fowsm 02stm23UTXow 2,2MITs j

3/4.2.3 NUCLEAR INTuMPY NOT S_"NEL FACTOR - F,$

1

]

LIMITING CCerDITIcet FOR OPERATION 3.2.3 N a shall be liatted by the following relattenships i

\\

r#

2 h 5 1.70 (1 + 0.3 (1 - P1) for VANTAGE 5 fuel and j

uo l

NFg g 4,g-[1, o,3 (1 - 7)) for 14 PAR fuel J

where P=

j bp. f'$

i i

APPLICASILITY:

ICDE 1 ACTION

?

b with F g exceeding its limit:

N a.

Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> either 4

1.

N j

Restere F g to within the aheve 11atts and demeastrate j

through in-core aspying that F g is within its 11 alt within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of exceeding the limit, or e

i, 2.

Reduce TEENNAL POWER to less than 504 et RATED TEEftlhL POW j

and reduce the Power Range Neutrea Flum - Eigh Trip Setpoints to 5 SSt of RATED TEE 30thL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and l

b.

}

Demonstrate through ia-sere aspping, if not previously perferned Per N

{

a.1 above, that F g is withis its limit withia 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> af ter j

exceeding the limit er reduce TEEfethL POWER to less than 5% of RATED TEEppthL POWER within the mest 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and Identify and correct the cause of the out of limit condition prior c.

l to increasing TIEfBqhL POWER above the reduced limit required by a or b, eheves subsequent POWER OPERATI0ff any proceed provided that F g N

is demonstrated through la-sere mapping to be within its 11 alt at a

{

nominal 50% of RATED TIBIDAL POWER prior to escoediaq this TIE 30thL POWER, at a naminal 75% of RATED TIEIDELL POWER prior to exceeding j

this TREfDqhL POWER and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> af ter attaining 95% or i

greater RATED TERfDEhL POWER.

4

}

i 1

4

{

FARLEY - UEIT 1 3/4 2-0 AMENEfqENT 300. 20.37.04,92,109 i

i

I

• able 3.7-1 MAXIMUM ALL0wA3tt PCwtR RANGt NEV*RCN FLUX MIGN stT INCptRA3Lg STEAM L:NE SAFETY VALVES SURING ) LOCP OPERATION ~

Maximum Number of Ineperable Safety Valves on Any Maximum Allowable Fewer Range operatine Steam Generator Neutron Flux Nigh Setpoint iPercent of RATED TMERMAL powtai 1

99 @ '"*

1 4 93 3

ee-if TAsts 3.7-1 MAXIMUM ALLCWAstE PowtR RANGE NEUTRCN TLUX 1%fA SETPCINT W INOPERASLE STEM LINE SAFETY VALVES DURIN47 WF OPERATION Maximum Number of Ineperable Safety Valves en Any Masimus Allowable Power Range l

Operatine Steam Generater*

Neutrea Flux Migh Setpoint (Percent of RATED TME?. 'J. PowtR) i 1

2 3

\\

1

• At least two safety valves shall be ofERASLR on the sea-operating steam generates.

"These values left blank pendlag NRC approval of 2 leep operation.

/4/ 55 U-->

rARLEY-UNIT 1 3/4 7-2 AMENnMENT No. 26.112 1

INSERT page 3/4 7 2 For plant operation approaching end of cycle (i.e., core average burnup a 14,000 MWD /MTU), with one inoperable safety valve on any steam generator, the maximum allowable Power Range Neutron Flux setpoint may be increased from 60% to 87% RTP, l

l l

t I.

h

3/4. 2 PO'JER DISTRIBUTION LIMITS 1

l BASES i

The specifications of this section provide assurtnce of fuel integrity i

during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (a) seeting the DNB design criterion during normal operation and in short ters transients, and (b) limiting the fission gas release, fuel pellet j

temperature and 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.

1 7

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

these specifications are as follows:

1 F,(Z)

Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation 2 divided j

by the average fuel rod heat flux, allowing for manufacturing i

i tolerances on fuel pellets and rods and measurement uncertainty.

l 4

F"'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 avert.ge rod power.

f O

o p e it in o

plane e

i j

3/4.2.1 AXIAL FLUX DIFFERENCE i

The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the F bound envelope of 2.45 for VANTAGE 5 and 2.32 for LOPAR times the,(Z) upper normalized axial peaking factor is not exceeded during either normal operation or in the event of xenon redistribution following power changes.

s e et flux difference is determined at equilibrium xenon conditions.

full long may be positioned within the core in accordance vi r

j' respective insert s and should be inserted near thei position for steady state operation ver levels. Th e of the target flux i

difference obtained under these con y the fraction of RATED

ogggyp, 1

THERMAL POVER is the target flux diff at THEMAL POVER for the i

associated core burnup condit arget flux differe r other THERMAL POVER levels are obtai multiplying the RATED THERMAL P0 by the j

appropriate fr TEERMAL POWER level. The periodic updating of t target iference value is necessary to reflect core burnup erations.

i

)

i I

FARLEY - UNIT 1 8 3/4 2 1 AMENDMENT NO.

25. 73.

92

POWER DISTRIBiffION LIMITS BASES AXIAL FLUX DIFFERENCE (Continued) ough it is intended that the plant vill be operated with the within th 5)% target band about the target flux difference, d plant THERMIL reductions, control rod motion vill cau rapid e AFD to deviate outside of arget band at reduced TIERMAL levels. This deviation vill not affect xenon redistributi ficiently to change the y envelope of peaking factors wh y be re on a subsequent return to RATED TIERMAL POWER (with the AFD v e target band) provided the time duration of the deviation is lia Ac ingly, a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> penalty deviation limit cumulative during th vious 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ovided for operation outside of the target band bu in the limits of Figure

-1) while at TIERMAL POVER levels be 50% and 90% of RATED TIERMAL POVER.

TRERMAL POVER levels bet 5% and 50% of RATED THERMAL POWER, deviations e AFD outs the target band are less significant. The penalty of 2 al time reflects this reduced significance.

Provisions for monitoring the AFD on an automatic basis are derived from 15 the plant process computer through the AFD Monitor Alars.

The computer determines the one minute average of each of the OPERABLE excore detector outputs and provides an alars message immediately if the AFD for 2 or more OrzaABLE excore channelswee outside the tr;n 'rIgnd the TIERMAL POVER

!;. ; h in u:: }^* r f ^^" r f 5: :r: r ' * ? -- f

• ^! ".".......- "..'.' "'"

is greater than ^C J RATED TEIRMAL POVER.

ri ; 1 r ni:: : t--"."""

e 3...

......u__

.6.

- -i.

2 M

n; _a. r:WJ_................ z n z. z.. a..,.. - ~ ' ' '

.._......u_....;

~ ~ ~ ~ ~ - ~ ~ - - - - - ~ ~ ~

- - ~

,g fi;r: 5 ?!* 2 1 -Mr:

tr-i cl ---*M r 'e : t

'--f.

8[ ')

y 3/4.2.2 and 3/4.2.3 HEAT FLUI B0T CIANNEL FAC"f0R, NUCLEAR iini=L 90ense w par. as u I 80T CHANNEL FACTOR The limits on heat flux hot channel factor, and nuclear enthalpy rise hot channel factor ensure that 1) the design limit on peak local power density is not exceeded, 2) the DNS design criterion is set, and 3) in the event of a LOCA the peak fuel clad temperature vill not exceed the 2200'F ECCS acceptance criteria limit.

Each of these is sensurable 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 limits are maintained provided:

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

insertion differing by more than 12 steps, indicated, from the group demand position.

~

b.

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

The control rod insertion limits of Specifications.3.1.3.5 and c.

3.1.3.6 are maintained.

f d.

The axial power distribution, expressed in terms of AXIAL FLUX l

DIFFERENCE, is maintained within the limits.

FARLEY - UNIT 1 8 3/4 2 2 AMENDMENT NO. 28, 92

- - - - - - - - - - ~ - - - -

]

$$sk W*W 95 et j

son 755 eft NE est aft NE 10 5 55 10E 8

+tet 6

+5B NEDeCATW AMIAL PWK ORPM PIpse 8 SM St TYPICAL 18eceCATE ANIAL PWN OsPPEASIOS VEIIRS menas4L Pousa 1

EY - UNIT 1 8 3/4 2 3 AMDEDWIT 2, 26 l

\\

\\

t PCutR t!STRI Sff"!ON 1!MI?3

{

SAsts i

1. g will be malatained within its limits provided senditions a j

above are maintained.

allows changes la the radial power shape for all permis i

i limits.

3 and manufacturing tolerance must be made.when an Fg measurement is ta j

8 a full core map taken with the Lacere detector f142 asAn allowenee of St is appropriate for j

allowance as appropriate for asaufacturiaq tolerance pping system and a 30 j

/N5847-+

when r#g is measured, esperiasatal errer must be allowed for and 44 is the appropriate allowance for a full core map taken with the ineer system.

The specified limit for # g eestains sa et allowenee fore detection j

uncertainties.

The et allowance is based on the following seasiderations:

l Abnormal perturbations in the radial power shape, such as from rod a.

j masalignment, affect # g more directly than Fg, i

j b.

Although red sevement has a direct influence upon limiting Fg to within j

its limit, such control is met readily available to limit # g, and

{

c.

Errors in predictica for sentrol power shapa detected during startup physics tests can be compensated for la Fg by reatracting axial flux

{

distribution. This compensation for #g is less readily available J

j If # g exceeds its limit, the unit will be allowed 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to restore N

F g to within its limits.

l This restoration asy, for emagle, involve l

realigning any miss11gned reda er reducing power enough to bring F g within its power dependent limit.

When the #g limit is emeeeded, the DNSR 11mLt is not likely violated in steady state operation, because events that could i

significantly perturb the #g value, e.g., static centrol red misalignment, i

considered in the safety analyses.

are l

DNS 11alting event occurs while F g is above its limit.Bewever, the DNSA limit may The increased allowed action time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> provides an acceptable time to resters # g to within its i

limits.without allowing the plant to remain in an unacceptable condition for an extended period of time.

4

)

Once corrective setion has been takes, e.g., realigament of misa11gned i

rods or reduction of power, an insere flus map must be obtained and the esasured l

value of F g verified met to emeeed the allowed limit.

Twenty additional hours are provided to perferus this task above the four hours allowed by Action 2

statement 3/4.2.3.a.

The eempletion time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is aseeptable because of i

l the low probability of havtag a DNS 11att*y event withis this 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period and, is the event that power is reduced, at taeresse in DNS margia is obtained 4

at lower power levels.

Additionally, operating emperience has indicated that this completion time is sufficient to obtala the incere fium asp, perform the required calculattens, and evaluate F g.

W 5

i FARLEY - UNIT 1 8 3/4 2*4 AMENEDIENT WO. IIeN *II' i

(

1 i

INSERT page B 3/4 2 4 The heat flux hot channel factor F (Z) is measured periodically and increased by a o

cycle and height dependent power factor appropriate to RAOC operation, W(Z), to provide assurance that the limit on the heat flux hot channel factor Fo(Z) is met.

W(2) accounts for the effects of normal operational transients within the AFD tnm+ Apgg and was determined from expected power control maneuvers over the full range of burnup conditions in the core.

I j

POUER OfSTRIBtTTION LXMITS BASES j

Tn king factor F, (Z), is measured periodically to provid assurance that teun SM factor, F (Z), remain wi ^ i.. a mit. The F d

gg4 limit for RATED THERMAL POVER (

-h6 Radial Peaking Factor "

limit report per S e

.1.11 was detern ected power j

p r. - ers over the full range of burnup conditions in the co 3/4.2.4 OUADRANT 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 i

distribution measurements are made during startup testing and periodically l

during power operation.

l The limit of 1.02, at.which corrective action is required, provides DNB and l

linear heat generation rate protection with x-y plane power tilts.

i the two hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of i

a dropped or misaligned control rod. In the event such action does not correct the tilt, the margin for uncertainty on F is reinstated by reducing the maximum i

i allowed power by 3 percent for each perce,nt of tilt in excess of 1.0.

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

The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles. The two sets of four symmetric thimbles is a unique j

set of eight detector locations. These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, and N-8.

i 3/4.2.5 DNB PARAMETERS i

j The limits on the DNB related parameters assure that each of the parameters are i

maintained within the normal steady state envelope of operation assumed in the i

transient and accident analyses. The limits are consistent with the initial i

FSAR assumptions and have been analytically demonstrated adequate to meet the DNB design criterion throughout each analysed transient. The indicated T j

value of 580.7'F is based on the average of two control board readings an8,In indication uncertainty of 2.5'F.

The indicated pressure value of 2205 psig is i

based on the average of two control board readings and an indication l

uncertainty of 20 psi. The indicated total RCS flow rate is based on one elbov i

tap measurement from each loop and an uncertainty of 2.4% flow (0.1% flow is included for feedvater venturi fouling).

i The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> surveillance of Tavg and pressurizer pressure through the control board readings are sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

The 18 month surveillance of the total RCS flo'v rate l' a precision measurement that verifies the RCS flow requirement at the beginning of each fuel cycle and ensures correlation of the flow indication channels with the measured loop flows. The monthly surveillance of the total RCS flow rate is a reverification of the RCS flow requirement using loop elbow tap measurements that are correlated to the precision RCS flow measurement at the beginning of the fuel cycle. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> RCS flow surveillance is a qualitative verification of significant flow degradation using the control board indicators and the lobp elbow tap measurements that are correlated to the precision RCS flow measurement at the beginning of each fuel cycle.

FARLEY - UNIT 1 B 3/4 2-5 AMENDMENT NO. I/ 92

i j

l

}

ADMINISTRATIVE CONTROL 5 1

i i

l 4

1 l

1 I

5 l

MONTHtY OpfRATINC Rfp0RT i

i 6.9.1.10 including documentation of all challenges to the be submitted on a monthly basis to the Commission, pursuant to Q

s, shall ne later than the 15th of each month following the calendar j

the report.

l 1

/ ".^.^:2 PtAKfist Fact 0A LIMIT Rf^Giii l 4..

The F RTP limit for Rated Thermal Power (Fry l

• control rods and all unredded cbre) planesfor all core p1 containing established and dociassated in the Radial Peaking Facto Report before

'to 10 CFR 50.4, upo(n issuance.each releed eyele prior tiNOBE 1) a ggptut la tlw Commission, pursuant w,m at the limit would be submitted 4

fasur at some other time during care t will bersubmitted spes issuance, unless

) othe mise exempted by ission.

- s.

' Any in en needed te support F will be by request from the d

i insleded la this report.

i 1

afress J-Aftm att IAnttITY DATA irimmi 4.9.1.12 The ember of tests (valid er invalid) and the number of fa start en demand for each diesel generater shall be submitted to the NRC 4

annually.

This report shall cantale the infomaties identified la Regulatory j

Position C.3.b of WC Regulatory Guide 1.100, Revision 1,1977.

i 1

i FARLEY-LalIT 1 6-19 AMDOWif NO. W9.82.n 4

i

{

4 J

i

INSERT page 6-19 6.9.1.11 The cycle dependent function, W(Z), and the burnup dependent F C(Z)

Penalty Factors, required for calculation of F c(Z) specified in LCO).2.2, " Heat a

a Flux Hot Channel Factor F (Z)", shall be documented in the fleWhp Peaking Factor a

Limit Report in accordance with the methodology in WCAP-10216-P A, " Relaxation of Constant Axial Offset Control FQ Surveillance Technical Specification", Rev.1, i

February 1994 (W Proprietary).

/

The4hedie& Peaking Factor Limit Report shall be provided to the Commission, pursuant to 10 CFR 50.4, upon issuance prior to each reload cycle (prior to MODE 2). In the event that the limit would be submitted at some other time during core life, it will be submitted upon issuance, unless otherwise exempted by the Commission.

I e

l l

t l

~. - - -

l e

Q INDEX ADMINISTRATIVE CONTROLS SECTION PAGE Review...................................................

6-10 Audits...................................................

6-11 l

t Authority................................................

6-12 Records...........*********..*...........................

6-12 6.5.3 TECHNICAL REVIEW AND CONTROL Activities...............................................

6-12 Records..................................................

6-13 6.6 REPORTABLE EVENT ACTION.....................................

6-14 6.7 SAFETY LIMIT VIOLATION..................................... 6-14 6.8 PROCEDURES AND PROGRAMS.....................................

6-14 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS Startup Report.......................................... 6-15a Annual Report............................................

6-16 Annual Radiological Environmental Operating Report.......

6-17 Annual Radioactive Effluent Release Report...............

6-17 Monthly Operating Report.................................

6-19 Peaking Factor Limit Report..............................

6-19 l

Annual Diesel Generator Reliability Data Report..........

6-19 Annual Reactor Coolant System Specific Activity Report... 6-20 Annual Sealed Source Leakage Report......................

6-20 6.9.2 SPECIAL REPORTS...........................................

6-20 6.10 RECORD RETENTION............................................ 6-20 l

6.11 RADIATION PROTECTION PROGRAM................................

6-21a i

6.12 HIGH RADIATION AREA.........................................

6-22 FARLEY-UNIT 1 XIX AMENDMENT NO.

i 680

~

UNACCEPTABLE 660 2440 psia OPERATION 2250 psia 640-2000 psia F

1840 psia 600 ACCEPTABLE

~

OPERATION 580

~

560 O.0 0.2 0.4 0.6 0.8 1.0 1.2 POWER (FRACTION OF RATED THERMAL POWER)

Figure 2.1-1 Reactor Core Safety Limits Three Loops in Operation FARLEY-UNIT 1 2-2 AMENDMENT NO.

I TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION Note 1: Overtemperature AT k

3 r

3 r+rs r

3 1

I 1+rs s

T

- T' + K (P-P')- f (AI) i e

AT s AT, K, - K 3

i 2

s l + r ss (1 + r SJ k l + r ss

~

2 o

3 Measured AT by RTD instrumentation; where:

AT

=

Indicated AT at RATED THERMAL POWER and reference Tavg; AT

=

O Averege temperature, *F; T

=

Reference T at RATED THERMAL POWER (s 577.2*F);

T

=

ayg Pressurizer pressure, poig; P

=

2235 peig (nominal RCS operating pressure);

P'

=

1+tsy The function generated by the lead-lag controller for T,yg dynamic compensation;

=

y 3,,,

on t1 &T2 Time constants utilized in the lead-lag controller for T,yg, t1 = 30 see,

=

t2 = 4 sec; 1+ts The function generated by the lead-lag controller for AT dynamic compensation;

=

1+ts$

Time constants utilized in the lead-lag controller for AT, t4 = 0 sec, 25 s 6 see; T4 E v5

=

1+t, Lag compensator on measured T,yg; 6

16 Time constant utilized in the measured T,yg lag compensator, 26 s 6 see; l

=

-1; Laplace transform operator, sec s

=

a Operation with 3 loops Operation with 2 loops g

K3 = 1.17; Ky = (values blank pending l

c

.O K2 = 0.017; K2 = NRC approval of l

K3 = 0.000825; K3 = 2 loop operation) l

TABLE 2.2-1 (Continuadi REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

• g i

NOTATION fContinued1 to M:

and ft (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 w

d tests such that w

(1) for qt - 9b between -23 percent and +15 percent, f1 (AI) = 0 (where qt and qb are percent RATED l

THERMAL POWER in the top and bottom halves of the core respectively, and qt + 9b is total THERMAL POWER in percent of RATED THERMAL POWER);

(ii) for each percent that the magnitude of (qt - 9b) exceeds -23 percent, the AT trip setpoint shall be automatically reduced by 2.48 percent of its value at RATED THERMAL POWER; and (iii) for each percent that the magnitude of (qt - 9b) exceeds +15 percent, the AT trip setpoint shall be automatically reduced by 2.05 percent of its value at RATED THERMAL POWER.

Note 2:

Overpower AT u

s e

~

3 8

m

~

z + 7*,

7',

g g

AT 5; A T. K-K T

- Kc T

- T"

- f (AI) 3 2

(1 + r s; (1 + r s; (1 + r s; (1 + r ss 3

3 o

where: AT Measured AT by RTD instrumentation;

=

ATO Indicated AT at RATED THERMAL POWER and reference Tavg;

=

T Average temperature, F;

=

T' Reference T at RATED THERMAL POWER (5 577.2*F);

l

=

ayg K4 1.10; l

=

5g K5 0.02/*F for increasing average temperature and O for decreasing average

=

g temperature; h

K6 0.00109/*F for T > T", K6 = 0 for T s T";

l

=

T3" h hh www W W e W mmh h T yg @& mwn&m

=

1+T s3

TABLE 2.2-1 (Continued)

• 4 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION (Continuedl i

E Time constant utilized in the rate lag controller for T,yg, t T3

=

3 = 10 see; 1 + r4s I + r$s The function generated by the lead-lag controller for AT dynamic compensation;

=

l 4 & ts

= Time constants utilized in the lead-lag controller for AT, t T

4 = 0 sec, is s 6 see; l

I 1+ rgs

= Lag compensator on measured T,9; 6

= Time constant utilized in the measured T t

ayg lag compensator, T6 s 6 see; l

u a

d

= Laplace transform operator,

-1; sec f (AI)

= 0 for all AI.

2 Note 3:

The channel's maximum trip point shall not exceed its computed trip point by more than 0 4 percent AT span.

Note 4: Pressure value to be determined during initial startup testing.

Pressure value of s 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 ccaputed trip point by more than O.4 percent AT span.

5 5=

H 8

(

6 l

l SAFETY LIMITS I

l BASES The curves of Figures 2.1-1 and 2.1-2 are based on the most limiting result usinganenthalpyhotchannelfactor,F$H, of 1.70 for VANTAGE 5fuelandanF$Hof1.30forLOPARfuelandareferencecocinewitha l peak of 1.55 for axial power shape.

An allowance is included for an increase inF$ Hat reduced power based on the expression:

F$H= 1.70 [1 + 0.3 (1 - P)) for VANTAGE 5 fuel and F$H=1.30(1+0.3 (1 - P)) for LOPAR fuel 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 1 (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 delta T 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 j

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, pressurizer and the reactor coolant system

. piping and fittings 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 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 FARLEY-UNIT 1 B 2-2 AMENDMENT NO.

LJMITING SAFETY SYSTEM SETTINGS BASES Overoower AT The overpower delta T reactor trip provides assurance of fuel integrity (e.g., no fuel pellet melting) under all possible overpower conditions, 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 distribution, changes in density and heat capacity of water with temperature, and dynamic compensation for transport, thermowell, and RTD response time delays from the core to RTD output indication.

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

Pressurizer Water Level The Pressurizer High Water Level trip ensures protection against Reactor coolant System overpressurization by limiting the water level to a volume sufficient to retain a steam bubble and prevent water relief through the pressurizer safety valves. No credit was taken for operation of this trip in the accident analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System.

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

Above 10 percent of RATED THERMAL POWER, an automatic reactor trip will occur if the flow in any two loops drops below 90% of nominal full loop flow.

Above 36% (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 FARLEY-UNIT 1 B 2-5 AMENDMENT NO.

l

s _. _ __ _ _. -

.__._m

{

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE (AFD1 LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the limits specified in Figure 3.2-1.*

APPLICABILITY:

MODE 1 above 50% of RATED THERMAL POWER **

l l

ACTION:

a.

With the indicated AXIAL FLUX DIFFERENCE outside of the limits specified in Figure 3.2-1:

1.

Either restore the indicated AFD to within the limits within 15 minutes, or 2.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes.

i SURVEILLANCE REQUIREMENTS l

4.2.1 The indicated AXIAL FLUX DIFFERENCE shall be determined to be within its limits by:

l a.

Monitoring the indicated AFD for each OPERABLE excore channel:

1.

At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and 2.

At least once per hour with the AFD Monitor Alarm inoperable. l i

• The indicated AFD shall be considered outside of its limits when at least 2 OPERABLE excore channels are indicating the AFD to be outside its limits.

• • See Special Test Exception 3.10.2.

FARLEY-UNIT 1 3/4 2-1 AMENDMENT NO.

120 100

(-12, 100)

(+9,100)

Unacceptable Unacceptable Operation l

Operation 80 I

x i

O Acceptable a.

Operaton J

60 5

(-30, 50)

(+24, 50)

O W

40 8

20 0

-50

-40

-30

-20

-10 0

10 20 30 40 50 Axial Flux Difference (Delta 1)%

42 3 FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER FOR RAOC FARLEY-UNIT 1 3/4 2-2 AMENDMENT NO.

1 POWER DISTRIBUTION LIMITS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F g LIMITING CONDITION FOR OPERATION 3.2.2 Fg(Z) shall be limited by the following relationships:

Fg( Z ) 5 [2.,,_4.1) [K(Z)) for P > 0.5 for VANTAGE 5 fuel P

Fg(Z) 5 (4.9) (K(Z)) for P 5 0.5 for VANTAGE 5 fuel and Fg(Z) 5(M) (K(Z)) for P > 0.5 for LOPAR fuel P

Fg(Z) 5 [4.64) (K(Z)) for P 5 0.5 for LOPAR fuel 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 l

ACTION:

With Fg(Z) exceeding its limits a.

Reduce THERMAL POWER at least 1% for each 1% Fg(Z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Satpoints 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 delta T Trip Setpoints have been reduced at least 1% for each 1% Fg(Z) exceeds the limit.

l b.

THERMAL POWER may be increased provided Fg(Z) is demonstrated through incore mapping to be within its limit.

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

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

l a.

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

FARLEY-UNIT 1 3/4 2-3 AMENDMENT NO.

.=.

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) b.

Determining the computed heat flux hot channel factor Fg (Z), as follows:

Increase the measured Fg(Z) obtained from the power distribution map by 34 to account for manufacturing tolerances and further increase the value by 5% to account for measurement uncertainties.

Verifying that Fg (Z), obtained in Specification 4.2.2.2b above, c.

satisfies the relationship in Specification 3.2.2.

d.

Satisfying the following relationship RTP F

x K (Z ) fo r P C

O Fn (Z ) s

> 0.5 P x VV (Z )

F ** T ' x K (Z ) fo r P C

Fn (Z ) s s 0.5 0.5 x vv (Z )

C Where Fg (Z) is obtained in Specification 4.2.2.2b above, Fq is the Fg limit, K(Z) is the normalized Fg(Z) as a function of core height, P is the fraction of RATED THERMAL POWER, and W(Z) is the cycle dependent function that accounts for power distribution transients encountered during normal operation.

Fg

= 2.45 (VANTAGE 5 fuel)

= 2.32 (LOPAR fuel)

K(Z) provided in Figure 3.2-2 W(Z) provided in the Radial Peaking Factor Limit Report e.

Measuring Fg(Z) according to the following schedules 1.

Upon achieving equilibrium conditions after exceeding by 20%

or more of RATED THERMAL POWER, the THERMAL POWER at which Fg(Z) was last determined *, or 2.

At least once per 31 Effective Full Power Days, whichever occurs first.

• During power escalation after each fuel loading, power level may be increased until equilibrium conditions at any power level greater than or equal to 50% of RATED THERMAL POWER have been achieved and a power distribution map obtained.

FARLEY-UNIT 1 3/4 2-4 AMENDKENT No.

i POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) f.

With measurements indicating

'F C(Z)'

n maximum i

K(Z),

over(Z)

(

has increased since the previous determination of Fg (Z) either of the following actions shall be taken:

C 1)

Increase Fg (Z) by the Fg (Z) penalty factor specified in the Peaking Factor Limit Report and verify that this value satisfies the relationship in Specification 4.2.2.2d, or j

C 2)

Fg (Z) shall be measured at least once per 7 Effective Full Power Days until two successive maps indicate that

'F C(Z)'

o maximum is not increasing.

K(Z)s over (Z) s g.

With the relationships specified in Specification 4.2.2.2d above not being satisfied:

1)

Calculate the percent Fg(Z) exceeds its limits by the following expression:

C Fn (Z) x W (Z) m axim um

- 1 x 100 for P > 0.5 over Z F,7, q

P

.)

C Fo (Z) x W (Z) m axim um

- 1 x 100 for P s 0.5, an d over Z F,7, n

. 0.5

.)

2)

The following action shall be taken:

Within 15 minutes, control the AFD to within new AFD limits j

which are determined by reducing the AFD limits specified in LCO 3.2.1, Axial Flux Difference, by 14 AFD for each percent FQ(Z) exceeds its limits as determined in Specification 4.2.2.2g.1.

Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, reset the AFD alarm setpoints to these modified l

limits.

FARLEY-UNIT 1 3/4 2-5 AMENDMENT NO.

- ~ _ - _ _ _. ~

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) h.

The limits specified in Specification 4.2.2.2c are applicable in all core plane regions, i.e.,

O - 100%, inclusive.

1 1.

The limits specified in Specifications 4.2.2.2d, 4.2.2.2f, and 4.2.2.2g above are not applicable in the following core plane regions:

1)

Lower core region from 0 to 15%, inclusive.

2)

Upper core region from 85 to 100%, inclusive.

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

s FARLEY-UNIT 1 3/4 2-6 AMENDMENT NO.

j POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEARENTHALPYHOTCHANNELFACTOR-F$H i

LIMITING CONDITION FOR OPERATION 3.2.3 F5Hshallbelimitedbythefollowingrelationships

]

F$Hs1.70(1+0.3(1-P)] for VANTAGE 5 fuel and F5H $ 1.30 (1 + 0.3 (1 - P)) for LOPAR fuel l

where P =

RATED THERMAL POWER APPLICABILITY:

MODE 1 ACTION:

WithF$Hexceedingitslimits a.

Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> either:

I 1.

RestoreF$H to within the above limit; and demonstrate throughin-coremappingthatF$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 /> of exceeding the limit, or 2.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setpoints to n 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 />, and b.

Demonstrate through in-core mapping, if not previously performed per a.1 above, thatF$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 l

c.

Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above the reduced limit required by a or b, above; subsequent POWER OPERATION may proceed provided that F$H is demonstrated through in-core mapping to be within its l

limit at a nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THERMAL POWER prior i

to exceeding this TKERMAL POWER and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after l

attaining 95% or greater RATED THERMAL POWER.

t i

i 4

FARLEY-UNIT 1 3/4 2-8 AMENDMENT NO.

IABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH 'sETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING 3 LCOP OPERATION Maximum Number of Inoperable Maximum Allowable Power Range Safety Valves on Any Neutron Flux High Setpoint ODeratino Steam Generator (Percent of RATED THERMAL POWER) 1 60***

l 2

43 l

3 24 l

IABLE 3.7-2 E IMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE JJJG LINE S AFETY VALVES DURING 2 LOOP OPERATION Maximum Number of Inoperable Maximum Allowable Power Range Safety Valves on Any Neutron Flux High Setpoint Operatino Steam Generator *

(Percent of RATED THERMAL POWER) 1 2

3

• At least two safety valves shall be OPERABLE on the non-operating steam generator.

• • These values left blank pending NRC approval of 2 loop operation.

• • • For plant operation approaching end of cycle (i.e.,

core average burnup 2 14,000 MWD /MTU), with one inoperable safety valve on any steam generator, the maximum allowable Power Range Neutron Flux setpoint may be increased from 60%

to 87% RTP.

FARLEY-UNIT 1 3/4 7-2 AMENDMENT NO.

(

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) meeting the DNB design criterion during normal operation and in short term transients, and (b) limiting the fission gas l

release, fuel pellet temperature and cladding mechanical properties to within assumed design criteria.

In addition, limiting the peak linear power l

density during Condition I events provides assurance that the initial l

conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200*F is not exceeded.

l l

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

l Fg(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 manufacturing tolerances on fuel pellets and rods and measurement uncertainty.

F[g Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the s

highest integrated power to the average rod power.

l 3/4.2.1 AXIAL FLUX DIFFERENCE The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the Fg(Z) upper bound envelope of 2.45 for VANTAGE 5 and 2.32 for LOPAR times the normalized axial peaking factor is not exceeded during either normal operation or in the event of xenon redistribution following power changes.

Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer determines the one minute average of each of the OPERABLE excore detector outputs and provides an alarm message Lamediately if the AFD for 2 or more OPERABLE excore channels is outside the allowed AI operating space for RAOC operation specified in Figure 3.2-1 and the THERMAL POWER is greater than 50% RATED THERMAL POWER.

l l

l l

[

l FARLEY-UNIT 1 B 3/4 2-1 AMENDMENT NO.

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

+

POWER DTSTRIBUTION LIMITS BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR. NUCLEAR ENTHALPY HOT CHANNEL FACTOR l

The limits on heat flux hot channel fe.ctor, and nuclear enthalpy rise hot channel factor ensure that 1) the design limit on peak local power density is not exceeded, 2) the DNB design criterion is met, and 3) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200*F ECCS acceptance criteria limit.

Each of these is 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 limits are maintained provided:

Control rods in a single group move together with no individual a.

I rod insertion differing by more than i 12 steps, indicated, from the group demand position, i

b.

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

The control rod insertion limits of Specifications 3.1.3.5 and c.

3.1.3.6 are maintained.

d.

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

F$gwillbemaintainedwithinitslimitsprovidedconditionsa.through

d. above are maintained. TherelaxationofF%gasafunctionofTHERMALPOWER allows changes in the radial power shape for all permissible rod insertion limits.

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

The heat flux hot channel factor Fg(Z) is measured periodically and increased by a cycle and height deperaent power factor appropriate to RAOC operation, W(E), to provide assurar.ca that the limit on the heat flux hot channel factor Fg(E) is met.

W(E) accounts for the effects of normal operational transients within the AFD limits and was determined from expected power control maneuvers over the full range of burnup conditions in the core.

FARLEY-UNIT 1 B 3/4 2-2 AMENDMENT NO.

4 8

PAGE INTENTIONALLY LEFT BLANK l

l l

l FARLEY-UNIT 1 B 3/4 2-3 AMENDMENT NO.

POWER DISTRJBUTTON LIMfTS BASES l

WhenF$H is measured, experimental error must be allowed for and 4% is the appropriate allowance for a full core map taken with the incore detection system. The specified limit for Fh contains an 8% allowance for uncertainties. The 8% allowanee is based on the following considerations:

Abnormal perturbations in the radial power shape, such as from rod a.

misalignment, af fect F[H more directly than Fg, i

b.

Although rod movement has a direct influence upon limiting Fg to within its limit, such control is not readily available to limit FfH, and Errors in prediction for control power shape detected during startup c.

physics teste can be compensated for in Fg by restricting axial flux distribution. ThiscompensationforF$H is less readily available.

If F$H exceeds its limit, the unit will be allowed 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to restore Th to within its limits.

This restoration may, for example, involve realigning any misaligned rods or reducing power enough to bring F$H within its power dependent limit.

When the F$H limit is exceeded, the DNBR limit is not likely violated in steady state operation, because events that could significantly perturb the F[H value, e.g.,

static control rod misalignment, are considered in the safety analyses. However, the DNBR limit may be violated if a DNB limiting event occurs while F[H is above its limit. The increased allowed action time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> provides an acceptable time to restore F[H to within its limits without allowing the plant to remain in an unacecpcable condition for an extended period of time.

Once corrective action has been taken, e.g.,

realignment of misaligned rods or reduction of power, an incore flux map must be obtained and the measured value of F$H verified not to exceed the allowed limit.

Twenty additional hours are provided to perform this task above the four hours allowed by Action Statement 3/4.2.3.a.

The completion time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is acceptable because of the low probability of having a DNB limiting event within this 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period and, in the event that power is reduced, an increase in DNB margin is obtained at lower power levels. Additionally, operating experience has indicated that this completion time is sufficient to obtain the incore flux map, perform the required calculations, andevaluateFh.

FARLEY-UNIT 1 B 3/4 2-4 AMENDHENT NO.

POWER DISTRIBUTTON LIMITS BASES 3/4.2.4 OUADRANT POWER TILT RATIO The quadrant power tilt ratio limit assures that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during startup testing and periodically during power operation.

The limit of 1.02, at which corrective action is required, provides DNB and linear heat generation rate prot =ction with x-y plane power tilts.

The two hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a l

dropped or misaligned control rod.

In the event such action does not correct the l

tilt, the margin for uncertainty on Fo is reinstated by reducing the maximum allowed power by 3 percent for each N rcent of tilt in excess of 1.0.

l For purposes of.Tonitoring QUADRANT POWER TILT RATIO when one excore detector is I

inoperable, the movable incore detectors are used to confirm that the normalized symriiric power distribution is consistent with the QUADRANT POWER TILT RATIO.

The ancore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles. The two sets of four symmetric thimbles is a unique set of eight detector locations.

These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, and N-8.

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

The indicated T value of 580.7'F is based on the average of two control board readings and kn indication uncertainty of 2.5'F.

The indicated pressure value of 2205 psig is based on the average of two control board readings and an indication uncertainty of 20 psi.

The indicated total RCS flow rate is based on one elbow tap measurement from each loop and an uncertainty of 2.4% flow (0.1% flow is included for feedwater venturi fouling).

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> surveillance of T and pressurizer pressure through the control av board readings are sufficient t$ ensure that the parameters are restored within their limits following load changes and other expected transient operation.

The 18 month surveillance of the total RCS flow rate is a precision measurement that verifies the RCS flow requirement at the beginning of each fuel cycle and ensures correlation of the flow indication channels with the measured loop flows. The armthly surveillance of the total XE flow rete is a reverification of the RCS flow requirement using loop elbow tap measurements that are correlated to the precision RCS flow measurement at the beginning of the fuel cycle. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> RCS flow surveillance is a qualitative verification of significant flow degradation using the control board indicators and the loop elbow tap measurements that are correlated to the precision RCS flow measurement at the beginning of each fuel cycle.

l FARLEY-UNIT 1 B 3/4 2-5 AMENDMENT NO.

._.m.

t l

ADMINISTRATIVE CONTROLS MONTHLY OPERATING REPORT 6.9.1.10 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORV's or safety valves, shall' be subiaitted on a monthly basis to the commission, pursuant to 10 CFR 50.4, no later than the 15th of each month following the calendar month covered by the report.

PEAKING FACTOR LIMIT REPORT l

C 6.9.1.11 The cycle dependent function W(E) and the burnup dependent Fg {g)

C yenalty factors, required for calculation of Fg (E) specified in LCO 3.2.2, " Heat Flux Hot Channel Factor - Fg(Z)," shall be documented in the Peaking Factor Limit Report in accordance with the methodology in WCAP-10216-P-A, " Relaxation of Constant Axial Offset Control FQ Surveillance Technical Specification," Rev. 1, February 1994 (M Proprietary).

The Peaking Factor Limit Report shall be provided to the Commission, pursuant to 10 CFR 50.4, upon issuance prior to each reload cycle (prior to MODE 2).

In the event that the limit would be submitted at some other time during core life, it will be submitted upon issuance, unless otherwise exempted by the Commission.

ANNUAL DIESEL GENERATOR RELIABILITY DATA REPORT l

6.9.1.12 The number of tests (valid or invalid) and the number of failures to start on demand for each diesel generator shall be submitted to the NRC annually.

This report shall contain the information identified in Regulatory Position C.3.b of NRC Regulatory Guide 1.108, Revision 1, 1977.

4 i

1 FARLEY-UNIT 1 6-19 AMENDMENT NO.

l F N P U nit 2 Technical Speci6 cations Channed Pagt1 Unit 2 Bevision XIX Replace Page 2-2 Replace Page 2-8 Replace Page 2-9 Replace Page 2-10 Replace Page.B 2-2 Replace Page B 2-5 Replace Page 3/4 2-1 Replace Page 3/4 2-2 Replace Page 3/4 2-3 Replace Page 3/4 2-4 Replace Page 3/4 2-5 Replace Page 3/4 2-6 Replace Page 3/4 2-8 Replace Page 3/4 7-2 Replace Page B 3/4 2-1 Replace Page B 3/4 2-2 Replace Page B 3/4 2-3 Replace Page B 3/4 2-4 Replace Page B 3/4 2-5 Replace Page 6-19 Replace

IEE A S IN! mlATIVE CONTROLS j

M M

Aevies.....................................................

6 10 Agadits.....................................................

$.1]

Authority..................................................

$.12 i

M...................................................

O.II 6.5.3 TEDm! CAL MV15 AM CW1ER, i

l Acttvities.................................................

6 12 t'

Records....................................................

6 13 a.a apostanti rynff E Tfou......................................

6 14 1

a.7 Sa N ff LIMIT V1GLAT1du.......................................

6 16 I

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i.

4.9.1 M BTIM M P E TS

(

1 Startup esport.............................................

6 18e Annual Aspert..............................................

4 16 a

a i

Assual Radielegtsal Eartrasmastal gerettag Aspart.........

6 17 4

Amamal Andteestive Effleest Releens assert.................

6 17 l

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i.

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6 19 Amment stesel ammerater anitamitty Faa esport............

s.ts Amamal Anaster Caelast % stas bestfie Asttvity esport.....

65 Aument Sealed Seures I,aakege esport........................

s.as

4. 9.3 WEIM. WWI5...........................................

45 aan aussa mraman............................................s.m 6.stal

. i.

.m 4.-

8 12 EXE M M.........................................

6 22 Fatf. WIT I III 6 5.88 81.85

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N.

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A i

Rea h * * "8"'"f"#'**"'

i 1

670" l.

660. N

\\

/

j UNACCEPTABLE

/

OPERA 110N

{

2440 psia 650"%

i s

4

640, 1250 psia

{

)

/

i 630 "

I 1

W 2000 psia a

i y 620-

/

e i

1875 psia

)

610 '

1840 pela J

\\

/

~

600'

/

i

{

$90' ACC ASLE s

i O

TION i

see, N,

\\\\\\\\

0.

.1

.2

.3

.4

.5

.4

.7

.6

.9 1.

1.1 1.2\\

POWER (PRACTION OP RATED THERMAL POWER)

Figure 2.1 1 Reactor Core Safety Limits Three Loops in Operation FARLEY - UltIT 2 22 AMENDNINT NO. 27. II.

79, 85

o f ) M F/ 6 s4 A E 2.) ~/

680

~

UNACCEPTABLE 660 2440 pse OPERATION 2250 psia 640 2000 psia iL

• w 1840 psia 600 ACCEPTABLE OPERATION 560 0.0 0.2 0.4 0.6 0.8 1.0 1.2 POWER (FRACTION OF RATED THERMAL POWER)

Figure 2.1-1 Reactor Core Safety Limits Three Loops in Operation

A

- ~ ~

1 T

TABLE 2.2 -1 (Cont inued)

REACTOR TRIP SYSTEN INSTRistENTATION TRIP SETPOINTS

• g s

NOTATION

.E 4 Note 1: Overteeperature M g

K (1 e T, s ) $ N, ( K, - K, ( 1. T, s )

(T (

I ) - T*) e K, (P - P') - t, (al)]

y (1 + T,s)

(1 +T,s)

T,s n

ubere R = Ileasured K by RfD insarueenaaaion, i-awwl rabene E, = Indicated M at RATED Tueensat poggs; A

T Average temperature 'F;

=

P T' ' 5??. ^7 '"12 Reference T.., at RATED TIIERNAL POV (6 6771 b

• F Pressurizer pressesre, psig;

=

P' - 2235 psig inal RCS opetating pressure);

[

y I*Tsa

- The function genesated by the lead-lag controllet for T,,, dynamic e_ompensation; 1 +

T, s T, & T,

= Time constants utilized in the lead-lag contaoller for T, r, - 30 sec, r, - 4 see;

= The function generated by the lead-lag controller for AT dynamic compensation; I + T,s rg = 4 SEC-i l

T, & T,

- Time coastants utilized in the lead-lag controller for AT, T,

O sec;

=

A i

. Lag compensator on measured 7,,,;

I * *s *

( $ (osee l

T,

= Time coastant utilized in the measured T,,, las compensator, T,

yec;

- Laplace transform operator, sec~';

s operation with 3 loops g

I,IT operation with 2 loops K,

M;

$G O.017 K, - (values blank pending l

K, O. :;2M;;

K,

. NHC approval of l

3 0.00082S K,

- " "" ' J "; ;

K, 1 loop opes.st iosi) l

1 I

l i

TABLE 2.1-1 (Continued)

REACTOR TRIP SYSTEN INSTRIBGENTATION TRIP SETPOINTS i

n E

NOT& TION (continued) i i

w i

and f ( AI) nuclea,r ten eh==hers; with gains to be selected based on measured instrumentis a funct g

A tests such thats response during plant startup d

-2.3 N

4/5 (1) for q -

between 4 percent and *44-percent, f ( AI) = 0 (where q and q ase percent RATED l

l i

M in the top and bottom halves of the c, ore respectively, a,nd q,,, q, is total TERANAL FOWER is percent of RATED THERNAL POWER);

-13 (11) for'ench percent that the magnitude of (q, - q,) exceeds @ percent, the AT trip setpoint shall be automatically reduced by 4T99 percent of its value at RATED THERNAL POWER; and

{

1.96 gjg (iii) for each percent that the magnitude of (q, - q,) exceeds w44 percent, the AT trip seapoint shall i

l be automatically reduced by +:4+ percent of its value at RATED THERNAL POWER.

2.oS Note 2: Overpower M R (1, T,s) f M, lK,- K, (

Ts ) (

1

)T-K, (T (

1

) - T") - t, ( AI ) ]

3 (1, T,s)

I

• T s 1 +

T, s I +

T, s 3

where:

AT - Neasured K by RTD instrumenIatlong enal ytf6fentA T**)

M,. Indicated M at RATBD THERNAL FOUER; 4

i T = Average temperature, 'F; i

i T* = Reference T at RATED THERNAL F0WER (C: lit:: != :-, rrature fer r ! etr- - '

i 577.2*F);'"

l.10 K, - 4,47; l

K, - 0.02/*F for increasing average temperature and 0 for decreasing average temperature; f

Q.00104 K, - 4:4NHM/*F for T > T", K, O for T f 7";

l E

Ts

- The function generated by the rate lag controller for T, dynamic compensation; g

a 1 + T, s 1

0l%

• I

%4

i TABLE 2.2-1 (Continued) m N

E REACTOR TRIP SYSTEN INSTRUNENTATION TRIP SETPOINTS 4

NOTATION (Continued) s E4 w

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

3 10 see; T -

3 I*Ts

- The function generated by the lead-lag controller for AT dynamic compensation; e

I+Ts3 Osa.,

tg T, & T, - Time constants utilized in the lead-lag controller for AT, T,

T, sec; I

- Lag compensator on measured T,,,;

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

T sec; s - Laplace transform operator, sec '*;

f 2( AI) - O for all al.

g,4 gg.,

j Note 3:

The channel's maximum trip point shall not exceed its computed trip point by more than percent,.

l Nate 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.

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

l A

A I

SAFETY LZRZTS

, - J. 3 0 BASES l

The cutres of Figures 2.1-1 and 2.)-2 are based on the most limiting result using an eithalpy hot channel factor. F",,. of 1.65 for VANTAGE 5 fuel and an F",, of 44M-for LOPAR fuel and a reference cosine with a pgak of 1.55 for axial power shape.

based on the expressionsAn allowance is included for an increase in F,, at reduced power F",, - 1.65 [1 + 0.3 (1-P)} for VANTAGE $ fuel and I.30 j

F",, = -1 :-H- [ 1

• 0. 3 ( 1-P ) ] f or LOPAR f uel j

vhere P is the fraction of RATED THERMAL POVER.

These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allovable control 1

rod insertion assuming the axial power imbalance is within the limits of the f.

(delta I) function of the Overtemperature trip.

is not within the toler&nce, the axial power imbalance effect on theWhen the axial pow i

Overteeperature delta T trips vill 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 i

Coolant System from overpressurization and thereby prevents the release of 1

radionuclides contained in the reactor coolant from reaching the containment i

atmosphere.

i The reactor pressure vessel, pressurizer and the reactor coolant systes piping and fittings 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.

design criteria and associated code requirements.The Safety Limit of 2735 psig j

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

FARLEY - UNIT 2 B 2-2 AMENDMENT NO.

35

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

1

)

LIMITING SAytTY SYSTEM SETTINGS i'

BASES l

Oversover er

{'

The Overpower delta T reactor trip provides assurance of fuel integrity (e.g., no fuel pellet selting) under all possible overpower conditions.

limits the required range for Overtemperature delta T protection, and provides l I

a backup to the Eigh Neutron Flux trip.

i h setpoint includes corrections for axial power distribution, changes in density and heat capacity of water 1

i with temperature, and dynamic compensaties for transport, thereowell, and RTD response time delays from the core to RTD output indication. I to credit was en for operation this trip la tae teent analyseal hows its fume capability at specified trip ting is required by DE"T specif1 en to amhamea the es11 reliabili l

Srstea.

the Reactor Protect j

Pressuriser Pressure The Pressuriser Eigh and Low Pressure trips are provided to limit the pressure range la which reactor operation is pareitted.

N Eigh Pressure trip is backed up by the pressuriser code safety valves for RCS everpressure protection, and is therefore set lower than the set pressure for these valves (2445 peig).

m Low Pressure trip provides protection by trippias the reactor in the event of a loss

{

of reactor coelaat pressare.

l pressuriser Water Level i

h Pressuriser Righ Water Level trip ensures protecties agatast. Reactor Coolaat System overpressurianties by 11alting the ester level to a volens sufficient to retala a steam bubble and prevent unter relief threegh the i

presseriser astery valves. No credit was taken for speesties of this trip in the i

accident analyses: heuwwer, its functiemal capability at the specified trip i

setting is required by this specificaties to enhance the overall relish 111ty of the Reacter Protecties System.

l Less of ylow h Less of Flev trips provide core protecties to prevent IEW in the event of a loss of een er more reacter eselaat peeps.

i Above 10 percent of RATED TERENAL POWER, an antenatic reacter trip will j

occur if the flev ta any two leeps drop belov 908 of naataal full leep flow.

i Above 36E (P-4) of RATED TERRNAL p0WER, auteentic reactor trip will occur if l

the flev la any single loop drops below 90E of assiaal full loop flow. This l

l

}

j pdNEET - Wir 2

~

5 2-5 N NO. 85 i

1 1

1 5

3/4.2 POWER OISTRIBUTION t.IMITS 3/4.2.1 AXIAL FLUX DIFFERENCE TM LIM nT3 %PEcineoD LIMITING CON 0! TION FOR OPERATION IA) FA nnE 3.2-l.+

3.2.1 The indicated AXIAL FLUX OIFFERE,NCE (AFO) shall be mainta,ined within-u _a_.,,.....ra....,.m a. a. s......._____

..a...,

.u r

\\

I APPLICA811,ITY: MODE 1 above 50% of RATED THERMAL POWER **-

\\

ACT10N' j.lhITs.SPEcsFrED pg p/4ud e 3.2 -1 With the indicated AXIAL FLUX OIFFERENCE outside of the_ '" : g a.

m__2

_m...

__..,.____..m

.s..._2..,,._._.___.__._ _. _ a. 6_. vue. _..

__2 aauca.

1.

an.. -.. a,en vue__._ anura m4. 6:_ 3a._.........._-__.

.m.._

-et-Either restore the indicated AFD to within the ter ;; M..

V u,wa 7 11*jit Of

,,,,,,a j

. ~

....n

..n..

2.

";^ z x **" :..' ^^* d==.vec tur n anwen.

.....m

__am_.

v....___.

,e m_m

.m_

u.._

m_.

2.._,_.

.m u u....

.a

,a.,.._._A__a___ -.r TL

.k,_

.L_

1,_,._

_L O:-d.!buce THERMAL POWER to 11ess than 505 of RArr.o 1 W AL POWER within 30 einutes.

.. _ ~ _ _.... _._ _.....

.....___....,... ~...._

.u

.m__

er. W.....

.W e.

T,

.lF

.u. -. s anuem

.u

.. A

u.._

....,m..k..

u

__.,,..s.i._._.__,......,,.u.___m____u.....ri~.

_..__..__..__._u.., _ _ - _. -. s. a s.. a. _-

,.____,___u.

......ug_ _ _..,.

F.,

p.i...

_7_

e.. e.

3..

..J...J

_2 2_2

.L_

. a,..._2 gen m..,.

.s.>_

,u.,

.a u.. u...._

_____... _ _.. u,s

_..a a 6

.m

.......u.

., ou

._......g

..m_m

. m... m.,=

..=,e m.._,__

.. u...

._._,_.._..,m 2..4.. 4...

___.i. -_.

' ". " - ^ ^. ^. '.

^~_~ll 2.-!I. ; M i

,_m.__m2

.-....m,

. a.. m.

m__m

__2

.-.n n...

.m

_.. s,,,,__..u.

u__

.. _ _ _..... a.. a.,. a_.

'" "See Special Test Exception 3.10.2 w3ERT Nha arg N.m N5 t T PM FAALEY-UNIT 2 3/4 2-1

r 1

i POWER DISTRIBUTION LIMITS l

ACTION (Continued) c.

POWER shall not be increased above 50% of RATED TH I

POWER un es k Jndicated AFD has not been out e 25% target i l

band for more than 1 h3ar penaity cumulative during the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

POWER do noLxegtrTr'eleing within the target blinttPowWses ect M penalty deviation is not violated.

ed the i

i SURVEILLANCE REQUIREMENTS i

4.2.1 The indicated AXIAL FLUX DIFFERENCE shall be determined its limits d;.ria; PO' " 0"C."' TION ;M a 1 2 ;f "4TE" T".""A POU:F, by:

1

{

Monitoring the indicated AFD for each OPERA 8LE excore channel:

a.

(

1.

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

{

and s

2.

At least once per hour f: tM fir;t '" Mur; ; fur r;;t:r'n;

}"

w.g the AFD Monitor Alara O ^^ "4",L: ;th;. Inor g ble.

N t iitoring and logging the indicated AXIAL FLUX 01FFER N r fe,, each i

0PIIXBt2 c -

i channel at least once per M"- TurTfirst 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and at least once

.. -3nu+ r t;,iTaiter, when the AXIAL FLUX

}

DIFFERENCE Mon 'a- "'..n a s i?,5 dei logged values of the 1

i i

L FLUX DIFFERENCE shall be ass interval crecedina each locoina.

t during the

,The indicated AFD shall be considered outside of its _ Z tra t had tiers i

when at least 2 OPERA 8LE excore channels are indicating the AFD to be outside in W l

tt tr ter-&pt Mad.

.:::?ty drei:ti:n :; tid: OftM O trgtn;d;MPR us ueg.

• M - : tf= M:f: Of; i

mesro

{

mva as a.

i nahs One minute penalty deviation for each one minute of POWER OPERA ide of the target band at THERMAL POWER levels equal t l

me above 505 o TED THERMAL POWER, and b.

One-half a penalty deviation for each one ute of POWER OPERATION outsibf the target band at L POWER levels between 155 and 505 of RATEDNRMAL POWER.

i N

j 4.2.1.3 The target flux difference o '

OPERABLE excore channel shall be I

i determined tey seasurement at leas e per Effective Full Power Days. Thel provisions of Specification are not applic h.

e i

4.2.1.4 The tar ux difference shall be updated at i

j 31 Effective F once per i

ower Days by either detemining the target difference l i

pursuant

.2.1.3 above or by linear interpolation between the i

meas recently value and o percent at the end of the cycle life.

The provis of j

ification 4.0.4 are not applicable.

1 FARLEY-UNIT 2 3/4 2-2 i

i, 4

Q RWLACE WoTd NEa) W<E

\\

/

kt l

!!1 RATED THEAMAL POWER. %

8 l

~

NI

1-iF i

i t.i:

.i j,i ogACCEPTAeLE

-l'

}

1. '

UNACCEPTAeLE' OP95AT10N OPERATION l

l l

(-ti.es l- -. t :. (it.se se

_.._. 1 !

I/

i i

k!

E._

i i

I i

/

\\

5 I

- le

_.. i. -

F/!

i i-

!\\

l I

i i

, / i..

J

!. 1 I

i g

i; i/

i.

\\

i t

/

I t.

i N

i L

!i

/!

i

\\

i se

'l

(' 1:

I

\\

'l t

Se

(-31.se l ACCEPTAett

! (31.8 8 i: }

.ll l

OPERATION j

j i

E.i

i l.

I

-~

i

.~

~

~

~,~. 3e E.!

c l'.

.h I

~ ~ T. - ' ::. a l

-I l

i 8

i j

s

~

I-I i

l, i

a i

se

" /

l.

t

'I l

I MP f:..

i

, l..

J J

..j.

J._,.

4e

-3e

-2e

-le e

te to se 4e se s PL 8)X DIPPERENCE (J H %

\\

Fipne 3.21 Axial Mux Difference Limits as a Function of Reasd Thermal Power FARLEY-UNIT 2 3/4 1-3

pugw FKa alt 3 :')

4 120

+

4

(-12,100)

(+9,100) 100 I

i

/

1

~

UNACCEPTABLE

(

OPERATION

\\

j UNACCEPTABLE

[

\\

OPERATION e

tu 3:

/

\\

o 80 1

n.

T L

i J

ACCEPTABLE

\\

a

{

/

OPERATION

{

E

/

\\

o 60 i

/

\\

1 x

j

\\

u.

(-30, 50) r

\\

(+24,50)

]

O 40 1

1 l

i

,a 20 i

a 4

L 60 50

-40

-30

-20

-10 0

10 20 30 40 50 60 l

AXIAL FLUX DIFFERENCE (DELTA I)%

Figure S 3 d'/

]

Axial Flux Difference Limits as a Function of Rated Thermal Power for RAOC f

I t

4 POWER DISTRISITfION LIMITS 3/4.2.2 BEAT FLUI BOT CHANNEL FACTOR - F;( Q LIMITING CONDITION FOR OPERATION 3.2.2 F,(Z) shall be limited by the following relationships:

F,(2) < [T) [K(2)j for P > 0.5 for VANTAGE 5 fuel l

~ 2.45 F,(2) ! [4.9) [K(Z)) for P f 0.5 for VANTAGE 5 fuel and l

F,(Z) < [T) [K(2)] for P > 0.5 for LOPAR fuel l

2.32

~

F,(Z) f [4.64) [K(2)) for P f 0.5 for LOPAR fuel l

vhere P = THERMAL F0VER 6

7/

and K(Z) is the function obtained from Figure 3.2-for a given core height location.

AFFLICABILITY: MODE 1 ACTION:

Vith F,(Z) exceeding its limits Reduce TEERMAL POWER at least 1% for each 1Z F (Z) exceeds the limit a.

within15minutesandsimilarlyreducethePovIrRangeNeutron Flux-Bigh 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 Overpoor delta T Trip Setpoints have been reduced at least 1% for each 1% F (Z) uceeds the limit. E ^:n;n n f:lt: T g

? i; 5:t--int ::f;: tier r 111 E W C rith '

::::: 1. :: 1--

509-48e1985, b.

'.f-tit -f:-et'e

! _n.u..

2....,e,f litit :: fiti:, ;-in :;,

a

.....n..

. ---u n mu..

.u....

aa.

.--..>.,u..

a-TIIRMAL POVER may.4hes> be increased provided F,(Z) is demonstrated through incore mapping to be within its limit FARLET - UNIT 2 3/4 2-4 AMENDNENT NO. 1J, H.

85

j j

l j

POWER DISTRIBUTION LIMITS i

3 SURVEILLANCE REQUIREMENTS 1

i 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

Fm(s) 4 W

j 4.2.2.2 g shall be evaluated to determine if p is within its limit by:

i Using the movable incere detectors to obtain a power distribution a.

sap at any THERMAL POWER greater than 5% of RATED THEPMAL POWER.

4 i

i Increasing the measured F component of the power distribution map j

by 35 to account for manuNcturing tolerances and furtt.tr increasin he value by 5% to account for measurement uncertainties.

c.

C ring the F,y computed (F ) obtained in b, above to:

f

/ME 1.

The{,y limits for RATED THERMAL POWER (F P) for th ppropriate i

measui%d core planes given in e and f below, and i

\\

l

/

)

2.

The relationship:

/

\\

F'

= F (1+0.2(1-P)]

j xy xy s

where F ' is the li'ait for fractiona i

"Y THERMAL POWER operation j

expressed as a function of F,RTP a P is the fraction of RATED l

y 1

i THERMAL POWER at which F3y,was

asured, i

x d.

Remeasuring F according to foH owing schedule:

i XY

/

j-

/

\\

i 1.

When F isgreaterNhantheFRTP

{

0 liett for the appropriate I

i xy xy g

esasured cove p ne but less than the F',S relationship, additional j

power distri, tion maps shall be taken andf C compared to F,RTP

{

[

y and F

's a) ither within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding by 2 'af RATED i

N C

THERMAL POWER or greater, the THERMAL POWER at wh h F,y was last detamined, or i

j b)

At least once per 31 EFP0, whichever occurs first.

\\

i l

FARLEY-UNIT 2 3/4 2-5 i

l i

J

turE47 - PME 3 H it - G Determining the computed heat flux hot channel factor F b.

C follows:

Q (g),,,

Increase the measured F (Z) obtained from the power oistribution g

mao by 3% to account for manufacturing tolerances and further increase the value oy 5% to account for measurement uncertaintias, C

c.

Verifying that Fg (Z), obtained in Specification 4.2.2.2b

above, satisfies the relationship in Specification 3.2.2.

d.

Satisfying the following relationship:

C Fg (g) g RTP, g(g)

P x W(Z)

C Fg (7) g RTP, g(g) for P

0. 5 0.5 x W(I)

~

C Where Fg (Z) is obtained in Specification 4.2.2.2b above, F RTP g

is the Fg limit, K(Z) is the normalized F (Z) as a function of core q

height, P is the fraction of RATED THERMAL POWER, and W(Z) is the cycle dependent function that accounts for power distribution transients encountered during normal operation.

Measuring F (Z) according to the following schedule:

e.

q 1.

Upon achieving equilibrius conditions after exceeding by 20% or more of RATED THERMAL POWER, the THERMAL POWER at which F (Z) was last detarsined*, or q

2.

At least once per 31 Effective Full Power Days, whichever occurs first.

• During power escalation after each fuel loading, power level say be increased until equilibrius conditions at any power level greater than or equal to 50%

of RATED THERMAL POWER have been achieved and a power distribution map obtained.

ppP

= 2.45 (Vantage 5 fuel)

= 2.32 (LOPAR fuel)

K(Z) provided in Figure 3.2-2 W(Z) provided in the Radial Peaking Factor Lirnit Report

, ~

.: s.

emu." r 4 m f.

With measurements indicating

/

maximum F, *(Z) \\

over 2

( K(Z) /

has increased since the previous determination of F *(Z) either of the following actions shall be taken:

1)

Increase F, 8(Z).by the r/ca.) P-a 74<*

/

specified in WAN pru,we, 7,uror A,m r de4aerand verify that this value satisfies the r'elationship in Specification 4.2.2.2d, or F*Z 2)

P,we(r ) Days until two successive maps indicate that o

F, c(Z)\\

is not increasing.

maximus over 2

\\ K(Z)

With the relationships specified in Specification 4.2.2.2d above not g.

being satisfied:

1)

Calculate the percent F,(Z) exceeds its limits by the following expression:

[maximusF[(2)rWII)

]

hx100forP>0.5 over Z. F,

• xK(Z)

-1

.r

[maximusF 't2) x utz)

]

l over Z F. "

x K(Z)

-1 x 100 for P s 0.5, and

.T 2)

The fellowing action shall be taken:

Within 15 minutes, control the AFD to within new AFD limits which are detemined by reducing the AFD limits specified in 4 w 1x 8,M * = 8 ww, exceeds its limits a;s delers.by 15 AFD for each percent F (Z) ined in Specification 4.2.2.2g.l.

Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, reset the AFD alam setpoints to these modified limits.

h.

The limits specified in Specification 4.2.2.2c are applicable in all core play regions, i.e., 0 - 1005, inclusive.

1.

The limits specified in Specifications 4.2.2.2d, 4.2.2.2f, and 4.2.2.2g above are not applicable in the following core plane regions:

1)

Lower core region from 0 to 155 inclusive.

t 2)

Upper core region from 85 to losz, inclusive.

a POVER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 2.

When the F

• is less than or equal to the FRTP limit for the appropriatEmeasuredcoreplane,addignalp5vegdistribution aps hall be taken and F,y' compared to F,y and F,,

at least ce per EFPD.

i e.

The F lim for RATED THERMAL POVER (FRTP) shall be vided for all coreplanesco aining bank "D" control rbds and all nrodded core planes in a Radi Peaking Factor Limit Report pe pecification MM 6.9.1.11.

w ars IUNi f.

The F,fegions as measured in limits of e, abov are not appl le in the following core plane ercent o core height from the bottom of l

the fuel:

f 1.

Lover core region from 0 15%

elusive.

2.

Upper core region fr 85 to 100%, inc ive.

3.

Grid plane re ns within + 2% of core heigh around the midpoint of the grid I

4.

Core e regions within + 2% of core height (+ 2.8 nches) about th demand position of the bank "D" control rods.

\\

g.

V F

• exceeding F ' the effects of F on F (Z) shall be valua 6d to determinF if F, (Z) is withiE its l}mits.

j nM MEnse. 'M6 ^^GM****** * *!'M'c"**A 4.2.2.3 Vhen F (Z) is seasured for%other th d: : M rr t z:. an overall measured F, (Z),shall be obtained from a power

' M ribution map and increased by i

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

I FARLEY-UNIT 2 3/4 2-6 AMENDMENT NO.

O, 74

i e

i POVER DISTRIBUTION LIMITS 3/4.2.3 NUCLEAR ENTHALPY HOT CHANNEL FACTOR - F",,

LIMITING CONDITION FOR OPERATION i.

F",, shall be limited by the following relationship:

3.2.3 1

i F",, f 1.65 [1 + 0.3 (1-P)] for VANTAGE 5 fuel and 1.30 l

F",, 3 +r n ll + 0.3 (1-P)] for LOPAR fuel 4

i l

where P = THERMAL POVER RAA w TutRMAL POWER j

APPLICABILITY: MODE 1 i

j ACTION:

]

T----

Vith P",, exceeding its limit:

i

\\

Reduce THERMAL POWER to less than 50% of RATED TIERMAL POVER within 2 a.

hours and reduce the Power Range Neutron Flux-Bigh Trip Setpoints 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 />, 3

b.

Demonstrate through in-core mapping that P",, THERMAL POWER t is within its limit within i

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 l

5% of RATED THERMAL POWER vithin the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and o less than i

)

c.

Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above the reduced limit required by g or b, above subsequent POWER OPERATION may proceed provided that F", is demonstrated through in-core marping to be within its limit at,a nominal

}

50% of RATED TIERMAL POWER prior to exceeding this THERMAL POWER, at a j

nominal 75% of RATED TRERMAL POVER prior to exceeding this THERMAL POVElt 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 POVER.

a l

J i

(

ij FARLEY - UNIT 2 3/4 2-8 AMENDMENT NO. 27, 37 85

.. ~

- -.... - - ~

Iae;e 3.a.1 max:KH ALLCWASLE POWER RANGE NEUTRON Ft'lX MIGM gg?pc;NT

NCPERASLE $7 TAP L

NE SAf t*Y val'.TS DUR:NG ) L0cp OpgAATION w;?g Maximum Number of Inoperable Safety Valves on Any Maximum Allowable Power Range 1

Cperating Steam Generator Neutron Flum High Setpoint (Per-_ent of RATED THERMAL PCWERJ_

1 ep. l C 2

-ve 43 3

-te.14 i

TABLE 3.7-2 max:KY ALLOWAELE_F0WER RANGE Nt*J"RCN FLUX WIGN S INOPERAALE f ? TAN ' :NE SAftTY VM'.18 DURING 2 LectlPfl%7 0N' Maximua Number tsf Inoperable safety valvas en Any Maximum Allowable Power Range Operatane Steam Generator Neutron Flum Migh Setpoint

.tfercent of DATED TMERMAL P0ertR 1

2 3

s

• At least two safety valves shall be OPERABLE en the non-operating steam generater.

"These values left bleak pending NRC approval of 2 leep operation.

/A/5ERJ4 FAALEY-UNIT 2 3/4 1-2 AMENDMENT NO.

103

I INSERT page 3/4 7-2 For plant operation approaching end of cycle (i.e., core average burnup a 14,000 MWD /MTUL with one inoperable safety valve on any steam generator, the maximum allowable Power Range Neutron Flux setpoint may be increased from 60% to 87% RTP.

4 l

l

3/t. 2 POVER DIS *RIBtT!!CN LIMITS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Freqaency) events by: (a) meeting the DN5 design criterion during normal operation and l

in short ters transients, and (b) limiting the fission gas release, fuel pellet temperature and cladding mechanical properties to within assumed design criteria.

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

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

F,(Z)

Heat Flux Rot 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 manufacturing tolerances on fuel pellets and rods and seasurement uncertainty.

F(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.

MJTS F"

ial Fe Factor, i efined as )ffe ratio of powerdejdity o averag power densi in the hor Wontal plane core elevation Z.

3/4.2.1 AIIAL FLUE DIFFERENCE The limits on AIIAL FLUX DIFFERENCE (AFD) assure that the F Z upper bound envelope of 2.45 for VANTAGE 5 and 2.32 for LOFAR times th$(no)rmalized axial peaking factor is not exceeded during either normal operation or in the event of xenon redistribution following power changes.

Target flux difference is determined at equilibrium xenon conditions.

full e a any be positioned within the core in accordance vi e

respective inse loits and should be inserted near the 1 position for steady state operat high power levels.

of the target flux DEJ.frg. difference obtained under these ions d the fraction of RATED THERMAL POWER is the target flux diff RATED TEIRMAL POVER for the associated core burnup conditi arget flux ces for other THERMAL F0VER levels are obtai

" multiplying the RATED value by the appropriate fr TEIRMAL POWER level. The periodic updat he target forence value is necessary to reflect core burnup erations.

FARLET - UNIT 2 5 3/4 2-1 AMENDMElff N0.

13. H. d5

j J

J i

POVER DISTRIBUTION LIMITS

\\

l BASES 1

l AXIAL FLUX DIFFERDCE (Continued) i though it is intended that the plant vill be operated with the A j

vithin t

.(5)% target band about i

plant THERMA the target flux difference du rapid deviate outside ofKtarget band at reduced THERMAL ROUER lev i

envelope of peaking factors' deviation vill not affec M he xenon redistribution s This DF4E7E. RATED THERMAL POWER (with the AFD. any be reachetf on a subsequent return to i

duration of the deviation is lia he target band) provided the time i

i limit cumulative during the

, M edingly, a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> penalty deviation vious 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />' Q rovided for operation outside i

of the target band bu in the limits of Figure %L2-1 1

POVER levels bet 50% and 90% of RATED THERMAL POVI1;') while at THERMAL levels betv or THERMAL POWER

{

5% and 50% of RATED THERMAL POVER, deviation outsi the target band are less significant.

the AFD 4

The penalty of al time reflects this reduced significance.

rs is s the plant process computer through the AFD Monitor Alars. Provis j

determines the one minute average of each of the OPERABLE excore detector The computer tours and provides an alara message immediately if the AFD for 2 or mo i

OPERABLE excore channels' p outside the

. n M ead the THERNAL POWER i re e

grea,tertha]n^^^";fRATEDTIERMALPOWER.

3 s

... i ;'.;; ; M..

,___u_..___

-_2 a-u...___

__2

==

.___m, m._.__.

m_

q 4

,,g 1

.>_._.,a n,

w

__.____-_--..._>___,__.a,... _-__.

u--,

3/4.2.2 and 3/4 2.3 e3PERAMG SPACK pag, ggec, opgumN L

spge.worow F4MALS2,I CHAmusL FACTOR HEAT FLUI 80T CEANNEL FACTOR. NUCLEAR DPfEALFY 50 The limits on heet flux hot channel factor, and nuclear enthalpy rise hot channel factor ensure that 1) the design limit on peak local power density is i

not exceeded, 2) the DNP design criterion is set, and 3) in the event of a LOCA the peak fuel clad temperature vill not exceed the 2200'F ECCS acceptance criteria limit.

Each of these is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3.

surveillance is sufficient to insure that the limits are maintained provided:

This periodic l

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

insertion differing by more than

• 12 steps, indicated, from the group dessad position.

b.

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

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

are maintained.

d.

The axial power' distribution, expressed in teras of AZIAL FLUI DIFFERDCE, is maintained within the limits.

FARLET - UNIT 2 8 3/4 2-2 AMENDNINT NO.

85

am m2-.s ma%ma s

---

w---.-

_a_4aa

_,.,a4wum,i.

.__e

1. m u-s4adaienams,h

.4 e.

.s.mme.

8.-hee._m.J

-med.J---L

_eAa-,,.._.Ja_ _ _ _ - - -

.m

,_F DELET hb 5%

8%

a_,I,

\\

W r.-. _.. - - _ _

~ ~~- ~ ~~ r =

v b=

,r I

I

.--.___..__;^=__.__=._:--_-._-

-==.

bh ((9_33

.:= = = = g:E=_

= ;=-

g_== g :g =.=:== _ -,

^ ::- C:

=

-~~

.2

--- : := = -

K V *a b = ;==i5--===j=== r' ~C= 2====- W_' ~'Lrm_':== ~ :

=

=== --

2W==g=i~=~

1 -..

~W- %=2

. -f = - _T 2

=1~*~^

_ g. ;_ _~^. = _ T _ ;;i =_:;: G G ^~ E 55M -5= -n M:_ lS LX 72X ~

_ -

= = _ - = _. _

= =3-

x5==g==

r==~ =

;

::. 3-q g.==3=35 = 53 3-555 =;_ ;

y.

-; _ - _; _ =

~:E:@5 55MT-5 5 3 =-- 3 q :

y-

_3-l E'? "&_ li 155 k55 ? & i&S?lM'I iEEE552 'N =lk"~'~2~~

=

~51lW [j

-;RMg-Q;f=JE EE? Es&W E =:f~; W: t[ &

c yps ;f : gpWw : z.: - r := 7-+== +== w +E gg m ;;_;= g-x-g]m 3_ y =x:

Q:--.- : =gg::. :

-=

=

W =EHE W F # M --" T* b M M C N N ii g g g' ;=3 EEEEE# M E 5 = Fi M 5N@==+== % MM M55 E

- *: d9-73 h m.. # s M t = #, Q-u e & = E = - -

gr

-. 2-

---

2

^

--. _ R +

_~z 1 hfi[

Q -2 :l_:l.

3 3:;;7 e_p=

. ; : ::. ~

y =p=== g _ q p;;&

= casu_=e

g ;= 7:; 73 3;- 7 5 3: g;;;5 ; :;:-: a :1 E t :. E.EE E E E + = E = = = - 7

3.

- 3 x._ _ - -

x --

3 g:p - g - ;g E.

=

d p __ g ; -- y -

. ry7 -7 2

.-, q -

r 40s 1

.(

I

^

20% i

[

' ~ ~ _

I.5 $

.,_ y

=w _ _

~

' 5 '. g.g[:_

l g

.gg

.33

. ten s

• 10%

'NE M

18ectCATIO AMIAL PLUX OlPPER$NC8 Pigwe 3 3/4 31 TYPtCAL INDICATED AXIAL PLUX OIPPERENC8 VERSUS TwannsAL PowtR FARLEY-UNIT 2 R 3/4 2-3

POSER D2STR25UT20N L2M2Ts l

BASES F" H vill be maintained within its limits providgd conditions a.

I through'd. above are maintained.

The relaxation of F I

THERNAL POVER allows changes in the radial power shape,8 as a function of rod insertion limits.

for all permissible

\\.

Vhen an F sessurement errorandmanulacturingtolerancemustbemade.is taken, an allowance for both experime An allowance of 5% is appropriate for a full core map taken with the incere detector flux sapping system and a 3% allowance is appropriate for manufacturing tolerance.

j INSEAT4 Vhen F" the appropria,H is measured, experimental error must be allowed for and 4% is te allowance for a full core aug taken with the incere detection system.

The specified limit for F 5 contains an 8% allowance for i

uncertainties.

The 8% allowance is based on he following considerations:

l t

Abnormal perturbations in the radial power shape, such as from rod

\\

a.

j misalignment, affect F",8 acre directly than F,,

4 b.

Although rod movement has a direct influence upon limiting F to vjthinitslimit, suchcontrolisnotcoadilyavailabletol}ait F,H, and Errors in prediction for control power shape detected during startup c.

physics tests can be compensated for i DF by restricting axial flux distribution.

This compensation for T,5,is less readily available.

i 4

2 i

I i

i l

FARLEY - WIT 2 B 3/4 2-4 AMENDMENT N0.

57, 85

INSERT page B 3/4 2-4 The heat flux hot channel factor F (Z) is measured periodically and increased by a o

cycle and height dependent power factor appropriate to RAOC operation, W(Z), to provide assurance that the limit on the heat flux hot channel factor Fo(Z) is met.

W(Z) accounts for the effects of normal operational transients within the AFDber+k.;fy and was determined from expected power control maneuvers over the full range of l

burnup conditions in the core.

l l

l l

l l

l 1

l l

l l

l l

POWER DISTRIBUT!ON L:MITS BASES b.9 nanking factor F,E(l factor. F,(Z).

Z). is measured periodically to provida -m ; wC assurance that thhv i 9-an remain = M ait. The F in the Radial Peaking Factoc '#

lb E LET E -- h>

limit for RATED THERMAL POVER e

limit report on 6.

1.11 was determ

-- svoect.d cover aneuvers over the full range of burnup conditions in the h--

3/4.2.4 OUADRANT POVER TILT RATIO The quadrant power tilt ratio limit assures that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during startup testing and periodically during power operation.

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

The two hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned control red. In the event such action does not correct the tilt, the margin for uncertainty on F is reinstated by reducing the maximum allowed power by 3 percent foreachperce$t of tilt in excess of 1.0, For purposes of monitoring QUADRANT POVER TILT RATIO when one excore detector is inoperable, the movable incere detectors are used to confirm that the neraalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO.

The incore detector monitoring is done with a full incore flux sep or two sets of four symmetric thimbles. The two sets of four symmetric thimbles is a unique set of eight detector locations. These locations are C-8, E-5. E-11, 6-3. H-13.

L-5. L-11. and N-8.

3/4.2.5 DNB PARANETERS The limits on the DNS 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 meet the DN8 design criterion throughout each analysed transient. The indicated T value of 580.7'F is based on the average of two control board readings an8,In indication uncertainty of 2.5'F.

The indicated pressure value of 2205 psig is based on the average of two control board readings and an indication uncertainty of 20 psi. The indicated total RCS flow rate is based on one elbow tap sessurement from each loop and an uncertainty of 2.4% flow (0.1% flow is included for feedvater venturi fouling).

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> surveillance of Tavs and pressurizer pressure through the control board readings see sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

The 18 month surveillance of the total RCS flow rate is a precision measurement that verifies the RCS flow requirement at the beginning of each fuel cycle and ensures correlation of the flow indication channels with the seasured loop flows. The monthly surveillance of the total RCS flow rate is a reverification of the RCS flow requirement using loop elbow tap measurements that are correlated to the precision RCS flow seasurement at the beginning of the fuel cycle. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> RCS flow surveillance is a qualitative verification of significant flow degradation using the control board indicators and the loop elbow tap sessurements that are correlated to the precision RCS flow seasurement at the beginning of each fuel cycle.

FARLEY - UNIT 2 3 3/4 2-5 AMENDMENT NO.

52. 35 v

1:

a l

i j

ADMINISTRATIVE CONTROLS k

)

i i

I 1

j 4

MONTHtY OPERATINC Rfp0RT 4.g.1.10 including documentation of all challenges to the be submitted on a monthly basis to the Coenission, pursuant ves, shall later than the 15th of each month following the calendar m

., no report.

e i

i i

i j

.V n in PtArime racTOs LIufY Rf;GET l

b 1

The Fxg limit for Rated Thermal Power (Fry ) for all core el RTP l

contain tank. D' centrol reds and all unredded cbre planes j

, established and d6cumented in the Radial Peaking Facter.Listi Repor i

i ~each reload cycle prior to 180L1) and provide 64e-Die Ceanission, p

' to 10 CFR 50.4. upo(n issuance.

g,vg l at sees other time during core Itft'~1t will~ Der subettte In the'ewest'that the limit would be substtted

,gsar i

jotherwiseexemptedby,theCanaission.

1 i

Any inf RTp en needed to support Fay will be by request free the Nhc lasladed la this report.

l i

r m> Bfns E-=&i6m attIAa!LITY DATA ef;6ei 6.9.1.11 The number of tests Kvalid er tava114) and the numb i

start en demand for each disse generator shall be submitted to the IRC annually. This report shall contata the infeenetten identified in Regulatory

{

Positten C.3.b of IRC Regulatory Guide 1.100. Revision 1. 1977.

I fag (gy. WIT 2 41g m m.gg,s u s,85, 4

INSERT page 6-19 6.9.1.11 The cycle dependent function, W(Z), and the burnup dependent Fo (Z)

C Penalty Factors, required for calculation of Fo (Z) specified in LCO).2.2, " Heat C

Flux Hot Channel Factor - Fo(Z)", shall be documented in thew Peaking Factor Limit Report in accordance with the methodology in WCAP-10216-P-A, " Relaxation of Constant Axial Offset Control FQ Surveillance Technical Specification", Rev.1, February 1994 (W Proprietary).

/

TheWPeaking Factor Limit Report shall be provided to the Commission, pursuant to 10 CFR 50.4, upon issuance prior to each reload cycle (prior to MODE 2). In the event that the limit would be submitted at some other time during core life, it will be submitted upon issuance, unless otherwise exempted by the Commission.

INDEX ADMINISTRATIVE CONTROLS SECTION g

Review...................................................

6-10 Audits...................................................

6-11 Authority................................................

6-12 Records..................................................

6-12 6.5.3 TECHNICAL REVIEW AND CONTROL Activities...............................................

6-12 Records..................................................

6-13 6.6 REPORTAELE EVENT ACTION.....................................

6-14

..................................... 6-14 6.7 SAFETY LIMIT VIOLATION 6.8 PROCEDURES AND PROGRAMS.....................................

6-14 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS

.......................................... 6-15a Startup Report Annual Report............................................

6-16 Annual Radiological Environmental Operating Report.......

6-17 Annual Radioactive Effluent Release Report...............

6-17 Monthly Operating Report.................................

6-19 Peaking Factor Limit Report..............................

6-19 l

Annual Diesel Generator Reliability Data Report..........

6-19 Annual Reactor Coolant System Specific Activity Report... 6-20 Annual sealed source Leakage Report......................

6-20 6.9.2 SPECIAL REPORTS...........................................

6-20 6.10 RECORD RETENTION............................................ 6-20 6.11 RADIATION PROTECTION PROGRAM................................ 6-21a 6.12 HIGH RADIATION AREA........................................

6-22 FARLEY-UNIT 2 XIX AMENDMENT NO.

680

~

UNACCEPTABLE 660 2440 psia OPERATION 2250 psia 640 2000 psia

_b cn620

~

1840 psia 600

~

ACCEPTABLE

~

OPERATION 580 560 O.0 0.2 0.4 0.6 0.8 1.0 1.2 POWER (FRACTION OF RATED THERMAL POWER)

Figure 2.1-1 Reactor Core Safety Limits Three Loops in Operation FARLEY-UNIT 2 2-2 AMENDMENT NO.

I

l

~

TABLE 2 8-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION N

Note 1: Overtemperature AT e

3

~

r 3

(

3 r+rs 1

1+rs 1

w T

- T' + K (P-P')- f,(AI) i H

AT s AT, K, - K 3

2 (1 + r S;

( I + T S>

I + I S; u

2 A

6 3

Measured AT by RTD instrumentation; where AT

=

Indicated AT at RATED THERMAL POWER and reference Tavgi AT

=

O Average temperature, "F; T

=

Reference T,yg at RATED THERMAL POWER (s 577.2"F);

T

=

Pressurizer pressure, psig; P

=

2235 psig (nominal RCS operating pressure);

P'

=

1+t"i The function generated by the lead-lag controller for T,yg dynamic compensation; u

=

3#7, a

t1 &T2 Time constants utilized in the lead-lag controller for T,yg, r1 = 30 see,

=

T2 = 4 sec; 1+Ts4 The function generated by the lead-lag controller for AT dynamic compensation;

=

1+,,

Time constants utilized in the lead-lag controller for AT, t4 = 0 sec, t5 s 6 see; 4 &T5 T

=

1+T, Lag compensator on measured T,yg; 6

6 Time constant utilized in the measured T,yg lag compensator, T6 s 6 see; l

T

=

-1; Laplace transform operator, sec s

=

ag Operation with 3 loops operation with 2 loops Ki = 1.17; Ky = (values blank pending l

zO K2 = 0.017; K2 = NRC approval of l

K3 = 0.000825; K3 = 2 loop operation) l

~

TABLE 2.2-1 (ContinuQd)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS aus NOTATION fContinued)

E and fy (AI) is a function of the indicated difference between top and bottom detectors of the power-range 3

nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup d

tests such thats u

(1) for qg - gb between -23 percent and +15 percent, ft (AI) =0 (where gt and qb are percent RATED l

THERMAL POWER in the top and bottom halves of the core respectively, and qt + 9b is total THERMAL POWER in percent of RATED THERMAL POWER);

(ii) for each percent that the magnitude of (qt - 9b) exceeds -23 percent, the AT trip setpoint shall be automatically redaced by 2.48 percent of its value at RATED THERMAL POWER; and (iii) for each percent that the magnitude of (qt 9b) exceeds +15 percent, the AT trip setpoint shall be automatically reduced by 2.05 percent of its value at RATED THERMAL POWER.

Note 2:

Overpower AT u

E

' l + r*s' r's l

1 AT s; A T,

K, - K T

- K. < T

- T"

- f (AI) 3 2

(1 + r s; (1 + r s; (1 + r s; (1 + r ss 3

3 o

where: AT = Measured AT by RTD instrumentation; Indicated AT at RATED THERMAL POWER and reference Tavg; l

ATo

=

T = Average temperature, 'F; Reference T at RATED THERMAL POWER (s 577.2*F);

l l'

=

ayg K4

= 1.10; l

5 K5 0.02/*F for increasing average temperature and O for decreasing average g

=

g temperature; N

K6

= 0.OO109/*F for T > T",

K6 = 0 for T 5 T;

l ta The function generated by the rate lag controller for T,yg dynamic compensation;

=

y,

TABLE R.2-1 (ContinuQd) 9 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION (Continued) k w

d 3

= Time constant utilized in the rate lag controller for Tavg* T3 = 10 see; 1

u 1+r s 4

I + r5s The function generated by the lead-lag controller for AT dynamic compensation;

=

Time constants utilized in the lead-lag controller for AT, T4 = 0 sec, T5 s 6 see; l

4 & v5 T

=

Lag compensator on measured T,,g;

=

6 = Time constant utilized in the measured T,yg lag compensator, T6 s 6 see; l

T

-1; w

s

= Laplace transform operator, sec o

f (AI)

= 0 for all AI.

2 Note 3: The channel's maximum trip point shall not exceed its computed trip point by more than O.4 percent AT span.

Note 4: Pressure value to be determined during initial startup testing.

Pressure value of 5 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 O.4 percent AT span.

I E

5 8

I SAFETY LIMITS BASES The curves of Figures 2.1-1 and 2.1-2 are based on the most limiting result using an enthalpy hot channel factor, F$H, of 1.65 for VANTAGE 5fuelandanF$Hof1.30forLOPARfuelandareferencecosinewitha peak of 1.55 for axial power shape.

An allowance is included for an increase inF$Hatreducedpowerbasedontheexpression:

F$H=1.65 [1 + 0.3 (1 - P)) for VANTAGE 5 fuel and F$H=1.30[1+0.3(1-P)] for LOPAR fuel 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 1 (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 delta T 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, pressurizer and the reactor coolant system piping and fittings 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 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 peig, 125% of design pressure, to demonstrate integrity prior to initial operation.

FARLEY-UNIT 2 B 2-2 AMENDMENT NO.

LIMITING SAFETY SYSTEM SETTINGS BASES overoower AT The overpower delta T reactor trip provides assurance of fuel integrity (e.g., no fuel pellet melting) under all possible overpower conditions, 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 distribution, changes in density and heat capacity of water with temperature, and dynamic compensation for transport, thermowell, and RTD response time delays from the core to RTD output indication.

l Pressuriter Pressure The Pressurizer High and Low Pressure trips are provided to limit the pressure range in which reactor operation is permitted. The High Pressure trip is backed up by the pressuriser code safety valves for RCS overpressure protection, and is therefore set lower than the set pressure for these valves (2485 psig).

The Low Pressure trip provides protection by tripping the reactor in the event of a loss of reactor coolant pressure.

gressurizer Water Level The Pressurizer High Water Level trip ensures protection against Reactor coolant System overpressurization by limiting the water level to a volume sufficient to retain a steam bubble and prevent water relief through the pressurizer safety valves. No credit was taken for operation of this trip in the accident analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System.

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

Above 10 percent of RATED THERMAL POWER, an automatic reactor trip will occur if the flow in any two loops drops below 90% of nominal full loop flow.

Above 36% (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 FARLEY-UNIT 2 8 2-5 AMENDMENT NO.

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE (AFD)

LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the limits specified in Figure 3.2-1.*

APPLICABILITY:

MODE 1 above 50% of RATED THERMAL POWER **

l ACTION:

With the indicated AXIAL FLUX DIFFERENCE outside of the limits a.

specified in Figure 3.2-1 1.

Either restore the indicated AFD to within the limits within 15 minutes, or 2.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes.

SURVEILLANCE REQUIREMENTS 4.2.1 The indicated AXIAL FLUX DIFFERENCE shall be determined to be within its limits by:

l Monitoring the indicated AFD for each OPERABLE excore channel:

a.

1.

At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and 2.

At least once per hour with the AFD Monitor Alarm inoperable. l l

• The indicated AFD shall be considered outside of its limits when at least 2 OPERABLE excore channels are indicating the AFD to be outside its limits.

• • see special Test Exception 3.10.2.

FARLEY-UNIT 2 3/4 2-1 AMENDMENT NO.

1

i l

l 120 l

l

\\

I 100 1

(-12,100)

(+9,100) l Unacceptable Unacceptable Operation Operation l

80 l

k(

Acceptable Q.

Operaton l

J 60

$ew h

(-30, 50)

(+24, 50)

O W

40 u.

O 20 l

0

-50

-40

-30

-20

-10 0

10 20 30 40 50 Axial Flux Difference (Delta l)%

an.3 FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER FOR RAOC FARLEY-UNIT 2 3/4 2-2 AMENDMENT NO.

I l

1 POWER DISTRIBUTION LIMITS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F g LIMITING CONDITION FOR OPERATION 3.2.2 Fg(Z) shall be limited by the following relationships:

Fg(Z) s [2211] (K(Z)) for P > 0.5 for VANTAGE 5 fuel P

Fg(Z) s [4.9) [K(Z)] for P S 0.5 for VANTAGE 5 fuel and Fg(Z) S [2212) [K(Z)) for P > 0.5 for LOPAR fuel P

Fg(Z) s (4.64) (K(Z)) for P 5 0.5 for LOPAR fuel 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 ACTION:

With Fg(Z) exceeding its limits Reduce THERMAL POWER at least 1% for each 1% Fg(Z) exceeds the a.

limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days <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 delta T Trip Setpoints have been reduced at least 1% for each 1% Fg(Z) exceeds the limit.

l b.

THERMAL POWER may be increased provided Fg(Z) is demonstrated through incore mapping to be within its limit.

SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of specification 4.0.4 are not applicable.

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

l Using the movable incore detectors to obtain a power distribution a.

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

FARLEY-UNIT 2 3/4 2-3 AMENDMENT No.

POWER DISTRIBUTION LIMITS l

SURVEILLANCE REQUIREMENTS (Continued) b.

Determining the computed heat flux hot channel factor Fg (Z), as follows:

l Increase the measured Fg(Z) obtained from the power distribution map by 3% to account for manufacturing tolerances and further increase the value by 5% to account for measurement uncertainties, C

c.

Verifying that Fg (Z), obtained in Specification 4.2.2.2b above, satisfies the relationship in Specification 3.2.2.

d.

Satisfying the following relationship:

F "" x K (Z )

C Fq (Z ) s fo r P > 0.5 P x W (Z )

F "" x K (Z )

C Fn (Z ) s for P s 0.5 0.5 x W (Z)

E Where Fg (Z) is obtained in Specification 4.2.2.2b above, Fg is the Fg limit, K(Z) is the normalized Fg(Z) as a function of core height, P is the fraction of RATED THERMAL POWER, and W(Z) is the cycle dependent function that accounts for power distribution transients encountered during normal operation.

TP Fg

= 2.45 (VANTAGE 5 fuel)

= 2.32 (LOPAR fuel)

K(Z) provided in Figure 3.2-2 W(Z) provided in the Radial Peaking Factor Limit Report Measuring Fg(Z) according to the following schedules e.

1.

Upon achieving equilibrium conditions after exceeding by 20%

or more of RATED THERMAL POWER, the THERMAL POWER at which Fg(Z) was last determined *, or 2.

At least once per 31 Effective Full Power Days, whichever l

occurs first.

l

• During power escalation after each fuel loading, power level may be i

increased until equilibrium conditions at any power level greater than or equal to 50% of RATED THERMAL POWER have been achieved and a power distribution map obtained.

FARLEY-UNIT 2 3/4 2-4 AMENDMENT NO.

l

M R DISTRIBUTTON 2,IMTTS SURVEILLANCE REQUIREMENTS (Contint4ed )

f.

With measurements indicating

' F C(Z)'

n maximum over (Z)

K(Z)j s

has increased since the previous determination of Fg (Z) either of the following actionts shall be taken:

C 1)

Increase Ig (Z) by the Fg (Z) penalty factor specified in the Peaking Factor Limit Report and verify that this value satisfies the relationship in Specification 4.2.2.2d, or C

2)

Fg (Z) shall be measured at least once per 7 Effective Full Power Days until two successive maps indicate that

' F C(Z)'

n ma).imum is not increasing.

K(Z)j over (Z)

(

g.

With the rolationships specified in specification 4.2.2.2d above not being satisfied:

1)

Calculate the percent Fg(Z) exceeds its limits by the following expression:

C Fn (Z) x W (Z) m axim um

- 1 x 100 fo r P > 0.5 over Z F,1, n

(

P C

Fn (Z) x W (Z) m axim u:n

- 1 x 100 fo r P s 0.5, and over Z F g 7, n

. 0.5

.)

2)

The following action shall be taken:

Within 15 minutes, control the AFD to within new AFD limits which are determined by reducing the AFD limits specified in LCo 3.2.1, Axial Flux Difference, by 1% AFD for each percent Fg(Z) exceeds its limits as determined in Specification 4.2.2.2g.1.

Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, reset the AFD alarm setpoints to these modified limits.

FARLEY-UNIT 2 3/4 2-5 AMENDMENT No.

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) h.

The limits specified in Specification 4.2.2.2c are applicable in all core plane regions, i.e.,

0 - 100%, inclusive.

1.

The limits specified in Specifications 4.2.2.2d, 4.2.2.2f, and 4.2.2.2g above are not applicable in the following core plane regions:

1)

Lower core region from 0 to 15%, inclusive.

2)

Upper core region from 85 to 100%, inclusive.

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

FARLEY-UNIT 2 3/4 2-6 AMENDMENT NO.

i i

l l

POWER DISTRIBUTION LIMITS l

3/4.2.3 NUCLEARENTHALPYHOTCHANNELFACTOR-F{H LIMITING CONDITION FOR OPERATION 3.2.3 F$Hehallbelimitedbythefollowingrelationships F$Hs1.65(1+0.3(1-P)} for VANTAGE 5 fuel and F$Hs1.30(1+0.3(1-P)) for LOPAR fuel l

where P =

RATi.D THERMAL POWER APPLICABILITY:

MOL E i ACTION:

WithF$Hexceedingitslimits l

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER a.

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 j

Setpoints to 5 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-coremappingthatF$H is within its limit j

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 />, l

l and c.

Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above the reduced limit required by a or b, above; subsequent POWER OPERATION may proceed provided that F$H 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 l

attaining 95% or greater RATED THERMAL POWER.

1 l

l l

FARLEY-UNIT 2 3/4 2-8 AMENDMENT NO.

....~

TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING 3 LOOP OPERATION Maximum Number of Inoperable Maximum Allowable Power Range Safety Valves on Any Neutron Flux High Setpoint Operatino Steam Generator (Percent of RATED THERMAL POWER) 1 60***

l 2

43 l

3 24 l

j TABLE 3.7-2 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING 2 LOOP OPERATION Maximum Number of Inoperable Maximum Allowable Power Range safety Valves on Any Neutron Flux High Setpoint Operatino Steam Generator

• iPercent of RATED THERMAL POWER) 1 2

3

• At least two safety valves shall be OPERABLE on the non-operating steam generator.

• • These values left blank pending NRC approval of 2 loop operation.

• • • For plant operation approaching end of cycle (i.e., core average burnup 2 14,000 mwd /MTU), with one inoperable safety valve on any steam generator, the maximum allowable Power Range Neutron Flux setpoint may be increased from 60% to 87% RTP.

FARLEY-UNIT 2 3/4 7-2 AMENDMENT NO.

l o

l 3/4.2 POWER DISTRIBUTION Z,IMITS BASES t

The specifications of this section provide assurance o'i fuel integrity during condition I (Normal Operation) and II (Incidents of hoderate Frequency) events by:

(a) meeting the DNB design criterior. during normal operation and in short term transients, and (b) limiting the U::sion gas release, fuel pellet temperature and 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 3

conditions assumed for the LOCA analyses are met and the ECCS acceptance i

criteria limit of 2200*F is not exceeded.

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

these specifications are as follows:

j I

Fg(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 l

l divided by the average fuel rod heat flux, allowing for

)

i manufacturing tolerances on fuel pellets and rods and measurement uncertainty.

F$

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.

l l

3/4.2.1 AXIAL FLUX DIFFERENCE The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the Fg(Z) upper 4

bound envelope of 2.45 for VANTAGE 5 and 2.32 for LOPAR times the normalized axial peaking factor is not exceeded during either normal operation or in i

the event of xenon redistribution following power changes.

l Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer determines the one minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for 2 or more OPERABLE excore channels is outside the allowed AI operating space for RAOC operation specified in Figure 3.2-1 and the THERMAL POWER is greater than 50% RATED THERMAL POWER.

FARLEY-UNIT 2 B 3/4 2-1 AMENDMENT NO.

e POWER DISTRIBUTION LIMITS BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR. NUCLEAR ENTHALPY HOT CHANNEL FACTOR The limits on heat flux hot channel factor, and nuclear enthalpy rise hot channel factor ensure that 1) the design limit on peak local power density is not exceeded, 2) the DNB design criterion is met, and 3) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200*F ECCS acceptance criteria ilmit.

Each of these is 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 limits are maintained provided:

Control rods in a single group move together with no individual a.

rod insertion differing by more than i 12 steps, indicated, from the group demand position.

b.

Control rod banks are sequenced nith overlapping groups as described in Specification 3.1.3.6.

The control rod insertion limits of Specifications 3.1.3.5 and c.

3.1.3.6 are maintained.

d.

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

F%gwillbemaintainedwithinitslimitsprovidedconditionsa,through

d. above are maintained.

The relaxation of F[g as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion Aimits.

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

The heat flux hot channel factor Fg(Z) is measured periodically and increased by a cycle and height dependent power factor appropriate to RAOC operation, W(2), to provide assurance that the limit on the heat flux hot channel factor Fg(Z) is met.

W(2) accounts for the effects of normal operational transients within the AFD limits and was determined from expected power control maneuvers over the full range of burnup conditions in the core.

FARLEY-UNIT 2 8 3/4 2-2 AMENDMENT NO.

PAGE INTENTIONALLY LEFT BLANK FARLEY-UNIT 2 B 3/4 2_3 M NDHENT NO.

o POWER DISTRIBUTION LIMITS BASES l

WhenF%gismeasured,experimentalerrormustbeallowedforand4% is the appropriate allowance for a full core map taken with the incore detection system.

The specified limit forF%gcontainsan8% allowance for uncertainties.

The 8% allowance is based on the following considerations:

a.

Abnormal perturbations in the radial power shape, such as from rod misalignment, affectF%gmoredirectlythanFg, b.

Although rod movement has a direct influence upon limiting Fg to within its limit, suchcontrolisnotreadilyavailabletolimitF[g,and Errors in prediction for control power shape detected during startup c.

physics tests can be compensated for in Fg by restricting axial flux distribution.

ThiscompensationforF$gislessreadilyavailable.

l l

k 1

i FARLEY-UNIT 2 B 3/4 2-4 AFLNDHENT NO.

POWER DISTPIBUTION LIMITS BASES 3/4.2.4 OUADRANT POWER TILT RATIO The quadrant power tilt ratio limit assures that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during startup testing and periodically during power operation.

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

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

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

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

The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles. The two sets of four symmetric thimbles is a unique set of eight detector locations.

These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, and N-8.

3/4.2.5 DNB PARAMETERS l

The limits on the DNB related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in the translent and accident analyses.

The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to meet the DNB design criterion throughout each analyzed transient. The indicated T value of 580.7'F is based on the average of two control board readings andyhn indication uncertainty of 2.5'F.

The indicated pressure valuw of 2205 psig is based on the average of two control board readings and an indication uncertainty of 20 psi.

The indicated total RCS flow rate is based on one elbow tap measurement from each loop and an uncertainty of 2.4% flow (0.1% flow is included for feedwater venturi fouling).

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> surveillance of T and pressurizer pressure through the control ay board readings are sufficient t$ ensure that the parameters are restored within their limits following load changes and other expected transient operation.

The 18 month surveillance of the total RCS flow rate is a precision measurement that verifies the RCS flow requirement at the beginning of each fuel cycle and ensures correlation of the flow indication channals with tha man =wed loop flows. The mmthly suzwillanos of the total X:s flow rate is a reverification of the RCS flow requirement using loop elbow tap measurements that are correlated to the precision RCS flow measurement at the beginning of the fuel cycle. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> RCS flow surveillance is a qualitative verification of significant flow degradation using the control board indicators and the loop elbow tap measurements that are correlated to the precision RCS flow measurement at the beginning of each fuel cycle.

FARLEY-UMIT 2 B 3/4 2-5 AMENDMENT NO.

j ADMINISTRATIVE CONTROLS 4

MONTHLY OPERATING REPORT 6.9.1.10 Routine reports of opterating statistics and shutdown experience, including documentation of all challenges to the PORV's or safety valves, shall be submitted on a monthly basis co the Connaission, pursuant to 10 CFR 50.4, no later than thr 15th of each month following the calendar month e

covered by the report.

I PEAKING FACTOR LIMIT REPORT l

6.9.1.11 The cycle dependent function, W(Z), and the burnup dependent Fg (3) penalty factors, required for calculation of Fg (E) specified in LCO 3.2.2, " Heat Flux Hot Channel Factor - Fg(Z)," shall be documented in the Peaking Factor Limit Report in accordance with the methodology in WCAP-10216-P-A, " Relaxation of Constant Axial Offset Control-FQ Surveillance Technical Specification," Rev. 1, February 1994 (H Proprietary).

The Peaking Factor Limit Report shall be provided to the Commission, pursuant to 10 CFR 50.4, upon issuance prior to each reload cycle (prior to MODE 2).

In the event that the limit would be submitted at some other time during core life, it will be submitted upon issuance, unless otherwise exempted by the Commission, l

ANNUAL DIESEL GENERATOR RELIABILITY DATA REPORT 6.9.1.12 The number of tests (valid or invalid) and the number of failures to start on dennand for sach diesel generator shall be submitted to the NRC annually.

This report shall contain the information identified in Regulatory Position C.3.b of NRC Regulatory Guide 1.108, Revision 1, 1977.

1 FARLEY-UNIT 2 6-19 AMENDMENT NO.

ATTACHMENT II Westinghouse letter CAW-96-968, dated May 23,1996, " Application For Withholding Proprietary Information From Public Disclosure," with the following enclosures: Affidavit, Proprietary Information Notice, and Copyright Notice.

Westinghouse Report NSD-NT-OPL-96-152, Revision 2, (Proprietary Class 2C), " Joseph M. Farley Nuclear Plant Units 1 & 2 Licensing Report for Technical Specification Changes Associated With Revised Core Limits, Revised OTAT/OPAT Trip Setpoints and Inclusion ofRAOC Control Strategy."

Westinghouse Report NSD-NT-OPL-96-158, Revision 2, (Proprietary Class 3), " Joseph M. Farley Nuclear Plant Units 1 & 2 Licensing Report for Technical Specification Changes Associated With Revised Core Limits, Revised OTAT/OPAT Trip Setpoints and Inclusion ofRAOC Control Strategy."

4