Text

Amends 55 to License NPF-37 & NPF-66,respectively,revising Plant TS Re ESFAS Instrumentation
	ML20057F197
Person / Time
Site:	Byron
Issue date:	10/04/1993
From:	Dyer J; Office of Nuclear Reactor Regulation
To:	; Commonwealth Edison Co
Shared Package
ML20057F198	List: ML20057F197; ML20057F199;
References
	NPF-37-A-055, NPF-66-A-055 NUDOCS 9310140281
	Download: ML20057F197 (31)
	v • d • e

_.

_m...

b-p otop

+, -

UNITED STATES

'h^i

.. j

- NUCLEAR REGULATORY COMMISSION l

/

. j

\\, df.....f l

WASHINGTON, D.C. 20565-0001 g

COMMONWEALTH EDIS0N COMPANY-DOCKET NO. STN 50-454

!l BYRON STATION. UNIT NO. 1 l

AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 55 l

License No. NPF-37 1.

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

j t

A.

The application for amendment by Commonwealth Edison Company (the licensee) dated August 5, 1992, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth.in i

10 CFR Chapter I; i

B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the i

Commission, t

C.

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

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, 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; j

and i

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.

2.

Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-37 is hereby amended to read as follows:

l t

9310140281 931004 PDR ADOCK 05000454 P

PDR t

. = _..

I

. (2)

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 55 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, 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 the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION M4O $ -

James E. Dyer, Director Project Directorate III-2 Division of Reactor Projects - III/IV/V Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

October 4, 1993 I

s

. p* **%

4 UNITED STATES

. y.k _

j NUCLEAR REGULATORY COMMISSION

.i j

21(

d, /

WASHINGTON, D.C. 20 tie 64101 s

COMMONWEALTH EDISON COMPANY DOCKET NO. STN 50-455-i BYRON STATION. UNIT NO. 2 i

AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 55 l

License No. NPF-66 i

1.

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

A.

The application for amendment by Commonwealth Edison Company (the-licensee) dated August 5, 1992, complies with the standards-and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; r

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 i

by this amendment can be conducted without endangering the health l

and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; 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 1

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.

2.

Accordingly, the license is amended by changes to the Technical Specifi-l cations as indicated in the attachment to this license amendment, and i

paragraph 2.C.(2) of Facility Operating License No. NPF-66 is hereby amended to read as follows:

f f

i a,

I I

i (2)

Technical Specifications The Technical Specifications contained in Appendix A (NUREG-1113),

as revised through Amendment No. 55 and revised by Attachment 2 to NPF-66, and the Environmental Protection Plan contained in Appendix B, both of which were attached to License No. NPF-37, dated February 14, 1985, are hereby incorporated into this license. Attachment 2 contains a revision to Appendix A which is hereby incorporated into this license. The licensee shall operate t

the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION

$1Q'L {.

24'

/

James E. Dyer, Director Project Directorate III-2 Division of Reactor Projects - III/IV/V Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

October 4, 1993 t

--- -~

I l

I ATTACHMENT TO LICENSE AMENDMENT NOS. 55 AND 55 FACILITY OPERATING LICENSE NOS. NPF-37 AND NPF-66 DOCKET NOS. STN 50-454 AND STN 50-455 l

Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages.

The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.

Pages marked with an asterisk are provided for convenience.

Remove Paaes Insert Paaes

+3 2-7

B 2-7 8 2-8 B 2-8 8 2-9 B 2-9 3/4 3-2 3/4 3-2 i

3/4 3-3 3/4 3-3 i

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 3-6a 3/4 3-6a 3/4 3-9 3/4 3-9 3/4 3-10 3/4 3-10 5

3/4 3-11 3/4 3-11 3/4 3-12 3/4 3-12 3/4 3-12a 3/4 3-12a 3/4 3-15 3/4 3-15 3/4 3-17 3/4 3-17 3/4 3-19 3/4 3-19 3/4 3-20 3/4 3-20 3/4 3-21 3/4 3-21 3/4 3-22 3/4 3-22 3/4 3-34 3/4 3-34 3/4 3-35 3/4 3-35 3/4 3 3/4 3-36 3/4 3-37 3/4 3-37 3/4 3-38 3/4 3-38 l

B 3/4 3-1 B 3/4 3-1 f

l k

s

i LIMITING SAFETY SYSTEM SETTINGS I

BASES Reactor Coolant Flow The Low Reactor Coolant Flow trips provide core protection to prevent DNB by mitigating the consequences of a loss of flow resulting from the loss of one or more reactor coolant pumps.

On increasing power above P-7 (a power level of approximately 10% of RATED THERMAL POWER or a turbine impulse chamber pressure at approximately 10%

of full power equivalent), an automatic Reactor trip will occur if the flow in more than one loop drops below 90% of nominal full loop flow.

Above P-8 (a power level of approximately 30% of RATED THERMAL POWER) an autumatic Reactor trip will occur if the flow in any single loop drops below 90% of nominal full loop flow.

Conversely on decreasing power between P-8 and P-7 an automatic Reactor trip will occur on low reactor coolant flow in more than one loop and below P-7 the trip function is automatically blocked.

Steam Generator Water Level The Steam Generator Water Level Low-Low trip protects the reactor from loss of heat sink in the event of a sustained steam /feedwater flow mismatch resulting from loss of normal feedwater.

The specified Setpoint provides allowances for starting delays of the Auxiliary Feedwater System.

Undervoltage and Underfrequency - Reactor Coolant Pump Busses The Undervoltage and Underfrequency Reactor Coolant Pump Bus trips provide core protection against DNB as a result of complete loss of forced coolant flow.

The specified Setpoints assure a Reactor trip signal is generated before the Low Flow Trip Setpoint is reached.

For undervoltage, the delay is set so that the time required for a signal to cause a reactor trip after the Under-voltage Trip Setpoint is reached shall not exceed 1.5 seconds.

For under-frequency, the delay is set so that the time required for a signal to cause a reactor trip after the Underfrequency Trip Setpoint is reached shall not exceed 0.6 second.

On decreasing power the Undervoltage and Underfrequency Reactor Coolant Pump Bus trips are automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, reinstated automatically by P-7.

i i

BYRON - UNITS 1 & 2 B 2-7

LIMITING SAFETY SYSTEM SETTINGS BASES Turbine Trio j

A Turbine trip initiates a Reactor trip. On decreasing power the Turbine trip is automatically blocked by P-8 (a power level of approximately 30% (P-8) of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 30% (P-8) of full power equivalent);' and on increasing power, reinstated automatically by P-8.

Safety In.iection Input from ESF If a Reactor trip has not already been generated by the Reactor Trip System instrumentation, the ESF automatic actuation logic channels will initiate a Reactor trip upon any signal which initiates a Safety Injection.

The ESF instrumentation channels which initiate a Safety Injection signal are shown in Table 3.3-3.

Reactor Coolant Pump Breaker Position Trio The Reactor Coolant Pump Breaker Position trips are anticipatory trips which provide core protection against DNB. The Open/Close Position trips assure a Reactor trip signal is generated before the Low Flow Trip Setpoint is reached.

No credit was taken in the accident analyses for operation of these trips. Their functional capability at the open/close position settings is required to enhance the overall reliability of the Reactor Trip System. Above P-7 (a power level of approximately 10% of RATED THERMAL POWER or a turbine impulse chamber pressure at approximately 10% of full power equivalent) an automatic Reactor trip will occur if more than one reactor coolant pump breaker is opened.

Below P-7 the trip function is automatically blocked.

BYRON - UNITS 1 & 2 B 2-8 Amendment No. 55 i

f

t LIMITING SAFETY SYSTEM SETTINGS BASES i

Reactor Trio System Interlocks The Reactor Trip System Interlocks perform the following functions:

P-6 On increasing power, P-6 allows the manual block of the Source Range Reactor trip (i.e., prevents premature block of Source Range trip),

provides an automatic backup block for Source Range Neutron Flux doubling, and the manual block that de-energizes the high voltage to the Source Range detectors. On decreasing power, Source Range Level trips and Neutron Flux doubling circuits are automatically reacti-vated and high voltage restored.

P-7 On increasing power, P-7 automatically enables Reactor trips on low l

flow in more than one reactor coolant loop, more than one reactor coolant pump breaker open, reactor coolant pump bus undervoltage and underfrequency, pressurizer low pressure and pressurizer high level.

l On decreasing power, the above listed trips are automatically blocked.

P-8 On increasing power, P-8 automatically enables Reactor trips on low flow in one or more reactor coolant loops and Turbine trip.

On de-creasing power, the P-8 automatically blocks the single loop low

. flow trip and Turbine trip.

P-10 On increasing power, P-10 allows the manual block of the Intermediate Range Reactor trip and the Low Setpoint Power Range Reactor trip; and automatically blocks the Source Range Reactor trip and provides an automatic backup function to de-energize the Source Range high voltage power. On decreasing power, the Intermediate Range Reactor trip and the Low Setpoint Power Range Reactor trip are automatically reactivated and Source Range high voltage to the detectors is restored if power decreases below the P-6 setpoint.

Provides input to P-7.

P-13 Provides input to P-7.

1 r

P k

BYRON - UNITS 1 & 2 B 2-9 Amendment No. 55 l

6

gg TABLE 3.3-1

.o EE REACTOR TRIP SYSTEM INSTRUMENTATION i

c; MINIMUM

^;

TOTAL NO.

CHANNELS CHANNELS APPLICABLE CI FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION e-1.

Manual Reactor Trip 2

1 2

1, 2 1

m 2

1 2

3*, 4*,

5*

10 2.

Power Range, Neutron Flux a.

High Setpoint 4

2 3

1, 2 2

b.

Low Setpoint 4

3 1###, 2 2

3.

Power Range, Neutron Flux 4

2 3

1, 2 2

High Positive Rate U

4.

Power Range, Neutron Flux, 4

2 3

1, 2 2

Y High Negative Rate m

5.

Intermediate Range, Neutron Flux 2

1 2

1###, 2 3

6.

Source Range, Neutron Flux a.

Startup 2

1 2

2##

4 b.

Shutdown 2

1 2

3,4,5 5-7.

Overtemperature oT 4

2 3

1, 2 6

8.

Overpower oT 4

2 3

1, 2 6

9.

Pressurizer Pressure-Low

_ gi (Above P-7) 4 2

3 1

6 E

b

n TABLE 3.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL N0.

CHANNELS CHANNELS APPLICABLE FUNCTION UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 10.

Pressurizer Pressure-High 4

2 3

1, 2 6

11.

Pressurizer Water Level-High 3

2 2

1 6

(Above P-7) 12.

Reactor Coolant Flow-Low a.

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

6 operating loop operating loop b.

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

'6 below P-8) operating loops operating loop

13. Steam Generator Water Level-4/stm. gen.

2/stm. gen. in 3/stm. gen.

1, 2 6

l Low-Low any operating each operaing stm. gen.

sim. gen.

14. Undervoltage-Reactor Coolant 4-1/ bus 2

3 1

6 Pumps (Above P-7)

15. Underfrequency-Reactor Coolant 4-1/ bus 2

3 1

6 Pumps (Above P-7)

16. Turbine Trip (Above P-8) l a.

Emergency Trip Header 3/ Train 2/ Train 2/ Train 1

6 Pressure b.

Turbine Throttle Valve 4

4 1

1 6

Closure 17.

Safety Injection Input from ESF 2

1 2

1, 2 9

?

BYRON - UNITS 1 & 2 3/4 3-3 AMENDMENT NO. 55

--r-.

u TABLE 3.3-1 (Continued)

~

REACTOR TRIP SYSTFM JNRT""vP!TATION MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE FUNCTION UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 18.

Reactor Coolant Pump Breaker 1/ breaker 2

1/ breaker per 1

11 Position Trip Above P-7 operating loop 19.

Reactor Trip System Interlocks a.

Intermediate Range Neutron 2

1 2

2!#

8 Flux, P-6 b.

Low Power Reactor Trips Block, P-7 P-10 Input 4

2 3

1 8

or P-13 Input 2

1 2

1 8

c.

Power Range Neutron Flux, 4

2 3

1 8

P-8 d.

Power Range Neutron Flux, 4

2 3

1, 2 8

P-10 e.

Turbine Impulse Chamber 2

1 2

1 8

Pressure, P-13 20.

Reactor Trip Breakers 2

1 2

1, 2 12 2

1 2

3*, 4*,

5*

10

21. Automatic Trip and Interlock 2

1 2

1, 2 9

9' 2

1 2

3*, 4*, 5*

10 22.

Reactor Trip Bypass Breakers 2

1 1

0,*

13 BYRON - UNITS 1 & 2 3/4 3-4 AMENDMENT NO. 55

F TABLE 3.3-1 (Continued)

~

t TABLE NOTATIONS I

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

l

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

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

0Whenever the Reactor Trip Bypass Breakers are racked in and closed for bypassing a Reactor Trip Breaker.

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 STANDBY 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:

I a.

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

Either, THERMAL POWER is restricted to less than or equal 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 with 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.

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)

Setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint; and b.

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

Setpoint but below 10% of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 10% of RATED THERMAL POWER.

I BYRON - UNITS 1 & 2 3/4 3-5 AMENDMENT NO. 55 l

t j

TABLE 3.3-1 (Continued)

ACTIONS STATEMENTS (Continued) j i

ACTION 4 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement suspend all operations involving j

positive reactivity changes.

i ACTION 5 - 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 within the next hour open the reactor trip breakers, suspend all-operations involving positive reactivity changes, and verify valves CV-111B, CV-8428, CV-8439, CV-8441, and CV-8435 are closed and secured in position. With no I

channels OPERABLE verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.l.1 or 3.1.1.2, as applicable, and take the actions stated above within-1 hour and verify compliance at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

j ACTION 6 - With the number of OPERABLE channels one less than the Total l

Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied i

a.

The inoperable channel is placed in the tripped condition t

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

i b.

The Minimum Channels OPERABLE requirement is met; however, the inoperable channel amy 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 i

Specification 4.3.1.1.

r ACTION 7 - Deleted j

ACTION 8 - With less than the Minimum Number of Channels OPERABLE, within I hour 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.

ACTION 9 - With the number of OPERABLE channels one less than the Minimum l

Channels OPERABLE requirement, restore the inoperable channel to j

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 STANDBY l

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 for j

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 10 - 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 open the Reactor trip breakers l

within the next hour.

l ACTION 11 - With the number of OPERABLE channels less than the Total Number of Channels, operation may continue provided the inope able channels are placed in the tripped condition within 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

l 1

i BYRON - UNITS 1 & 2 3/4 3-6 AMENDMENT NO. 55

-r py--

f w

w v.

,i-4----e T-ee=

m-P C

%m 7

P

TABLE 3.3-1 (Continued)

ACTIONS STATEMENTS (Continued)

ACTION 12 - a.

With one of the diverse trip features (Undervoltage or Shunt Trip attachment) inoperable, restore it to OPERABLE status with 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or declare the breaker inoperable and apply the requirements of b below. The breaker shall not be bypassed while one of the diverse trip features is inoperable, except for the time required for performing maintenance and testing to restore the diverse trip feature to OPERABLE status.

b.

With one of the Reactor Trip Breakers otherwise inoperable, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one Reactor Trip Breaker 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 4.3.1.1, provided the other Reactor Trip Breaker is OPERABLE.

ACTION 13 - With the Reactor Trip Bypass Breaker inoperable, restore the Reactor Trip Bypass Breaker to OPERABLE status prior to using the Reactor Trip Bypass Breaker to bypass a Reactor Trip Breaker.

If the Reactor Trip Bypass Breaker is racked in and closed for bypassing a Reactor Trip Breaker and it becomes inoperable, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Restore the Bypass Breaker to OPERABLE status within the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the Bypass Breaker within the following hour.

l I

i i

I L

BYRON - UNITS 1 & 2 3/4 3-6a AMENDMENT NO. 55 s

)

TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE RE0VIREMENTS TRIP ANALOG ACTUATING MODES FOR CHANNEL DEVICE WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST IFST LOGIC TEST IS REOUIRED 1.

Manual Reactor Trip N.A.

N.A.

R(14)

N.A.

1, 2, 3*, 4*,

5*

2.

Power Range, Neutron Flux a.

High Setpoint S

D(2,4),

Q N.A.

N.A.

1, 2 M(3, 4),

Q(4, 6),

R(4, Sa) b.

Low Setpoint S

R(4)

Q N.A.

N.A.

1*",

2 3.

Power Range, Neutron Flux, High N.A.

R(4)

Q N.A.

N.A.

1, 2 Positive Rate 4.

Power Range, Neutron Flux, High N.A.

R(4)

Q N.A.

N.A.

1, 2 Negative Rate 5.

Intermediate Range, Neutron Flux S

R(4, Sa)

Q N.A.

N.A.

1*",

2 6.

Source Range, Neutron Flux S

R(4, 5b)

Q(9)

N.A.

2",3,4,5 7.

Overtemperature AT S

R(13)

Q N.A.

N.A.

1, 2 8.

Overpower al S

R Q

N.A.

1, 2 9.

Pressurizer Pressure-Low S

R Q

N.A.

1 (Above P-7)

10. Pressurizer Pressure-High S

R Q

N.A.

1, 2

11. Pressurizer Water Level-High S

R Q

N.A.

1 (Above P-7)

?

BYRON - UNITS 1 & 2 3/4 3-9 AMENDMENT NO. 55 i

n TABLE 4.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEllLANCE REQUIREMENTS TRIP ANALOG ACTUATING MODES FOR CHANNEL DEVICE WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED

12. Reactor Coolant Flow-Low S

R Q

N.A.

1

13. Steam Generator Water Level-Low-S R

Q N.A.

N.A.

1, 2 I

Low

14. Undervoltage-Reactor Coolant N.A.

R N.A.

Q(10)

N.A.

1 l

Pumps (Above P-7) s

15. Underfrequency-Reactor Coolant N.A.

R N.A.

0(10)

N.A.

1 Pumps (Above P-7)

16. Turbine Trip (Above P-8) l

a. Emergency Trip Header N.A.

R N.A.

S/U(1,10)

N.A.

1 Pressure

b. Turbine Throttle Valve N.A.

R N.A.

S/U(1,10)

N.A.

1 Closure

17. Safety Injection Input form ESF N.A.

N.A.

R N.A.

1, 2

18. Reactor Coolant Pump Breaker N.A.

N.A.

R N.A.

1 Position Trip (Above P-7)

19. Reactor Trip System Interlocks

a. Intermediate Range Neutron N.A.

R(4)

R N.A.

N.A.

2" Flux, P-6

b. Low Power Reactor Trips N.A.

R(4)

R N.A.

N.A.

1 l

Block, P-7

c. Power Range Neutron Flux, P-8 N.A.

R(4)

R N.A.

N.A.

1 BYRON - UNITS 1 & 2 3/4 3-10 AMENDMENT NO. 55

--,e

,r, n

-,e w--

---n

TABLE 4.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP ANALOG ACTUATING MODES FOR CHANNEL DEVICE WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED

19. Reactor Trip System Interlocks (Continued)

d. Low Setpoint Power Range N.A.

R(4)

R N.A.

N.A.

1, 2 l

Neutron Flux, P-10

e. Turbine Impulse Chamber N.A.

R R

N.A.

1 l

Pressure, P-13

20. Reactor Trip Breaker N.A.

N.A.

M(ll)

N.A.

1, 2, 3*, 4*,

5*

21. Automatic Trip and Interlock N.A.

N.A.

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

5*

Logic

22. Reactor Trip Bypass Breakers N.A.

N.A.

(15),R(16)

N.A.

1, 2, 3*, 4*,

5*

BYRON - UNITS 1 & 2 3/4 3-11 AMENDMENT NO. 55

TABLE 4.3-1 (Continued)

TABLE NOTATIONS

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

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

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

(1)

If not performed in previous 31 days.

l (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)

The initial single point comparison of incore to excore AXIAL FLUX DIFFERENCE following a refueling outage shall be performed prior to exceeding 75% of RATED THERMAL POWER. Otherwise the single point comparison of incore to excore AXIAL FLUX DIFFERENCE shall be performed 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.

For the purposes of this surveillance, monthly shall mean at least once per 31 EFPD. The 24 nour completion time provisions of Specification 4.0.3 are not applicable.

(4)

Neutron detectors may be exc.luded from CHANNEL CALIBRATION.

(Sa)

Initial plateau curves shall be measured for each detector. Subsequent plateau curves shall be obtaine, evaluated and compared to the initial curves.

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

(5b) With the high voltage setting varied as recommended by the manufacturer, an initial discriminator bias curve shall be measured for each detector.

Subsequent discriminator bias curves shall be obtained, evaluated and compared to the initial curves.

(6)

Incore - Excore Calibration, above 75% of RATED THERMAL POWER. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

For the purposes of this surveillance, quarterly shall mean at least once per 92 EFPD.

(7)

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

(8)

Not used.

(9)

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

BYRON - UNITS 1 & 2 3/4 3-12 AMENDMENT No. 5.5

T8BLE 4.3-1 (Continued)

TABLE NOTATIONS l

(10) Setpoint verification is not applicable.

(11) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall be performed such that each train is tested at least every 62 days on a STAGGERED TEST BASIS and following maintenance or adjustment of the Reactor Trip Breakers and shall include independent verification of the OPERABILITY of the Undervoltage and Shunt Trip Attachments of the Reactor Trip Breakers.

(12) Not used.

I (13) CHANNEL CALIBRATION shall include the RTD bypass loops flow rate.

(14) Verify that the appropriate signals reach the Undervoltage and Shunt Trip Relays, for both the Reactor Trip and Bypass Breakers from the Manual Trip Switches.

(15) Manual Shunt Trip prior to the Reactor Trip Bypass Breaker being racked in and closed for bypassing a Reactor Trip Breaker.

(16) Automatic Undervoltage trip.

t 1

BYRON - UNITS 1 & 2 3/4 3-12a AMENDMENT NO. 55

TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 1.

Safety Injection (Reactor Trip, Feedwater Isolation, Start Diesel Generators, Containment Cooling Fans, Control Room Isolation, Phase "A"

Isolation, Turbine Trip, Auxiliary Feedwater, Containment Vent Isolation, and Essential Service Water).

s a.

Manual Initiation 2

1 2

1,2,3,4 18 b.

Automatic Actuation Logic 2

1 2

1,2,3,4 14 and Actuation Relays c.

Containment Pressure-High-1 3

2 2

1,2,3 19 l

d.

Pressurizer Pressure-Low 4

2 3

1, 2, 3' 19 (Above F-11) e.

Steam Line Pressure-Low 3/ steam line 2/ steam line 2/ steam line 1, 2, 3" 19 l

(Above P-11) any steam line 2.

Containment Spray a.

Manual Initiation 2 pair 1 pair 2 pair 1, 2, 3, 4 18 b.

Automatic Actuation Logic 2

1 2

1, 2, 3,.4 14 and. Actuation Relays c.

Containment Pressure-High-3 4

2 3

1,2,3 16 7

BYRON -' UNITS 1 & 2 3/4 3-15 AMENDMENT NO. 55

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

....,,.,n

TABLE 3.3-3 (Continued)

~

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE i

FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 4.

Steam Line Isolation a.

Manual Initiation 1)

Individual 1/ steam line 1/ steam line 1/ operating 1,2,3 23 steam line 2)

System 2

1 2

1,2,3 22 b.

Automation Actuation Logic 2

1 2

1, 2, 3 21 and Actuation Relays c.

Containment Pressure-High-2 3

2 2

1, 2, 3 19 l

d.

Steam Line Pressure-Low 3/ steam line 2/ steam line 2/ steam line 1, 2, 3' 19 l

(above P-11) any steam line e.

Steam Line Pressure-Negative 3/ steam line 2/ steam line 2/ steam line 3"

19 l

Rate-High (below P-11) any steam line 5.

Turbine Trip & Feedwater Isolation a.

Automatic Actuation Logic 2

1 2

1, 2 24 and Actuation Relays b.

Steam Generator Water Level-4/stm. gen.

2/stm. gen. in 3/stm. gen. in 1, 2 19 High-High (P-14) any operating each operating sim. gen, sim. gen.

c.

Safety Injection See Item 1. above for all Safety Injection initiating functions and requirements, r

il BYRON - UNITS 1 & 2 3/4 3-17 AMENDMENT NO. 55

-- m

.-%e

-vr-+ - -

-w-2

--* -e-

-'1w e em ver r1=Mw etg='w*-+--

r-es-w 94---

gne+>e

-ww

+-%

m

-,.,w.e-e w

m w

p ir*ww

i TABLE 3.3-3 (Continued)

JGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO.

CHAsinEL5 CHANNELS APPLICABLE FUNCl?0NAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 6.

Auxiliary Feedwater (Continued) 9 Auxiliary Feedwater Pump 1/ Train 1/ Train 1/ Train 1, 2, 3 15a Suction Pressure-Low (Transfer to Essential Service Water) 7.

Automatic Opening of Containment Sump Suction Isolation Valves a.

Automatic Actuation Logic 2

1 2

1,2,3,4 14 and Actuation Relays b.

RWST Level -' Low-Low 4

2 3

1,2,3,4 15 Coincident With Safety Injection See Item 1. above for Safety Injection initiating functions and requirements.

8.

Loss of-Power a.

ESF Bus Undervoltage 2/ Bus 2/ Bus 2/ Bus 1, 2, 3, 4 25a b.

Grid Degraded Voltage 2/ Bus 2/ Bus 2/ Bus 1, 2, 3, 4 25b l

l l

BYRON - UNITS 1 & 2 3/4 3-19' AMENDMENT NO. 55 l

~

.(

i

_ TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO.

CHANNELS CHAi4NELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 9.

Engineered Safety Features Actuation System Interlocks a.

Pressurizer Pressure, P-ll 3

2 2

1,2,3 20 b.

Reactor Trip, P-4 4-2/ Train 2/ Train 2/ Train 1, 2, 3 22 Low-Low T,y, P-12 4

2 3

1,2,3 20 c.

6 BYRON - UNITS 1 & 2 3/4 3-20 AMENDMENT NO. 55 ii

.. ~. -

4 TABLE 3.3-3 (Continued)

TABLE NOTATIONS

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

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

i ACTION STATEMENTS ACTION 14 - 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 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 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ; however, one channel may be bypassed or 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 performance of the next required ANALOG CHANNEL OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within 6 l

hours.

ACTION 15a - With the number of OPERABLE channels one less than the Total Number of Channels, declare the associated pump IN0PERABLE and take the ACTION required by Specification 3.7.1.2.

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

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.

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 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 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:

BYRON - UNITS 1 & 2 3/4 3-21 AMENDMENT NO. 55 i

s

~

i TABLE 3.3-3 (Continued)

ACTION STATEMENTS (Continued) 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 4.3.2.1.

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.

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 STANDBY 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 nueber of OPERABLE channels one less than the Total Number of Chunels, 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 .

i 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 3.7.1.5.

ACTION 24 - 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 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 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 25 - a.

With the number of OPERABLE channels one less than the Minimum Number of Channels, STARTUP and/or POWER OPERATION may proceed 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 . 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 the OPERABLE channel per Specification 4.3.2.1.

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

BYRON - UNITS 1 & 2 3/4 3-22 AMENDMENT N0. 55 s

a

TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP ANALOG ACTUATING MCCES CHANNEL DEVICE ACTUATION MASTER SLAVE FOR WHICH CHANNEt.

CHANNEL OPERATIONAL OPERATIONAL LOGIC RELAY RELAY SURVEILLANCE FtCCTIONAL UNIT CMECK CALIBRATION TEST TEST TEST TEST _

TEST

- IS REQUIRED 1.

Safety injection (Reactor Trip, Feedwater Isolation, Start Diesel Generators, Contalment Cooling Fans, Control Room Isolation, Phase "A" Isolation, Turbine Trip, Auxilibry Feedwater, Contalment Vent Isolation and Essentist i

Service Water) a.

Manual Initiation N.A.

N.A.

R N.A.

M.A.

N.A.

1,2,3,4 b.

Automatic Actuation Logic N.A.

N.A.

M(1)

Q 1,2,3,4 and Actuation Relays c.

Containment Pressure-High-1 S

R Q

N.A.

W.A.

N.A.

1,2,3 l

d.

Pressurizer Pressure-Low S

R Q

N.A.

1,2,3 l

(Above P-11) e.

Steam Line Pressure-Low S

R Q

N.A.

M.A.

N.A.

1,2,3 l

(Above P-11) 2.

Containment Spray a.

Manual Initiation N.A.

N.A.

R N.A.

N.A.

1,2,3,4 b.

Automatic Actuation Logic N.A.

N.A.

M(1)

Q 1,2,3,4 and Actuation Relays c.

Containment Pressure-High 3 S

R Q

N.A.

1,2,3 l

3.

Contalment Isolation s.

Phase "A" Isolation 1)

Manual Initiation N.A.

N.A.

R N.A.

N.A.

1,2,3,4 2)

Automatic Actuation N.A.

N.A.

M(1)

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

?

BYRON - UNITS 1 & 2 3/4 3-34 AMENDMENT NO. 55

6 TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP ANALOG ACTUATING MODES CHANNEL DEVICE ACTUATION MASTER SLAVE FOR WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL LOGIC RELAY RELAY SURVEILLANCE FUNCTIONAL 174tf CHECK CAllBRATION TEST TEST TEST TEST IgSL.

IS REQUIRED 3.a. Phase "A" Isolation (continued) 3)

Safety Injection See item 1. ab3ve for att safety injection Surveltlance Requirements.

b.

Phase "B" Isolation 1)

Manual Initiation M.A.

N.A.

R N.A.

N.A.

1,2,3,4 2)

Automatic Actuation N.A.

N.A.

M(1)

M(1) 0 1,2,3,4 Logic Actuation Relays l

3)

Contat. e t Pressure-S R

Q N.A.

N.A.

1,2,3 Nigh-3 c.

Containment Vent Isolation 1)

Automatic Actuation N.A.

N.A.

M.A.

M(1)

M(1) 0 1,2,3,4 Logic and Actuation Relays 2)

Manuel Phase "A" See Item 3.a.1 above for att manual Phase "A" Isolation Surveittance Requirements.

Isolation 3)

Manual Phase "B" See Item 3.b.1 above for all manual Phase "B" Isolation Surveillance Requirements.

Isolation 4)

Safety Injection See Item 1. above for all Safety Injection Surveillance Reg;Irements.

4.

Steam Line Isolation a.

Manuel Initiation N.A.

N.A.

R N.A.

N.A.

1,2,3 b.

Automatic Actuation Logic N.A.

N.A.

M(1)

M(1) 0 1,2,3 and Actuation Relays c.

Containment Pressure-High-2 S

R 0

N.A.

1,2,3 l

d.

Steam Line Pressure Low S

R Q

N.A.

1,2,3 l

(Above P-11)

?

BYRON - UNITS 1 & 2 3/4 3-35 AMENDMENT NO. 55

TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS i

TRIP ANALOG ACTUATING MODES CHANNEL DEVICE ACTUATION MASTER SLAVE FOR WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL LOGIC RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CAllBRATION TEST TEST TEST TEST Ig ST_,

15 REQUIRED 4.

Steam Line Isolation (continued) e.

Steam Line Pressure -

S R

Q N.A.

N.A.

3 Negative Rate

Migh (Below P-11) 5.

Turbine Trip and feedwater Isolation a.

Automatic Actuation Logic N.A.

N.A.

M(1)

Q 1, 2 and Actuation Relay b.

Steam Generator Water Levet-S R

S N.A.

M(1)

M(1) 0 1, 2 l

High-Migh (P-14) c.

Safety Injection See item 1. above for att Safety Injection Surveillance Requirements.

6.

Auxillary feedwater a.

Manual Initiation N.A.

N.A.

R N.A.

M.A.

N.A 1,2,3 b.

Automatic Actuation Logic N.A.

N.A.

M(1)

Q 1, 2, 3 and Actuation Relay c.

Steam Generator Water Level-S R

0 N.A.

N.A.

1,2,3 l

Low-Low d.

Undervoltage-RCP Bus N.A.

R N.A.

Q(3)

N.A.

1, 2 e.

Safety injection See Item 1. atnve for alt Safety injection Surveillance Requirements.

f.

Division 11 for Unit 1 N.A.

R N.A.

M(2, 3)

N.A.

1,2,3,4 (Division 21 for Unit 2) ESF Bus Undervoltage g.

Auxiliary Feedwater Pwp S

R M

N.A.

1,2,3 Suction Pressure-Low

?

BYRON - UNITS 1 & 2 3/4 3-36 AMENDMENT NO. 55 ee es em 4-w-

u-w-e weew*

e'w-a e

e-mW'N-w-

-v---

a ---

r-m.

m-ev+se mo

O A

TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP ANALOG ACTUATING MODES CHANNEL DEVICE ACTUAfl0N MASTER SLAVE FOR WHICM CHANNEL CHANNEL OPERATIONAL OPERAfl0NAL LOGIC RELAT RELAT SURVE!LLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST TEST TEST TEST _

15 RE0VikED_

F.

Automatic Opening of conteirenent Se p Suction Isolation Valves i

s.

Automatic Actuettor Logic W.A.

N.A.

M(1)

Q 1,2,3,4 and Actuation Retsys l

b.

RWST Levet-Low-Low S

R Q

N.A.

1,2,3,4 Coincident With Safety Injection See item 1. above for ett safety injection Surveittance Requirements G.

Loss of Power a.

ESF Bus Undervoltage N.A.

R W.A f8(2, 3)

N.A.

1,2,3,4 b.

Grid Degraded Voltage N.A.

R N.A.

M(3)

N.A.

1,2,3,4 9.

Engineered Safety Feature Actuation System Interlocks l

e.

Pressuriter Pressure, P-11 N.A.

R Q

N.A.

1,2,3 l

b.

Reactor Trip, P-4 N.A.

N.A.

R N.A.

N.A.

1,2,3 c.

Low-Low T P-12 N.A.

R Q

N.A.

1,2,3 l

l TABLE NOTATION (1)

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

(2) Undervoltage relay operability is to be verified independently. An 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 the OPERABLE channel per Specification 4.3.2.1.

(3)

Setpoint verification is not applicable.

?

BYRON - UNITS 1'& 2 3/4 3-37 AMENDMENT NO. 55

9 INTENTIONALLY LEFT BLANK

?

BYRON - UNITS 1 & 2 3/4 3-38 AMENDMENT NO. 55

3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION The OPERABILITY of the Reactor Trip System and the Engineered Safety Features Actuation System instrumentation and interlocks ensures that: (1) the associated ACTION and/or Reactor trip will be initiated when the parameter monitored by each 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 OPERABILITY 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 safety 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 i

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 Engineered Safety Features Actuation System Instrumentation Trip Setpoints specified in Table 3.3-4 are the nominal values at which the bistables are set for each functional unit. A Setpoint is considered to be adjusted consistent with the nominal value when the "as measured" Setpoint is within the band allowed for calibration accuracy.

To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which Setpoints can be measured and calibrated, Allowable Values for the Setpoints have been specified in Table 3.3-4.

Operation with Setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error.

BYRON - UNITS 1 & 2 B 3/4 3-1 AMENDMENT NO. 55

..

ML20057F197

Text

Attachment:

Attachment:

Navigation menu

Search