Text

Amends 165 & 147 to Licenses NPF-9 & NPF-17,respectively, Revising TS to Increase STIs & AOTs for Reactor Trip System & ESFAS
	ML20097G824
Person / Time
Site:	McGuire, Mcguire
Issue date:	02/15/1996
From:	Berkow H; NRC (Affiliation Not Assigned)
To:	;
Shared Package
ML20097G826	List: ML20097G824; ML20097G831;
References
	NUDOCS 9602220025
	Download: ML20097G824 (24)
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4 h

UNITED STATES j,

NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20606-0001

.....,o DUKE POWER COMPANY DOCKET N0. 50-369 McGUIRE NUCLEAR STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.165 License No. NPF-9 1.

The Nuclear Regulatory Commission (the Commission) 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 January 13, 1995, as supplemented by letter dated August 30, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (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 provisions of the Act, 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 amendment 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.

9602220025 960215 PDR ADOCK 05000369 p

PDR

9

. 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.

165, are hereby 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 and shall be implemented within 30 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION l

Herbert N. Berkow, Direct r Project Directorate II-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Technical Specification Changes Date of Issuance:

February 15, 1996

i

. pung g

i UNITED STATES g

j NUCLEAR REGULATORY COMMISSION 2

WASHINGTON, D.C. 20086 4 001

\\,...../

DUKE POWER COMPANY DOCKET NO. 50-370 McGUIRE NUCLEAR STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.

147 License No. NPF-17 1.

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

A.

The a;, plication for amendment to the McGuire Nuclear Station, Unit 2 (the facility), Facility Operating License No. NPF-17 filed by the Duke Power Company (licensee) dated January 13, 1995, as supplemented by letter dated August 30, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (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 provisions of the Act, 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 amendment 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.

-4 l 2.

Accordingly, the license is hereby amended by page changes to the l

Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Facility Operating License No.

NPF-17 is hereby amended to read as follows:

Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 147, are hereby 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 and shall be implemented within 30 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

{

5f erbert N. Berkow, Di e tor Project Directorate II-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Technical Specification Changes Date of Issuance: February 15, 1996 i

l

W l

1 l

ATTACHMENT TO LICENSE AMENDMENT NO. 165 FACILITY OPERATING LICENSE NO. NPF_1 DOCKET NO. 50-369 l

)

ANQ TO LICENSE AMENDMENT NO. 147 FACILITY OPERATING LICENSE NO. NPf-ll DOCKET NO. 50-370 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 Paoes 3/4 3-3 3/4 3-3 3/4 3-4 3/4 3-4 3/4 3-5 3/4 3-5 3/4 3-6 3/4 3-6 3/4 3-7 3/4 3-7 3/4 3-11 3/4 3-11 3/4 3-12 3/4 3-12 3/4 3-13 3/4 3-13 3/4 3-14 3/4 3-14 3/4 3-22 3/4 3-22 3/4 3-23 3/4 3-23 3/4 3-24 3/4 3-24 3/4 3-29 3/4 3-29 3/4 3-34 3/4 3-34 3/4 3-3,5 3/4 3-35 3/4 3-36 3/4 3-36 3/4 3-37 3/4 3-37 3/4 3-38 3/4 3-38 8 3/4 3-1 B 3/4 3-1 l

Ne TABLE 3.3-1 (Continued) 5

~

N REACTOR TRIP SYSTEN INSTRUNENTATION E

NININUN Q

TOTAL NO.

CHANNELS CHANNELS APPLICABLE i

FUNCTIONAL UNIT OF CHAMELS TO TRIP OPERA 8LE MODES ACTION

(

7.

Overpower AT 4

Four Loop Operation 4

2 3

1, 2 6

Three Loop Operation

(**)

(

8.

Pressurizer Pressure-Low 4

2 3

1 6

l 9.

Pressurizer Pressure--High 4-2 3

1, 2 6

~

10. Pressurizer Water Level--High 3

2 2

1 6

Y t

11. Low Reactor Coolant Flow j

t w

1 O

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

5 t

a.

any oper-each oper-ating loop ating loop t

b.

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

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

((

ating loops ating loop i

ss

((

12. Steam Generator Water 4/sta. gen.

2/sta. gen.

3/sta. gen.

1, 2 J

l

==

Level--Low-Low in any oper-each oper-l

((

ating sta.

[

op gen.

gen.

l

~

+

(

t l

8m

(

on 4."

ew l

2

E TABLE 3.3-1 (Continued) m 7

REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUN g

TOTAL NO.

CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF. CHANNELS TO TRIP OPERABLE MODES ACTION k

13. Undervoltage-Reactor Coolant Pumps (above P-7) 4-1/ bus 2

3 1

6 m

14. Underfrequency-Reactor Cool g t Pumps (above P-7) 4-1/ bus 2

3 1

6

15. Turbine Trip a.

Low Fluid Oil Pressure 3

2 2

1 6

b.

Turbine Stop Valve Closure 4

4 1

1 11 R

16. Safety Injection Input from ESF 2

1 2

1, 2 7

l

17. Reactor Trip System Interlocks a.

Intennediate Range Neutron Flux, P-6 2

1 2

2,,

8 b.

Low Power Reactor Trips Block, P-7

>k P-10 Input 4

2 3

1 8

g

==

or kk P-13 Input 2

1 2

1 8

oa

((

c.

Power Range Neutron

,o o Flux, P-8 4

2 3

1 8

E$

d.

Low Setpoint Power Range Neutron Flux, P-10 4

2 3

1, 2 8

EE e.

Turbine Impulse Chamber hh Pressure, P-13 2

1 2

1 8

~~

ww s

M i

E TABLE 3.3-1 (Continued)

REACTOR TRIP' SYSTEM INSTRUMENTATION I

e f

U=

MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE

[

FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION k

18. Reactor Trip Breakers 2

1 2

1, 2 9, 12 2

1 2

3*, 4*, 5*

10 y

19. Automatic Trip and Interlock 2

1 2

1, 2 7

l Logic 2

1 2

3*, 4*, 5*

10 l

?

l U.

t i

b l.

aa 25

%a

.? ?

hsm b

bh 30 s,

M TABLE 3.3-1 (Continued)

TABLE NOTATION

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

- Values left blank pending NRC approval of three loop operation.

I II elow the P-6 (Intermediate Range Neutron Flu'x Interlock) Setpoint.

8 IIIBelow the P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.

ACTION STATEMENTS ACTION 1 -

With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, 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 HOT STAND 8Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 2 -

With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

The inoperable channel is placed in the tripped condition a.

within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , b.

The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels per Specification 4.3.1.1, and Either, THERMAL POWER is restricted to less than or equal c.

to 75% of RATED THERMAL POWER and the Power Range Neutron Flux Trip Setpoint is reduced to less than or equal to 85%

of RATED THERMAL POWER within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; or, the QUADRANT POWER TILT RATIO is monitored 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 and 2 3/4 3-6 Amendment No.165 Unit 1)

Amendment No.147 Unit 2)

4 TABLE 3.3-1 (Continued)

ACTION STATEMENTS (Continued) l ACTION 3 -

With the number of channels OPERABLE one less than the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:

a.

Below the P-6 (Intermediate Range Neutron Flux Interlock) i Setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 l

Setpoint, and i

i b.

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

Setpoint but below 10% of RATED THERMAL POWER, restore the l'

inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 10% of RATED THERMAL POWER.

I ACTION 4 -

With the number of OPERABLE channels one less than the Minimum i

Channels OPERABLE requirement suspend all operations involving i

positive reactivity changes.

ACTION 5 -

With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1 or i

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 j

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

F ACTION 6 -

With the 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 j

within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and b.

The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels per Specification 4.3.1.1 and Specification 4.3.2.1.

ACTION 7 -

With the' number of OPERABLE Channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least H0T STANDBY s

within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed i

for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per' Specification 4.3.1.1, provided the other channel is OPERABLE.

ACTION 8 -

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.

McGUIRE - UNITS 1 and 2 3/4 3-7 Amendment No. 165 Unit 1)

Amendment No.147 Unit 2)

TABLE 4.3-1 W@

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS g

m TRIP ANALOG ACTUATING MODES FOR 3

CHANNEL DEVICE WilCH p

CHANNEL CHANNEL ~

OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CAllBRATION TEST TEST LOGIC TEST IS REQUIRED

{

l.

Manual Reactor Trip N.A.

N.A.

R (11)

N.A.

1, 2, 3 *, 4 *, 5

l 1

N 2.

Power Range, Neutron Flux -

High Setpoint S

MI,;3,4, DI 2, 4,

Q N.A.

N.A.

1, 2 l'l Q(4. 6,

Rt4, 5 Low Setpoint S

R(4)

S/U(1)

N.A.

1#II, 2 l'

3.

Power Range, Neutron Flux, N.A.

R(4)

Q N.A.

N.A.

1, 2 l

High Positive Rate w

I 4.

Intermediate Range, S

R(4,5)

S/U(1)

N.A.

IIII, 2 l

~

Neutron Flux 5.

Source Range, Neutron Flux S

R(4,5)

S/U(1),Q(9)

N.A.

2II, 3, 4, 5 EN gg 6.

Overtemperature AT S

R(15)

Q N.A.

N.A.

1, 2 gg 7.

Overpower AT S

R(15)

Q N.A.

N.A.

1, 2 en g@

8.

Pressurizer Pressure--Low S

R Q

N.A.

1 i

2;;;

9.

Pressurizer Pressure--High S

R Q

N.A.

1, 2 sm 10.

Pressurizer Water Level--High S R

Q N.A.

N.A.

1 M

~

N.A.

I

5. 3.

11.

Low Reactor Coolant Flow S

R Q

en U

~

t z

9, TABLE 4.3-1 (Continued) 5 M

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIRENENTS e

TRIP

-5 ANALOG ACTUATING MODES FOR d

CHANNEL DEVICE WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE

~

FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED E

12. Steam Generator Water Level-- S R

Q N.A.

N.A.

1, 2 l

m Low-Low

13. Under,ltage - Reactor N.A.

R N.A.

Q N.A.

1 l

Coolant Pumps

14. Underfrequency - Reactor N.A.

R N.A.

Q N.A.

1 l

Coolant Pumps w2

15. Turbine Trip

a. Low Fluid Oil Pressure N.A.

R N.A.

S/U(1,10)

N.A.

1 w'

b. Turbine Stop Valve Closure N.A.

R N.A.

S/U(1,10)

N.A.

I N

16. Safety Injection Input from N.A.

N.A.

R N.A.

1, 2 ESF

17. Reactor Trip System Interlocks j

RR

a. Intermediate Range 22 Neutron Flux, P-6 N.A.

R(4)

N.A N.A.

N.A.

2II RE

b. Power Range Neutron z2 Flux, P-8 N.A.

R(4)

N.A N.A.

N.A.

1 23,

~u b""

nn

z E

TABLE 4.3-1 (Continued) 5

~

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS j%

c TRIP 5

ANALOG ACTUATING MODES FOR d

CHANNEL DEVICE WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE

~

=

FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED E

l

c. Low Setpoint Power Range m

Neutron Flux, P-10 N.A.

R(4)

N.A N.A.

N.A.

1, 2

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

R 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 w2 Interlock Logic N.A.

N.A.

M (7) 1, 2, 3*,

4*, 5*

w

20. Reactor Trip Bypass Breakers N.A.

N.A.

M(13),R(14)

N.A.

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

at 25 aa

.?.?

w bEE ne O

l TABLE 4.3-1 (Continued)

TABLE NOTATIO_H With the Reactor Trip System breakers closed and the Control Rod Drive System capable of rod withdrawal.

Below P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.

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

Eetpoint.

(1)

If not performed in previous 31 days.

(2)

Comparison of calorimetric to excore power indication above 15% of RATED THERMAL POWER. Adjust excore channel gains consistent with calorimetric power 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 dif-ference is greater than or equal to 3%. The provisions of Specifi-cation 4.0.4 are not appli' cable 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 Intemediate Range and Power Range Neu-tron Flux channels the provisions 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 Specifiation 4.0.4 are not applicable for entry into MODE 2 or 1.

(7)

Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.

(8)

Deleted.

l (9)

Quarterly surveillance in MODES 3*, 4* and 5* shall also include l

verification that permissives P-6 and P-10 are in their required state for existing plant conditions by observation of the permissive annunciator window. Quarterly surveillance shall include l

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. 165 Unit 1)

Amendment No. 147 Unit ?)

L g

TABLE 3.3-3 (Continued) e ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION m

e MINIMUM i

g TOTAL'NO.

CHANNELS CHANNELS APPLICABLE q

FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION m

7.

Auxiliary Feedwater (continued) f.

Station Blackout (Note 1)

Start Motor-Driven Pumps m

and Turbine-Driven Pump

1) 4 kV Loss of Voltage 3/ Bus 2/ Bus 2/ Bus 1, 2, 3 19 Either Bus

2) 4 kV Degraded Voltage 3/ Bus 2/ Bus 2/ Bus 1, 2, 3 19 Either Bus g.

Trip of All Main Feedwater Pumps w

Start Motor-2 Driven Pumps 2-1/MFWP 2-1/MFWP 2-1/MFWP 1, 28 27 w0 8.

Automatic Switchover to Recirculation RWST Level 3

2 2

1,2,3 15b 9.

Loss of Power a.

4 kV Loss of Voltage 3/ Bus 2/ Bus 2/ Bus 1, 2, 3, 4 15a EE gg b.

4 kV Degraded Voltage 3/ Bus 2/ Bus 2/ Bus 1, 2, 3, 4 15a an 55

10. Engineered Safety Features 5E Actuation System Interlocks gg a.

Pressurizer Pressure, 3

2 2

1,2,3 20 P-11 b.

Low-Low T, P-12 4

2 3

1,2,3 20 Reactor T, rip, P-4 2

2 2

1, 2, 3 22 0$

c.

g Level, P-14 2/sta gen.

2/sta gen.

1, 2, 3 20 d.

Steam Generator 3/ste gen.

in any in each ga operating operating sta gen.

sta gen.

-a 30 A

'l TABLE 3.3-3 (Continued)

TABLE NOTATION

.j Trip function may be blocked in this MODE below the P-11 (Pressurizer Pressure Interlock) Setpoint.

j Trip function automatically blocked above P-11 and may be blocked below P-11 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 Mackout 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 STAND 8Y within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-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 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 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 perfonnance of the next required OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

l ACTION 15a With the number of OPERABLE channels less than the Total Number of Channels, operation may proceed until performance of the next i

required OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . With more than one channel inoperable, enter Specification 3.8.1.1.

l ACTION 15b With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed until perfonnance 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 .

ACTION 16 With the number of OPERABLE channels one less than the Total Number of Channels, omration may proceed provided the inoperable channel is placed in tie bypassed condition and the Minimus Channels OPERABLE requirement is met. One additional channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.2.1.

l ACTION 17 With less than the Ninimum 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.165 Unit 1 Amendment No.147 Unit 2

.j TABLE 3.3-3 (Continued)

ACTION STATEMENTS (Continued)

ACTION 18 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, 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 STANOBY 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 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ACTION 19 - With the 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:

i The inoperable channel is placed in the tripped condition a.

within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and l

b.

The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for sur-l veillance testing of other channels per Specification 4.3.1.1 and Specification 4.3.2.1.

ACTION 20 - With less than the Minimum Number of Channels OPERABLE, within I hour determine by observation of the associated pemissive annun-ciator window (s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3.

ACTION 21 - With the number of OPERABLE Channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable Channel to OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT STANOBY within the next 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 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 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 ino)erable and take the action required by Specification 3.7.1.2.

Witi the channels associated with more than one auxiliary feedwater pump inoperable, immediately declare 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. 165 Unit 1 Amendment No.147 Unit 2

TABLE 3.3-4 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEN INSTRUNENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES

[

8.

Automatic Switchover to Recirculation 2

3 RWST Level 2 90 inches a 80 inches

[

9.

Loss of Power

s

[

Unit 1 a) 4 kV Loss of Voltage 3174

45 volts with a a 3122 volts 8.5

0.5 second time delay b) 4 kV Degraded Voltage a 3678.5 volts with s li a 3661 volts second with SI and s 600 second without SI time delays g

Unit 2 a

y a) 4 kV Loss of Voltage 3157 i 45 volts with a 2 3108 volts 8.5

0.5 second time delay N*

b) 4 kV Degraded Voltage a 3703 volts with s 11 2 3685.5 volts second with SI and s 600 second without SI time delays

((

10.

Engineered Safety Features Actuation gg System Interlocks 55 gg a.

Pressurizer Pressure, P-11 s 1955 psig s 1%5 psig y[

b.

T,, P-12 2 553'F 2 551*F

[*j c.

Reactor Trip, P-4 N.A.

~

N.A.

d.

Steam Generator Level, P-14 See Item Sb. above for_ all Trip Setpoints and Allowable l

Q Values.

7,

Note 1:

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

2

i y

TABLE 4.3-2 E

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION m

SURVEILLANCE REQUIREMENTS i

g TRIP i

ANALOG ACTUATING MODES m

CHANNEL DEVICE-MASTER SLAVE FOR WHICH t

CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE ~

g FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED a

1. Safety Injection, Reactor m

Trip, Feedwater Isolation, Component Cooling Water, Start Diesel Generators, and Nuclear Service Water a.

Manual Initiation M.A.

N.A.

R N.A.

N.A.

1, 2, 3, 4 b.

Automatic Actuation M.A.

N.A N.A N.A.

M(1)

Q 1, 2, 3, 4 Logic and Actuation w

Relays a

r w

c.

Containment Pressure-S R

Q N.A.

N.A.

1, 2, 3 i

High

?

w*

d.

Pressurizer Pressure-S R

Q N.A.

N.A.

1, 2, 3 l

Low-Low e.

Steam Line S

R Q

N.A.

1, 2, 3 Pressure--Low g

2. Containment Spray E.

a.

Manual Initiation M. A.-

N.A.

R

~ N. A. -

N.A.

1, 2, 3, 4 I

b.

Automatic Actuation

'N.A.

N.A.

N(1)

M(1)

Q 1, 2, 3, 4 5

Logic and Actuation

=

Relays

?

c.

Containment Pressure-- S R

Q N.A.

High-High

~N.A.

N'.A.

N.A.

1, 2, 3

.l l

.m i

.Ac I

S Y

O

.a

i g

TABLE 4.3-2 (Continued) b ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION i"4 SURVEILLANCE REQUIREMENTS j

e TRIP 5

5 ANALOG ACTUATING MODES U

' CHANNEL DEVICE MASTER SLAVE FOR WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL. ACTUATION RELAY RELAY SURVEILLANCE g

FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED a.

m

3. Containment Isolation a.

Phase "A" Isolation

1) Manual Initiation N.A.

N.A.

R N.A.

N.A.

1, 2, 3, 4

2) Automatic Actua-N.A.

N.A.

M(1)

Q 1, 2, 3, 4 j

tion Logic and Actuation Relays

3) Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.

j w

w b.

Phase "B" Isolation L

1) Manual Initiation N.A.

N.A.

R N.A.

N.A.

1, 2, 3, 4

2) Automatic Actua-N.A.

N.A.

M(1)

Q 1, 2, 3, 4 i

tion Logic and Actuation Relays

3) Containment S

R Q

.N.A.

N.A.

1, 2, 3 Pressure-High-High 2

E c.

Purge and Exhaust

)

Isolation i

re

1) Manual Initiation N.A.

N.A.

R N.A.

N.A.

1, 2, 3, 4 z

?

2) Automatic Actua-N.A.

M.A.

N.A.

N. A -

M(1)

Q 1, 2, 3, 4

.g tion Logic and Actuation Relays

=

3) Safety Injection See Ites 1. above for all Safety Injection Surveillance Requirements.

i l

{

a

I jq:

TABLE 4.3-2 (Continued)

~

m h

Q GJNEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS m

g TRIP ANALOG ACTUATING MODES U

CHANNEL DEVICE MASTER SLAVE FOR WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE:

g FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED a.

N

4. Steam Line Isolation a

a.

Manual Initiation M.A.

N.A.

R N.A.

N.A.

1,2,3 b.

Automatic Actuation M.A.

N.A.

M(1)

Q 1,2,3 Logic and Actuation Relays lL c.

Containment Pressure-- S R

Q N.A.

N.A.

1, 2, 3 High-High i

w d.

Negative Steam Line S

R Q

.N.A.

N.A.

3 l'

A Pressure Rate-High w

e.

Steam Line S

R Q

N.A.

1, 2, 3 l,

g Pressure--Low

5. Turbine Trip and Feedwater Isolation a.

Automatic Actuation N.A.

N.A.

M(1)

Q 1, 2

>g Logic and Actuation 55 Relay E. g.

b.

Steam Generator Water S R

Q N.A.

M(1)

Q 1, 2, 3 l!

25 Level-High-High (P-14)

L 55 c.

Doghouse Water S

N.A N.A R

N.A.

I, 2 2 :m Level-High

??

(FeedwaterIsolation

~

r 25 Only) i sm

6. Containment Pressure Control System g

Start Permissive /

S R

M N.A.

N.A.

1, 2, 3, 4

=t =,,

Tereination ae j.

I g

TABLE 4.3-2 (Continued)

E I

E ENGINFFRED SAFETY FEATURES ACTUATION SYSTEM INST!"_ MENTATION

~

SURVEILLANCE R,EQUIREMENTS g

TRIP

~

ANALOG ACTUATING MODES m

CHANNEL DEVICE NASTER SLAVE FOR WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE!

FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED '

7. Auxiliary Feedwater N

a.

Manual Initiation M.A.

N.A.

R N.A.

N.A.

1, 2, 3 b.

Automatic Actuation M.A.

N.A.

M(1)

N(1)

Q 1, 2, 3 Logic and Actuation Relays j

t w

c.

Steam Generator Water S R

Q N.A.

N.A.

N.A N.A.

1, 2, 3 Level--Low-Low l

w k

d.

Auxiliary Feedwater N.A.

R N.A.

N.A.

1, 2, 3 Suction Pressure-Low i

e.

Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements f.

Station Blackout N.A.

N.A.

N.A R

N.A.

1, 2, 3 g.

Trip of Main Feedwater N.A.

N.A.

R N.A.

N.A.

N.A 1, 2 g

Pumps SN

8. Automatic Switchover to

=z Recirculation

??

RSWT Level S

R M

N.A.

1, 2, 3 gg Nm

9. Loss of Power g

a.

4 kV Loss of Voltage N.A.

R N.A.

M N.A.

N.A.

N.A 1, 2, 3, 4 l

=a b.

4 kV Degraded Voltage N.A.

R N.A.

N N.A.

N.A.

N.A 1, 2, 3, 4 DL~.

?

.h

l t

TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEN INSTRUNENTATION 5

SURVEILLANCE REQUIREMENTS E

TRIP c

ANALOG ACTUATING MODES 5

CHANNEL DEVICE MASTER SLAVE FOR WHICH d

CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCI FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED!

o, E

10. Engineered Safety n

Features Actuation System Interlocks a.

Pressurizer N.A.

R Q

N.A.

1, 2, 3 l!

Pressure, P-11 b.

Low, Low T,, P-12 N.A.

R Q

N.A N.A.

N.A.

1, 2, 3 ll c.

Reactor Trip, P-4 N.A.

N.A.

R N.A.

N.A.

I, 2, 3 Y

M d.

Steam Generator 1

Level, P-14 See Item Sb for all surveillance requirements.

aa RR aa

??

OO 22 11 en i

A

4' 3/4.3 INSTRUMENTATION i

BASES i

3/4.3.1 and 3/4.3 2 REACTOR TRIP AND ENGINEERED SAFETY FEATURES AC10ATION 2

SYSTEM INSTRUMENIATION The OPERABILITY of the Reactor Trip and Engineered Safety Features Actuation System instrumentation and interlocks ensure that:

1 the associated ACTION and/or Reactor trip will be initiated when the param(e)er monitored by each t

channel or combination thereof reaches its Setpoint (2) the specified coincidence logic and sufficient redundancy is maintained to permit a channel to be out-of-service for testing or maintenance consistent with maintaining an appropriate level of reliability of the Reactor Protection and Engineered Safety Features Instrumentation and (3) sufficient system functions capability is available from diverse parameters.

The OPERA 8ILITY of these systems is required to provide the overall reliability, redundancy, and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions. The integrated operation of each of these systems is consistent with the assumptions used in the accident analyses. The Surveillance Requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability.

Specified surveillance intervals and surveillance and maintenance outage times-have been determined in accordance with WCAP-10271, " Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System," and supplements to that report. Surveillance intervals and out of service times were determined based on maintaining an appropriate level of reliability of the Reactor Protection System and Engineered Safety Features instrumentation. The NRC Safety Evaluation Reports for the WCAP-10271 series were provided in letters dated February 21, 1985 from C. O. Thomas (NRC) to J.

J. Sheppard (WOG), February 22, 1989 from C. E. Rossi (NRC) to R. A. Newton (WOG), and April 30, 1990 from C. E. Rossi (NRC) to G. T. Goering (WOG).

The measurement of response time at the 'specified frequencies provides assurance that the Reactor trip and the Engineered Safety Feature actuation associated with each channel is completed within the time limit assumed in the accident analyses. No credit was taken in the analyses for those channels with response times indicated as not applicable. Response time may be demonstrated by any series of sequential, overlapping, or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either: (1) in ' place, onsite, or offsite test measurements, or (2) utilizing replacement sensors with certified response times.

The Engineered Safety Features Actuation System senses selected plant parameters and determines whether or not predetermined limits are being exceeded. If they are, the signals are combined into logic matrices sensitive to combinations indicative of various accidents, events, and transients. Once the required logic combination is completed, the system sends actuation signals to those Engineered Safety Features components whose aggregate function best serves the requirements of the condition. As an example, the McGUIRE - UNITS 1 and 2 B 3/4 3-1 Amendment No.16 Unit 1 Amendment No.14 Unit 2

---

ML20097G824

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