Text

Amends 130 & 112 to Licenses NPF-9 & NPF-17,respectively, Supporting Unit 2 Cycle 8 Reload with B&W Fuel & Making Unit 2 TS Identical to Previously Reviewed & Approved Unit 1 TS
	ML20090K738
Person / Time
Site:	McGuire, Mcguire
Issue date:	03/06/1992
From:	Matthews D; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20090K746	List: ML20090K738; ML20090K758;
References
	NUDOCS 9203190279
	Download: ML20090K738 (80)
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NUCLEAR REGULATORY COMMISSION 5

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00KE POWER COMPANY DOCKET NO. 50-369 McGUIRE NUCLEAR STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.130 License No. NPF-9 1.

The Nuclear Regulatory Commission (the Comission) has found that:

A.

.The application for amendment to the McGuire Nuclear Station, Unit 1 (the facility), Facility Operating License No. NPF-9 filed by the Duke Power Company (licensee) dated December 18, 1991, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulttions as set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Aci, and the rules and regulations of the Commission; C.

There is-reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this amenchat will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

9203190279 920306 DR ADOCK 05000369 PDR

,4 -

I 2.-

Accordingly,- the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Facility Operating License No. NPF-9

-is hereby amended to read as follows:

Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.130, are hereoy incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance.

FOR THE NOCLEAR REGULATORY COMMISSION j

]

, 7 # # /D gy g,'DavidB.Matthews,Dirctor L

Project Directorate 11-3 Division of Reactor Projects-I/II Office of Nuclear Reactor Regulation

Attachment:

Technical Specification Changes Date of Issuance:

March 6, 1992

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-E-NUCLEAR REGULATORY COMMISSION

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WASWNGTON,0 C 2M65 ~

DUKE POWER COMPANY DOCKET NO. 50-370 McGUIRE NUCLEAR STATION. UNIT 2 AMEN 0 MENT TO FACILITY OPERATING llCENSE Amendment No. 112 License No. NPF-17

- 1.

- The. Nuclear Regulatory Commission'(the Commission) has found that:

A.

The application for amendment to the:McGuire Nuclear Station, Un'it 1:

(the facility), Facility Operating License No. NPF-17 filed by the-

- Duke Power Company (licensee) dated December 18l, 1991, complies with the standards and requirements of the Atomic Energy.Act f 1954, as amended 1(the Act), and the Commission's rules and regulations as set

- forth in 10 CFR Chapter I;

- B.

The. facility will-operate in conformity with the application, the A.;

provisions of the-Act, and the: rules and regulations of the Commission; C.-

.There is reasonable assurance-(i)1that the activities authorized by

- this' amendment can be conducted without_ endangering-the health and safety of the. public, 'and (ii) that such ' activities.will be conducted in: compliance' with the Commission's: regulations set forth in_10 CFR Chapter I; D.

The issuance ofLthis amendment.will-not be inimical-to the' common l

- defense and:securityLor to the health and safety-of the public;-and E

- E.

- The issuance of:this amendment is in accordance with 10 CFR Part 51 of:the Commission's regulations and all applicable requirements' have been satisfied.

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Accordingly, the license is nereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Facility Operating License No. NPF-I7

-is hereby amended to read as follows:

" Technical SDecifications The-Technical Specifications contained in Appendix A, as revised n eough Amendment No. 112, are hereby incorporated into this ense.

The licensee shall operate the facility in accordance with

c Technical Specifications and the En'vironmental Protection Plan.

(

h73 license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

.1

[, $f lI t.D h David B. Matthews, Director Project Directorate 11-3 Division of Reactor Projects-I/II Office of Nuclear Reactor Regulation

Attachment:

Technical Specification Changes Date of Issuance:

-March 6, 1992 l-

ATTACHMENT TO LICENSE AMENDMENT NO,130 FACIllTY 0PERATING LICENSE NO. NPF_Q DOCKET NO. 50-369 i

AND TO LICENSE AMENDMENT N0,112 FACILITY OPERATING LjtENSE NO. NPF-17 DOCKET NO. 50-370 l Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages, The revised pages are identified by Amendment number and contain vertical-. lines indicating the areas of change.

Remove Paaes Insert Paaes

~

Index_

III III V

V VI VI XII XII XVI-XVI

.A_poendix A 2-A1 2-1 2-A2 2-2 2-A4 2-4 2-A5 2-5 2-A6 2-6 2-A7 2-7 2-A8 2-8 2-A9 2-9 2-A10 2-10 2-All 2-11 B1 B11 3/4 A2-1 3/4 2-1 3/4 A2-la 3/4_2-la 3/4.A2-6 3/4 2-6 3/4 A2-7

-3/4 2-7

-3/4 A2-8 3/4 2-8 3/4 A2-8a 3/4 2-8a 3/4 A2-9 3/4 2-9 3/4 A2-9a 3/4 2-9a 3/4 A2-14 3/4 2-14 3/4 A2-14a 3/4 2-14a 3/4 A2-15 3/4 2-15 3/4 A2-15a 3/4 2-15s 3/4 A2-19 3/4 2-19

-f:

.?.

s' (ggntinued)

Remove Paces Insert Paaes 3/4 A2-20 3/4 2-20 3/4 A2-21 3/4 2-21 3/4 A2-22 3/4 2-22 3/4 A2-22a 3/4'2-22a 3/4 A2-23 3/4 2-23 3/4 A2-24 3/4 2-24 3/4 B2 3/4 82-23 3/4 A3-1 3/4 3-1 3/4 A3-2 3/4 3-2.

3/4 A3-3 3/4 3-3 3/4 A3-4 3/4 3-4

-3/4 A3-5 3/4 3-5 3/4 A3-6 3/43-6 3/4-A3-7 3/4 3-7 3/4 A3-8 3/4 3-8 3/4 A3-9 3/4 3-9 3/4 A3-10 3/4 3-10

'3/4 A3-ll 3/4 3-11 3/4 A3-12 3/4 3-12 3/4 A3-13 3/4 3-13 3/4 A3 3/4 3-14 3/4 A3-14a 3/4 3-14a 3/4 3-23 3/4 3-23 3/4 3-24 3/4 3-24 3/4.3-24a-3/4 3-24a 3/4 A3-15 3/4 3-15 3/4 A3-25 3/4 3-25 3/4 A3-26 3/4 3 3/4 A3-21 3/4 3-27 i

3/4 A3-28 3/4 3-28 3/4 A3-29 3/4 3-29 3/4 A3-30 3/4 3-30 3/4 A3-31 3/4 3-31 3/4 A3-32 3/4.3-32 3/4 A3-33 3/4 3-33 3/4 83-1 3/4 B3-2 3/4 B3-9 3/4 B3-11 3/4 B3-15 3/4 B3 -----

3/4 B3-30 3/4 A7-8 3/4 7-8 3/4 B7-8 B A2-1 8 2-1 B A2 B 2-4 B B2-1 B B2-4 B 3/4 A2-1 B 3/4 2-1 B 3/4 A2-2 8 3/4 2-2

.\\ -

(continued)

Remove Paaes insert Pacel B 3/4 A2-3 8 3/4 2-3 B 3/4 A2-4 8 3/4 2-4 8 3/4 A2-5 B 3/4 2-5 8 3/4 A2-Sa B 3/4 2-Sa B 3/4 B2 B 3/4 B2-25 B 3/4 A4-1 B 3/4 4-1 B 3/4 B4-1 h

et 9

u.

'l INDEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY LIMl_TS l

2.1.1 REACTOR CORE..

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

2-1 i

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

2-2 l

FIGURE 2.1-2 (BLANK).....

2-3 l

2.2 LIMITING SAFETY SYSTEM SETTINGS l-2.2.1 -REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS..............

2-4

[

TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS...

2-5 l

BASES SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE.............................................

B 2-1 2.1.2 -REACTOR COOLANT SYSTEM PRESSURE.

B 2-2 I.2.2' LIMITING SAFETY SYSTEM SETTINGS 2.- 2.1 REACTOR TRIP SYSTEM-INSTRUMENTATION SETP0lNTS..............

B 2-3 McGUIRE - UNITS 1 and 2 III Amendment No.130 (Unit 1)

Amendment No.112 (Unit 2)

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

3/4 1-21 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE.....

3/42-1 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - Fq (X,Y,Z)..

3/4 2-6 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 Dh8 PARAMETERS........................................

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

3/4 2-23 FIGURE 3.2-1 REACTOR COOLANT FLOW V5 RATED THERMAL POWER...........

2/4 2-24 3/4.3 INSTRUMENTATION

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

3/4 3-1 i

i McGUIRE - UNITS 1 and 2 v

Amendment No.130 (Unit 1)

Amendment No.112 (Unit 2)

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS

-SECTION PAGE TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION.

3/4 3-2 I

r TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES......

3/4 3-9 TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS...............

3/4 3-11 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION....

3/4 3-15 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...................

3/4 3-25 i

TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE-TIMES............

3/4 3-30 TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SUR!EILLANCE REQUIREMENTS..........

3/4 3-34 3/4.1.3 MONITORING INSTRUMENTATION RADIATION MONITORING FOR PLANT OPERATIONS........................................

3/4 3-40 TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION FOR PLANT 0PERATIONS.........................................

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.130 (Unit 1)

Amendment No.112 (Unit 2)

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE Ice Bed Temperature Monitoring System.

3/4 6-36 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 Seal......

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 OPERATION...........

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 FE,dwater Sys t7em...

3/4_7-4 Speci fic-Activity........

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

3/4 7-7 Main Steam Line Isolation Valves......................

3/4 7-8 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 3/4.7.4 NUCLEAR SERVICE WATER SYSTEM..

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

3/4 7-11a l-l McGUIRE - UNITS 1 and 2 XII Amendment No.130(Unit 1)

Amendment No.112(Unit 2) l

i INDEX BASES SECTION PAGE 3/4.0 APPLICABILITY..............

B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1-BORATION CONTROL.........

B 3/4 1-1 3/4.1.2 BORATION SYSTEMS.........

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

B 3/4 1-3 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE..................................

B 3/4 2-1 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR.......................

B 3/4 2-3 3/4.2.4.

QUADRANT POWER TILT RAT10.................................

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

B 3/4 2-5 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.............

B 3/4 4-1 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 1 and 2 XVI Amendment No.130 (Unit 1)

Amendment No.112 (Unit 2)

i 5

SECTION 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS J

1'.

a

a i

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS l

2.1_ W ETY LIMITS REACTOR CORE 2.1.1 Tra combination of THERMAL POWER, pressurizer pressure, and the highest 1

operating loop coolant temperature (T,yg) shall not exceed the limits shown in Figures 2.1-1 and 2.1-2 for four and three loop operation, respectively.

l APPLICABILITY:

MODES 1 and 2 l

i ACTION:

Whenaver the point defined by the combination of the highest operating loop average temperature and THEbiAL POWER has exceeded the apprcpriato pressurizer pressure line, be in HOT STANOBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the is7uire-ments 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.

l ACTION:

MODES 1 and 2 Whenever the Reattor Coolant System pressure has exceeded 2735 psig, be in HOT STANDGY w th the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the requirements of Specification 6.7.1.

MODES-3, 4 and 5 Whenover the Reactor Coolant System pressure has exceeded d.s5 psig, redet the Reactor Coolant System.oressure to within its limit within 5 minutes, and comply with the requirements of Specification 6.7.1.

l McGUIRE - UNITS 1 and 2 2-1 Amendment No.130 (Unit 1) l Amendment No 112 (Unit 2)

- _ - = _ _

Figure 2.1-1 Reactor Core Safety Limits -

Four Loops in Operat'an 665 :

FLOW PER LOOP - 96250 GPl.1 i

i 660 5 0

s55 5 650 :

%,.N 2455 psia UNACCEPTABLE OPERATION 645 2 2400 psia 640 :

635 2280 psia m -b

=

1

~ 625 2 2100 psia G* j N

615 :

610 :

1945 psia 605 5 600 :

595 5 4

3 ACCEPTABLE 590 :

OPERATION 585 i gg)

I 4

4 0.00 0.20 0.40 0.60' O.80 1.00 1.20 Fraction of Rated Thet-mal Power McGUIRE - UNITS 1 and 2 2-2 Amendment No.130 (Unit 1) l Amendment !10 112 (Unit 2)

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS l

2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The Reactor Trip System Instrumentation and Interlocks Setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.

l APPLICABILITY:

As showr for each cha.nel in Table 3.3-1.

l ACTION:

With a Reactor Trip Syste? Instrumentation or Interlock Setpoint less conser-vative than the value shown in the Allowable Values column of Table 2.E-1, l

declare the channel inoperable and apply the applicable ACTION ste.:ement requirement of Specification 3.3.1 until the channel is restored to CPERABLE status with its Trip Setpoint adjusted consistent with the Trip Setpoint value.

r t

F McGUIRE + UNITS 1 and 2 2-4 Amendment No.130 (Unit 1)

Amendment No.112 (Unit 2)

2 TABLE 2.2-1

'E

Si REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINIS R;

i FUNCTIONA! UNIT TRIP SETPOINT ALLOWABLE VALUES C

55

1. Manual Reactor Trip N.A.

N.A.

i

2. Power Range, Neutron Flux tow Setpoint -< 2S% of RATED THERMAL POWER ~

Lew Setpoint

-~ 26% of RATED r.

IHERMAL POWER 5l High Setpoint

$ 109% of RA1ED High Setpoint

_ 110% of RAllD n3 TiiERMAL POWER IHERMAL POWER

3. Power Range, Neutron Flux,

< 5% of RATED THERMAL POWER with

< S.S% of RAIED THERMAL POWER 2

i High Positive Rate i time constant 2 2 seconds with a time constant 2 2 seconds 4.

Intermediate Range, Neutron 1 25% of RATED THERMAL POWER 1 30% of RAILD IHERMAL POWER Flux

5. Source Range. tieutron flux

< 10L counts per second

1. 3 x 105 counts per second 1

6. Ove: temperature AT See Note 1 See Note 3

7. Overpower AT See Note 2 See Note 4

8. Pressurizer Pressure--tcw

> 1945 psig g 1935 psig 3rg"

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

2 i

515E gg

10. Pressurizer Water Level--High 1 92% of instrueent span 5 93% of instrument span

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11. Low Reactor Coolant. Flow 3 90% of minimum measured 3 88.8% of minimum measured

_. x flow per loop

tiow per loop

PP I* bJ 4

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Minimum measured flow is 96,250 gpm per loop.

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g TABLE 2.2-1 (Continued) c15 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SEIP0lNTS M

h FUNCTIONAL UNIT TRIP SETPOINI AtLOWABt[ val UI S

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c.

Power Range Neutron Flux, P-8,

< 48% of RATED

< 4'3% of RAftD Low Reactor Coolant Loop Flow, THERMAL POWER TiiERMAL POWER and Reactor Coolant Pump Breaker o,

Position N

i d.

Low Setpoint Power Range Neutron 10% of RATED

> 9%, < 11*. o f RA IL D Flux, P-10, Enable' Block of THERMAL' POWER THLRMAL POWER Source Intermediate and Power Range Reactor Trips e.

Turbine Impulse Chamber Pressure.

P-13. Input to 1ow Power Reactor 10% RIP Turbine 11% RIP iurtiine Trips Block P-/

Impulse Pressure Impulse Pressure i

7 Equivalent tquivalent w

18.

Reactor Trip Breakers N.A.

N.A.

19.

Automatic Trip and Interlock logic N.A.

N.A.

NN RE i *M

en IFF SS l ;: ;;

IOU N

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2 TABLE 2.2-1 (Continued)

[

' S

_ 55 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINIS NOTAIION (Continued)

C

~ 35 NOTE 2:

OVERPOWER AT

. _4

^^

1 + 1,5 1

tyS 1

1

((

A T (1 + IzS) (1 + T 5) 3 AT

{K -Kg (1 + 1 5 (1 + tuS} I ~ 6bI l + r.,5)~ I"3 ~ '2Iu II 3

g 4

7 oc Where:

AT As defined in Note 1,

=

3

{

= As defined in Note 1

= As defined in Note 1 T i. 12 1

=

.S defined in Note 1,

,,,g u

AT As defined in Note 1,

=

o K

1.0809, 4

Kg 0.02/ F for increasing average temperature red 0 for decreasing average

=

temperature, 1 S 7

IjI(

p

,3 The function generated by the rate-lag controller for I, dynamic

=

g[ p, compensation,

.as S$

T7 Time constant utilized in the rate-lag controller for Iavy, 1

=

7 5 sec,

, re r lE? E 1

As defined in Note 1,

=

1 + tsS

- us As defined in Note 1,

=

is SEE I E 3.

K<

=

0.001239/ F for T > I" and K = 0 for T <

I",

r+ e 6

6 lEst:

L I

-m-

~

c TARLE 2.2-1 (Continued)

' 'E l

{j RrACTOR TRIP SYSIEM INSTRUMENfATION TRIP SCTPOINIS m

NOTATION (Continued)

E

. Gl T

As defined in Note 1,

=

w T"

=

< 538.2"F Refeience T at RATED THERMAL POWER, o

aug a

. ro S

=

As defined in Nete 1. and f (AI)-is a function of the indicated differet.9 between top and bottom detectors 2

of the power range nuclear ion chan.bers; with mites to be selected based on measured instrument response during plant startup tests such that:

(i) for q, - o between -3S% and +35% AI; f..(AI) = 0, where "t

""d "I are percent RAlfD1HERMAt-Pb,wEPinthetopandbottomhalvesofthecorerespectiv'ely,and q,

q + q s total ~fitERRAL POWER in percent of RAl[D IHLRMAL POWLR; t

b (ii) for each percent imbalance that the magnitude 6t 4 q

is move negative than t

-35% a!, the AT Trip Setpaint shall be automatically reduced la,y 7.0% of alo; and (iii) for each percent imbalance that the magnitude of y y

is more positive than

+3E%al,thealTrip5etpointshallbeautomaticallyreducedby7.0%ofaio.

[ >

Note 3:

The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 3.6% of Rated Thermal Power.

i oo cz. a 2g Note 4:

The channel's maximum Trip Setpoint shall not exceed its computed frip Setpoint by more than 4.2%

.. ' j' j

of Rated Thermal Power.

??

2C co o C CN

' Of t; i

1r w-t w w

. _. _ _ - =

.3/4.2 POWER CISTRIBUTION LIMITS l

3/4.2.1 AXIAL FLUX OIFFERENCE (AFO)

LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFO) shall be maintained within the acceptable limits as specified in the Core Operating Limits Report (COLR).

i APPLICABILITY:

MODE 1 above 50% of RATED THERMAL POWER

l 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 THERMA'l 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 .

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.

See.5pecial Test Exception 3.10.2.

McGUIRE - UNITS 1 and 2 3/4 2-1 Amendment No.134 Unit 1)

Amendment No.11RUnit 2)

I PCWER OISTRIBUTION LIMITS 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 Alaro is

4. operable.

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.

McGdlRE - UNITS 1 AND 2 3/4 2-la Amendment No. 130 (Unit 1) l Amendment No. 112 (Unit 2)

POWER DISTRIBUTION LIMITS i

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

LIMITING C030!T10N FOR OPERATION 3.2.2 F (X,Y,Z) shall be limited by imposing the following relationship:

9 F

(X,Y,Z) $ F K(Z) for P > 0.5 P

MA Fq (X,Y,Z) i F K(Z) for P i 0.5

0. 5 Where F

= the F limit at RATED THERHAL 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

in the COLR for trc propriate fuel type, and FMA(X,Y,Z) = the measured heat flux hot channel factor FM (X,Y,Z) with the adjustments specified in 4.2.2.3 APPLICABILITY:

MODE 1.

l ACTION:

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

9 Redure THERMAL POWER at least 1% for each 1% F"A(X,Y,Z) exceeds the a.

9 limit within_15 minutes 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 , and b.

Control the AFD to within new AFD limits w5ich are determined by reducing the allowable power at each point along the AFD limit j

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 OPERATION may proceed for up to a total.of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION may proceed provided the Overpower aT Trip Setpoints (value of K ) have been reduced at least 1% (in ai span) for each 1%

4 F A(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 requirkd by ACTION a.,

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

dtmonstrated through incore mapping to oe within its limit.

l McGUIRE - UNITS 1 AND 2 3/4 2-6 Amendment No.130 -(Unit 1) l Amendment No.112 (Unit 2)

POWER DISTRIBUTION LIMITS I

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

4.2.2.2 f (X,V,2)(

shall be evaluated to detaraine whether F (X,Y,?) is g

within its limit b :

/

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

map at any THERMAL POWER greater tnan S*6 of RATED THERMAL POWER.

b.

Mear sring F (X,Y,l) at the earliest of:

1.

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

Upon reaching equilibrium conditions after exceeding by 10%

or more of RATED THERMAL POWER, the THERMAL POWER at whic5 F"(X,Y,Z) was last determined

), or 3.

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

i (1)

No additional uncertainties are required in the following equations for F (X,Y,Z) 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.

McGUIRE - UNITS 1 AND 2 3/4 2-7 Amendment No.130 -(Unit 1) 1

e

~

POWER DISTRIBUTION LIMITS l

SURVEILLANCE REQUIREMENTS (Continued) c.

Performing the following calculations:

1.

For each core location, calculate the % margin to the maximum l

allowable design as follows:

M

% Operational Margin = ( 1 -

F (X,Y,Z) g

) x 100%

[F (X,Y.?))0P

'Y'2)

% RPS Margin = ( 1 -

Q

) x 100%

[Fh(X,Y,Z)]#5 where[Fh(X,Y,Z))0P and[Fh(X,Y,Z)]RPSare the Operational and RPS design peaking limits defined ir. the COLR.

2.

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

If any margin is less than zero, then either of the following actions shall be taken:

(a) Within 15 minutes:

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

(3)

(AFD Limit)

= (AFD Limit) C

- W GIN uc t e n gative 0

(3) c (AFD Limit) re

= (AFD Linit)

- MARGIN 9

e s tive where MARGIN "P is the minimum margin from 4.2.2.2.c.1, N

0 and (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 Specifical on 3.2.2, treatirig the margin violation in 4.2.2.2.c.1 above as the A

amount by which F is exceeding its limit.

( ) Defined and specified in the COLR per Specification 6.9.1.9.

.McGUIRE - UNITS 1 AND.2 3/4 2-8 Amendment No.130 (Unit 1) l' Amendment No.112 (Unit 2)

POWER DISTRIBUTION LIMITS l

l SURVEILLANCE REQUIRtMENTS (Continued) l 3.

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

If any margin 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 Ki value for OTM by:

g, adjusted, gt (4) - [KSLOPE(3) x Margin m h)tbsolute value RPS where MARGIN i s the minimum margin fi nm 4.2. '.2.c. l.

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

Power Days-beyond the most recent _ measurement and if:

N

[F (X,Y,Z)] (extrapolated)3[Fh(X,Y,Z)]0P(extrapolated)',and

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

> [F (X,Y,Z)]

N n

[

OP

[

OP

[F (X,V,Z)]

(extrapolated)

(f (X,Y,Z)]

9 q

or N

[F (X,Y,2 M (extrapolated)1[Fh(X,Y,Z)]RPS(extrapolated),and O

[F"(X,Y,Z)]

.. L y ' (extrapolated)

L RPS L

RPS

[F (X,/,;)]

(extrapolated)

[F (X,Y,Z)]

9 9

either of the following actions shall be taken:

1.

F (X,Y,Z) shall be increased by 2 percent over that specified in 4.2.2.2.a, and the calculations of 4.2.2.2.c repeated, or

) Defined and specified in the COLR per Specification 6.9.1.9.

l I)K value from Table 2.2-1.

3

( ) Extrapolation of FM for the initial flux map taken after reaching equili-q brium conditions is not required since the initial flux map establishe, the baseline measurement for future trending.

Also, extrapolation of F limits are not valid for core locations tnat were previously rodded, 'or for core locations that were previously within 127, of the core height abcut the demand position of the rod tip.

McGUIRE - UNITS 1 AND 2 3/4 2-8a Amendment No.130 (Unit 1)

I Amendment No.112- (Unit 2)

m. _ _ - _ _..

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Cortinued) 2.

A movable incore detector power distribution map shall be cbtained, and the calculations c' 4.2.2.2.c.1 shall be performed no later than the time at which the margin in 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 of core height from the bottom of the fuel:

1.

Lower core region from 0 to 15%, inclusive.

2.

Upper core region from 85 to 100%, inclusive.

t t

l L

McGUIRE - UNITS 1 AND 2 3/4 2-9 Amendment No.130 (Unit 1) l L

Amendment No.112 (Unit 2)

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 4.2.2.3 When a ful! core power distribution map is taken for reasons other M

than meeting the requirements of Specification 4.2.2.2, an overall F (X,Y,Z) 9 shall be determined, then incree.ed by 3% to account for manufacturirg toler-ances, further increased by 5% to account for measurement uncertainty, and further increased by the radial-local ceaking factor to obtain a max' mum local peak.

This salue shall be compared to the limit in Specification 3.2.2.

a McGUIRE - UNITS 1 AND 2 3/4 2-9a Amendment Ho,130 (Unit 1) 1 Amendment No.112 (Unit 2)

s POWER OISTRIBUTION LIMITS 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - F g(X,Y)

LIMITING CONDITION FOR OPERATION i

3.2.3 FaH(X,Y) shall be limited by imposing the following relationsnip:

F'"g(X,Y)1Pg(X,Y)h 3

where:

F H(X,Y) - the measured radial peak.

[F g(X,Y)]LCO - the maximum allowable radial peak as defined in Core Operating Limits Report (COLR).

APPLICABILITY:

MODE 1.

i l

2 ACTION:

With FaH(X,Y) exceeding its limit:

a.

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 MH%( ) for each 1% that F H(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 F H(X,Y) to within the limit of Specification 3.2.3 for RATED THERMAL POWER, or 2.

Reduce the Power Range Neutron Flux-High Trip Setpoint in Table 2.2-1 at least RRH% for each 1% that F g(X,Y)

.eeds that limit.

l and 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, either:

M 1.

Restore FAH(X,Y) to within the limit of' Specification 3.2.3 for RATED THERMAL POWER,-or 2,

Perform the.following actions:

()

RRH is the amount of THERMAL POWER reduction required to compensate for M

each 1% that F3g(X,Y)exceedsthelimitofSpecification3.23,provided in the COLR per Specification 6.9.1.9.

McGUIRE - UNITS 1 AND 2 3/4 2-14 Amendment No.130 (Unit 1) l Amendment No.112 (Unit 2)

POWER DISTRIBUTION LIMITS i

3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - Fdg(X,Y)

LIMITING CONDITION FOR OPERATION ACTION:

(a)

Reduce the OTAT Kg term in Table 2.2-1 by at least TRH( )

l for each l'. that F H(X,Y) exceeds the limit, and (b) Verity through incore mapping that F H(X,Y) 's restored to within the limit for the reduced THERMAL POWER allowed by ACTION a, 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 .

i l

l 1

l l

l

()

TRH is the amount of OTM K setpoint reduction required to compensate for 3

each 1% that Fh(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 2-14a Amendment No.130 (Unit 1) l Amendment No.112 (Unit 2)

POWER DISTRIBUTION LIMITS l

LIMITING CONDITION FOR OPERATION ACTION:

(Continued) d.

Identify and correct the cause of the out-of-limit conditior orior to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION a. and/or c.2 atove; subsequent POWER OPERATION may proceed provided that F H(X,Y) is demonstrated, through incore flux mapping, to be within the Limit specified in the COLR orior to exceeding the following THERMAL POWER levels:

1.

50% of RATED THERMAL POWER, 2.

75% of RATE 3 THERMAL POWER, and 3.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of attaining greater than or equal to 95% of RATED THERMAL POWER.

SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable.

M 4.2.3.2 FaH(X,Y) shall be evaluated to determine whether F3g(X,Y) is within its limit by:

a.

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

b.

Measuring F H(X,Y) according to the following schedule:

1.

Upon reaching equilibrium conditions after exceeding 10%

or more of RATED THERMAL POWER, the THERMAL POWER at which

)

F g(X,Y) was last determined

, or 2.

At least once per 31 Ef fective Full Power Days, or 3.

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

l c.

Performing the following calculations:

l-1.

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

C') During power escalation at the beginning of each cycle, THERMAL POWER may be increased until a power level for extended operation has been achieved I

and a power distribution map obtained.

l McGUIRE - UNITS-1 AND 2 3/4 2-15 Amendment No.130 (Unit 1) l l

Amendment No.112 (Unit 2)

!.m..,

POWER DISTRIBUTION LIMITS l

SURVEILLANCE REQUIREMENTS

%F Margin = (1 -

) x 100%

H

[FfH(X,Y)]"

No additional uncertainties are required for F3g(X,Y), because

[F H(X,Y)]3"IV includes uncertainties.

2.

Find the minimum margin of all locations examined in 4.2.3.2.c.1 above.

If any margin is less thar zero, comply with the ACTION requirements of Specification 3.2.3 as if [FfH(X,Y)]surv is the same as [F H(X,Y)] LCO Extrapolating ( ) at least two measurements to 31 Effective Full Power d.

Days beyond the most recent measurement and if:

g(X,Y)(extrapolated)>[FfH(X,Y)]surv (extrapolated) and F

FN (X',Y)

(extrapolated)

F fX'Y)

H H

[FfM(X,Y)]surv(extrapolated)

[F'H(X,Y)]surv i

either of the following actions shall be taken:

F"g(X,Y) shall be increased by 2 perc.ent over that specified 1.

3 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 performed no later than the time at which the margin in 4.2.3.2.c is extrapolated to be equal to zero.

()

M Extrapolation of F for the initial flux map taken after reaching equili-3g brium conditions is not required since the initial flux map establishes the baseline measurement for future trending, i

McGUIRE - UNITS 1 AND 2 3/4 2-15a Amendment No.130 (Unit 1)

Amendment No.112 (Unit 2)

~

i POWER DISTRIBUTION LIMITS I

3/4.2.4 QUADRANT POWER TILT RATIO LIMITING CONDITION FOR OPERATION I

3.2.4 The QUADRANT POWER TILT RATIC shall not exceed 1.02.

APPLICABILITY:

MODE 1 above 50% of RATED THERMAL POWER * **

l ACTION:

a.

With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but i

less than or equal to 1.09:

1.

Calculate the QUADRANT POWER TILT P.ATIO 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 tne QUADRANT POWER TILT RATIO to within its limit, or b)

Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each Ut of 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 fetpoints to less than or equal to 55% of RATED THERHAL POWER within the ney.t 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% cf 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 acceptable at 95%

or greater RATED THERMAL POWER.

i-

See Special Test Exception 3.10.2.

l.

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

1 McGUIRE - UNITS 1 AND 2 3/4 2-19 Amendment No.130 (Unit 1)

Amendment No.112 (Unit 2)

.~

-.. ~ _.

I POWER DISTRIBUTION LIMITS l

LIMITING CONDITION FOR OPERATION ACTICN: (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 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.

Reduce DERMAL 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.

terify that the QUADRANT. COWER 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 5etpoints 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 dbove 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verifioj 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 determined to exceed 1.09 due to causes other than the misalignment of either 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 te 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.130 (Unit 1) l Amendment No. 112(Unit 2)

f POWER DISTRIBUTION LIMITS I

LIMITi"G CONDITION FOR OPEPATION 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 1

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 p?r 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 .

McGUIRE - UNITS 1 AND 2 3/4 2-21 Amendment No.130 (Unit 1)

Amendment No.112 (Unit 2)

]

e 1

i ByERDISTRIBUTIONLIMITS l

3/4.2.5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB related parameters shall be maintained within the limits shown on Table 3.2-1.

l a.

Reactor Coolant System T b.

Pressurizer Pressure, and c.

Reacter Coolant System Total Flow Rate.

APPLICABILITY:

MODE 1.

I ACTION:

a.

With either of the parameters identified in 3.2.5a. and b. above exceeding its limit, restore L e parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 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 System total flow rate and THERMAL POWER within the region of restrictc.d 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 Flux-Higt. Trip Setpoint to below the nominal setpoint by the same amount (% RTP) as the power reduction required by Figure 3.2-1.

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

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 POWER to within the region of permissible operation, b)

Restore the combination of Reactor Coolant System total flow rate and THERMAL POWER to within the region of 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 Rcnge Neutron Flux - High Trip Setpoint to less tnan or equal tn b5% 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 operation specified in Figure 3.2-1, verify that the com-bination of THERMAL POWER and RCS total flow rate are restored 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 .

McGUIRE - UNITS 1 AND 2 3/4-2 22 Amendment No.130 (Unit 1) l Amendment No.112 (Unit 2)

POWER DISTRIBUTION LIMITS l

3/4.2.5 DNB PARAMETERS SURVEILLANCE REQUIREMENTS 4.2.5.1 Each of the parameters of Table 3.2-1 shall be measured by averaging 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 CAllBRATION at least once per 18 nonths.

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

l l

l 1

1 McGUIRE - UNITS 1 AND 2 3/4 2-22a Amendment No.130 (Unit 1)

Amendment No.112 (Unit {}

POWER DISTRIBUTION LIMITS l

TABLE 3.2-1 DNB PARAMF,TERS

OPERABLE PARAMETER INDICATION CHANNELS LIMITS

Indicated Reactor Coolant System T meter 4

<590.5'F

V9 meter 3

[590.2"F computer 4

<591.0*F computer 3

[590.8'F Indicated Pressurizer Pressure **

meter 4

>2226.5 psig meter 3

[2229.8psig computer 4

12221.7 psig computer 3

32224.2 pstg Reactor Coolant system Total-Flow Rate Figore 3.2-1 t

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.

Amendment No.130 McGUIRE - UNITS 1 AND 2 3/4 2-23 Amendment No.112 ((Unit 1) l Unit 2)

n POWIL ' STRI9VTION - lit'.,S

' Figure 3.2 1.

Reactor Coolant System Total Flow Rate Versus Rated Thermal Power - Four Locps in Operation t

L ry,e

.w for undetected foeowster Pemitssible 9

e-j a rnessurernent uncenatrary Ooorsten

- m ircruced in tnre figure.

(98,385000) 3g3oco 5

I (96.341150)

I381150" Restncted Operation 2

d Aegron (94.377300)

C 377300 E3

=_g Prohtkttse (92.373450)

Opersson Regen 373450 u

y 4..

(90.369800) 460400 D

3E5750 p

361900!

86 83.

90 92 94 96 98 100 1

(1 it l-l l _

Fraction of Rated Thermal Power 1

l J.'

!McGUIRE - UNITS 1 AND_2 3/4 2-24 Amendment No.130 (Unit _1) l Amendment No.112 (Unit _2)

. -.~

TABLF 3.3-3 (Continued)

ACTION STATEMENTS (Continued)

ACTION 25 - With one of the two trains of doghouse water level instrumenta-tion inoperable (less than the minimum required number of chanriels operable), restore the inoperable train to operable status in 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , After 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months with one train inoperable, or within one hour with 2 trains inoperable, monitor doghouse water level in the affected doghouse continuously until both t.ains are restored to operable status.

ACTION 26 - Wit % any of the eight channels inoperable, place the Inoperable channel (sj in the start permissive mode within one hour and apply the applicable action statement (Containment Spray - T.S. 3.6.2, Containment Air Return / Hydrogen Skimmer - T.S. 3.6.5.6).

ACTION 27 - With the number of OPERABLE channels one less than the total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the inoperable channel is placed in the tripped condition within I hour.

McGUIRE - UNITS 1 & 2 3/4 3-24a Amendme.it No.130 (Unit 1) l Amendment No.112 (Unit 2)

3/4.3 INSTRUMENTATION l

3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION I

3.3.1 As a minimum, the Reactor Trip System Instrumenta* ion channels and interlocks of Table 3.3-1 (Unit 1) shall be OPERABLE with RESPONSE TIMES as shown in Table 3.3-2.

APPLICABILITY:

As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1.

I SURVEILLANCE REQUIREMENTS

-_4.3.1.1 Each Reactor Trio System Instrumentation cha1rel a.id interlock shall be demonstrated OPERABLE by the. performance of the Reactor Trip System Instrumentation Surveillance Requirements spu:lfied in Table 4.3-1.

I 4.3.1.2 The REACTOR TRIP SYSTEM RESPONSE TIME of each Reactor trip function shall be demonstrated to be within its limit at least once per 18 months.

Each test'shall-include at least one train such that both trains are tested at least once per 36 months and orie channel per function such that all channels are, tested at-least once-every N times 18 months where N is the total number of redundant channels in a specific Reactor trip function as shown in the

" Total No. cf Channels" column of Table 1.3-1.

l

4. 3.1. 3 The response time of R'Os associatea with the Reactor Trip System shall be demonstrated to be within thei. limits (see note 2 to Table 3.3-2)

{

at least once per 18 months.-

F 1

McGUIRE - UNITS 1 and.2 3/4 3-1 Amendment No.130 (Unit 1)-

l Amendrent No.112 (Unit 2)

~

at

. TABLE 3.3-1 S

55 REACTOR TRIP SYSTEM INSTRUMENTATION A

MINIMUM c:

TOTAL NO.

CHANNELS CHANNELS APPLICABLE 25 -FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERAB_L_E

'HODES ACTION y

- v, 1.

Manual Reactor Trip 2'

1 2

1, 2 1

s.

2 1

2 3*,

4*,

S*

10 E-

' s, 2.

Power Range, Neutron Tlux - High 4

2 3

1, 2 2

Setpoint ggg Low 4

2 3

1

,2 2

Setpoint 3.

Power Range, Neutron Flux 4

2 3

1, 2 2

liigh Positive Rate 4.

Intermediate Range. Neutron flux 2

1 2

1

,2 3

{

y 5.

Source Range, Neutron Flux 2,

4 g

N a.

Startup

.2 1

2 b.

Shutdown 2

1 2

3*,4*,5*

10 c.

Shutdown 2

0 1

3, 4, and 5 5

6.

Overtemperature AT l[![

Four Loop Operation 4

2 3

1, 2 6

Three Loop Operation

(**)

=sRR aa se ZZY?

~

OC NO El ne

~<

+

TABLE 3.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTAil0N m

MINiciUM 10TAL NO.

CHANNELS CHANNELS APPLICABLE E.' FUNCTIONAL UNil 0F CHANNELS TO TRIP OPERABLE MODES ACTION d

'7.

Overpower WT g

E Four Loop Operation 4'

2 3

1, 2 6

[

Three Loop Operation

(**)

('^)

(**)

8.

Pressurizer Pressure-Low 4

2 3

1 6

(***)

9.

Pressorizer Pressure--High 4

2 3

1, 2 6

(.4.)

10.

Pressurizer Water Level--High 3

2 2

1 6

w 11.

Low Reactor Coolant flow 0

a.

Single Loop (Above P-8) 3/ loop 2/ loop in 2/ loop in 1

6 any oper-each oper-ating loop ating loop b.

Two Loops (Above P-7 and 3/ loop 2/isop in 2/ loop 1

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

((

ating loops ating loop oo@E 12.

Steam Generator Water 4/stm. gen.

2/stm. gen.

3/stm. gen.

1. 2 6

Level--Low-Low in any oper-each oper-(***)

ating stm.

[f gen.

gen.

CC NO CC hA SC

. m

.x

.2 m

3,,.

i' l_.,..

7

-TABLE 73.3-1 (Continued)

]

c> -

!:.i t

^

3 REACTOR' TRIP' SYSTEM INSTRUMENTATION

-MINIMUM.

'i-e 10TAL NO.

CHANNELS

' CHANNELS'

' APPLICABLE' t E. FUNCTIONAL UNIT-

OF CHANNELS-TO TRIPL OPERABLE-MODES' ACTION

_.m n, v,.

.13.. Undervoltage-Reactor Coolant 4 r Pump (above.P-7) 4-1/ bus

2-3 1

6

,. 3

s.

J n14.

Underfrequency-Reactor Coolant Pumps-(above P-7)

4-1/ bus 2

3 1

6 L15.

turbine Trip a.

iLow Fluid Oil' Pressure 3

,2 2

1 6

i b.

Turbine Stop Valve Closure-4 4

1 1

11

16.

Safety injection Input.-

y-from ESF

-2' 1

2 1, 2 9

l y

17.

Reactor. Trip System Interlocks a.

Intermediate Range, Neutron Flux,LP : 2.~

1 2

2,,

8 i

b.'

Low Power Reactor Trips Block, P-7:

P-10. 'I npu t '

4 --

2-3-

1 8

..kk or:

=3

.P-13 Input 2

1 2

1 8

ikk igg c.

Power. Range Neutron Flux, P-8 4.'

2 3

1 8

!* ?

:0 d.

Low Setpoint Power

' re Range Neutron Flux, P-10 4 -'

2 3

1, 2 8

%nc c--

s=

- e.

Turbine Impulse Chamber

<+ e Pressure, P-131 2..

1.

v2 1

8

- %, v

\\

d t

y

- S

-s

=

w

,w s

+v, e.--

m,s w

c m,--+

4 v-M-

o

' TABLE 3.3-1 (Continued) l 3e REACTOR TRIP SYSTEM INSTRUMENTATION iE MINIMUM 10iAL NO.

CHANNELS CHANNELS APPLICABLE

. C:

25 FUNCTIONAL UNIT OF CHANNELS 10 TRIP OPERABLE MODES ACTION d

18.

Reactor Trip Breakers

.2 1

2 1, 2 9, 12 g, -

2 1

2 3*,

4*, $*

10 g

E 19.

Automatic 1 rip and Interlock 2

_1 2

1, 2 9

3, Logic 2

1 2

3*, 4*,

S*

10 M.

u, aM

.t a

'aa er

' w" NO C C no i.-

50 4

TABLE 3.3-1 (Continued)

TABLE N01ATION With the Reactor Trip System breakers in the closed position, the Control Rod Drive System cacable of rod withdrawal.

Values left blank pending NRC approval of three loop cperation.

- =

Comply with the provisions of Specification 3.3.2 for any portion of the channel requirad to be OPERABLE by Specification 3.3.2.

1. Below the P-6 (intermediate Range Neutron Flux Interlock) Setpoint.

1. 1. Below the P-10 (Low Setpoint Power Ganga Neutron Flux Ir<terluck) 54tpoint.

ACTION STATEMENTS ACTION 1 With the number of ODERABLE channels one less than the Minimum Channels OPERABLE requirement, restort the inoperable channel to OPERABLE states withiri 48 haurs or be in HOT STANDBY within the next 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

AC.TICN 2' with the numbev of OPERABLE charmels one less than the Total

- Nvinber of Char.nels, STARTUP. and/cr POWER OPERAfl0N may proceed

.provided th6 Lfollowing condititns. are satisfied:

-a.

The inope able channel is placed in the tripped coriaition wit.hin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

.b.

The Miniinum Channelt OPERABLE requirement is met; however,

. the trtJper& Die chaDnel may D6 bypJ5 Sed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for strveillance tat' ting of other ch.inpels, per 3pecification

4. 3.1.1', and c-Eithee, TEiERMAL PON5P. ;is iwtricted to len then or equal tc'7 % of RATED THE9 %L 70VER and the 9cwer Range Neutron Flux Trip =Setooint is rtdaced to le;s toms or equal to fM% of RATED THERMAL P1)WER.vf triin 4 haurs; or, t.he QUADRANT POWEA TILT RATIO is-monitorod1 at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months per Specification 4,2.4.2.

~.

McGUiRE - UNITS 1 & 2 3/4 3-6 Amendment No.130 (Unit 1) l Amendment No.112 (Unit 2)

' ' JQR Q1& M 5 '"

" h " W '* " L 2^^'-

e, n..'

9 l1 p.

a h

i 7 A815 3. 3-1 Gortfrjued}

l ACTION STATEMENTSICy,1tinued) 5 ACTICH 3 - With the number of cNrnels OPERA 6LE one less than the Minirnum Chen m !c GPERAELE requie nent and with the TH BMAL POWER hvel; Below the P-G (latermedicte Range Neutrco Flux Interlocr.)

a.

Setpoint, res, tore the inoperable channel to OPERASLE status prict to increasing THERt1AL power abose ttie P fi setpoint, and i

l b.

Atmve the P-6 (Intermediate Range Neutron Flux Interlock)

]

Setpcint but below 10% of RATED THERNAL POWER, restore the inoperable channel to OPERABuE ststus prior to increar,ing THERMAL POWER above 10% of RATED THERMAL POWER.

ACTION 4 - With the number of OPERABLE channels one :ess than the Minimum Channels 09ERABLE requirement suspc.d all ocerations involving positive reactivity changes.

ACTION 5 - With the number of OPERASLE channels one less than the Minimum Chanr.els OPERABLE requirement, verify compliance with the SHUT 0OWN MARGIN requirements of 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 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months tnereafter.

ACTION 6 - With the nurnber of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWE9 OPERATION may proceed provided the followir.g canditions are satisfied:

a.

The inoparaole channel is olaced in the tripped conditicn witnin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and b.

The tiinimum Channah OPERABLE requiremer:t is met; howver, the inopatsblo channe) 3ay be byp3ssed for up to 4 u urs li for survv!1ance tetting of cther channels per specification 4.3.1.1 and Specificaticn 4.3.2.1 ACTION 7-Delete ACTION 8 - With less than the Minio2m Number of Chancels OPERABLE, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months determine by chservatica of +.ha associated permissive annunciator window (:-) that the interlock is in its required state for the axisting pleut condition, or apply Speci"icati n 3.0.3.

McGUIRE - ilh1T51 & 2 3/4

'J-7 Amendraent No.130 (Unit 1) l Amenament No.112 (Unit 2)

M f

TABLE 3.3-1 (Continued) l ACTION STATEMENTS (Continued)

ACTION 9 - With the number of OPERABLE channels one less than the Minimum Channels OPERA 3LE requirement, be in at least HOT STANDBY ithin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to w

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.1.1, provided the other channel is OPERABLE.

ACTION 10 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to CPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the Reactor trip breakers within the next hour.

ACTION 11 - With the number of OPERABLE channels less than the Total Number of Channels, operation 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 features (Undervoltage or shunY.

trip attachment) inoperable, restore it to OPERABLE. status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or declare the breaker inoperable and apply ACTION 9.

The breaker shall not be bypassed while one of the diverse trip features is inoperable except for the time required for performing maint(eance to restore the breaker to OPERABLE status.

McGUIRE - UNITS 1 and 2-3/4 3-8 Amendment No.130 (Unit 1)

{

Amendment No.112 (Unit 2)

3c-TABLE 3.3-2 S~

C5

' REi.CTOR TRIP SYSTEM INSTRUMENTATION RESPONSE IIMES El FUNCTIONAL UNIT RESPONSE TIME

n w-j 1.

Manual Reactor Trip N.A.

r:.

2.

Power Range, Neutron Flux

-<0.5 second (1) na 3.

Power Range,. Neutron Flux, n,

High Positive Rate N.A.

4.

Intermediate Range, Neutron Flux N.A.

5.

. Source Range, Neutron Flux N.A.

6.

Overtemperature AT 110.0 seconds (1)(2)

R 4'

?.

Overpower AT

$10.0 seconds (1)(2)

T to 8.

Pressurizer Pressure--Low

<2.0 seconds 9.

Pressurizer Pressure--High

<2.0 seconds 10.

Pressurizer Water Level--High N.A.

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

Response time of the neutron flux signal portion gg.

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

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

e+ re zz lb.

^

. NO S. a len L

e

m-a-

O

'I TABLE 3. 3-2 (Coritinued) 2 no5 REACTOR TRIP-SYSTEM _jMTRUMEh1ATION RESPONSF TIMES IE FUNCTIONAL UNI l RESPONSE 1IME

-3 5

~

d 11.

Low keactor Coolant, flow-w

'a.

Single Loop (Above P-8) 11.0 second a

b.

Two Lucps (Above P-7 and below P-8) 11.0 second m

12.

Steam Generator Vater Level--Low-Low

$2,0 (Unit 1), 3 S (l' nit 2) seconds 13.

Undervoltage-Reactor Coolant Pumps (1.5 seconds 14.

Underfrequency-Reactor Coolant Pumps

<0.6 second 15.

Turbine Trip m1 a

Low Iluid Oii Pressure N.A.

w; b.

Turbine 5 top Valve Closure N.A.

o 16.

Safety Injection input from ESF N.A.

N.A.

17.

Reactor Trip System Interlocks H.A.

18.

Reactor Trip Breakers

, N,a 3

19.

Automatic Trip and Interleck Logic N.A.

nn

! =& &

nn e-P W5 wo C

3. 2.

\\se

~

.e a-

l. '

TABLE 4.2-l'

.X.

O

. REACTOR TRIP SYSTEM!!NSTRUMENTAl' ION-SURVEltLANCE REQUIREMEN!S 5

TR!P ANALOG

'ACIUATING MOUfS FOR z

Cl!ANNEL DEVICE WillCH

--4 CHANNEL-CHANNEL OPERATIONAL OPERAT10NAL ACiUAll0N SURVEILLANCE

[ FUNCTIONAL UNIf CHECK CAtlBRATION 1Ed s.

lEST LOGIC ILSI IS REQUIRED

(

1.

Marual Reactor Trip.

N.A.

R (11)

N.A.

1, 2, 3 *, 4 *,5*

2.

Power Range, Neutron Flux High Setpoint S

D(2, 4),

M N. A.

H.A.

1, 2 M(3, 4),

Q(4, 6),

R(4, 5) gg, Low Setpoint S

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 C

4.

Intermediate Range, S

R(4, 5)

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

N.A.

l

,2 Neutron Flux 5.

Saurce Range, Neutran flux 5

R(4, 5)

S/U(1),M(9)

N.A.

2

, 3, 4. S 6.

Overtemperature AT S

R H

N.A.

1, 2

:2 $

7, Overpower AT S

R H

N.A.

.4. A.

1, 2 l$j 8.

Pressarizer Pressure--Low S

R M

N.A.

1

E,E 9.

P.ressurizer Pressure--High 5

R M

N.A.

1, 2 10.

Pressurizer Water Level--High S

R H

N.A.

1 nn EE 11.

Low Reactor Coolant Flow S

R H

N.'A.

N.A.

1

,X2 l 3C l-

l TABLE 4.3-1 (Continued) x E

E5 REACTOR 1 RIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMfNIS E

TRIP ANALOG ACTUATING MODES FOR 2:

CliANNEL-DEVICE WHICH h!

CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACIDAT10N SURVEILLANCE

-FUNCTIONAL UNii CHECK CALIBRATION TEST TEST-LOGIC TEST IS REQUIRED r.

l k

12.

Steam Generator Water Level--

S R

M N.A.

H.A.

1, 2 Low-Low no f

13.

Undervo!tage - Reactor Coolant N.A.

R N.A.

M N.A.

1 Pumrs 14.

Underfrequency - Reactor N.A.

R N.A.

M N.A.

1 Coolant Pumps u>

];

15.

Iurbine Trip y

a.

Low E1uid Oil Pressure N.A.

R N.A.

5/U(1, 10)

N.A.

1 E3 b.

Turbine Stop Ve.lve Closure N. A.

R N.A.

S/U(1, 10)

N.A.

I 16.

Safety Injection Input from N.A.

N.A.

N. A.

R N.A.

1. 2 ESF 17.

Reactor Trip System Interlocks 3,

a.

Intermediate Range jg gg Neutron Flux, P-6 N.A.

R(4)

M N. A.

N. A.

2 SE b.

Low Power Reactor l((((

Trips Block, P-7 N.A.

R(4)

M (8)

N.A.

1 PP c'.

Power Range Neutron

C Flux, P-8 N.A.

R(4)

H (8)

N.A.

1 na o CC

!. E

<+ c

.y

^ ~ ' '

T& LE 4.3-1 (Continued)

E

~ REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS j

TRIP ANALOG ACIUATING MODES FOR E'

CilANNEL DEVICE WillCil 7

CHANNEL' CHANNEL OPERAT10NAL OPERAIIONAL AC1UAI10N SURVEIL 1ANCE

[

FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST-10GIC.IEST IS REQUIRED d.

Low Setpoint Power Range

[.

Neutron Flux, P-10 N.A.

R(4)

M (8)

N.A.

1, 2 e.

Turbine Impulse Chamber Pressure, P-13 N.A.

R M (8)

N. A.

N.A.

1 18.

Reactor Trip Breaker N.A.

N.A.

M (7, 12)

N.A.

1, 2, 3*,

4*, 5*

19.

Automatic Trip and

{

Interlock Logic N.A.

N.A.

N. A.

M (7) 1, 2, 3 *, 4 *, 5

T 20.

Reactor Trip Bypass U

Breakers N.A.

N.A.

M ( 13), R ( 14 )

N.A.

I, 2, 3 *, 4 *,. S a i

aa II aa E E.

22 bh SC

s I

,T,ABLE 4.3-1 (Continued) l TABLE NOTATICN With the Reactor Trip System breakers closed and the Contral Rod Drive System capable of rod withara.va!.

Celow P-6 (Intermediate Range Neutron flux Interlock) Setpoint.

Below P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.

(1)

If not performed in previous 7 days.

(2) -

Comparison of calorimetric to excore power indication above 15% of RATED THERMAL POWER.

Adjust excore cnannel gains consistent with calorimetric pcwer if absolute difference is greater than 2%.

The provisions of Specification 4 0.4 are not applicable for_ entry into MODE 2 or 1.

(3)

Single point comparison of incore to excore axial flux difference above 15% of RATED THERMAL POWER.

Recalibrate if the absolute difference is greater than or equal to 3%.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(4)

Neutron detectors may be excluded from CHANNEL CALIBRATION.

(5)

Detector plateau curves shall be obtained, evaluated, and compared to manufacturer's data.

For the Intermediate Range and Power Range Neutron Flux channels the provisio.1s of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(6)

Incore - Excore Calibration, above 75% of RATED THERMAL POWER.

The provisions of Scecification 4.0.4 are not applicable for entry into MODE 2 or 1.

(7)

Each train shall b? tested at least every 62 days on a STAGGERFD TEST BASIS.

(8) -

With power greater than or equal to the interlock Setpoint the required operational test shall consist of verifying that the interlock is in the required state by observing the permissive annunciator window.

(9)

Monthly surveillance in MODES 3*, 4* and 5* snail also include verification that permissives P-6 and P-10 are in their required state for existing plant conditions by observation of the permis-sive annunciator window.

Monthly surveillance shall include verification of the High Flux at Shutdown Alarm Setpoint of less than or equal to five times background.

(10) -

Setpoint verification is not required.

McGUIRE - UNITS 1 and 2 3/4 3-14 Amendment No.130 (Unit 1) l Amensamnt No. 112 (Unit 2)

i l

TABLE 4.3-1 (Continued) l TABLE NOTATION (11) -

The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independentiy verify the OPERABII.ITY of the undervoltage and shunt trip circuits for the Manual Reactor Trip Function.

(12) -

The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the OPERA 81LITY of the undervoltage and shunt trip attachmentr of the Reactor Trip Breakers.

(13) -

Prior to placing breaker in service, a local manual shunt trip shall be performed.

-(14) -

The automative undervoltage trip capability shall be verifiea

operable, i

l l-l l

McGUIRE UNITS 1 and 2 3/4 3-14a Amendment No.130 (Unit 1) l Amendment No. 112 (Unit 2)-

- 9e O

INSTRUMENTATION l

3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2 The Engineered Safety Features Actuation System (ESFAS) Instrumentation channels and interlocks 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 and with RESPONSE TIMES as shown in Table 3.3-5.

l APPLICABILITY:

As shown in Table 3.3-3.

g ACTION:

a.

With an ESFAS Instrumentation channel or interlock Trip Setpoint less conservative than the value shown in the Allowable Values 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 OPERABLE status with the Trip Setpoint adjusted consistent with the Trip Setpoint value.

b.

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

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 ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be demonstrated to be within the limit at least once per 18 months.

Each test shall include at least one train such that both trains are tested at le6st once per 36 months and one channel per function such that all channels I

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.

f MCGUIRE - UNITS 1 AND ?.

3/4 3-15 Amendment No.130 (Unit 1) l Amendment No.112 (Unit 1)

TABLE 3.3-3 (Continued) l TABLE NOTATION

Trip function may be blocked in this MODE below the P-ll (Pressurizer Pressure Interlock) Setpoint.

1. Trip function automatically blocked above P-11 and may be blocked below P-ll when Safety Injection on low steam pressure is not blocked.

- These values left blank pending NRC approval of three loop operation.

Note 1:

Turbine driven auxiliary feedwater pump will not start on a Dlackcut signal coincident with a safety injection signal.

ACTION STATEMENTS ACTION 14 With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANOBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ; however, one channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2,1, provided the other channel is OPERABLE.

ACTION 15 With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed until performance of the next requirea OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 15a With the number of OPERABLE channels less than the total Number of Channels, operation may proceed until performance of the next required OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . With more than one channel

-inoperable, enter Specification 3.8.1.1.

h ACTION 16 With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the inoperaole channel is placed in the bypassed condition and the Minimum

' Channels OPERABLE requirement is met.

One additional channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2.1.

ACTION 17 With less than the Minimum Channels OPERABLE requirement, operation may continue provided the containment purge supply and exhaust valves are maintained closed.

McGUIRE - UNITS 1 and 2 3/4 3-23 Amendment No.130 (Unit 1) l Amendment No.112 (Unit 2)

t i

TABLE 3.3-3 (Continued) t ACTION STATEMENTS (Continued)

ACTION 18 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, re> tore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 ho'FS.

ACTION 19 - With tne number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied; a.

The inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and b.

The Minimum Channeis OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing of other channels per Specifica*

tion 4.3.11 and specification 4.3.2.1.

ACTION 20 - With less than the Minimum Number of Channels OPERABLE, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months determine by observation of the associated permissive annunciator window (s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3.

t ACTION 21 - With the number of OPERABLE Channels one less than the Minimum Channels OPERABLE requirer 9nt,.be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least-HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2.1 provided the other channel is OPERABLE.

ACTION 22 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 23 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or declare the associated valve inoperable and take the action required by Specification 3.7.1.4.

ACTION 24 - With the number of OPERABLE channels less than the Total Number of Channels, restore the inoperable channel to OPERABLE ~ status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or declare the associated auxiliary feedwater pump. inoperable and.take the action required by Specifit a-tion 3.7.1.2.

With the channels associated with more than one auxiliary feedwater pump inoperable, immediately dqclare the associated. auxiliary feedwater pumps inoperable and'take the action required by Specification 3.7.1.2.

McGUIRE - UNITS 1 and 2 3/4 3-24 Amendment No. 130(Unit 1) l Amendment No. 112(Unit 2)

O l.

TABLE 3.3-4

?

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINIS E

FUNCTIONAL UNIT-TRIP SEIP0lNT All0WABLE VALUES E

31.

Safety Injection,. Reactor Trip, Feedwater Isolation, Component Cooling m

Water,-Start Diesel Generators, and

,g Nuclear Service Water.

a.

Manual Initiation N.A.

N.A.

b.

Automatic Actuation Logic N.A.

N.A.

and Actuation Relays c.

Containment Pressure--High 1 1.1 p<29 1 1.2 psig j

d.

Pressurizer Pressure--Low-tow

> 1845 psig

> 1835 psig e.

Steam Line Pressure - Low

> 775 psig

755 psig 2.

Containment Spray a.

Manual Initiation-N.A.

N.A.

b.

Automatic Actuation Legic H.A.

N.A.

l@

and Actuation Relays

{y c.

Containment Pressure--High-High

$ 2.9 psig 1 3.0 psig 55

.E E NO CC hh S t' A "

e

7 TABLE '3.3-4 (Cc.ntinut:d)

,..g 5

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTAIION TRIP SETPOINTS A

FUNCTIONAL UNIT TRIP SETP0lhT ALLOWABLE VAtUFS 3.

Containment Isolation w

a,-

a.

Phase "A" Isolation EJ 1)

Hanual 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 Irip Setpoints and Allowable Values w1 b.

Phase "fi" Isolation w

4 1)

Manual Initiation N.A.

N.A.

m 2)

Automatic Actuation Logic N.A.

N.A.

and Actuation Relays 3)

Containment Pressure--High-High 1 2.9 psig i 3.0 psig fk c.

Purge and Exhaust Isolation an kk 1)

Manual Initiation N.A.

N.A.

aa r+ r*

2)

Automatic Actuation Logic N.A.

N.A.

22PP and Actuation Relays dd No 3)

Safety Injection See item 1. above for all Safety Injection Trip Setpoints gg and Allowable Values S. S.

ee

..x-..

.6 x

TABLE 3.3-4 (Continued)-

E

_5 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINIS W

[ ' FUNCTIONAL UNIT TRIP SElPOINT AtLOWABLE VALULS 5g 4.

Steam Line Isolation

.. w a.

Manual Initiation N. A.

N. A.

.E b.

Automatic' Actuation Logic

. N. A.

N.A.

m and Actuation Relays c.

Containment Pressure--High-High 1 2.9 psig 5 3.0 psig d.

Negative Steam Line i 100 psi with a i 120 psi with a Pressure Rate - High rate / lag function rate / lag function time constant time constarit gg 3 50 seconds 3 50 seconds Steam Line Pressure - Low

> 775 psig 3 755 psig e.

5.

Turbine Trip and feedwater Isolation a.

Automatic Actuation Logic N.A.

N.A.

and' Actuation Relays

[S E

b.

Steam Generator Water level--

< 82% of narrow range

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

Instrument span each steam Instrument span each steam o

{(

generator generator oo c.

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

Containment Pressure Control System nn EE Start Pern.issive/ Termination 0.3 < SP/T < 0.4 PSIG

~

33 (SP/T)

~

0.25 < SP/T < 0.45 PSIG SU

R R

E E

W W

s oO oO tP o0R tP o0R.

t n

t 0E t 0L i

oL 1

W 0L 1

W o

A y O

p mMl tP mMl tP t

oR ra oR ra e

S rE a L

rE a L

S l

enA t H enA E

f i nam T nam p

V T

n pR n

pR i

l i

i A

a0l sE aDl sE r

s V

pf l

pE l

T t

l l

sl gf I

sT gf i l

E Ano A no n

o S

L f Ri D

f Ri D g o

v T

B o

s%E o

s%E i i

N A

f a0T f a0T s

t 0

W

%oe

.A

%oe

.A p

c 0

1 6

O A.

1 r9R 1

r9R e

2 P

L A

1%c3 1%c3 1

j 3

A T

L 0 n f

0 n f

n E

A N N 33 i > o 33i ;o >-

3 N

I S

y P

t I

e R

f T

a S

N R

R O

E E

l W

W I

l T

oO O

A tP o0R P o0R.

a a

T t0E o

t0E r

h e N

0L 1 W tL 1 W os t m E

Ay O

A y O

f e ii M

mMl tP 0Ml tP u wt U

oR ra R ra el

)

.R rE a L

mE a L

va sd d

T fH enA oH enA oV t n e

S T nam rT nam b

l o u

N T

n i pR f

i pR ae oc n

I N

aDl sE Dl sE l

ve

'i I

pE i

aT gfT 1 a 3 l

nE H

.b s

t M

O sT gfl n

E P

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-TABLE 3.3-4 (Continued) n C1' c5 ENGINEERED SAFETY TEA!URES ACTUATION SYSTEM INSTRUMENTATION TRIP SEIPOINTS E

[

FUNCTIONAL UNIT TRIP SEIPOINT ALLOWABLE VALUES z

[]

8.

Automatic Switchover to Recirculation

[

RW5T Level 3 90 inches 3 80 inches a

9.

Loss of Power y

4 kV Emergency Bus Undervoltage-3464 1 173 volts with a 3 3200 volts Grid Degraded Voltage 8.5 i 0.5 second time delay 10.

Engineered Safety Features Actuation System Interlocks u,

s 4>

a.

Pressurizer Pressure, P-Il

$ 1955 psig S 1965 psig U$

b.

Tavg, P-12 3 553"f

S51 I c.

Reactor Trip, P-4 N.A.

fl. A.

d.

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

sa Note 1:

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

a&

8r c.

ZZ??

20 NO CC

==

t

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ft fP 6

s J

s,@ )%,'.

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. TABLE 3.3-5 i-

y SNGINEERED SAriTY FEATURES RESPONSE TIMES 7,.,

qM

^

. INITIATING SIGNAL AND FUNCTION' RESPONSE TIME ~IN SECONDS T

. 1.

' Manual M.

"i a.

Safety-Injection-(ECCS)

N.A.

[p [n$

.o.

. Containment Spray ~

N.A.

s p'

-c..

Containment.Isolatit>

Phise "A Isolation N.A.

s

-Phase.'"B" Isolation-N.A.

Purge and Exhaust Isolation, N. A.

U

-d.

Steam Line~ Isolation N. A.

'e.--

Feedwater Isolation N.A.

aj,

.f.

Auxiliary Feedwater N.A.

f,

-g.

Nuclear Service Water N.A.

h.

Component Cooling Water N.A.

(la Reactor Trip (from SI)

N.A.

"2j.

istart'0iesel' Generators N.A.

g.

Containment Pressure-High-Sa fety.; Injection - (ECCS).

5 27(1)

-a.

'b.'

Reactor; Trip (from~SI) 1 2--

c.

Feedwater Isolation-

< 12

~

? Containment' Isolation-Phase "A"(2)

_ 18(3)/28(4)

^d.

.e.,

-Containment Purge and Exhaust _ Isolation 14 2

ff.

LAuxiliary Feedwater N.A.

_ g.

- Nuclear ~ S'ervice' Water-

< 65(3)/76(4)

U I h'.-

LComponent-Cooling Water.

65(3)/76(4)

/'

i.

Start -Diesel ; Generators :

< 11 1

1

~ m '.

(_ Q ?

3 s

e-(D &

,r!h I

'McGUIRE --UNITS 1 AND 2-3/4 :-30 Amendment No. t30 (Unit 1) l Amendment No.112 (Unit 2)

>t-

< v w

%,, :.y, '

m-

'(

.;y

'. p j

s g

gy

' TABLE 3.3-5 (Continued) l-

$ m,-

~

ENGINEERED SAFETV FEATURES RESPONSE TIMES su, >

s

's LINITIATING SIGNAL >AND FUNCTION' RESPONSE TIME IN SECON3S 3.

Pressurizer-Pressure-Low-Low y,*

Ea; l Safety Injection (ECCS) 1 27(1)/12(3)'

1

b:

Reactor. Trip;(from SI)

-. <2 o

ce -Feedwater Is'olation

< 12 f

i 1 Containment -Isolation-Phase A"(2b 18

/28 j

I

~

e; ; Containment Purge and Exhaust Isolation

=< 4 i f.

Auxiliary Feedwater(5) y,4,

g.

1 Nuclear Service' Water System

< 76(1)/65(3)

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

[

Li.

' Start: Diesel Gerierators

~

~< 11 3

4. z ; Steam Line Pressure-Low-Safety (I'njection~(ECCS).

1 12(3)/22(4)

^a.

7

~

y P

-. b ;

iReactor Trip!(from SI)?

.< 2

(.

c; ifeedwater' Isolation-

.< 12 i.

Containmentilsolation Phase "A"(2)

{18(3)/28(4)

Y d

2 a

e..

Containment. Purge and Exhaust Isolation-3 4-f.

Auxiliary Feedwater(5)

N.A.

>j g.

. Nuclear Service Water ~

$ 65(3)/76(4)

): h.

! Steam Line. Isolation

< 10 65(3)/76(4) f.

-ComponentJCooling Water

~

"ji TStart Diesel. Generators.

5 11 Si l Containment' Pressure-High-Highs i

Coc+.:.inmentCSpray

-1 45 M, _

^

15 team O ne Isolation "Lt.

Containment Isolatio'n-Pi:ue "B

N.A.

_ - 10 c.

'[#

[6.3 - StAam Generator Water Level-High-High-

=a.

5 Turbine Trip l N.A.

,('-

l b'. - =Feedwater IsolationL 5 12' f.

}Q N

u v

LMcGUIREs-UNITS 15AND'2-3/4 3-31 Amendment No.130 (Unit 1)

[

Amendment-No.112 (Unit 2) n.

e ot;L p.

, a.

TABLE 3.3-5 ' Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTICN RESPONSE TIME IN SECONDS j

7.

Steam Generator Water Level Lo* Low 1

a.

Motor-driven Auxiliary Feedwater Pumps

~

$ 6C b.

Turbine-driven Auxilia,

Feedwater Pumps

60 8.

Negative Steam Line Pressere Rate - High Steam Line Isolation 1 10 9.

Start Permissive Containment Pressure Control Sy s t em N.A.

10, Termination Containment Pressure Control System N.A.

11.

Auxiliary Feedwater Suction Pressure - Low Auxiliary Feedwater Pumps ($uction Supply Automatic Realignment)

$ 13 IL RWST Level Automatic Swi'.: hover to Recirculation 1 60 13.

Station Blackout a.

Start Motor-Driven Auxiliary Feedwater Pumps 1 60 c.

51.et Turbine-Oriven Auxiliary Feedwater Pump (6) 1 60 14.

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

Lor.s of Power 4 kV Emergency Bus Undervoltage-1 11 Grid Degraded Voltage i

McGUIRE - UNITS 1 AND 2 3/4 3-32 Amendment No.130 (Unit 1)

Amendment No.112 (Unit 2) l

e.

TABLE 3.3-5 (Continued) l TABLE NOTATION

-(1) Diesel generator starting and-sequence loading delays included.

Response-time limit includes opening of valves to establish Safety Injection path and attainment of discharge pressure for centrifugal charging pumps,

$afety Injection and RHR pumps.

(2) Valves IKC305B and 1KC3158 for Unit 1 and Valves 2KC305B and 2KC315B for

. Unit 2 are exceptions to the response times listed in the table.

The I

following response times in seconds are the required values for these valves for the initiating signal and function indicated:

2.d

< 30(3)f4g(4) 3)

7 30((3)/40 3.d l4) 530 4.d (3) Diesel generator starting and sequence loading delays not included.

Offsite power available.

Response time limit includes opening of valves to establish Safety Injection path and attainment of discharge pressure for centrifugal charging pumps and Safety Injection pumps.

(4) Diesel generator starting and sequence loading delays included.

Response

time limit includes opening of valves to establish Safety Injection path and attainment of discharge pressure for centrifugal charging pumps and Safety Injection pumps.

.(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 path and attainment of discharge pressure for auxiliary feedwater pumps.

(6) The turbine driven pump does not start on a blackcut signal coincident with a safety injection signal.

McGUIRE - UNITS 1 and 2 3/4 3-33 Amendment No.130 (Unit 1) l Amendment No.112 (Unit 2)

PLANT SYSTEMS:

MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.7'1.4 Eacn main-steam line isolation valve (MSLIV) shall be OPEHABLE.

APPLICABILITY:

MODES 1, 2, and 3.

l ACTION:

MODE 1 - With one MSLIV inoperable but open, POWER GPERATION 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 nours.

MODES 2 - With one MSLIV inoperable, subsequent ope *ation in MODE 2 or 3 may and 3 proceed proviced; a.

The isolation valve is maintained closed, and b.

The provisions of Specification 3.0.4 are not applicable.

Otherwise, be in H01 STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN 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.

s McGUIRE - UNITS 1 AND 2 3/4 7-8 Amendment No.130 (Unit 1) l Amendment No.112 (Unit 2)

- w 2.1 SAFETY LIMITS l

e BASES m

2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission

-products to the reactor coolant.

Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface 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 depacture 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 developad to predict the DNB flux and the lccation of DNB for axially uniform and nonuniform heat flux distributions.

The local DNB heat flux catio (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 tne minimum DNBR of the limiting rod during Condition I and II events is greater than or equal to the ONBR limit of the ONB 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-relation are considered 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

_ parameters without unc:rtainties.

The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure,:and average temperature belnw 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 curve's are based on a nuclear enthalpy rise hot channel factor, Fg, of 1.50 and a reference-cosine axial power shape with a peak of 1.55.

An allow-ante is included for an increase in at reduced power based on the expression:

H N

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

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

McGUIRE - UNITS 1 AND 2 8 2-1 Amendment No.130 (Unit-1) l Amendment No.112 (Unit 2).

.s LIMITING SAFETY SYSTEM SETTINGS l

1 BASES-Power Range, Neutron Flux (Continued) o The Low Setpoint 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 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 ejectior m partial power.

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 suberitical condition.

These trips provide redundant protection to the Low Setpoint trip 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 cnannels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active.

\\-

l.

McGUIRE - UNITS 1 AND 2 B 2-4 Amendment No.130 (Unit 1) l Amendment No.112 (Unit 2) l

3/4.2 POWER O!ST.RIBUTION LIMITS BASES The specifications of this section provide assurance of fuci 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 tilat the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria are not exceeded.

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

T (X,Y,Z) Heat Flux Hot Channel Factor, is defined as the local heat flux 0

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 H(X,Y) Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along a rod at core location X,Y to the average rod power, K(z) is defined as the normalized F (X,Y,Z) limit for a given core height.

g

-3/4.2.1 AXIAL FLUX DIFFERENCE The limits on AXIAL FLUX DIFFERENCE (AFD) ensure that F (X,Y,Z) and n

ekg(X,Y)limitsspecifiedintheCOREOPERATINGLIMITSREPORY(COLR)arenot F

deeded during either normal operation or in the event of xenon redistribution "ollowing power changes.

The AFD envelop specified in the COLR has been adjusted for measurement uncertainty.

l I

l i

I McGUIRE - UNITS 1 AND 2 8 3/4 2-1 Amendment No.130 (Unit 1) l Amendment No.112 (Unit 2)

POWER DISTRIBUTION LIMITS f

3 i

BASES i

AXIAL FLUX DIFFERENCE (Continued)

Tne computer determines the one sinute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are outside the allowed al power operating space during normal power operation.

These alarms are active when power is greater than 50% of RATED THERMAL DOWER.

l l:

l:

l l

McGUIRE - UNITS 1 AND 2 B 3/4 2-2 Amendment No.130 (Unit 1) l Amendment No.112 (Unit 2)

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

POWER DISTRIBUTION LIMITS

-BASES-

.3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY RISE l

HOT CHANNEL FACTOR The limits on heat flux hot channel f actor, 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 acceptance criteria are not exceeced.

The peaking limits are specified in the CORE OPERATING LIMITS REPORT (COLR)-per Specification 6.9.1.9.

The heat flux hot channel factor and auclear enthalpy rise hot channel factor are each measurable, but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3.

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

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; 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 I

3.1.3.6 are maintained; and 1

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

F3g(X,Y) will be maintained within its limits provided Conditions a. through d.

above are maintained.

The limits on the nuclear enthalpy rise hot channel factor, FaH( ' I' are specified in the COLR as Maximum Allowable Radial Peaking (MARP) limits, 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 ONBR for the limiting transient tht is equivilent to the DNBR calculated with a design Fg(X,Y) value of 1.50 and a limiting reference axial power shape.

For transition 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 ONBR effects.

The relaxation of Fg(X,Y) as a function of THERMAL POWER alloc for a change in the radial pwoer shape for all permissible control bank insertion limits.

This relaxation is implemented by tne 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-

t McGUIRE - UNITS 1 AND 2 B 3/4 2-3 Amendment No.130 (Unit 1) l Amendment No.112 (Unit 2)

POWER DISTRIBUTION LIMITS 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 g

factor, F H(X,Y), are measured periodically to verify that the core is operating as designed.

F (X,Y,Z) and F g (X,Y) are compared to allowable limits to provide reasonable assurance that limiting criteria will not be exceeded for operation within the Technical Sperification limits of Sections 2.2 (Limiting Safety Systems Settings), 3.1.3 (Movable Control Assemblies),

3.2.1 (Axial Flux Difference), and 3.2.4 (Quadrant Power Tilt Ratio).

A peaking margin calculation is performed to provide the basis for decreasing the width of the AFD and f(.11) limits and for reducing THERMAL POWER.

When an F (X,Y,Z) measu ement is obtained from a full-core map in accordance with surveillance requirements of Specification 4.2.2, no uncertainties are applied to the measured pt.ak since a measurement uncertainty of 5.0% and a

'nanufacturing tolerance of 3.0% are included in the peaking limit.

When M

F (X,Y,Z) is measured for reasons other than meeting the requirements of q

Specification 4.2.2, the measured peak is increased by the radial-local peaking factor and appropriate allowances for measurement uncertainty and for manufacturing tolerances.

When an F g(X,Y) measurement is 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.

McGUIRE - UNITS 1 AND 2 8 3/4 2-4 Amendment No.130(Unit 1)

Amendment No.112(Unit 2).

- ~. _

POWER DISTRIBUTION LIMITS BASES 3/4.2.4 QUADRANT POWER TILT RATIO 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.

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

The peaking increase that corresponds to a QUADRANT p0WER TILT RATIO of 1.02 is included in the generation of the AFD limits.

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 g

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

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

The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimb!es.

3/4.2.5 DNB PARAMETERS The limits on the DNB-related parameters assure that each of the para-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 adequate to maintain a design limit DNBR throughout each analyzed transient.

As noted on Figure 3.2-1, RCS flow rate and THERMAL POWER may be " traded off" against one another (i.e., a low measured RCS flow rate is acceptable if the-power level is decreased) to ensure that the calculated DNBR will not be below the design DNBR value.

The relationship defined on Figure 3.2-1 remains valid as long as the limits placed on the nuclear enthalpy rise hot channel factor, Fg(X,Y), in Specification 3.2.3 are maintained.

The indicated T,yg values and the indicated pressurizer pressure values correspord to analytical limits of 592.6 F-and 2220 psia respectively, with allowan:e for indication instrumen-tation measurement uncertainty, When RCS flow rate is measured, no additional allowances are-necessary prior to comparison with the limits of Figure 3.2-1 since a measurement error of 1.7% for RCS total flow rate has been allowed for in-determination of the design DNBR valt The measurement error for RCS total flow rate is based upon performing a precision heat balance and using the result to calibrate the RCS fjow rate indi-cators.

Potential fouling of the feedwater venturi which might not be detected could bias the result from the precision heat balance in a non-conversative r

McGUIRE - UNITS 1 AND 2 B 3/4 2-5 Amendment No.130 (Unit 1) l Amendment No.112 (Unit 2)

POWER DISTRIBUTION LIMITS l

BASES 3/4.2.5 DNB PARAMETERS (Continued) manner.

Therefore.a penalty of 0.1% for undetected fouling of the feedwater venturi is included in Figure 3.2-1.-

Any fouling which might bias 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 heat balance mesurements, i.e.,

either the effect of the fouling shall be quantified and compensated for in the RCS flow rate measurement or the venturi shall be cleaned to eliminate the fecling.

The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits fo11owing load changes and other expected transient operation.

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

4 4

}

McGUIRE - UNITS 1 AND 2 B 3/4 2-Sa Amendment No.130 (Unit 1)

Amendment No i12 (Unit 2)

~

3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION 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 STANDBY-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 capability for removing decay heat; however, single failure considerations require that three_ loops be OPERABLE.

Also, the uncontrolled bank withdrawal from zero power or subcritical assumes three reactor coolant loops in operation.

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

'In MODE 5 with reactor coolant loops not filled, a single RHR loop provides sufficient 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 coolant pump (RCP) or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System.

The-reactivity change rate associated with boron reduction 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 will be protected against overpressure transients and will not exceed _the limits of Appendix G by either:

(1) restricting the water volume in the pressurizer 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.

t McGUIRE - UNITS 1 AND 2 B 3/4 4-1 Amendment No.130 (Unit 1)

Amendment No.112 (Unit 2)

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