Text

Proposed Tech Specs Deleting Sections 2.1.1 Re Reactor Core, 2.1.2 Re RCS pressure,2.2.1 Re Reactor Trip Sys Instrumentation Setpoints & Table 2.2-1b Re Reactor Trip Sys Instrumentation Trip Setpoints
	ML20086R558
Person / Time
Site:	Mcguire
Issue date:	12/18/1991
From:	; DUKE POWER CO.
To:	;
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i Attachment I Technical Specification Changes Introduction This Attachment 3s divided into three parts.

Attachment Ia contains the Technical Specification (TS) markups, Attachment Ib contains the Technical Justification, and Attachment Ic contains a No Significant flazards Analysis for the changes, is.

g ku fDR f5DOCV O'_m..r' g _: 70 e

pr.re

s Attachment Ia Technical specification Markups

--_-__________________-__-______-____________-______--___-_____-_____________-_a

l l

I INDEJ SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY LIM QS wo:! '.k 2.1.1 REACTOR CORE............

2-[1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE.......

2-f1 FIGURE 2.1-1kUNITS 1 and 2 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION......

2-/2 l

J FIGURE 2.1-2 (BLANK)..

2-3 4:

ca um 5 2.1.

REACTOR CORE.......

2-0 2.1.2 CTOR COOLANT SYSTEM r' SSUPE...........

2-B1 l

f FIGURE 2.1. UNITS 1 and 2 REACT 0 mRE SAFETY LIMIT - F0 s OOPS

\\

' OPERATION...

2-B2 l

FIGURE 2.1-2 ( B L A. ' )......

2-3 l

LIMITINGkAFETYSYSTEMSEThrCGS(UNIT 2)

P. 2 i 2.1 REACTOR Tk - SYSTEM INSTRUME TION SETPOINTS.............

2-B4 k

REACT 0kTRIPSYSTEMINSTR%ENTATIONTRIPSET TA C 2.2-lb INTS....

-B5 l

L 2.2 LIMITING SAFETY SYSTEM SETTING 5

.I 2.2.1 REACTOR TRIP SYSTEM INSTRUMENT ATION SETPOINT5...............

2-[%

TABLE 2.2-1/ REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS..2-/,5 n

BASES SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE r :' 1).........

B/2-1 0: 1

~ ~.uk Lunt w,:T )

..m 2.1.2 REACTOR COOLANT SYSTEM PRESSURE.,

B 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP01hVS....

B 2-3 l

McGUIRE - UNITS 1 and 2 III Amendment No.

(Unit 1)

Amendment No (Unit 2)

i INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE Control Rod Insertion Limits......

3/4 1-21 3/4.2 POWER DISTRIBUTION LIMITS C s' 1

3/4.2.1 AXIAL FLUX OIFFERENCE...................................

3/4/2-1 3/4/2-6 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - Fq ( k,7, Z ).......

3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR................

3/4/2-14 3/4.2.4 QUADRANT POWER TILT RATIO...........

3/4/2-19 3/4.2.5 DNB PARAMETERS.........................

3/4 (2-22 TABLE 3.2-1/- DNB PARAMETERS..................

3/4/2-23 FIGURE 3.2-1 REACTOR COOLANT FLOW VS RATED THERMAL POWER..........

3/4/2-24 3/4h POWER DISTRIBUTI0A LIMIT 5 (UNIT 2)

~ ~ ~ '

1

^

.3/4.2.

AXIAL FLUX DIFFER CE..............

3/4 B2-1 3/4.2.2 HEAT FLUX HOT CHANN FACTOR - Fg (Z).

/4 B2-6 3/4.2.3 R

FLOW RATF AND NUCL,%R ENTHALPY RISE i T CHANNEL FAC R..................

3/4 2-14

-3 2.4 QUADRA'T POWER TILT RATIO.

3/4 B 19 3/4.2.'

ONB PARA TERS...............

3/4 B2-2 TABLE 3.6 lb DNS PA METERS............

3/4 82-23 1

3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION.......................

3/4 3-1 Mc'.iUIRE - UNITS I and 2 V

Amendment No.

(Unit 1)

Amendment No.

(Unit 2)

i 4

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 3.3-1)

REACTOR TRIP SYSTEM INSTRUMENTATION To':' 1).

3/4/3-2 TABLE 3.3-2\\

REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES (% :7 'r........

3/4/3-9 TABLE 4.3-1)

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENT S (4""

3/4/3-11 i

i

(

' IABLE '.3-lb REACTOR i P SYSTEM INSTN MENTATION (UNI 2)...

y un i

TABLE 3.3 'b REACTOR TRIP ~YSTEM INSTRUMEJTATION RESPONS TIMES (UNIT t 3/4 B3 TABLE 4.3-lb LEACTOR TRIP SYS M INSTRUMENTATA N SURVEILLANCE RE0VIREMFNT< (!!N T ?t 3/4 B3-11 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION J 1,....

3/4%3-15 rey : 3 y-g g g, % y, g ; ;;,

1

' :n ::TlJ::'

'"""M a,,

0; -

TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION......

3/4 3-16 TABLE 3.3-4[

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS (m;;T 1-)........

3/4[3-25

% a.a

. }

{," E T CS inmuMtn uu iva inih

,a TABLE 3.3-Sh ENGINEERED SAFETY FEATURES RESPONSE TIMES ( C :T 1)...

3/4 A3-30

-,n re7e roen orew 7;;gr,;; r,;:py:: v:o : (t 7 :).

,, ;; n TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS..........

3/4 3-34 3/4.3.3 MONITORING INSTRUMENTATION RADIATION MONITORING FOR PLANT OPERATIONS.......................

3/4 3-40 TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS......

3/4 3-41 TABLE 4.3.3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE REQUIREMENTS....

3/4 3-43 McGUIRE - UNITS 1 and 2 VI Amendment No. ( (Unit 1) l Amendment No.

(Unit 2)

)

b 4

INDEX LIMITIf,'G CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE Ice Bed Temperature Monitoring System.,...................

3/4 6-;6 Ice Condenser Doors...................

3/4 6-37 Inlet Door Position Monitoring System...........

3/4 6-39 Divider Barrier Personnel Access Doors and Equipment Hatches................................

3/4 6-40 Containment Air Return and Hydrogen Skimmer System........

3/4 6-41 Floor Drains..........................

3/4 6-42 Refueling Canal Drains............................

3/4 6-43 Divider Barrier Sea 1................

3/4 6-44 TABLE 3.6-3 DIVIDER BARRIER SEAL ACCEPTABLE PHYSICAL PROPERTIES....

3/4 6-45 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE Safety Valves............................................

3/4 7-1 TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING FOUR LOOP 0PERATION.......................

3/4 7-2 TABLE 3.7-2 (BLANK)..............................................

3/4 7-2 TABLE 3.7-3 STEAM LINE SAFETY VALVES PER L00P...................

3/4 7-3 Auxiliary Feedwater System................................

3/4 7-4 Specific Activity...............................

3/4 7-6 TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM..............................

3/4 7-7 Main Steam Line Isolation Valves (CG! 1}................

3/4/7-8

'n r :t::: 'J c I::!:ti:r " ' :- /"MTT ?)

1, ;7 4?

l 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION..........

3/4 7-9 3/4.7.3 COMPONENT COOLING WATER SYSTEM...........................

3/4 7-10 L

3/4.7.4 NUCLEAR SERVICE WATER SYSTEM.............................

3/4 7-11 FIGURE 3/4 7-1 NUCLEAR SERVICE WATER SYSTEM......................

3/4 7-11a i

l l

l McGUIRE - UNITS 1 and 2 XII AmendmentNog.1(Unit 1)

Amendment Ho M (Unit 2)

= - - - - -

)

INDEX BASES

_SECTION PAGE 3/4.0 APPLICABILITY.............................................

B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION CONTR0L.....................................

B 3/4 1-1 3/4.1.2 BORATION SYSTEM 5.................

B 3/4 1-2 3/4.1.3 MOVABLE CONTROL ASSEMBLIES........................

B 3/4 1-3 3/4.2 POWER DISTRIBUTION LIMITS sva.:., '

3/4.2.1 AXIAL FLUX DIFFERENCE...............

B3/4/2-1 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR

.ENTHALPY RISE HOT CHANNEL FACTOR.....

B3/4/2-3 3/4,2.4 QUADRANT POWER TILT RATIO...........................,.

B 3/4 /2-5 3/4.2.5 DNB PARAMETERS............................................

B 3/4 /2-5 E^gt2.. te,;;:::-::cT::

Pac : neii t; A

3/4.2.1 AX ' FLUX DIFFERENCE.

B 3/4 82-3

2. 2 and 3

.3 HEAT FLUX H0

.HANNEL FACTOR and FLOW RATE A NUCLEAR ENTHALPY SE HOT CHANNEL FAC B

4 B2-2

\\

3/4.2.9 QUADRANT P0

, TILT RATIO.......

B 3/4'~2-5

(

3/4.2.5 8 PARAMETERS.

B 3/4 8 3/4.3 INSTRUMENTATION 3/4.3.1 and 3/4.3.2 REACTOR TRIP and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION........................

B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION................................

B 3/4 3-2 3/4.3.4 TURBINE OVERSPEED PROTECTION..............................

B.3/4 3-5 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION (;:GT 1)....

B3/4/4-1

/ n:

A

: : re t ov,aaer n: :::te"* ::::; Ariu, tun 2.

e o,, ;; ;

i 3/4.4.2 SAFETY VALVES.............................................

B 3/4 4-2 3/4.4.3 PRESSURIZER.......................

B 3/4 4-2 3/4.4.4 RELIEF VALVES.............................................

B 3/4 4-3 3/4.4.5 STEAM GENERATORS........................................

B 3/4 4-3 McGUIRE - UNITS I and 2 XVI Amendment No.

(Unit 1)

Amendment No.

(Unit 2) w.

y..,.g e

l SECTION 2.0 SAFETY l.IMITS AND LIMITING SAFETY SYSTEM SETTINGS

.c

2. 0 SAFETY LIMITS-AND LIMITING SAFETY SYSTEM SETTINGS q

-2.1 SAFETY LIMITS REACTOR CORE 2.1.1-The combination of THERMAL-POWER, pressurizer pressure, and the highest

-operating loop coolant temperature (Tavg) shall not exceed the limits shown in Figures 2.1-1/and2.1-2forfourandthreeloepoperation,respectively.

l-APPLICABILITY: MODES 1 and 2.

JHtI c'M

. ACTION:

'Whenever the point defined-by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the require-monts of. Specification 6.7.1.

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

APPLICABILITY:

MODES 1, 2, 3, 4,_and 5.

tuni n.1 2'r

{

ACTION:

MODES 1 and 2

Whenever the Reactor Coolant System pressure has' exceeded 2735 psig, be in; HOT STANDB_Y with-the Reactor Coolant System pressure within its limit within l' hour, and comply.with the requirements of Specification 6.7.1.

MODES'3,14and'5-Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to.within its limit within

[

'5 minutes, and comply with the requirements of Specification 6.7.1.

I i

L l'

L l

2-f1 Amendment No.

(Unit 1)

-McGUIRE - UNITS 1 and 2 d

Amendment No.

(Unit 2) y

4 N

\\

Figure 2,1-1/ Reactor Core Safety Limits -

Four Loops in Operation (F 't a j

665 3

FLOW PER LOOP = 96250 GPM N

660 y 1

655 ',

2455 psia UNACCEPTABLE 650_,

OPERATION e.,

t 645

400 paia 640 635 630 2 1

625 a

1 Pala 5 620 e

615 i 610 5 1945 psia 605 i 600 5 0

595 5 4

d 590 -

ACCEPTABLE OPERATION 585 5:

550 l

~

i 0.00 0.20 0.40 0.60 0.80 1,00 1.20 Fraction of Rat.ed Thermal Power 2-[2 McGUIRE - UNITS 1 and 2 Amendment No (Unit 1)

Amendment No (Unit 2)

'AFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS a:' ;-

l 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The Reacter Trip System Instrumentation and Interlocks Setpoints shall besetconsistentwiththeTripSetpointvaluesshowninTable2.2-lh.

y...,, g.

APPLICABILITY:

AsshownforeachchcnnelinTable3.3-1/m;T1);

l ACTION:

With a Reactor Trip System Instrumentation or Inter'ock Setpoint less conser-vative than the value shown in the Allowable Values column of Table 2.2-1 declarethechannelinoperableandapplytheapplicableACTIONstatement),

requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its Trip Setpoint adjusted consistent with the Trip Setpoint value.

l McGUIRE UNITS 1 and 2 2-f4 Amendment No.

(Unit 1)

Amendment No.-

(Unit 2)

- ala,'

f TABLE 2.2-1/

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

[

FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES z

3

1. Manual Reactor Trip N.A.

N.A.

[

2. Power Range, Neutron Flux Low Setpoint 1 25% of RATED Low Setpoint 1 26% of RATED THERMAL POWER THERMAL POWER

~

High Setpoint 1 109% of RATED High Setpoint 1 110% of RATED THERMAL POWER THERMAL POWER l

3. Power Range, Neutron Flux, 5 5% of RATED THERMAL POWER with 5 5.5% of RATED THERMAL POWER High Positive Rate a time constant > 2 seconds with a time constant > 2 seconds

4. Intermediate Range, Neutron i 25% of RATED THERMAL POWER i 30% of RATED THERMAL POWER Flux

5. Source Range, Neutron Flux i 10 counts per second i 1.3 x 105 counts per second 5

6. Overtemperature AT See Note 1 See Note 3

7. Overpower AT See Note 2 See Note 4

8. Pressurizer Pressure--Low

> 1945 psig

> 1935 psig gg i

hy

9. Pressurizer Pressure--High 1 2385 psig i 2395 psig l

.e 1

m.A5

10. Pressurizer Water Level--High 1 92% of instrument span i 93% of instrument span

11. Low Reactor Coolant Flow

> 90% of minimum measured

> 88.8% of minimum measured T10w per loop

Tiow per loop

22

" Minimum measured flow is 96,250 gpm per loop.

hb

$5

,,. =

1

?-

C?:" *: $5 8-

. TABLE 2.2'-14 (Continued)

.E.

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS.

1

]

FUNCTIONAL UNIT TRIP SETPOINT-ALLOWABLE VALUES

~

12. Steam Generator Water.

> 12%.of span from-0 to 307 of'

> 11% of span from 0 to-30% of Level--Low-Low:

RATED THERMAL POWER. increasing RATED THERMAL POWER; increasing linearly to > 40% of span at to 39.0% of span at 100% of

'100% of RATED THERMAL POWER RATED THERMAL POWER.

'13. Undervoltage-Reactor.-

2 5082 volts-each bus

> 5016. volts-each hus-Coolant Pumps 1:

14. Underfrequency-Reactor 2 56.4 Hz - each bus

> 55.9 Hz - each bus Coolant Pumps-7

15. Turbine. Trip m

a.

Low Trip System Pressure-l'45 psig 3 42 psig b.

Turbine Stop Valve i

Closure:'

> 1% open

16. Safety Injection Input N. A.

N.A.

from ESF

.E.>a

. &R 17.

Reactor Trip System Interlocks 55

'E a a.

Intermediate Range Neutron Flux, P-6,

> 1 x 10 20 amps 1 6 x 10 11 amps Enable Block Source Range Reactor. Trip 2z b.

Low Power. Reactor Trips Block, P-7 1)

P-10 Input' 10% of RATED

> 9%, < 11% of RATED

_y&

THERMAL POWER THERMAE POWER

??

2)

'P-13 Input

< 10% RTP Turbine

< 11% RTP Turbine m e_.

Tapulse Pressure Impulse Pressure Equivalent Equivalent a e -

-~

-,.m

?~

4 N

x g

07:1 ^.: "^':' ?; :

g3 TABLE 2.2-I). (Continued)

x

][

. REACTOR TRIP SYSTEM INSTRUMENTATION ~ TRIP SETPOINTS E

((

FUNCTIONAL UNIT.

TRIP SETPOINT-ALLOWABLE VALUES-c.

Power: Range Neutron Flux,1 P-8,

-< 48% of RATED

< 49%.of' RATED Lo'w Reactor Coolant. Loop. Flow,

. THERMAL POWER' THERMAL POWER

.and Reactor Coolant Pump Breaker

Position d.

, Low Setpoint. Power Range' Neutron 10% of. RATED

> 9%, < 11% of RATED Flux, P-10. Enable Block;of.

THERMAL POWER.

THERMAL POWER

-Source Intermediate and Power Range Reactor. Trips

.e.

Turbine Impulse Chamber Pressure.

'?

P-13, Input:to Low Power, Reactor

< 10% RTP Turbine

< 11% RTP Turbine "T7 Trips Block P-7 Impulse Pressure Impulse Pressure Equivalent Equivalent 4

18.

Reactor Trip Breakers N.A.

N.A.

19.

Automatic Trip and Interlock. Logic N.A.

N.A.

' Ef 3I

=

ea aa on

.C Q

22

=L E:

ne n

[

-.m

-e, r

sm-aw

wo i r r

TABLE 2.2-Lt (Continued)

REACTOR TRIP SYSTEM' INSTRUMENTATION TRIP SETPOINTS E

NOTATION Q

[

NOTE 1:

OVERTEMPERATURE AT l+t S 1

AT(f*I

)I (1 + toS)-T'] + K (P-P') - f (al)}

. t l+T 5) 3 AT M -K2 (1 + t S 3

g y

3 3

Where:

AT Heasured AT by Loop Harrow Range RID,

=

lead-lag compensator on measured AT,

=

T1, 12

= Time constants utilized in the lead-lag controller for AT, 13 > 8 sec., 12 5 3 sec.,

1 Lag compensator on measured AT,

=

y, 13 Time constants utilized in the lag compensator for AT, 13 $ 2 sec.*

=

Indicated AT at RATED THERMAL POWER, AT

=

g K

< 1.1958, 3

d gg 0.03143

[

K

=

2 q

rr 1+T 5

~ ~

4 Tt.e function generated by the lead-lag controller for T d "8"IC C *P'"5 li "*

=

Y 3, Tsg avg 14 15 Time constants utilized in the lead-lag controller for T

=

mm

9,

{

28 sec, is 5 4 sec.,

14 c> c>

Average temperature. *F, T

=

I Lag c mpensator on measured T

=

1 + ts5

3yg,

. p,

' TABLE 2.2-14 (Continued) 5, g-REACTOR TRIP SYSTEM; INSTRUMENTATION TRIP SETPOINTS

' NOTATION (Continued).

z 3

NOTE 1:

(Con'tinued) -

Time' constant-utilized in the' measured I,,g lag. compensator, Is < 2 sec e.

Is

=

T'

< 588.2*F Refere'nce T,yg'.at RATED. THERMAL POWER,

=-

K3 0.001405,

=

P

'=

Pressurizer' pressure, psig, P'

=.2235 psig (Nominal RCS operating pressure),

S

= Laplace transform operator, sec 8, and f3(al) is a' function of the indicated dif ference between top and bottom detectors e

of'the'pnwer range ~ nuclear ion chambers; with gains to.be selected based on measured instrument response during plant startup tests.such that:

(i) for q ~ U between -39% and +7.0%al; f (AI) = 0, where q and a a m percent RA M

[

t b

y t

b gy THERMAL POWER in the top and bottom halves of the core respectively, and q + q t

b is total THERMAL POWER in percent of. RATED THERMAL POWER; EE (ii) for each percent imbalance that the magnitude of q q' is more negative than -39%i1,

)

er t

b the AT Trip Setpoint shall be automatically reduced by 6.153% of ATo, and z2 P?

j (iii)' for each percent imbalance that the magnitude of q o is more positive than

' -+7.0%AI,the'AT.TripSetpointshallbeautomaticallyreNucedby1.511%ofAT.

^^

D3 Y h-

~w vv t

N

t m

M f

TABLE 2.2-1)(Contimmed) n E

M ItfAcles talP SYSIIM 18151ESENTATICII TR1r SETPolafis

[

NOIATI0li{Centinued) ad IETE 2:

DVERPOIER AT v15 1

1 H

[

A I (1, I (1 +

3) 1 AT, $ - g (1 + 1 53Il + 1.5) T -KgT(g, t 3)- W - f ISI)I 4

2 3

1 eo As defined la Note 1, idhere:

AT.

=

I * **I

= As defined in Itete 1 1 +1:S 13, tz

= As defined in Ilote 1 1

As defined in Note 1,

=

1+T53 T.

AT' As defined in flote 1

=

. it' 3

g 4

5 LM.

I K

' L~-

K

=

0.02/*T for increasing average tevrature and 0 for decreasing average 3

-2 temperature, kk ty g,

3 The f::mction pnerated by the rate-lag controller for I,g dynamic

.: pj

=

C"'pm.atwi, Z.,

==

er gg 1,

= Time constant ut.ilized in the rate-lag controller for I

, tr > $ sec, i

1 er, I'

3, s5 Ij EE ts As defined in IInte 1,

=

2 a 0.001239/*f for I > 1" and K6 = 0 for T $ P, l

K

=

y.

ee 6

je

i r

IRIIi 1 -

]

J

{'

1AKE2.2-1/(Cantloved) c F

i G-AEACIOR 1 RIP SYSTEM INSlaisEA1AlicII 1 RIP SETPetNTS l

m

!WTATISil(Costinued)

=c

'*4 As defined in ftste 1, T

=

e E

1"

-< 588.2*f Reference T at ItATED T E WWnt PO E R,

=

n aug As defined in Nota 1, and S

=

f (AI) is a function of the indicated difference between tap and bettes detectors 2

of the peser-range nuclear ies cWrs; with gases to be selected based en seasured

[

lastrument response eering plamt startg tests such that.

.I (i) for between -3!iE a::aa +3!iE AI; f (A1) = 0, tsher-? g and g are percent

.)

RATFD THE

. in the tap and bettaan halves of the care r4spectively, and i

g + g is total TESmL PSER in percent of RATER TIEalmL PlaER; (ii) ' for each pexent inhalance that the maritude of g g is more negattwe than

-35E AI, the AT Trip Setpoint shall be autenetically radhrad gy 7.3 of AI.; and l

'(

.~2 (iii) for each percent imbalamon that the empitude cf g -

is more positive than i

+35E AI, the AT Trip Setpoint shall be avtamstically redhced 7.5 of AT..

U kk plate 3:

The channel's maximum Trip Setpoint shall not emoeed its caguted Trip 5etpoint by more than 3.EK of Roted Thersel Power.

l}

l iM a a sente 4:

The channel's mexiuman Trip Setpoint. shall not exceed its coupstod Trip setpoint by more than 4.2%

'[

C of Rated 7ternal Power.

z2 oo E$

j h.:.

22 11 CC r:

-j d.

l i

fQ W Y

v 6L g

-2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS UNIT /

l 1

2.1 S ETY LIMITS REACTOR RE 2.1.1 The mbination of THERMAL POWER, pressurizer pressure, d the highest operating lo coolant temperature (T,yg) shall not exceed the imits-shown in figures 2.1-lb nd 2.1-2 for four and three loop operation, espectively.

l APPLICABILITY:

i ES 1 and 2. (Unit 2 only) l ACTION:

Whenever the poir.t de. ned by the combination of

,e highest operating loop average temperature an THERMAL POWER has excee d the appropriate pressurizer pressure line, be in H0 TANDBY within I hour and comply with the require-ments of Specification 6.

1.

REACTOR COOLANT SYSTEM PRESSU ~

2.1.2 The Reactor Coolant Syste pr sure shall not exceed 2735 psig.

APPLICABILITY:

MODES 1, 2, 3, 4, d 5. (Units 1 and 2) l ACTION:-

MODES 1 and 2 Whenever the Reactor oolant System p essure has exceeded 2735 psig, be in HOT STANDBY wit the Reactor Coolan System pressure within its limit within I hour, an comply with the requ ements of Specification 6.7.1, MODES 3, 4 and 5 Whenever t Reactor _ Coolant System pressure s exceeded 2735 psig, reduce th Reactor Coolant System pressure to

' thin its limit within 5 minute, and comply'with the requirements of ecification 6.7.1.

McGUIRE - UNITS I and 2 2-B1 Amendment No.128(Uni 1) l Amendment No. 110(Unit

3/4.2 POWER DISTRIBUTION LIMITS Uru i r-I 3/4.2.1 AXIAL FLUX DIFFERENCE (AFDJ LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the acceptable limits as specified in the Core Operating Limits Report (COLR).

j APPLICABILITY:

MODE 1 above 50% of RATED THERMAL POWER *.

,m,,', !b ACTION:

For operation with the indicated AFD outside of the limits specified

(

a.

in the COLR, 1.

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

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux -

High Trip setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

5 g?

b.

THERMAL POWER shall not be increased above 50% of RATED THERMAL

)

POWER unless the indicated AFD is within the limits specified in the COLR.

1

(

i l

l

See Special Test Exception 3.10.2.

fy s

3/4[2-1 McGUIRE - UNITS 1 AND 2 Amendment No.,', (Unit 1)

Amenoment No,

i. (Unit 2)

POWER DISTRIBUTION LIMITS "f:: T- ;

l SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 50% of RATED THERMAL POWER by:

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 for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after restoring the AFD Monitoring Alarm to OPERABLE status, b.

Monitoring and logging the indicated AFD for each OPERABLE excore channel at least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least once per 30 minutes thereafter, when the AFD Monitor Alarm is inoperable.

The logged values of the indicated AFD shall be assumed to exist during the interval preceding each logging.

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

limits, h

t l

l l

l McGUIRE - UNITS 1 AND 2 3/4[2-la Amendment No.

(Unit 1) l Amendment No.

0 (Unit 2)

. _ _. _ _ _ _ ~ _ _

m__.~.

POWER OISTRIBUTION LIMITS

-g 4

[

3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F (X,Y,2)

LIMITING CONDITION FOR OPERATION 3.2.2 F (X,Y,Z) shall be limited by imposing the following relationship:

yi 9

Q

)I K(Z) for P > 0.5 P

MA RTP Fq (X,Y,2) $ F Q K(Z) for P s 0.5 k

0.5 Where F TP = the F limit at RATED THERMAL POWER (RTP) specified g in the CORE OPERATING LIMITS REPORT (COLR),

p, THERMAL POWER RATED THERMAL POWER K(Z)_=

the normalized F (X,Y,Z) limit specified

(

9 5(

in the COLR for the appropriate fuel type, and N#

Fq (X,Y,Z) = the measured heat flux hot channel g

factor F (X,Y,Z)'with the adjustments

[

specified in 4.2.2.3 APPLICABILITY: _ MODE 1. '."V.! b l

ACTION:

With F (X,Y,Z) exceeding its limit:

/

9 a.

Reduce THERMAL POWER at least H for each 1% F9 (X,Y,Z) exceeds the

)

MA limit within 15 minutes-and similarly reduce the Power Range. Neutron

).

Flux-High Trip Setpoints within the ncxt 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , and.

i

)

b.-

Control the AFD to within new AFD limits which are' determined by

' reducing the-allowable power at each point along the AFD limit lines of Specification 3.2.1 at least 1% for each 1% F A(X,Y,Z) exceeds the limit within 15 minutes and reset the AFD alarm

(

setpoints to the modified limits within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , and c.

POWER 0PERATION may proceed for up to a total =of_72 hours; subsequent-y POWER OPERATION may proceed provided the Overpower AT Trip Setpoints i

(value of K ) have been reduced at least 1% (in AT span) for each 1%.

h 4

FMA(X,Y-Z) exceeds the limit, and d.

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

above; THERMAL. POWER may then be increased provided F (X,Y,Z) is 9

demonstrated through incore mapping to be within its limit.

l I

'McGUIRE - UNITS 1 ANO 2 3/4 2-6 Amendment No.\\yf (Unit 1) l 1-Amendment No M (Unit 2)

DOWER DISTRIBUTION LIMITS t;, ;-r I

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

F"(X,Y,Z)(1) shall be evaluated to determine whether F (X,Y,Z) is 4.2.2.2 9

9 within its limit by:

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

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

b.

Measuring F (X,Y,Z) at the earliest of:

1.

At least once per 31 Effective Full Power Days, or 2.

Upon reaching equilibrium conditions after exceeding by 10%

or more of RATED THERMAL POWER, the THERMAL POWER at which MF (X,Y,Z) was last determinedI)

(

, or 3.

At each time the QUADRANT POWER TILT RATIO indicated by the excore detectors is normalized using incore detector

[

. measurements.

\\

(1) No additional uncertainties are required in the following equations for F (X,Y,2) because the limits include uncertainties.

(2) During power escalation at the beginning of each cycle, THERMAL POWER may be increased until a power level for extended operation has been achieved and a power distribution map obtained.

q McGUIRE - UNITS 1 AND 2 3/4 2-7 Amendment No.

(Unit 1) l Amendment No.

(Unit 2)

POWER DISTRIBUTION LIMITS

^wp l

SURVEILLANCE REQUIREMENTS (Continued) c.

Derforming the following calculations:

1.

For each core location, calculate the % margin to the maximum allowable design as follows:

% Operaiional Margin = ( 1 -

(*'Y'2)

) x 100%

[F (X,Y,Z)]OP

(

M

% RPS Margin = ( 1 -

F (X,Y,Z)

) x 100%

[Fh(X,Y,Z)]

b where[Fh(X,Y,Z)]0P and[Fh(X,Y,Z))RPSare the Operational and RPS design peaking limits defined in the COLR.

2.

Find the minimum Operational Margin of all locations u amined f

in 4.2.2.2.c.1 above.

If any margin is less than zero, then I

either of the following actions shall be taken:

L h

(a) Within 15 minutes:

l (1) Control the AFD to within new AFD limits that are determined by:

(3)

(AFD Limit)

= (AFD Limit) C

-MARGINhN uc

,t e

n ative (3) e c U

(AFDLimit)[g

, = (AFD Limit) p ggy, - MRGIN I

N where MARGIN is the minimum margin from 4.2.2.2.c.1, x

and r

(2) Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , reset the AFD alarm setpoints to the modified limits of 4.2.2.2.c.2.a. or (b) Comply with.the ACTION requirements of Specification 3.2.2, i

treating the margin violation in 4.2.2.2.c.1 above as the

/

amount by which F is exceeding its limit.

)

>l

(

Defined and specified in the COLR per Specification 6.9.1.9.

/

McGUIRE - UNITS 1 AND 2 3/4 2-8 Amendment No.

(Unit 1) j Amendment No.

(Unit 2) 1 l

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

POWER DISTRIBUTION LIMITS w+

SURVEllLANCE-REQUIREMENTS (Continued) v 3.

Find the minimum RPS Margin of all locations examined in 4.2.2.2.c.1 above.

If any n.argin is less than zero, then the following action shall be taken:

Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , reduce the K 2 value for OTaT by:

adjusted, gx(4) - [KSLOPE(3) x Margin "RPS] absolute value I"

K 3 where MARGIN is tre minimum margin from 4.2.2.2.c.1.

/

Extrapolating (5) at least two measurements to 31 Effective Full kl d.

Pcwer Days beyond the trost recent measurtment and if:

M

[F (X,Y,Z)] (extrapolated)1[Fh(X,Y,Z))0P(extrapolated),and I :

[F (X,Y,Z)] (extrapolated)

[F (X,Y,Z, L L

OP L

OP

[F (X,Y,Z)]

(extrapolated)

[F (X,Y,2)]

q q

or l

k N

[F (X,Y,Z)] (extrapolated) 3 [F (X,Y,2))RPS (extrapolated), and q

[F"(X,Y,Z)]

X Y,Z)] (extrapolated) a L'

'RP5 L

RPS

[F (X,Y,'Z)]

.(extrapolated)

[F (X,Y,Z)]

I q

q either_of the following actions shall be taken:

l F"(X,Y,Z) shall be increased by 2 percent over that specified 1.

q in 4.2.2.2. a. and the calculations ' of 4.2.2.2.c repeated, or

( ) Defined and specified in the_COLR N r Specificat-lon 6.9.1.9.

( ) K value from Taole 2.2-1.

. i

( )L xtrapolation of Ff for-the initial-flux map taken af ter reaching equili-E brium' conditions is'not' required since the initial flux map establishes the baseline measurement for future' trending. 'Also, extrapolation of F limits are not valid for core -locations that were previously rodded, or for core locailons that were previously within ;1% of the core height about the demand position of the rod tip.

McGUIRE - UNITS 1 AND 2 3/4/2-Ba Amendment No.fji$ (Unit 1)

Amendment No.f M (Unit 2)

4 POWER DISTRIBUTION LIMITS 0,;T T SURVEILLANCE REQUIREMENTS (Continued) 2.

A movable incore detector power distribution map shall be obtained, and the calculations of 4.2.2.2.c.1 shall be nerformed no later than the time at which the margin in 4

4.2.2.2.c.1 is extrapolated to be equal to zero, The limits in Specifications 4.2.2.2.c and 4.2,2.2.d are not appli-e.

cable in the following core plane regions as measured in percent i

of core height from the tsottom of the fuel:

1.

Lower core region from 0 to 1%, inclusive.

j 2.

Upper core region from 85 to 100%, inclusive.

/

lI l

ll

\\)

/

\\

)

/

i l

l

/

\\

3/4[2-9 McGUIRE - UNITS 1 AND 2 Amendment No.

(Unit 1)

Aniendment No.

(Unit 2) i

~

POWER DISTRIBUTION LIMITS

,M c l

SURVEILLANCE REQUIREMENTS (Continued) 4.2.2,3 When a full core power distribution map is tak.en for reasons other than meeting the requirements of Specification 4.2.2.2, an overall F (y,y,7; M

q shall be determined, then increased by 3% to account for manufacturing toler-ances, further increased by 5% to account for measurement uncertainty, and fu ther increased by the radial-local peaking factor to obtain a maximum local

^

peak.

This.value shall be compared to the limit in Specification 3.2.2,

(

\\

k ll

/

I

\\

l

\\

l

/

\\

l

\\

l

/

l

\\

L 3/4/2-9a McGUIRE - UNITS 1 AND 2 Amendment No.

(Unit 1)

'i

[

Amendment No.

(Unit 2)

POWER Of$TRIBUT!04_ LIMIT $

49Pt 3/4.2,3 NUCLfAR ENTHALP1 Ri$1 %T CHANNEL FACTOR

Fg(X,0 LIMITIN3 CON 0!T!04 FOR OPERATION 3.2.3 Fg/X,Y) shall be ifGited by impostrg the following relatieaships 7lg(X,Y)f,(7h(X,Y))l00 Fh (X,Y) the sensured radial peak.

wnere:

(Fh (X,Y))L O the saxinus allowable radial peak as defined in Core Operating Limits Report (COLR).

i APPLICat!LITY: H00E 1.

tesey ACTION:

With 'j,Y) exceeding its 11eit:

X 4.

Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , reduce the allowable THERMAL POWER from RATED THERMAL POWER at least RRHM*) for each 1% that [g(X,Y) exceeds the limit, and b.

Within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months either:

1.

Restore U (X,Y) to within the limit of $pecification 3.2.3 for g

RATED THERMAL POWER, or 2.

Reduce the Power Range Neutron Flux High Trip setpoint in Table 2.22{stleastRRH4foreach1% that [g(X,Y) exceeds that limit, ana c.

Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of initially being outside the limit of specification 3.2.3, 41 W rt Restore Fh(X.Y) to within the 11ett of specification 3.2.3 for 1.

RATED ThERPML POWER, or 2.

Perform the following actions:

RRH is the amount of THERM L POWER reduction required to conpensate for each 1% that [g(X,Y) exceeds the limit of Specification 3.2.3, provided in the COLR per Specification 6.9.1.9.

McGUIRE - UNITS 1 AND 2 3/4{214 Amendment ho. Ng Unit 1 I

Amendment No. W Unit 2 y

c

. ~,.,

POWER DitTRIBilTION l! NITS det*-t

}/4,2 ) NUC(!)R (NTHALPV Ri$E HOT CW/AAEL FACTOR Fg(X,Y)r t!4!T!NG CONDITION FOR OPERATION d.iMEl I)

(4) Reduce the CTai Kg term in Table 2.2-[ cy at 1sait TRH foreach1%thatU(W,Y)exceedsthelimit,and g

(b) Verify tnrough inccre aspping that Fh(X,Y) 16 resterda to within the limit for the reduced THERMAL, POVER allowed by ACT'10N a, or reduce THERKAL POVER to less than 5% of RATED WERMAL P'NER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

5 O) TRH is the amount of CTAT Kg setpoint reduction required to cospansate for mach 1% that U (X,Y) exceeds the limit of Specification 3.2.3, provided in g

the COLR per Specification 6.9.1.9.

McGUIRE UNITS 1 AND 2 3/4it144 Mendment No.128 (Unit 1)

/\\

Mendment No.110 (Unit 2)

.;,: 4; li H a1 li.'i K'.fi

$Y d I) N'i'*I' l.

POWER 0?$T4160T!0NL! NIT $

l o; 6 Uu!Tik300N0!T!0NFOROsERATION

$JQ$1 (Continued) l o.

Identify and correct the cause of the out of limit condition prior to increasing T!iEML POWER above the roouced THtRMAL POWER limit required by ACTION a. and/or c.2 above subsequent POWER PERAT!0N Inay procabd provided that Ph(X,Y) is demonstrated, throu;h intore flux mpping, to be within the Litit soecified in the COLR prior to exceeding the following TH(RNAL PCVLR 1evels:

1.

50% of RATED THERMAL POWER, l

2, 75% of RATED THERMAL POWER, and l

3.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of attaining greater than or equal tn 9E% of

'!ATED THEIMAL POWEA.

$URVEILLAA'.E REQU!AF,MENTS 4.2.3.1 The provistens of Spc?tftcation 4.0.4 dra not applicable.

M 4.2.3.2 F3(X,Y) shall be evaluated to dettreine whether Fg(X,Y)iswithin its 1%it try e.

Using the movable incere detectors to obtain a power distributto1 map at any THERNAL POWER greater than 5% of RATED THERMAL PCWER.

Measuring Fh(X,Y) according to the following schedule b.

1.

Upon reaching equilibriis conditions after exceeding 10%

or more of RATED THERMAL POWER, the THCRMAL POWER at which Ffg(X,Y) was last determined (D, or 2.

At 1oest once per 31 Effactive Fell Pcwor D6/s or 3.

At asch time the QUADRAVT POWER TILT RATIO indicated by the exeore detectors is norgalized using incere detector m esurements, c.

Performing the following calculations:

1.

For each location, esiculate the % sargin to the maximus allowable deeign as follows U) Durtrg power escalation at the beginning of e6ch cycle, THERNAL POWER may De increased until a power level for axtended operation nas been achieved and a power distribution anp obtained.

McGUIRE - UNITS 1 AhD 2 3/4f215 Amendment No.

Unit 1)-

Amendment No.

Unit 2) 41 0

-t4 e

w e

l DOWER D1$TRINT!0N dMIT$

-m6a-t-

$L[RVEILLANCE Itf091REMEWT$

?

Nargin = (1 h(X,Y) ) x 100%

F

%fg (Fh(X,Y)) sun Noadditionaluntertaintiesarerequiredfor(g(X.Y),because (Ph(X,Y)) sun incluces uncertainties.

2.

Find the minimus margin of all 1ccations examined in 4.2.3.2.c.1 above. If any targin is less than zero, comply with the ACTICN reqeirements of specification 3.2.3 as if CFh(X,Y)]"" is tre someas(Fh(X.Y)]LCO, j

Extrasolating ') at least two ressurements to 31 Effective Full Power I

d.

Days beyond the most recent m asurement and ift Fh(X,Y) (extrapolated) > (Fh(X.Y)]sury (extrapolated) and Fh(X,Y)

F.s,(X.Y)

(axtrapolated)

(Fh(X,Y))'"" (extrapolated)

(Fh(X,Y))'*

either nf the following actions shall be taken:

Ffg(X,Y) shall be leeressed by 2 percent over that specified 1.

in 4.2.3.2.a. and the calculations of 4.2.3.2.c repeated, or 2.

A movable incore detector power distribution map shall be obtained, and the calculations of 4.2.3.2.c shall be perfomed no later than the time at which the margin in 4.2.3.2.c is extrapolated to be equal to zero.

() Extrapolation of Fh for the initial flux map taken after reaching equili-brium conditions ' not required since the initial flux mao establishes the b6eeline a>easurement for future trending.

McGUIRE UNITS 1 AND 2 3/4k21$a Amendsent No.

Unit 1)

Amenchent No.

Unit 2)

O W U U U U U5b 0A NY

. Ni..-

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

4 I

POWER OlSTRIBUTION LIMITS rr 3/4.2.4 QUADRANT POWER TIL1 RATIO LIMITING CONDITION FOR OPERATION 1

L l

3.2.4 The 00ADRANT POWER TILT RATIO shall not exceed 1.02.

APPLICABILITY:

MCDE 1 above 50% of RATED THERMAL POWER *.** w ow

'A l

4 -

ACTION:

a.

With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but less than or equal to 1.09:

1.

Calculate the QUADRANT POWER TILT RATIO at least once per hour

. until either:

a)

The QUADRANT POWER TILT RATIO is reduced to within its limit, or b)

THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.

2.

Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

a)

Reduce the QUADRANT POWER TILT RATIO to within its limit, or b)

-Reduce THERMAL POWER at-least 3% from RATED THERMAL POWER for each 1% t,i indicated QUADRANT POWER TILT RATIO in excess of 1.02 and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next.4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

3.

~ Verify that the QUADRANT POWER TILT RATIO is within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit or reduce-THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the. Power Range Neutron Flux-High Trip '

Setpoints tc tess than or equal to 55% of RATED THERMAL /0WER-within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; and 4.

Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION-

- above 50% of RATED THERMAL POWER may proceed provided that the 4

- QUADRANT POWER-TILT RATIO is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95%

or-greater RATED-THERMAL POWER.

See Special Test Exception 3.10.2.

- Not' applicable until calibration of the excore detectors is completed subsequent to refueling.

McGUIRE4 UNITS 1 AND 2 3/4/2-19 Amendment No. V ((Unit 1) l Amendment No. p N. Unit 2) s u-er 9 ~ pge

- g q-r-ym p s -+w. rw y

y

-r sp-5vyS 97w n -y

-s-g pg., W eeq - q.d w-gpay-w,. 4 y -rs-mg p,-wee-qv--

g p-q p.i-g.g.,w;q

,wa-r r

r 3:e.,y a.re.rmm.iw a-g g-n mim*a er %

ao+ew4a.m.w e-r

i POWER DISTRIBUTION LIMITS i

LIMITING CONDITION FOR OPERATION ACTION: (Continued) b.

With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to misalignment of either a shutdown or control rod:

1.

Calculate the QUAD l TANT POWER TILT RATIO at least once per hour until either:

a)

The 00ADRANT POWER TILT RATIO is reduced to within its limit, or b)

THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.

2.

Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.02, within 30 minutes;

/

3.

Verify that the QUADRANT POWER TILT RATIO is within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; and 4.

Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILi RATIO is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95%

or greater RATED THERMAL POWER.

c.

With the QUADRANT POWER TILT RATIO dete"

o exceed 1.09 due to causes other than the misalignment of eit a shutdown or control rod

1.

Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:

a)

The QUADRANT POWER TILT RATIO is reduced to within its limit, or b)

THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.

McGUIRE - UNITS 1 AND 2 3/4/2-20 Amendment No.

(Unit 1)

/

AmendmentNo.f (Unit 2)

s POWER DISTR 4BUT10N LIMITS V::' r i

LIMITING CONDITION FOR OPERATION ACTION:

(Continued) 2.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; and 3.

Identify and correct the cause of the Out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified at 95% or greater RATED THERMAL POWER.

d.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by:

a.

Calculating the ratio at least once per 7 days when the alarm is OPERABLE, and b.

Calculating the ratio at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during steady-state operation when the alarm is inoperable.

4.2.4.2 The QUADRANT POWER TILT RATIO shall be determined to be within the limit when above 75% of RATED THERMAL POWER with one Power Range channel inoperable by using the movable incore detectors to confirm that the normalized symmetric power distribution, obtained from two sets of four symmetric thimble locations or a full-core flux map, is consistent with the indicated QUADRANT POWER TILT RATIO at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

3/4[2-21 Amendment No.

(Unit 1)

I McGUIRE - UNITS 1 AND 2 Amendment No.

(Unit 2)

POWER DISTRIBUTION LIMITS "c" " L 3/4.2 d DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB related parameters shall be maintained within the limits shown on Table 3.2-1K /^'* ' t Reactor Coolant System T,yg, e.

b.

Pressurizer Pressure, and c.

Reactor Coolant System Total Flow Rate.

I wo, & S l

- APPLICABILIT,,Y:

MODE 1.

ACTION:

a-With either of the parameters identified in 3.2.5a. and b. above I

exceeding its limit, restore the parameter to within its limit within 2-hours or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , b.

-With the combination of Reactor Coolant Syst:r total flow rate and 9

5 THERMAL POWER within the region of restricted operation specified on Figure 3.2.1, within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months reduce the Power Range Neutron h Trip Setpoint to below the nominal ~setpoint by the same Flux-Hig% RTP) as the power reduction required by Figure 3.2-1.

amount (

c.

With the combination of RCS total flow rate and THERMAL POWER within the region of prohibited operation specified on Figure 3.2-1:

{

f 1.

Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

a)

Restore the combination of RCS total. flow rate and THERMAL r

POWER to within the region of permissible operatiUn, g

1 b)

Restore the combination of Reactor Coolant System total

[

flow rate and THERMAL POWER to within the region of s

restricted operation and comply with action b. above, or-c)

Reduce _ THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 55%-of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

2.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of initially being within the region of prohi-bited oper ation specified in Figure 3.2-1, verify that.the-com-i bination of THERMAL POWER and RCS total flow rate are restored 1

to within the regions of permissible or restricted operation, or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

l McGUIRE - UNITS 1 AND 2 3/4/2-22 Amendment No. ?; (Unit 1)

Amendment No.

')-(Unit 2)

POWER 0157RIBUT10N l.!MITS

,;; y

{

3/4.2.5 DNB PARAMETERS SURVEll.l ANCE REQUIREPENT5 4.2.5.1 Each of the parameters of Table 3.2-1 shall be measured by averaging j

the indications (meter or computer) of the operable channels and verified to be within their limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4. 2. 5. 2 The RCS total flow rate indicators shall be subjected to a CHANNEL cat.lBRATION at least once per 18 months.

4.2.5.3 The RCS total flow rate shall be determined by precision heat balance measurement at least once per 18 months.

l t

l i

l 3/4[2-22a Amendment No. V (Unit 1) l l

McGUIRE - UNITS 1 AND 2 Amendment No. /N (Unit 2')

__ ~

POWER DISTRIBUTION LIMITS

_ ;"; ' L TABLE 3.2-1l DNB PARAMETERS

OPERABLE PARAMETER INDICATION CHANNELS LIMITS

Indicated Reactor Coolant System T meter 4

<590.5'F avg meter 3

[590.2'F computer 4

<591.0'F computer 3

[$90.8'F Indic6ted Pressurizer Pressure **

meter 4

12226.$ psig meter 3

12229.8 psig computer 4

3 221.7 psig 2

computer 3

32224.2 psig Reactor Coolant System Total Flow Rate Figure 3.2-1

Limits applicable during four-loop operation.

- Limits not applicable during either a THERMAL POWER ramp in excess of 5% of RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10%

RATED THERMAL POWER.

McGUIRE - UNITS 1 AND 2 3/4 2-23 Amendment No.

(Unit 1) l Amendment No.

(Unit 2)

-. -.. - - ~. - -. ~. -

POWER DISTRIBUTION LIMITS l

va,_

i Figure 3.2 - 1.

Reactor Coolant System Total flow Rate Versus Rated Thermal Power - Fcur Loops in Operation 388850 A pensity of 0.1% for Undetecteo feeoweter Permissible ventun fortmg Ltd a meast;tement unconsinty Opersson M*'*"

of 1.7% for flow are inctwdoo in inte f>gure.

98 31500M l

3g$oco......................................

(95.381150)

~*

3811$0 Restneted Operation g

Region j

(94,377300) 377300 E.g j

Profubrted Opermn (92,373450)

Regen g 373450 u

3 T-(90,369500) 369600 365750 361900 i

i 86 Os 00 92-94 96 et 100 102 Fraction of Rated Thermal Power c

McGUIRE - UNITS 1 AND 2 3/4[-24 Amendment No.

(Unit 1)

Amendment No.

(Unit 2)

f

( /

s 3/4 POWER D!,,STRIBUTION LIMITS j

i 3/4.2 AXIAL FLUX O!FFERENCE (AFM i

$ h NW L

LIMITING 'ONDITION FOR OPERATION b

s 3.2.1 The i dicated AXIAL FLUX OlFFERENCE (AFD) shall be maintained within:

a.

the allow 1 operational space as specified in the CORE OPERATING LIM $

REPORT (00t ) for RA00 operation, or b.

the target ba d specified in the COLR about tne target flux dif rence during base lo<1 operation.

APPLICABILITY:

MODE 4 above 50% of RATED THERMAL POWER *.

(Uni 2 only) l ACTION:

a.

For RA00 operat,n with the indicated AFD outsi of the limits specified in tne COLR, 1.

Either restort the indicated AfD to wi iin the COLR lim!ts within minutes, or 2.

Reduce THERMAL P ER to less than

% of RATED THERMAL POWER within 30 minutes nd reduce the ower Range Neutron Flux -

High Trip setpoints to less tha or equal to 55% of RATED THERMAL POWER withir the next

hours, b.

For base load operation abt eA with the indicated AXIAL FLUX DIFFERENCE outside of the a.

cable target band about the target flux difference:

1.

Either restore the i. dica.d AFD to within the COLR specified target band limits ithin ; minutes, or NO 2.

Reduce THERMAL WER to less\\tnan APL of RATED THERMAL POWER and discontini Base t.oad ope ation within 30 minutes.

c.

THERMAL POWER s 1 not be increase. above 50% of RATED THERMAL POWER unless t indicated AFD is wi in the limits specified in the COLR.

5ee Special est Exception 3.10.2.

ND

- APL is ne minimum allowable (nuclear design) p er level for base load eperat4 sn and is specified in the CORE OPERATING L MITS REPORT per Spec ication 6.9.1.9.

l l

McGUlRE - UNITS 1 and 2 3/4 82-1 Amendment No.; ;(Unit 1)

Amendment No.' 1(Unit 2)

E

0 3/4.3 th514UWENTATION u :'

l 3/4.3.1 REACTOR Tt!P system 1%$TRUPENTATM N MMIT! W CON 0!?!CN F01 0P(RATICW the Reactor Trip System Instrunentation channels and 3.3.1 At a minitut,3.3 4 C;,,M O shall be OPERABLE with Resp 04sf i!Nis interlock of Table as shown in TsMe 3.3 7,A.

APPLIC A8tLITY: As shown in Table 3.31/ (1.1;.

l A?lls.$

AsshowninTable3.3-K.

SURVE!LMNCE REQUIRIWENTS 4.3.1.1 Each Reactor Tri) System InstrLeontation channel and inter 1nck shall be delhonstrated OPERA 8LE by the perfortaance of the Reactor Trip $ stam Instrumitation surveillance Requirements specified in Table 4.3*

I 4.3.1..t The R(ACTOR TRIP $Y$TfM RESPON$t TIME of each Reactor trip functier shall be damer.strated to be within its limit at 1satt once per 18 months.

Each test shall include at least one train such that both trains are tested at least once par 36 months and one channel por function such that all channels are tested at least once avsty N times le months where N is the totel numb?r of redundant channels in a specific Rasctor trip function at shewn in the l

" Total No. of Channels" colunn of Table 3.3-f.

4.3.1.3 The response time of RTOs associated with the Reactor Trip Syster shall be demonstrated to be within their limits (see note 2 to Table 3.3-y) l at least once per la montns.

McGUIRE - UNITS 1 ana 2 3/4p-1 An ndmant No i

Unit 1 l

As.endu nt No Unit 2

1

.m.w num nu n:-

1 "9' r E

TABLE 3.3-pf g

REACTOR 1 RIP SYSTEM INSTRUMENTATION c

MINIMUM z

TOTAL NO.

CHANNELS CHANNELS APPLICABLE g

FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

[

1.

Manual Reactor Trip 2

1 2

1. 2 1

y 2

1 2

3*,4*,5*

10 2.

Power Range, Neutron Flux - High 4

2 3

1, 2 2

Setpoint gg, Low 4

2 3

1

,2 2

Setpoint 3.

Power Range, Neutron Flux 4

2 3

1, 2 2

High Positive Rate

,s

)(

4.

Intermediate Range, Neutron Flux 2

1 2

1

,2 3

E 5.

Source Range, Neutron Flux a.

Startup 2

1 2

2,,

4 b.

Shutdown 2

1 2

3*, 4*, 5*

10 c.

Shutdown 2

0 1

3, 4, and S 5

gg 6.

Overtemperature t.T EE Four Loop Operation 4

2 3

1, 2 6

32 Three Loop Operation

(**)

(*")

(**)

ee OO na la no

4-TABLE 3.3-14 (Centfausd) e

}

1: ;

y E

HEACIDE 1 RIP SYSIEM IIBSIMEERHATION MNHest TDTAL 39.

NLS CMLS APPLICABLE k

itBICTIONAL 13817 OF CHAISELS TO 1 RIP erfamLE sesES ACTIsti d

i 7.

Overpower WT y

E Faer Leap Operation -

4 2

3 1, 2 6

[

Three Leap Operation

(**)

8.

Presserizer Pressure-Law 4

2 3

1 6

l (ama) 9.

Pressurizer Pressure-MigGo 4

2 3

1, 2 6

{

(***)

10.

Pressurizer h ter Level--Itigh 3

2 2

1 6

i

~ P.

T 11. Lesv Geoctor Coelent Flew Single Loop (h P-8) 3/leep 2/leep in 2/leep in 1

6 1

..C O

a.

880f ePer-

- Ch W l 2:

ating leap ating leap L L'.

b.

Two Leops (Keeve P-7 and 3/leep 2/leep in 2/leep 1

6 below P-8) two aper-each oper-

2-gg 55 ating leaps ating leep e

sa ae 55

12. Steae Generater Water 4/sta. gen.

2/sta. gen.

3/sta. gem.

1, 2 6

)

i M Level-Lew-Lew it. any oyer-each p r-(***)

l'f ating sta.

ating sim.

zm

??

see.

gen.

i i'i 22 l

11 n-9...

NM e 94 e

w

S

,.t.

i M

l

~--

taste 3.3-(Continued)

- 'l '

l x

5 g

RfACTOR TRIP SYSTDI IbMNIATIDef 5

l i%

MIN 1RBI

[

IDIAL ND.

CHAW EIS 00ldSE15 APPLICABtL OF CWWest15 10 1 RIP OPUEWLE MBES AC110if 5 FlasCTICIEL 1811T

't.

13 Undervoltage-Reactor Ceciant Pumps (alwwe P-7) 4-1/bes 2

3 1

6 l

g 5

Undertic;~ V aarter Coolant B

Pusups (above P-7) 4-1/ bus 2

3 1

{

14.

m l

15. Turbine Trip Low Fluid Oil Pressure 3

2 2

3 6

)

A 4

1 1

11 a.

b.

Turbine Step valve Closure I

16.

Safety Injection Impet 2

1 2

1. 2 9

l o}

free ESF 3

U Y 17.

Reacter Trip System Interlocks me a.

Intermediate Range Newtrea flux. P-6 2

1 2

2--

8

]

l, b.

Low Power Heector 1

S Trips Block, F7 P-It Ireest 4

2 3

1 8

gg L;

.c

,[

hh P-13 Irpest 2

1 2

1 8

l.

..EE Posser Range Neutrom 4

2 3

1 8

c.

gg Re, FB d.

Low Setpoint Power lC 4

2 3

1, 2 8

Range Neutron Flux P-10 1,.

l.-

IJ3 EE forbine Impulse Ctauber 2

1 2

1 8

e.

77 Pressure, P-13 l'"

05 1.'

L l

1 4N l

i l U

g L

eS e ?.

~

h s

5 6

ga

~-

-I ~.:i ai i

gilll{""""

E n

1 l s

2

=

b S

w Mf '>

A e4 N

NN NN g

g 5

5

~B 1i e

a

.5

.E a.

.h h

b A

u E

I 5.2 W

ag A

2.

b

~

McGUIRE UNITS 1 and 2 3/4p5 Anandeont No.

(Unit 1) l Amendment No.1 (Unit 2)

(? _i

.:'.!4;;!iiiI"13.'i X1 KiLI'!:

.'i:"!)_________,

l TABLE 3.3*1f(Continued) m l

TABLt NOTAf!ON

Witn the Reactor trip Systesi creakers in the closed position. tee

, Control Rod Drive System capatle of roc withdravel.

Values lef t blank panding hRC approval of three loop operation.

Cosply with the provisions of Sp6cification 3.3.2 for any portion of tne channel required to be CPIRABLF. by $pecificatien 3.3.2.

" Below the P 6 (!ntermediets Range Neutron Fl u Interlock) Satpoint.

elow the P-10 (Low Setpoint Power Range Neutron Flu Interlock) $stpoint.

8 ACTION $ TAT (MENTS ACTION 1 + Vith the number of OPERABLE channeln one less than tbs Minimum Channels OPEEILE requirement, restore the inoperable channel to CPERIiBLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STAN0BY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 2

With the number of CPERA8LE cnannels one less than the Total Hutber of Channels, STARTUP and/or POWER OP MTION may proceso providt.d the following concitione are satisfied a.

The innperable channel is pIrced in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , The Minium Chenrels OPERABLE requirnent is set;4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months

newever, b.

the Inoperatis channel say be bypan ed for up to for surveillance testing of other channels per Specification 4.3.1.1, and to 75%. THEP)(AL POWER is restricted to less than or equal Either c.

of RATED THElHAL POWER and the Power Range Neutron Flu Trip $stpoint is reduced to less than or ocual to 85% of RATED THERMAL POW 1R within 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s: ne, the QUAD MNT POWER TILT RATIO is monitored at least once par 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months per Sp1cification 4.2.4.2.

"cGUIRE - UNITS 1 & 2 3/4$3*6 Amendment No.

(Unit 1) l A

Amendment No.

(Unit 2)

'*~

h '. I

' ;e

.n e

e

..~

1 TABLE 3.3 2 (Continued) v4i ?

l ACTION $TATENENT$ (Continued)

ACTION 3 with the nurter of channels OPERABLE one less than the Minissn Channels OPERABLE requirement and with the THERMAL P0ktR 1evel:

a.

Belev the P 6 (!ntemodiate Range heutron F1LF Interlock)

Satpoint, ice to increasing THERMAL POWER above the P 6 restore the inoperable channel to OPERABLE status pr Setpoint, and 3.

Above the P 6 (Intermediate Range kautron Flux Inter 1cck) letooint but below 105 of RATIO THEML POWER restore the inocerab): channel to CPERAILE status prior to increasing THE M L POWER above 105 of AATED THERMAL POWfR.

ACTION 4

With the number of OPERA 8LE channels one less than the Minimum Channels OPERA 8LE requirement suspend all operations involving positive reactivity changes.

ACTION 5 - With the mnbar of OPERAILE channels one less then the Minimum Channels OPERA 8LE requirement, verify compliance with the SHlfT00WN MARGM requirements 01 Specification 3.1.1.1 or 3.1.1.2, as applicable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 11 hours1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months thereafter.

ACTION 8 With the number of OPERAILE channels one less then the Total STARTUP and/or POWER OPERATION may prcceed Number of Channels,ing conditions are satisfied provided the follow

a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , ard b.

The Minisus Channels OPERABLE mqvirement is met: however, the inoperable channel may es bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels per

$pecification 4.3.1.1 and $pecification 4.3.2.1.

ACT!CN 7-Delete ACTION 8 - With less than the Minimum Number of Channels OPERAILE within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months deterstre by observation of the associated perelssive annunciator window (s) that.the interlock is in its mquired stata for the existing plant conditien, or apply Specification 3.0.3.

Amendment No.

Unit 1 l

3/4f37 Mc0UIRE UNITS 1 & 2 Amandeent No.

Unit 2 i?!

!.!!! M ii: O!h HN E h* XY !? Q I M I+!!

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

- - - - ~

1 TA8tE 3.3=11 (ContinLed) gii i-l

)

ACT!ON STATLMENTS (Continued)

ACTION 9 - With the etaber of CPERAILE charnels one less than the Hiniu channels SPERABLE requirement, to in at least HOT $ TAN 00Y within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; however, or=e channel may be bypassed fer up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing :er Specification 4.3.1.1, provided the other channel is OPELABLE.

AC'J10N 10

With th naper of-OPEulLE channels one less than the Minime Channels CPERAILE requirement, restore the inoperable channel to CPERABLE status within 44 Seurs or open the Reacter trip breakers within the next hour.

ACTION 11 With the number of OPERA 8LE channels less than the Total Neber of f.hannels, cperation may continue provided the inoperable channels are placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 12 With one of the diverse trip featums (Undervolteile or shunt trip attacheent) incperable, restore it to CPERAB,E status within 44 hours5.092593e-4 days 0.0122 hours 7.275132e-5 weeks 1.6742e-5 months or esclare the breaker inoperable and apply ACTION 9.

The breaker sha11 not be typassed while one of the diverse trip features is troperable except for the time required for performing raintenance to restore the breaker to 0PERA4LE status.

McGU1RE UNITS 1 and 2 3/4I38 AtenMnt No.1IB (Unit 1) l G

Amendment No.110 (Unit t)_

A en 4

W e

[

TABLE 3.3-7/

s.

REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES

o 5

FUNCTIONAL UNIT RESPONSE TIME

[

1.

Manual Reactor Trip t'. A.

1.S second (1) 0 2.

Power Range, Neutron Flux m

3.

Power Range, Neutron Flux, High Positive Rate N.A.

f 4.

Intermediate Range, Neutron Flux N.A.

5.

Source Range, Neutron flux N.A.

[

6.

Overtemperature AT

$10.0 seconds (1)(2) 7.

Overpower AT 110.0 seconds (1)(2)

,I e

8.

Pressurizer Pressure--Low 12.0 seconds 9.

Pressurizer Pressure--High 12.0 seconds 10.

Pressurizer Water Level--High N.A.

yp eeEE BB$$

(1) Neutron detectors are exempt from response time testing.

Response time of the neutron flux signal portion l

of the channel shall be measured from detector output or input of first electronic component in channel.

EE (2) The < 10.0 second response time includes a 6.5 second delay for the RTDs mounted in thermowells.

l 2a bh vv l

IA8t.E 3 3-N (Continued}

m Ij

.i :;,

2n KACTOR TRIP SYSTDs UtSTEISENTATION EfSPWSF.7;25 c;

m atspea u IInE hFissCTIommtumII I

3

n. taw acacter Caetant Flow a.

sime:e toep (above P-s)

<1.0 second w

b.

Tum Emeps (Abese P-7 and below P-8) 11.0seceed

,,g

12. Steam Generator Ideter level-hism

<2.0 (Whit 1), 3.5 (Unit. 2) seconds l

e.

13. Underveltage-Reacter Caelant PuuPs

<l.5 seconds i

<0.6 second I

14. tenderfrequency-Seactor Caelant Pumps

15. Turbine Trip i

,s a.

Law Fluid Oil Pressure M. A.

s-h.

Turt1 2 Step Valee Closure M.A.

y N. A.

l

16. Safety injection Input free ESF N.A.

I U.

Reacter Trip System Interlocks

=

n.A.

g Reacter Trip areakers j

g 18.

j.

19. Autamatic 1 rip and Interlock Imgic 3.A.

n z.,

n:

gFg 33

$5 s

t l

l l

l 8

W 3

b TABLE 4.3-1/

.5g REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANrE REGUIREMENTS 5

TRIP ANALOG ACTUAIING MODES FOR U

CHANNEL DEVICE milch CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE

[

FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED

~

1.

Manual Reactor Trip N.A.

N.A.

R (11)

N.A.

1, 2, 3*, 4*, S*

2.

Power Range, Neutron Flux High Setpoint S

D(2, 4),

M N.A.

N.A.

1, 2 M(3, 4),

Q(4, 6),

R(4. 5)

Lew Setpoint 5

R(4)

M N.A.

N.A.

1*,2 ms(

3.

Power Range, Neut' n Flux, N.A.

R(4)

M N.A.

N.A.

1, 2 High Positive Rate w

(

4.

Intermediate Range, S

R(4, 5)

S/U(1),M N.A.

N.A.

1

,2 Neutron Flux 5.

Source' Range, Neutron Flux 5

R(4, 5)

S/U(1),M(9)

N.A.

N. A.

2, 3, 4. S ll 6.

Overtemperature AT S

R M

N. A.

N.A.

1, 2 aa jj 7.

Overpower AT S

R M

N.A.

1, 2 ee gg 8.

Pressurizer Pressure--Low 5

R H

N.A.

1 9.

Pressurizer Pressure--High 5

R H

N.A.

I, 2 10.

Pressurizer Water Level-High 5

R H

N.A.

1

3 11.

Low Reactor Coolant Flow 5

R H

N. A.

N.A.

I UU

3:9E4 i

taste 4.3 (Contired)

V

bb N

REAC102 TRIP SYSTEM INS 1153CMTATION SMEILLANCE REUI#IREENTS E

inir ANAL 3.'

ACI! Mills seats for CML DEVICE WIOR U

OWWEL CHUSBEL SPERATIGIIAL OPERATIC 8tAL ACTWATION S WWEILLAfs[E E

OECK CAEIBilATION TEST TEST (CGIC TESI 15 Pfqp3EES

[ ftseCTIOpedd. UNIT k

12. Stema Generator Wier icvel -

5 R

M R. A.

M.A.

1. 2 I~

L e tsw

13. t>wlerweltage - Reacter Coolant N.A.

R N.A.

M

u. A.

I g

Pu'P5 N.A.

R M.A.

M M.A.

1 l

14. Underfrwesency - Beactor Coolarit Pemips g

% IA larbine Trip N. A.

R N. A.

5/5(1, 19)

N. A.

1 tow Flu!d Oil Pressere N.A.

R N.A.

5/IKL 12)

K. A.

1 a.

h b.

Turbine Step Valve Closwre

'~

N.A.

M.A.

N.A.

R M.A.

1, 2 l

l

16. Safety Injectice ( W from y

j 17.

Reach Trip System InterlocLs I

N.A.

RI4)

N M. A.

N. A.

i r o ate na t.S I

sectron Fluz, P-6 M.A.

R(4)

N (8)

!!.A.

M.A.

I C..

55 b.

Esse Power Reacter Trips Block. P-7 xx oo N

M. A.

E(4)

N (8)

N.A.

M.A.

1 C.

I F1:a,P-8 m

3.3.

gp

.S$

M

W h

~~

I 5

2 h

h k

&@ C N

d A

A Rim d

J J

J fi 53 0

=

x x

v3 x

^,

O

'd !

3

".n

$

n S$

8-W 4 b W

O W

5 5W$IE d

d 3

2

-4 x4 x

x d d

a 6

a 3

W 5

A 5

5 i

d i

~

4

_a u

u a a

u 2

h a5 a

a y

4 4 4 4

4 5

s W 4 W

W b

E 5

4 S

L u

39 y

.Y. 5~ E}. 'u

=xj[gh&

O

y a :

E g

jj j g, I_,.

I.

Ija

~

NN N

An ncment No. 1 Unit 1 l

M:GUIRE UNITS 1 and 2 3/4q313 Amendment No. 1 Unit 2 III eiU'Td9IDihiEij IN d C 'I U ~I' f

4 TAllt4.3Il(Continuse) mrtt-t-l TABLE NOT4T!g i

d Witn the Reactor Trip Systes breakers closed and the Control Rod Drive System captile of rod withdrawal.

Balw P $ (Intersediate Range Neutron Flux Interlocx) Setpoint.

N Balow P 10 (Low Setpoint Pwer Range Neutron Flux Interlock) Setpoint.

$N (1)

If not perfomed in previous 7 days.

Cotparison of calorimetric to excore power indication above 15% of (2)

RATED THERMAL POWER. Adjustexcorechannelgainscorsistentwith calorimetric pcver if absoluta difference is greater then 24. The stovisions of $pocification 4.0.4 are not applicable for entry into

?00( 2 or 1.

Single wint concarison of incore to ancore axial flux d. *fewee (3) above NE of P.ATED TliERM L POWER. Racalibrate if the abe..ute difference is greater than or equal to 35 The provisions of

$;ecification 4.0.4 are not applicable for entry into MODE 2 or 1.

heut mn detectors may be excluded from CHAhWEL CAL!lRATION.

(4)

Detector plateau curves shall be obtained evaluated, and ccapared (5) to sanufacturer's data. FortheIntermedIntaRangeandPcworRange Neutron Flux channels t% provisions of $5,eciNcation 4.0.4 are not applicabl6 for entry into 400t 2 or 1.

Incom Excere Calibratien above 75% of RATIIs THERfiAL POWER. The (6) provisions of specification,4.0.4 are not applicaole for entry into MODE 2 or 1.

Each train shall be tested at least every 62 days on a STMGERED (7)

TEST BAS!$.

Vit5 power greator than or equal to the interlock $atpoint the (8) required operational test shall consist of verifying that the interlock is in the required state by observie.? the permissivJ annunciator windo.t.

4* and 6* 6he11 also include. '

Monthly turveillance in MODI $ 38 (9) verificationthatpermissivesPkandP10areinthstrrewired state for axisting plant conditions by observation of the persis-sive annunctator window. Monthly surveillance shall include verif testion of the High Flux at Shutdown Alare Setpoint of less than or equal to five times background.

(10)-

Satooint verification is not required.

Mc0UIRE - UNITS 1 and 2 35 3-14 heendment No.1h (Unit 1)

I' heendment No,t (Unit 2) r e

b i

e

'i s

e

l TABLE 4.318(ContijgsA u.,; -;-

l TA8M MTA,JJQN, (11)-

The TRIP ACTUAT!he OtVICE GPERAT!CkA'..'PT thall independently verify i

the OPERA 4!LITY of the 4.ndervoltage ard Shunt trip circuits for the Manuel Asecter Trip Function.

(11)-

ihe TRIP ACTUAT!w Oty!CE OPERATIONAL TE$T thall independently vari'y l

the CPERA4!LITY of the urWorvoltage and thunt trip attscheents of the Reactor Trip Bresurs.

1 (13)*

Prior to piscing bretter in service, a local renual shsnt trip shall be performed.

(14)*

TPe autorative undervoltage trip capability shall be verifled operable.

6 1

I i

McGU!RE

UNITS 1 an( !

3/4)(3*14e Ame W nt No.

(Unit 1)

Aseansent No.

(Unit 2) l l

fiWil? %RD'ik.): $ $ i ? bl:

INSTRUMENTATION

b i

3/4.3.2 ENGINEERED, SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION LIM' TING CONDITION FOR OPERATION 3.L 2 The Engineered Safety Features Actuation System (ESFAS) Instrumentation channels and inter'ocks shown in Table 3.3-3 shall be OPERABLE with their Trip Setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4; (" M '.; and with RESPONSE TIMES as shown in 1able 3.3 E; r...-

O, l

AlpLICABILITY:

As shown in Tabie 3.3-3.

(

ACTION:

1 i

With'an ESFA$ Anstrumentati9n channel or interlock Trip Setpoint

-a.

less conservative than the value shown in the Allowable Values I

column of Table 3.3-4, declare the channel inoperable and apply the applicable ACTION requirement of Table 3.3-3 until the channel is restored to OPERt.BLE status with the Trip Setpoint adjust 0d consistent with the Trip Setooint value.

j b.

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

4 SURVEILLANCE REQUIREMENTS 4.3.2.1 Each ESFAS Instrumentation channel and interlock and the automatic actuation logic and relays shall be demonstrated OPERABLE by the performance of the ESFAS Instrumentation Surveillance Requirements specified in Table 4.3-2.

4.3.2.2 The rNGINEERED SAFETY FEATURES RESPONSE TIME of each E5FAS function shall be demonstrated to.be within the limit at least once per 18 months.-

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

d L

l MCGUIRE UNITS 1 AND 2_

3/4[3-15 Amendment No.

'(Unit-1) l Amendment No (Unit 1)

[

i.u 3:

r-23 TABLE 3.3-4A 5

pg ENGINEERED SAFETY FEATURES ACTUATION SYSTEM IPiiRUMENTATION TRIP SETPOINTS

!k FUNC110NAL UNIT TRIP SETPOINT.

ALLOWABLE VALUTS

_y v'

1.

Safety Injection, Reactor Trip,

(

Feedwater Isolation, Component Cooling 6"

P*

Water, Start Diesel Generators, and to Nuclear Service Water.

a.

Manual Initiation N.A.

N.A.

b.

Autematic Actuation Logic N.A.

N.A.

and Actuation Relays c.

Containment Pressure--High i 1.1 psig

$ 1.2 psig t'

i d.

Pressurizer Pressure--Low-Low 3 1845 psig

-> 1835 psig d

N a

sg e.

Steam Line Pressur< - Lcw 3 775 psig

> 755 psig 2.

Containment Spray a.

Manual Initiation N.A.

N.A.

4 l[f[

b.

Automatic Actuation Logic N.A.

N.A.

g(g[

and Actuation Relays ss l[h[

c.

Containment Pressure--High-High 1 2.9 psig i 3.0 psig 55 22 E. E.

c> r+

s

bktt-t i

x S

{E*"E ' ':"' ??

l SE TABLE 3.3-d/ (Continued) 7n' ENGINEERED SATETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS E

h3 FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES h

3.

Containment Isolation

~

a.

Phase "A" Isolation 1)

Manual Initiation N. A.

N.A.

2)

Automatic Actuation Logic N.A.

N. A.

and Actuation Relays 3)

Safety Injection See Item 1. above for all Safety Injection Trip Setpoints u,

and Allowable Values

'a s

T b.

Phase "B" Isolation l

no 1)

Manual Initiation N.A.

N.A.

2)

Automatic Actuation Logic N.A.

N.A.

I i

and Actuation Relays EE gg 3)

Containment Pressure--High-High

$ 2.9 psig 5 3.0 psig aa ll c.

Purge and Exhaust Isolation es gp jp 1)

Manual Initiation N.A.

N.A.

h(

2)

Automatic Actuation Logic N.A.

N.A.

and Actuation Relays 4

~,.

E[![

3)

Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values er OO l

n 2

t" L g

TABLE 3.3-4k(Continued)

C ENGINEERED SAFETY FEATURES ACTUATION SYSTEH INSTRUMENTATION TRIP SETPOINTS E

FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES Z

[

4.

Steam Line Isolation

[

a.

Manual Initiation N.A.

N.A.

b.

Automatic Actuation Logic N.A.

N.A.

and Actuation Relays c.

Containment Pressure--High-High

$ 2.9 psig 5 3.0 psig d.

Negative Steam Line

$ 100 psi with a

-< 120 psi with a Pressure Rate - High rate / lag function ate / lag function T

time constant time constant 1 50 seconds 1 50 seconds g

e.

Steam Line Pressure - Low 1 775 psig 1 755 psig

[

5.

Turbine Trip and Feedwater Iso'ation a.

Automatic Actuation Logic N.A.

N.A.

: . 2:-

and Actuation Relays Po S k@R b.

Steam Ger.erator Water level--

< 82% of narrow range

< 83% of narrow rance High-High (P-14)

Instrument span each ste:m Instrument span each steam generator generator zz PP c.

Doghouse Water Level-High 12" 13" (Feedwater Isolation Only) 6.

Containment Pressure Control System mm

="

ao

%?

Start Permissive / Termination 0.3 < SP/T < 0.4 PSIG gg (SP/7)

~

0.25 < SP/T < 0.45 PSIG

d r

imi i r 9,.

(E"X "i "" Z) 5 iABLE3.3-44:(Continued)

Al ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOIN15

.E 3

FUNCTIONAL UNIT TRIP SETPOINT Alt 0WABLE VALUES w

7.

Auxiliary Feedwater ro a.

Manual Initiation N.A.

N.A.

b.

Automatic Actuation' Logic N.A.

N.A.

and Actuation Relays

}

c.

Steam Generator

. Water Level--Low-Low 1)

. Start Motor-Driven Pumps

> 12% of span from 0 to

> 11% of span from 0 to R

30% of RATED THERMAL POWER, 30% of RATED THERMAL POWER, increasing linearly to increasing linearly to

> 40.0% of span at 100%

> 39.0% of span at 100%

T if RATED THERMAL POWER.

of RATED THERMAL POWER.

\\

2)

Start Turbine-Driven Pumps 12% of span from 0 to

> 11% of span from 0 to 30% of RATED' THERMAL POWER, 30% of RATED THEAMAL POWER, increasing linearly to increasing linearly to

> 40.0% of span at 100%

> 39.0% of span at 100%

EF Hf RATED THERMAL POWER.

of RATED THERMAL POWER.

em EE d.

Auxiliary Feedwater

-> 2 psig

~> 1 psig 52 Suction Pressure - Low AA (Suction Supply Automatic gg Realignment) e.

Safety Injaction -

See Item 1. above for all Safety Injection Trip Setpoints Start Motor-Driven Pumps and Allowable Values o

f.

Station ' Blackout - Start 3464 1 173 volts with a

> 3200 volts

^^

EE Motor-Driven' Pumps and 8.5 1 0.5 second time j

33 Turbine-Driven Pump delay 3

(Note.1)-

g.

Trip of Main Feedwater Pumps -

N.A.

N. A.

f Start Motor-Driven Pumps s

,h b[; [ ^.'., 2^^P "

WE; LTABLE'3.3-4[(Continued)'

x:

ENCINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS

.t g:

t

.]~

FUNCTIONAL UNIT'-

TRIP'SETPOINT ALLOWABLE VALUES 8.'

Automatic Switchover to Recirculation-e

"~

90. inches

-> 80 inches RWST Level o

9.

Loss of. Power

'4 kV Emergency Bus,Undervoltage-'

3464 1 173 volts with a

~> 3200 volts Grid Degraded Voltage 8.5 1 0.5 second time l

delay t

i R

10.- Engineered Safety Features Actuation System Interlocks N

e

-< 1955'psig

-< 1965 psig 4

a.

Pressurizer Pressure,'P-11

>'553 F

'> 551*F T,yg, P-12 b.

c.

Reactor' Trip, P-4 N.A.

N.A.

]

22 d.

Steam Generator Level, P-14 See Item 5. above for all Trip Setpoints and Allowable RR Values.

RR 5S Note 1:

The turbine' driven pump will not start on a. blackout signal coincident with a safety injection signal.

sk M

.i 22 i

1 3.

t

<+ o h

r b

.: j

- ~,. -

i TABLE 3.3-5)

\\

ENGINEERED SAFETY FEATURES RESPONSE TIMES

, INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONOS 1.

Manual a.

Safety Injection (ECCS)

N.A.

b.

Containment Spray N.A.

c.

Containment Isolation Phase "A" Isolation N.A.

Phase "B" Isolation N.A.

Purge and Exhaust Isolation N.A.

d.

Steam Line Isolation N.A.

e.

Feedwater Isolation N.A.

f.

Auxiliary Feedwater N. A.

g, Nuclear Service Water N.A.

h.

Component Cooling Water N.A.

i.

Reactor Trip (from SI)

N.A.

j.

Start Diesel Generators N.A.

2.

Containment Pressure-High a.

Safety Injection (ECCS) 1 27(1) b.

Reactor Trip (from SI)

$2 c.

Feedwater Isolation i 12 4

d.

Containment Isolation-Phase "A"(2)

$ 18(3)/28(4) e.

Containment Purge and Exhaust Isolation

<4 f.

Auxiliary Feedwater(5)

N.A.

g.

Nuclear Service Water 1 65(3)/76(4) h.

Component Cooling Water 1 65(3)/76(4) 1.

Start Diesel Generators 1 11 McGUIRE - UNITS 1 AND 2 3/4[3-30 Amendment No.1 (Unit 1)

{

Amendment No.

I (Unit 2) i

l m

1 TABLE 3.3-Ek(Continued) l q

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATIgG SIGNAL AND FUNCTION RESPONSE TIME IN SECONOS 3.

Pressurizer Pret sure-Low-Low a.

Safety Injection (ECCS) 1 27(1)/12(3) b.

Reactor Trip (from SI)

<2

~

c.

Feedwater Isolation 1 12 g

d.

Containment Isolation-Phase "A"(2) e.

Containment Purge and Exhaust Isolation

~< 18(3)/28(4)

<4 f.

Auxiliary Feedwater(5) h.A.

g.

Nuclear Service Water System

< 76(1)/65(3) h.

Component Cooling Water 1 76(3)/65(3) i.

Start Diesel Generators

_ 11 4.

Steam Line Pressure-Low a.

Safety Injection (ECCS) i 12(3)/22(4) b.

Reactor Trip (from SI)

<2

~

c.

Feedwater Isolation

' 12 5

d.

Containment Isolation-Phase "A"(2) i 18(3)/28(4) e.

Containment Purge and Exhaust Isolation

<4 f.

Auxiliary Feedwater(

h.A.

g.

Nuclear Service Water 1 65(3)/76(4) h.

Steam Line Isolation i 10 4

i.

Component Cooling Water 1 6b(3)/76(4) j.

Start Diesel Generators

_ 11 5.

Containment Pressure-High-High a.

Containment Spray

,45 b.

Containment Isolation-Phase "B" N. A.

c.

Steam Line Isolation i 10 4

6.

Steam Generator Water Level-High-High a.

Turbine Trip N. A.

b.

Feedwater Isolation 5 12 i

McGUIRE - UNITS 1 AND 2 3/4[3-31 Amendment No. V4 (Unit 1)

Amendment No. /\\0 (Unit 2)

s se LWP f

TABLE 3.3-5)(Continued)

I ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 7.

S_ team Generator Water Level Lowlow a.

Motor-driven Auxiliary Feedwater Pumps 5 60 b.

Turbine-driven Auxiliary Feedwater Pumps 5 60 8.

Negative Steam Line Pressure Rate - Hioj Steam Line Isolation 5 10 9.

Start Permissive Containment Pressure Control System N.A.

10.

Termination Containment Pressure ;ontrol System N.A.

11.

Auxiliary Feedwater Suction Pressure - Low Auxiliary Feedwater Pumps (Suction Supply Automatic Realignment) 5 13 12.

RWST Level Automatic Switchover to Recirculation 5 60 13.

Station Blackout a.

Start Motor-Driven Auxiliary Feedwater Pumps 1 60 b.

Start Turbine-Driven Auxiliary Fet:dwater Pump (6) 5 60 14.

Trip of Main Feedwater Pumps Start Motor-Driven Auxiliary Feedwater Pumps 1 60 15.

Loss of Power 4 kV Emergency Bus Undervoltage-

< 11 Grid Degraded Voltage i

McGUIRE - UNITS 1 AND 2 3/4I3-32 Amendment No.

(Unit 1)

Amendment No.

(Unit 2)

t 4

TA9tf3.3-k(Continued)

TABLE NOTATION (1) Diesel generater starting and sequence leading delays included. Response tir.e 11 sit 1,cludes opening of valves to establish safety Injection path and attainment of discharge pressure for centrifugal charging cusps, Safety injection and RNA pumps.

(2) Valves 1KC3C5B and 1KC3158 for Unit 1 and velves 2KC305B and 2KC315B for Unit 2 are exceptions to the response times listed in the table. The following response times in seconds are the required values for these valves for the initiating signal and function indicated:

<35f3f/40(4) 2.d 3

7 30

3. d

4. d i30(3)/40(4)

(3) Diesel generator starting and sequence loading delays nq.) included.

Offsite power available. Response time limit includes opening of valves to establish Safety Injection cath and attainment of discharge pressure for centrifugal charging piaps and Safety Injection pumps.

5 (4) Diesel geneaator starting and sequence loading delays included. Response-time limit includes opening of valves to establish Safety Injection path and attainment cf discharge pressLre for centrifugal charging pumps and SafetyInjectionpumps.

(5) Response time for motor-driven auxiliary feedwater pumps on all safety Injection signal shall be less than or equal to 60 seconds.

Response time limit includes opening of valves to establish Safety Injection patn and attainment of discharge pressure for auxilis y feedwater pcass.

(6) The turbine driven pump does not start on a blackout signal coincident witn : safety injection signal.

McQJ1RE - UNITS 1 and 2 1/4/3'33 Amencmant No. \\?g (Unit 1)

{

kuencment No. Q0 (Unit 2)

/

e d

W

b m

l 3/4. 3____IWSTRUMERTAT!0N 3#.3.1 Rf. ACTOR TRIP SYST(M INSTRU,M(NTATION L!N 3 CONDITION FOR OPERATION

/

3.3.1 As minimuin, the )leactor Trip Systes Instruesntation channel ano interlocks Table 3.31gW.SOshallbeOPERABLEwithRESPONS IFES as l

3.32j.

shown in Tab APPLICA8:LITY:

s shown in Table 3.3

($ ". 21 g

ACTION:

As shown in Table 3. -k.

g SURVEILLANCE REQUIREME

\\

I 4.3.1.1 Each Reactor Trip stem Instrumentation nnel end interlock shall be demonstrated OPERABLE by a performance of tn Reactor Trip Sy_ stem 6

Instrumentation Surveillance quirerents specif inTable4.33 l

4.3.1.2 The REACTOR TRIP SYSTEM ESPONSE TIME f each Reactor trip function shall be deeanstrated to be within its limit least once per 18 months.

Each test shall incluce at least c train ch that both trains are tested at least once per 36 months and one cha. el p function such that all channels are testen at least once every N times enths where N is the total nuster of redundant channels in a specific Rea or trip function as shown in the

" Total No. of Channals" column of Tabl 3 %.

l

4. 3.1. 3 The response time of RTDs asociat with the Reactoe Trip System chall be demonstrated to be withi their lie s(seenote2toTabis3.3-2l)at g

least once per 18 months.

/

McGUIRE - UNITS 1 and 2 3/4$3*1 Amendtant No.128 (Unit 1) l Amendment Nc,110 (Unit 2)

'?!

W i ;iti: i.i u 'UI M.'li U ;?'IC IHI#

4 LTABLE 3.3-lb UhT 2 f

O REACTOR TRIP SYSTEM' INSTRUMENTATION

.[

E MINIMUM TOTAL NO.

CHANNELS CHANNELS APPL BLE cz FUNCTIONA NIT' 0F CHANNELS TO TRIP OPERAB'_E J DES ACTION

-5

[

1.

Manual Reac Trip' 2

1 2

1, ?

I 2

1 2

3*, 4 *, 5*

10 2.

Power Range, Neutron F

- - High 4

2 1, 2 2

i tpoint low 4

2 3

1

,2 2

Setpoin 3.

Power Range, Neutron Flux 4

3 1, 2 2

High Positive Rate

{

4.

Power Range, Neutron Flux, 4

2 3-1, 2 2

High Negative Rate 5.

Intermediate Range, Neutron Flux 2

1 2

1

,2 3

6.

Source Range, Neutron Flux y

a.

Startup 2

1 2

2,,

4 b.

Shutdown 2

1 2

4*, 5*

10 c.

Shutdown 2

0 1

3, and 5 5

BB

((

7.

Overtemp ture.tT r

r Loop Operation 4

2 3

1, 2 6

hree Loop Operation

(**)

- ) @

. ? ?'

.\\

Y ou 22 E. B.

e c>

A A

f TABLE'3.3-2b-l T2

- 2 8-5-

REACTOR TRIP SYSTEM INST'!UMENTATION RESPONSE TIMES

.A

[

' FUNCTIONAL UNIT' RESPONSE TIME g

1.

Manual Reactor Trip N.A.

2.

Power Range, Neutron Flux

10. 5 < ond (1)

{

ro 3.

Power-Range, Neutron Flux, High Positive Rate N.A.

4.

. Power' Range,-Neutron Flux, P

High Negative Rate 10.5 second-(1) 5.

Intermediate Range, Neutron Flux-N.A.

6.

Source Range, Neutron. Flux N. A,

' y 7.

Overtemperature AT 110.0 seconds (1)(2) m g

t 8.

Overpower AT 510.0 seconds (J)(2) g.

l 9.

Pressurizer Pressure--L

$2.0 seconds

[

10.

Pressurizer Press.

--High 12.0 seconds E,a

{ k 11 Pressurizer te

' --H i gh N.A.

3 <3 l

aa zz (1) ron detectors are exempt from response tisne testing.

Response time of the freutron flux signal portion

?P of the-channel shall be measured from detector output or input of first electronic component in channel.

CC

2) The < -10.0 second response time includes a 6.5 second delay for the RTOs mounted in thermowells.

c,.

CC S. E i

ec Nb vv i

I L

+

l TABLE 4.3-lb Ut 2

2 l

85 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS

l TRIP ANALOG ACTUATING MODES FOR c-Z CHANNEL DEVICE WHICH d

CHANNEL CHANNEL OPERATIONAL OPERATIO tL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED

~

o.

1.

Manual Reactor Trip N.A.

N.A.

R (11)

N.A.

1, 2, 3 *, 4 *, S

m 2.

Power Range, Neutron Flux High Setpoint S

D(2, 4),

M N.A.

N.A.

1, 2 M(3, 4),

Q(4, 6),

R(4. 5)

Low Setpoint 5

R(4)

M N.A.

N.A.

1 2

3.

Power Range, Neutron Flux, N.A.

R(4)

M N.A.

N.A.

1, 2 High Positive Rate w

h 4.

Power Range, Neutron Flux, N.A.

R(4)

M N.A.

N.A.

1, 2 High Negative Rate 5.

Intermediate Range, S

R(4, 5)

S/U(1),M N.A.

N.A.

1

,2 Neutron Flux 1

[ k 6.

Source Range, Neutron Flu 5

R(4, 5)

S/U(1),M(9)

N.A.

2

, 3, 4, 5 7.

Overtemperature AT S

R H

N.A.

N. A.

1, 2 A

j

,aa 8.

Overpower AT S

R H

N.A.

1, 2 zz 6

ee 3g 9.

Pressuri_

Pressure--Low 5

R H

N.A.

1 W

g--

D 1, 2 @ --- p s

om

}

gg 10.

Pre rizer Pressure--High 5

R H

N.A.

D

3. 3.

1 Pressurizer Water Level--High 5

R M

N. A.

I uw 12.

Low Reactor Coolant Flow S

R H

N.A.

1

. -.. _. ~ -

. ~. _ -.

INSTRUMENTATION UNIT 2 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATipd LIMITING CONDITION FOR OPERATION s

1 3.3.2 he Engineered Safety Features Actuation System (E AS) Instrumentation channels nd interlocks shown in Table 3.3-3 shall be OP ABLE with their Trip Setpoints t consistent with the values shown in the T ip Setpoint column of.

Table 3.3-4b nit 2) and with RESPONSE TIMES as show in Table 3.3-5b (Unit 2).

l APPLICABILITY:

shown in Table 3.3-3.

(Unit 2 o y)

ACTION:

a.

With an ESFAS trumentation channel r interlock Trip Setpoint less conservative han the value sho in the Allowable Values column of Table 3.

, declare the tannel inoperable and apply the applicable ACTION re irement of T le 3.3-3 until the channel is restored to OPERABLE s tus with e Trip Setpoint adjusted consistent with the Trip etpoi value.

J b.

With an ESFAS Instrumentatio hannel or interlock inoperable, take the ACTION shown in Table 3.

SURVEILLANCE REQUIRFMENTS 4.3.2.1-Each ESFAS Instrumenta on channel and inte ock and the automatic actuation logic and relays sha be demonstrated OPERA E by the performance of the ESFAS Instrumentation rveillance Requirements s cified in Table 4.3-2.

4.3.2.2 The ENGINEERED S ETY FEATURES RESPONSE-TIME of each SFAS function shallibe demonstrated to e within.the limit at least once per r' months.

Each test-shall include at least one train such that both trains-e tested at

- least once per 36 mont s and one channel per function such that al hannels are tested at least a ce per N times 18 months where N is the total mber of redundant channels i a specific'ESFAS function as shown in the " Total

o. of Channels" column of able 3.3-3.

MCGUIRE - UNITS 1 & 2 3/4 B3-15 Amendment No.128 (Unit 1)

Amendment No.110 (Unit 2)

'+,

- - r+ rr w

e e ww-e

=

+rw,

.w

-re ar

-me4-e-*-=u w

' - < - =

~

'4

_z

. TABLE 3.3-4b UNIT 2' I

o ENGINEERED' SAFETY FEA10RES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINT m

FUNCTIONAL UNIT TRIP 5ETPOINT ALLOWABL ALUES i-*

1.

. Safety Injection, Reactor Trip,

' '^

'feedwater Isolation, Component Cooling

-Water, Start Diesel Generators, ad L

Nuclear Service Water.

.ro a.

Manual Initiation' N.A.

N. A.

b.

Aetomatic: Actuation Logic N.A.

N.A.

ar.d Actuat. ion Relays c

Containment Pressure--High 51 psig 5 1.2 psig d.

Pressurizer Pressure--Low-Low 3 145 psig 3 835 psig

.m w

e.

Steam Line Pressure - Low 3 585 psig 1 565 psig m

A o'

2.

Containment Spray a.

.Mancal I.titiation-N. A.

N.A.

\\

b.

Automatic Act ion looic N.A.

N.A.

\\.

gy and Actuat*

Relays U

gg c.

Cont ent Pressure--High-High 5 2.9 psig 5 3.0 psig aa r-

?

1 ^.. /

L.S f

LA.

wa t

.?2 A

e.t:

(>

h

J 4' gg 3.3'o 5 b6L6<QUN T 2 TABLE 3.3 5b l

E_NGINEERED SAFETY FEATURES RESPONSE TIMES s

INIT TING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 1,

Ma 1

a.

fety injection (ECCS)

N.A.

b.

Con inment Spray N. A.

c.

Contai ent Isolation Phase "A Isolation N. A.

Phase "B" solation N.A.

Purge and E aust Isolation N. A.

d.

Steam Line Iso tion N.A.

e.

Feedwater Isolati N.A.

f.

Auxiliary Feedwater N.A.

g.

Nuclear Service Water N.A.

h.

Cornponent Cooling Water N.A.

i.

Reactor Trip (from SI)

N.A.

j.

Start Diesel Generators N.A.

2.

Containment Pressure-High a.

Safety injection (ECCS) 1 27(1) b.

Reactor Trip (from SI) 12 c.

Feedwater Isolation

<9 d.

ContainmentIsolation-Phase"A"(2)

18(3)/28(4) e.

Containment Purge and Exhaust Isolation 4

s f.

Auxiliary Feedwater(5)

N.

g.

Nucleat Service Water

< 65 )/76(4) h.

Component Cooling Water 1 65(3 6(4) i.

Start Diesel Generators-1 11 McGUIRE - UNITS 1 & 2 3/4 B3-30 Amendment No.128 Unit 1 Amendment No.110 Unit 2

PLANT SYSTEMS

,AIN STEAM LINE ISOLATION VALVES u n 3. -v.

M LIMITING CONDITION FOR OPERATION 3.7.1.4 Each main steam line isolation va'.ve (MSLIV) shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3. M U - " ' _ 1, )-

ACTION:

MODE 1 - With one MSLIV inoperable but open, POWER OPERATION may continue provided the inoperable valve is restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; otherwise, reduce power to less than or equal to 5% of RATED THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

MODES 2 - With one MSLIV inope able, subsequant operation in MODE 2 or 3 may and 3 proceed provided:

a.

The isolation valve is maintained closed, and b.

The provisions of Specification 3.0.4 are not applicable.

Otherwise, be in HOT STAN0BY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUT 00WN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.7.1.4 Each MSLIV shall be demonstrated OPERABLE by verifying full closure within 8 seconds when tested pursuant to Specification 4.0.5.

'$g 3/4f7-8 Amendment No.

(Unit 1)

McGUIRE - UNITS 1 AND 2 Amendr..cnt No (Unit 2)

kINSTEAMLINEISOLATIONVALVES UNIT 2 LI NG CONDITION FOR OPERATION s

i

\\

3.7.1.4

. ch main steam line isolation valve (MSLIV shall be OPERABLE.

APPLICABILIT MODES 1, 2, and 3, (Unit 0 --';)

{

ACTION:

MODE 1 - With one S'.IV incperable but open, 0WER OPERATION may continue provided t e inoperable valve is re tored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; ot rwise, reduce power less than or equal to 5% of RATED THERMAL OWER within 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ..

MODES 2 - With one MSLlv i operable, subs.quent operation in MODE 2 or 3 may and 3 proceed provided:

a.

The isolation ve is aintained closed, and b.

The provisions of 'pe fication 3.0.4 are not applicable.

Otherwise, be in HOT STAF BY within the next 6 bcurs and in HOT SHUTDOWN within the fol' w g 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS F

\\

4.7.1.4 Each MSLIV shal be demonstrated OPERA E by verifyi,79 full closure within 5 seconds when t ted pursuant to Specific < tion 4.0.5.

MCGU

- UNITS 1 and 2 3/4 B7-8 Amendment No.126 (U 't 1) l Amendment No.110 (Unit 2)

3 i

4 ~

2.1 SAFETY LIMITS M ;=

l BASES 2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible _cladoing perforation which would result in the release of fission products to the reactor coolant.

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

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

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

DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and reactor coolant temperature and pressure have been related to DNB.

This relation has been developed to predict the DNB flux and the location of DNB for axially uniform and nonuniform heat flux distributions.

The local DNB heat flux ratio-(DNBR), defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is

. indicative of the margin to DNB.

The-DNB design basis is as follows:

there must be at least a 95% proba-bility that the minimum DNBR of the limiting rod during Condition I and II events is greater than.or equal to the DNBR limit of the DNB correlation being used (the BWCMV correlation in this application). -The correlation DNBR' set

(

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

In meeting this design basis, uncertainties-in plant operating parameters, nuclear and thermal parameters, fuel fabrication parameters, and the CHF cor-h relation are consioered statistically such that there is at least a 95% con-fidence that the minimum DNBR for the limiting rod is greater than or equal to the DNBR limit.

The combined DNBR uncertainty is used.to establish a design DNBR value-which must be met in plant safety analyses using values of input

_ h u

parameters without uncertainties.

The curves of Figure 2.1-1 show the-loci of points of THERMAL POWER, Reactor Coolant System pressure, and average temperature below which the calculated DNBR is no less than the-design-DNBR.value or the average enthalpy at the vessel exit

-is less than the enthalpy of saturated liquid.

N The curves are based on a nuclear enthalpy rise not channel factor, F

, of 1.50 and a-reference cosine axial power shape with a peak of 1.55.

An allow-k

--anceisincludedforanincreaseinFfg at reduced power based on the expression:

F

= 1.50 [1 + (1/RRH) (1-P)]

i g-Where P is the fraction of RATED THERMAL POWER, and RRH is given in the COLR.

9 g

McGUIRE - UNITS 1 AND 2 B 2-1 Amendment No.1 (Unit 1) lendment No.

(Unit 2)

LIMITING 5AFETY SYSTEM SETTINGS MrrF9P BASES Power Range, Neutron Flux (Continued)

The Low Setooint trip may be manually blocked above P-10 (a power level of approximately 10% of RATED THERMAL POWER) and is automatically reinstated below the P-10 Setpoint.

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

Intermediate and Source Range, Neutron Flux The Intermediate and Source Range, Neutron Flux trips provide core protection during reactor startup to mitigate the consequences of an uncon-trolled rod cluster control assembly bank withdrawal from a subcritical condition.

These trips provide redundant protection to the Low Setpoint trip i

of the Power Range, Neutron Flux chgnnels.

The Source Range channels will initiate a Reactor trip at about 10 5 counts per second unless manually blocked when P-6 becomes active.

The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of l

RATED THERMAL POWER unless manually blocked when P-10 becomes active.

l l

l McGUIRE - UNITS 1 and 2 B/2-4 Amendment No.

(Unit 1)

(

Amendment No.

(Unit 2) v

l l

2.1 SAFETY LIMITS UNIT 2 l

BASES j

f 2 1.1 REACTOR CORE The restrictions of this Safety Limit preverit overheatin of the fuel and ossible cladding perforation which would result in the rele se of fission pr ucts to the reactor coolant.

Overheating of the fuel c 4dding is prev ted by restricting fuel operation to within the nuci. ate boiling regime where e heat transfer coefficient is large and the cle ing surface temperat e is slightly above the coolant saturation terr erature, Operat above the upper boundary of the nucleat boiling regime could result in exc ive cladding temperatures because of t e onset of departure from nacleate bu' ling (DNB) sr.d the resultant sharp r duction in heat transfer coefficient.

ONB not a directly measurable param er during operation and therefore THERMAL P ' R and reactor coolant tempera re and pressure have been related to DNB, This ' elation has been developed t predict the DNB flux and the location of DNB for xially uniform and nonuni orm heat flux distributions.

The local DNB heat flux r io (DNBR), defined as

e ratio of the heat flux that would cause DNB at a pa ticular core locati to the local heat flux, is indicative of the margin to DN The DNB design basis is as fo lows:

ther. must be at least a 95% proba-bility that the minimum DNBR of the imiting od during Condition I and 11 events is greater than or equal to th DNBR imit of the DNB correlation being used (the WRB-1 correlation in this app c. ion).

The correlation DNBR set such that there is a 95% probability with 5% confidence that DNB will not occur when the minimum DNBR is at the DN imit.

In meeting this design basis, unc.tainti in plant operating parameters, nuclear and thermal parameters, and f el fabrica on parameters are considered statistically such that there is at east a 95% ce fidence that the minimum DNBR for the limiting rod is greater th or equal to th. DNBR limit.

The uncertain-ties in the above plant parameter < are used to determ e the plant DNBR uncer-

tainty, This DN3R uncertainty, ombined with the corre ation DNBR limit, establishes a design DNBR valu which must be met in pla safety analyses using values of kput paramet s without uncertainties.

The curves of Figure 2

-1 show the loci of points of T MAL POWER, Reactor Coolant System pressure, d average temperature below which t calculated DNBR is no less than the desi DNBR value or the average enthalpy at the vessel exit is less than the enthal of saturated liquid.

The curves are b sed on a nuclear enthalpy rise hot channel fa +or, F q, of 1.49 and a referen cosine with a peak of 1.55 for axial power shape.

An allow ance is included or an increase in F at reduced power based on the e ression:

H N

F

= 1,

[1 + 0.3 (1-P)]

Where P is the fraction of RATED THERMAL POWER.

McGUIRE - UNITS 1 and 2 B B2-1 Amendment No.128 (Unit 1)

)

Amendment No.110 (Unit 2) l

- _ _ _ - _ - _ _ _ _ - _ - - _ _ - _ _ - - - - _ _ - _ _ = - _

-k 4

LIMITING SAFETY SYSTEM SETTINGS UNIT 2 l

BASES Pow Range, Neutron Flux (Continued)

-T e Low Setpoint trip may be manually blocked above P-10 (a power level of appr ximately 10% of RATED THERMAL POWER) and is automatical'y reinstated below th P-10 Setpoint, Power Range Neutron Flux, High Rates The Powe Range Positive Rate trip provides protecti n against rapid flux increases which are characteristic of rod ejection even from any power level.

Specifically, th trip complements the Power Range Ne ron Flux High and Low trips to ensure t t the criteria are met for rod eje ion from partial power,

.The Power Rang Negative Rate trip provides pr ection for control rod drop accidents. -At

' gh power, a rod drop acciden of a single or multiple.

rods could cause loca flux peaking which could c se an uncor.servative local DNBR'to exist, The Po r Range Negative Rate tr' will prevent this from occurring by-tripping t

reactor, No credit i taken for operation of the Power Range Negative Rate trip for those contr rod drop accidents for which DNBR's will be greater tha the design limit BR value.

Intermediate and Source Range Neutron Flu The Interme'iate and Sourc Range,- utron Flux trips provide core d

protection.during-reactor startu to mi gate the consequences of.an uncon-trolled rod cluster control assem yb k withdrawal from a subtritical condition.

These trips provide re n nt protection to the Low Setpoint trip of the Power Range, Neutron-Flux chg.els.

The Source Range channels will L

initiate a Reactor trip at about 10 counts per second unless manually blocked when.P-6 becomes active, e

termediate. Range channels will initiate a Reactor trip at a curr t le 1 equivalent to approximately 25% of

. RATED THERMAL POWER unless manua y bloc d when P-10 becomes active.

i i.

i I-McGUIRE - UNITS 1 and 2 B B2-4 Amendment No.128(Unit 1)

Amendment No.110(Unit 2)

. -. -. - ~.

o l

3/4.2 POWER DISTRIBUTION LIMITS es f

i BASES The_ specifications _of this section' provide assurance of fuel integrity during Condition I (Normal Operation) and II -(Incidents of Moderate Frequency) events by:

(1) maintaining the calculated DNBR in the core at or above the design limit during normal operation and in short-term transients, and. (2) limiting the fission gas release, fuel pellet temperature, and cladding mechanical prop-erties 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 are not exceeded.

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

F (X,Y,Z) Heat Flux Hot Channel Factor, is defined as the local heat flux g

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

-F g(X,Y) Nuclear Entha':py Rise Hot Channel Factor, is d6 fined as the r

. ratio of the integral of linear ;ower along a rod at core

d location X,Y to thL average rod power.

Y K(:) is defined-as the normalized F (X,Y,Z) limit for a given core height.

g

- 3 /4.' 2.1

-AXIAL FLUX OIFFERENCE-The limits on AXIAL FLUX DIFFERENCE (AFD) ensure that F (X,Y,Z) and 7

n F u(X,Y) limits-specified ln the CORE OPERATING LIMITS REPORT (COLR) are not

-ekBeededduringeither' normal-operationorinthe-eventofxenonredistribution

(

i following power p anges.

The.AFD envelop specified in the C01.R has been h

adjustedformeasur,'mentuncertainty.

I

%($

1 t-L l

L McGUIRE - UNITS 1 AND 2 B3/432-1 Amendment No.

(Unit 1)

Amendment No.

(Unit 2)

I

=A i

i.

POWER DISTRIBUTION LIMITS pp, BASES AXIAL FLUX OlFFERENCE (Continued) 3 k<

The computer detertaines the one minute average of each of the OPERABL E excore detector outputs and provides an alarm message immediately if the AFD for at least ? of 4 or 2 of 3 OPERABLE excore channels are outside the allowed I

al power operating space during normal power operation.

These alarms are active when power is greater than 50% of RATED THERMAL POWER, t

McGUIRE - UNITS 1 AND 2 B3/4[2-2 Amendment No.

?-

(Unit 1)

Amendment Nog 3 (Unit 2)

3 POWER DISTRIBUTION LIMITS l

Wr BASES 3/4.2.2 and 3/4.2.3 HEA1 FLUX HOT CHANNEL FACTOR, and RCS FLOW RATE AND i

NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR b., 3 l

The limits on heat flux hot channel factor, and nuclear enthalpy rise hot channel factor ensure that:

(1) the design limits on peak local power density and minimum DNBR are not exceeded, and (2) in the event of a LOCA the ECCS i

acceptance criteria are not exceeded.

The peaking limits are specified in the CORE OPERATING LIMITS REl' ORT (COLR) per Specification 6.9.1.9.

The heat flux hot etannel factor and nuclear enthalpy rise hot channel i

factor are each measurablto but will normally only be deterniined periodically I

as specified in Specifications 4.2.2 and 4.2.3.

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

a.

Control rods in a single group move together with no individual rod insertion differing by more than + 12 steps, indicated, from'the group demand position; 4

-b.

Control rod groups-are sequenced with overlapping groups as described in Specification 3.1.'3.6; c.

The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and d.

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

F3g(X,Y)will'bemaintainedwithinitslimitsprovidedConditionsa.thrt.ughd.3 above are maintained.

The limits on the nuclear enthalpy rise hot channel factor, F-f arespecifiedintheCOLRasMaximumAllowableRadialPeaking(MARd(X,Y),

limits,

- l obtained by dividing the Maximum Allowable Total: Peaking-(MAP) limit.by the-

{

axial peak [ AXIAL (X,Y)] for location (X,Y).

By definition, the Maximum Allow-able Radial Peaking limits will result in a DNBR fnr the limiting transient i

tht'is equivalent to the DNBR calculated with a design F f

ano a limiting reference axial power shape.

Fortransitku(X,Y)valueof1.50 0n cores, MARP limits may be applied to both MARK-BW and optimized fuel-types provided allowances for differences in DNBR are accounted for in the generation of MARP limits.

The MARP. limits specified in the COLR include allowances-for mixed core DNPR ef fects.

The relaxation'of F the radial pwoer shSpe(X,Y)-as a function of THERMAL POWER allows for a change i y

for-all permissible control bank insertion limits.

This relaxation is implemented by the application of the following-factors:

k = [1 + (1/RRH)-(1 - P)]

-where k = power factor multiplier applied to-the MAP limits p = THERMAL POWER / RATED THERMAL POWER RRH is given in the COLR McGUIRE'- UNITS 1 AND 2 B 3/4 2-3 Amendment No.

(Unit 1)

Amendment No.

(Unit 2) 4

POWER DISTRIBUTION LIMITS

' v41 %

{

BASES HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)

The hot channel factor F"(X,Y,Z), and the nuclear enthalpy rise hot channel f actor, F"H(X,Y), are measured periodically to verify that the core is operating as designed.

F"(X,Y,Z) and F H (X,Y) are compared to allowable i

limits te provide reasonable assurance thet limiting criteria will not be exceeded for operation within the Technical Specification limits of Sections 2.2 (Limiting Safety Systems Settings), 3.1.3 (Movable Control Assemblies),

j 3.2.1 (Axial Flux Dif ference), and 3.2.4 (Quadrant Power Tilt Ratio).

A I

peaking margin calculation is performed to provide the batis for decreasing h

the width of the AFD and f(AI) limits and for reducing THERMAL POWER.

M When an F (X,Y,Z) measurement is obtained from a full-core map in accordance with surveillance requirements of Specification 4.2.2, no uncertainties are applied to the measured peak since a measurement uncertainty of 5.0% and a manufacturing tolerance of 3.0% are included in the peaking limit.

When F"(X,Y,Z) is measured for reasons other than meeting the requirements of Specification 4.2.2, the mea;ured peak is increased by the radial-local peaking factor and appropriate allowances for measurement uncerteinty and for i

manufacturing tolerances.

When an F H(X,Y) measurement i-obtained from a full-core map, regardless of

)

the reason, no uncertainties are applied to the measured peak since the required uncertainties are included in the peaking limit.

i B3/4f2-4 Amendment No.1 - (Unit 1)

McGUIRE - UNITS 1 AND 2 Amendment No. ',1(Unit 2) l

J lc:

POWER'0I5TRIBUTION LIMITS w

BASES 3/4.2.4 QUADRANT POWER TILT RATI_0 The QUADRANT' POWER TILT RATIO limit assures-that the radial power distri-bution~ satisfies the design values used in the power capability analysis.

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

The limit of 1.02, at which corrective action is required provides DNB p

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

)

The peaking increase that corresponds to a QUADRANT POWER TILT RATIO of 1.02 is included in-the_ generation of the AFD limits.

j)

The 2-hour time allowance 'for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correc-tion of a_ dropped or misaligned rod.

In the event such action does not cor-rect the tilt, the margin for uncertainty on F (X,Y,Z) is reinstated by 9

reducing the power by 3%-from RATED THERMAL POWER for each percent of tilt in excess of 2.0%.

(

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

The incore detector monitoring is done with a full incore t"

flux _ map or two sets of four symmetric thimbles, 3f4.2.5 DNE PARAMETERS-The limits on the DNB related parameters assure that each of the para-I meters'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 i

adequate to maintain a design limit DNBR throughout each analyzed transient.-

iAs_noted on Figure 3.2-1, RCS flow rate and THERMAL POWER may be " traded off"

. 3 Jagainst one'another (i.e., a low measured RCS flow rate is acceptable if the q

power level'is decreased) to ensure that the calculateo DNBR will not be below

)

the design-DNBR value.

The relationship defined on dgure 3.2-1 remains valid as long:as the limits placed on the nuclear enthalpy rise hot channel factor, h

g (X,Y), in Specification 3.2.3 are maintained. 'The-indicated T,yg values

(

F and the indicated pressurizer pressure values correspond to analytical limits of 592.6*F and 2220 psia respectively, with-allowance for indication instrumen-tation measurement uncertainty, When RCS flow rate is measured, no additional S

allowances are necessary prior to comparison with the limits of Figure 3.2-1 I

since a-measurement error of 1.7% for RCS total flow rate has been allowed for-5 in determination of the design DNBR value.

J

'The measurement error for RCS total flow rate is based upon performing a I

precision heat balance and using the res' ult to calibrate the RCS flow rate indi-g cators.

Potential fouling of the feedwater venturi which might not be detected g

could bias the result from the precision heat balance in a non-conversative B3/4f2-5 McGUIRE - UNITS 1 AND 2 Amendment No.

(Unit 1)

(

Amendment No.1 (Unit 2)

I

POWER DISTRIBUTION J MITS xdi h I

B15E5 3/4.2.5 DNB PARAMETERS (Continued)

Therefore, a penalty of 0.1% for undetected fouling of the feedwater manner.

venturi is includcd in Figure 3.2-1.

Any fouling which might Dias the RCS flow rate measurement greater than 0.1% can be detected by monitoring and trending various plant performance parameters.

if detected, action shall be

(

taken before performing subsequent precision haat balance mesurements, i.e.,

J either the effect of the fouling shall be quantified and compensated for in

)

g the RCS ficw rate measurement or the venturi shall be claaned to eliminate the

fouling, d

The 12-hour periodic surveillance cf these parameters through instrument readout is suf ficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

Indication instrumentation measurement uncertainties are accounted for in the limits provided in Table 3.2-1.

McGUIRE - UNITS 1 AND 2 B 3/4 i,2-5a Amendment No.

(Unit 1) l Amendment No.

(Unit 2)

h 3/4.2_ POWER DISTRIBUTION LIMITS m

l bbk BAS ($

j 1

The specifications of this section provide assurance of fuel integrity during Condition 1 (Normal Operation) and 11 ( ncidents of Moderate Frequency)

(eventsby:

(1) maintaining the calculated DN in the core at or above the

\\' design limit during normal operation and in e ort t* arm transients, and (2) limiting e fission gas release, fuel pellet temper..ure, and cladding mechanical prop-er ' s to within assteed design criteria.

n addition, limiting the peak linea' power nsity during Condition _I events pr vides assurance that the initial conditio assumed for the LOCA analyses e met and the ECCS acceptance crit na limit of 2.

'F is not exceeded.

The defin ons of certain hot char iel and peaking factors as used in

-these specificati s are as follows:

F (Z)

Heat Flux H Channel facto, is defined as the maximum local 0

heat flux on surface of a fuel rod at core elevation Z divided by the. average 1 rod he t flux, allowing for manufacturing toler-ances on fuel pel ts and ods; F$g Nuclear Enthalpy Rise o Channel Factor, is defined as the ratio of the integral of linear

'wer along the rod with the highest integrated power to the average re war.

3/4.2.1 AXIAL FLUX DIFFERENCE The limits on AX1AL FLUX IFFERENCE ( D) assure that the F (Z) upper E

q bound envelope of--the F

-1 it specified in he CORE OPERATING LIMITS REPORT (COLR) times the nor, alized axial peakin facter is not exceeded during either normal opera on or in the event of encn redistribution following power changes, Target flux differ ice is determined at equilibri xenon conditions.

The full-length rods m be positioned within the core i accordance with their respective inse7 ion limits and should be inserted rh r their normal potition for steady-tate operation at high power levels.

value of the target flux differe e obtained under these conditions divide by the f* action of RATLD THERMAL P EF, is the target flux difference at RATED _ ERMAL POWER for the associate core burnup conditions. Target flux differene for other THERMAL POWER le is are obtained by multiplying the RATED THERMAL POWER value by the appropri e fractional THERMAL POWER level.

The periodic updating of L

the. target flu. difference value is necessary to reflect core burnup considerationt l

i 1

l l

McGUIRE

UNITS 1 and=2 8 3/4 B2-1.

Amendment No.128 (Uni _t 1)-

l Amendment No.110 (Unit 2)

L I

9 o

1 3/4.4 REACTOR COOLANT SYSTEM m

j i

BASES 1

3/a 4.1 REACTOR COOLANT LOOPS AND C00l. ANT CIRCULATION I

2 The plant is designed to operate with all reactor coolant loops in operation and maintain DNBR above the design limit during all normal operations and antici*

pated transients.

In MODES 1 and 2 with one reactor coolant loop not in oper-ation this_ specification requires that the plant be in at least HOT 51ANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

In MODE 3 two reactor coolant loops provide sufficient heat removal l

capability for removing decay heat; however, single failure considerations require that three loops be OPERABLE.

Also,-the uncontrolled bank withdrawai i

from zero power or_ subcritical assumes three reactor-coolant loops in k

operation.

In MODE 4, and in MODE 5 with reactor coolant loops filled, a single reactor coolant loop or RHR loop provides sufficient beat removal capability-for removing decay heat; but single failure considerations requito that at least two loops (either RHR or RCS) be OPERABLE.

In MODE 5 with reactor coolant loops not filled, a single RHR loop provides suf ficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat-removing component, require that at least two RHR loops be OPERABLE.

The operation _of one reactor ccolant pump (RCP) or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual i

reactivity changes during boron concentration reductions in the Reactor i

. Coolant System.

The reactivity change rate associated with boron reduction 1

will, therefore, be-within the capability of operator-recognition and control.

The restrictions on starting a reactor coolant pump with one.or more RCS cold legs _less than or equal to 300'F are provided to prevent RCS pressure transients, caused by energy additions from the Secondary coolant System, which could exceed the limits of Appendix G to 10 CFR Part 50.

The RCS wi!!

bel protected against overpressure transients and will not-exceed the limits of Appendix G by either; (1) restricting the water volume in the pressuriter and'thereby providing a volume for the reactor coolant to expand into, or (2) by restricting' starting of the RCPs to when the secondary water tempera

ture of each steam generator is less than 50*F above each of the RCS cold leg temperatures.

\\-

l L

i B3/4f4-1 Amendment No.

(Unit 1)

McGUIRE - UNITS 1 AND 2-Amendment No.

(Unit-2)

.--.2..

3/4.4 REACTOR COOLANT SYSTEM UNIT 2 l

HASES i

3/4.\\1 REACTOR COOLANT LOOPS-AND COOLANT CIRCULATION The lant is designed to operate with all reactor coolant loops in operation d.naintain DNBR above the design limit during all normal operations ai anticipated transients.

In MODES 1 and 2 wi one reactor coolant loop no in operation this specification requires iat the plant be in at least HOT STA Y within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

. In MODE 3, two r ctorcoolantloopsprovidesuff}ureconsiderations ient heat removal capability for removin ecay heat; however, single f-require that three loops e OPERABLE.

in MODE 4, and MODE 5 w'th reactor coulant lo ps filled, a single reactor coolant loop or RHR loop prov es sufficient heat removal capability for removing decay heat; but single ailure consider tions require that at least two loops (either RHR or RCS) be PERABLE.

In mode 5 with Mactor coolant Sops not filled, a single RHR loop orovides sufficient heat removal capa lity or removing decay heat; but single failure considerations, and the na i lability of the steam generators as a heat removing component, require th at least two RHR h ops be OPERABLE.

The operation of one reactor coola p p (RCP) or one RHR pump provides adequate flow to. ensure mixing. preven strat'ftcation and produce gradual reactivity changes during boron conce ration ductions in the Reactor Coolant System.

The reactivity chan e rate ass. -iated with boron reduction will, therefore, be within the capa liity of.oper tor recognition and control.

- The rectrictions on starting a reactor coolant ump with one or more RCS cold legs less than or equal to 00'F are provided t revent RCS pressure transients, caused by energy-ad itions-from the Secon ry Coolant System,

- which could exceed the limits f Appendix G to 10 CFR P t 50- The RCS will be protected against overpres ure transients and will r,o exceed tne limits of Appendix G by either:

(1) r stricting the water volume in he pressurizer ard thereby providing a volume or the reactor coolant to expan into, or (2) by restricting starting of t RCPs to when the secondary water mperature of each generator le less tt n 50*F-above each of the RCS cold le temperatures.

l McGuire - UNITS 1 and 2 B 3/4 B4*1 Amendment No.128 (Unit.1) l Amendment No.110(Unit 2) m

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Attachment I,b Technical Justification The administrativo changos presented in Attachment Ia reflect the J

application of a previously-approved generic methodology (soo

- References 1-3) to McGuiro Unit 2 Cyclo 8.

Thoso methods have boon previously used for McGuire Unit 1 Cycle 8.

As determined by tho approval-of identical Technical Specification changos for Unit 1 (soo Reference 4), the amended Toch Specs will provido a consistant level of safety relative to the superceded Unit 2 specifications.

It is not considered necessary to provido a detailed Technical Justification for the individual.ech Spec changes as-they apply to Unit 2; details of the. changes may be found in tho application for i

amendment to the Unit 1 - Specifications (soo Reference 5).

The changos and justification for the changes are identical.

- The application of this methodology to Unit i resulted in the affected Technical Specifications (mostly in the

" Power i

Distribution" section of Toch Sper") being different betwoon the two units.

Application of this mot)dology to Unit 2 will rostore uniformity to the units' Toch Specs.

i References 1 i

+

1) Safety Evaluation Report for DPC-NE-2004, T.

A.

Hood to H.

B.

. Tucker, November 15, 1991.

r

2) Safoty Evaluation Report for DPC-NE-3000, T. A. Reed to H. B.

Tucker, November 15, 1991.

3) Safety Evaluation Report for DPC-NE-3001, T. A.

Rood to H.

B.

Tucker, November 15, 1991.

F

4) License Amendment Number 128 to' Facility Operating Licensa j

NPF-9,_T. A.

Reed to T.

C. McMoekin, November 27, 1991.

5) Request f or License Amendment for McGuire Unit 1 Cycle 8 Heload, M.

S. Tuckman to USNRC, June 26, 1991.

I '

M 4

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.,y,,y

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.y

r ATTACHMENT Ic NO SIGNIFICANT HAZARD _S ANALYSIS The changes presented in Attachment la are administrativo in

nature, to roflect the application of a

previously approved methodology _to McGuiro Unit 2.

_The changes doloto references to specific _ units on individual Technical Specification (TS) pagos, and dolote superceded (previously Unit 2 only) pages.

The unit-specific references became necessary upon application of-a now safety analysis methodology for McGuire Unit 1

Cyclo 8,

and resulted in parallel, but difforent, sets of TSs; notably in the Power Distribution chaptor.

The analysis which mado the changos necossary in the Unit i

reload submittal is a generic one, applicable equally to both McGuire units and Duke's two Catawba

-units.

Thorofore, thoro is no now significant hazards consideration (SitC) which-will be raised by this amendment.

This-

.i dotormination-is in-koeping with staf f guidance which was publishod in the Federal Register (48FR14864) to assist in dotormining whether or not proposed amendmonts ara likely to raise an SitC.

This guidanco cites as an example of an amendment not likely to involve a

significant hazards considoration "a

purely administrative chango-to technical specifications: for examplo, a change to achieve consistoney..."

Since those changes are considered administrativo, no further analyis is required.

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