ML20133E680

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Proposed Tech Specs to Eliminate SI Signal on Low Steam Line Pressure.Unnecessary SI Actuation Causes Unnecessary Challenges to Plant Safety Sys,Unnecessary Thermal Transient & Increases Actuation Cycles
ML20133E680
Person / Time
Site: Catawba  Duke Energy icon.png
Issue date: 01/03/1997
From:
DUKE POWER CO.
To:
Shared Package
ML20133E673 List:
References
NUDOCS 9701130107
Download: ML20133E680 (27)
v • d • e


Text

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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, Phase "A" Isolation.

Feedwater Isolation, Control Room Area Ventilation Operation,  ;

Auxiliary Feedwater-Motor-Driven  !

Pump, Purge & Exhaust Isolation, ._

Annulus Ventilation Operation.

Auxiliary Building Filtered Exhaust Operation, Emergency Diesel Generator Operation, Component Cooling Water, Turbine Trip, and Nuclear Service Water Operation)

a. Manual Initiation 2 1 2 1,2,3,4 18
b. Automatic Actuation 2 1 2 1,2,3,4 14 '

1 Logic and Actuation Relays

c. Containment 3 2 2 1,2,3 15 Pressure-High
d. Pressurizer 4 2 3 1, 2, 3# 19 Pressure-Low
e. Steam Line Dyggggyg_ }/gtg37 ligg gfggg37 ]jpg gfgtg37 1jgg }, 3, 3m 15 -

L-ew -in any steam l 4ine- -

9701130107 970103 E' DR ADOCK 05000413 t PDR f CATAWBA - UNIT 1 3/4 3-15 Amendment No. 148

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TABLE 3.3-4 -

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE

1. Safety Injection (Reactor Trip, Phase "A" Isolation, Feedwater Isolation, Control Room Area Ventilation Operation, Auxiliary feedwater-Motor-Driven Pump, Purge & Exhaust Isolation, Annulus Ventilation Operation, Auxiliary Building Filtered ._

Exhaust Operation Emergency Diesel Generator Operation, Component Cooling Water, Turbine Trip, and Nuclear Service Water Operation)

a. Manual Initiation N.A. N.A.
b. Automatic Actuation Logic N.A. N.A.

and Actuation Relays

c. Containment Pressure-High s 1.2 psig s 1.4 psig
d. Pressurizer Pressure-tow 2 1845 psig 2 1839 psig
e. Ste - Line Pressure-Lee 1 775 psig 2 prig
2. Containment Spray l
a. Manual Initiation N.A. N.A. .
b. Automatic Actuation Logic N.A. N.A.

and Actuation Relays  ;

c. Containment Pressure-High-High s 3 psig s 3.2 psig A CATAWBA - UNIT 1 3/4 3-29 Amendment No. 148

TABLE 3.3-5 (Continued) .

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

4. Steam Line Pressure-Low
a. fety injeci. ion (ECC5) s Iz!3I 'zz!*I
1) eactor Trip s
2) Fee ater Isolation s 12
3) Phase " Isolationt2) s 18(3I/28(4I
4) Purge and haust Isolation s6 l
5) Auxiliary Feedw ter(5) s 60 1
6) Nuclear Service Wh er Op ation s 6 5(31/76(4I
7) Turbine Trip N.A. j
5) Component Cooling ater s 65(3I/76(4I i
9) Emergency Dies Generator .eration s 11
10) Control Ro Area Ventilation N.A. l l Operatio

{

1

11) Annu s Ventilation Operation s 23
12) xiliary Building Filtered Extjtust i Isolation .A.

i

23) Containment 5 ump i<ecircuiation N.A.

l

% Steam Line Isolation s 10

5. Containment Pressure-High-High
a. Containment Spray s 45
b. Phase "B" Isolation s 6 5(33/76(4I Nuclear Service Water Operation N.A.

I.

4

c. Steam Line Isolation s 10
d. Containment Air Return and Hydrogen s 600 Skimmer Operation
6. Steam Line Pressure - Negative Rate-High

. Steam Line Isolation s 10 I

i CATAWBA - UNIT 1 3/4 3-40 Amendment No. 148

TABLE 4.3-2 -

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION -

SURVEILLANCE REOUIREMENTS e

' v TRIP ANALOG ACTUATING MODES CHANNEL DEVICE MASTER SLAVE FOR WHICH CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS RE0VIRED

1. Safety Injection (Reactor Trip, Phase "A" Isolation, Feedwater  !

Isolation, Control Room Area  ;

Ventilation Operation, Auxiliary '

Feedwater-Motor-Driven Pump, Purge and Exhaust Isolation, Annulus Ventilation Operation, Auxiliary Building Filtered .

Exhaust Operation, Emergency '

Diesel Generators Operation, '

Component Cooling Water, Turbine Trip, and Nuclear i Service Water Operation) '

a. Nanual Initiation N.A. N.A. N.A. R N.A. N.A N.A. 1, 2, 3, 4
b. Automatic Actua- N.A. N.A. N.A. ., N.A. M(1) M(1) Q 1,2,3,4 tion Logic and Actuation Relays  ;
c. Containment S R Q N.A. N.A. N.A N.A. 1, 2, 3  ;

Pressure-High ,

d. Pressurizer S R Q N.A. N.A. N.A N.A. 1, 2, 3 I Pressure-Low '
c. Ste = Line S R Q N.^. M.^. N.* N.^. 1, 2, 2 n_m...... i m. i CATAWBA - UNIT 1 3/4 3-44 Amendment No. 148

, 5) l f ti --3 t___________ _ _ _ _ _

1

I 4 4 .g

. INSTRUMENTATION BASES REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued)

The Engineered Safety Features Actuation System interlocks perform the following functions:

P-4 Reactor tripped - Actuates Turbine trip, closes main feedwater valves on T,yg below Setpoint, prevents the opening of the main feedwater valves which were closed by a Safety Injection or High Steam Generator Water Level signal, allows safety injection block so that 1 components can be reset or tripped.

Reactor not tripped - prevents manual block of Safety Injection.

I P-11 Defeats the manual block of Safety Injection actuation on low pres-surizer pressure M lew sta= "9e Fa=re and defeats steam line isolation on negative steam line pressure rate. Defeats the manual block of the motor-driven auxiliary feedwater pumps on trip of main feedwater pumps and low-low steam generator water level.

4 P-12 On decreasing reactor coolant loop temperature, P-12 automatically t

blocks steam dump and allows manual bypass of steam dump block for O, the cooldown valves only. On increasing reactor coolant loop temper-U ature, P-12 automatically defeats the manual bypass of the steam dump block. .

P-14 On increasing steam generator level, P-14 automatically trips all feedwater isolation valves, pumps and turbine and inhibits feedwater control valve modulation.

Surveillances for the Reactor Trip Bypass Breakers are included in response to the NRC's Generic Letter 85-09, dated May 23, 1985.

3/4.3.3 MONIl0 RING INSTRUMENTATION 3/4.3.3.1 RADIATION HONITORING FOR PLANT OPERATIONS The OPERABILITY of the radiation monitoring instrumentation for plant operations ensures that: (1) the associated action will be initiated when the radiation level monitored by each channel or combination thereof reaches its Setpoint, (2) the specified coincidence logic is maintained, and (3) suffi-cient redundancy is maintained to pennit a channel to be out-of-service for testing or maintenance. The radiation monitors for. plant operations senses radiation levels in selected plant systems and locations 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 and abnormal conditions. Once the required logic combi-

.^,'

nation is completed, the system sends actuation signals to initiate alarms or automatic isolation action and actuation of Emergency Exhaust or Ventilation

[N../ Systems.

CATAWBA - UNIT 1 B 3/4 3-3

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TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION MINIMUM 3: TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

%et

1. Safety Injection (Reactor Trip, Phase "A" Isolation, Feedwater Isolation, Control Room Area Ventilation Operation, Auxiliary Feedwater-Motor-Driven Pump, Purge & Exhaust Isolation, Annulus Ventilation Operation, Auxiliary Building Filtered n Exhaust Operation, Emergency Diesel Generator Operation,

-@' Component Cooling Water, Turbine Trip, and Nuclear Service Water Operation)

a. Manual Initiation 2 1 2 1,2,3,4 18
b. Automatic Actuation 2 1- 2 1,2,3,4 14 L Logic and Actuation Relays
c. Containment 3 2 2 1,2,3 15 ,

Pressure-High

d. Pressurizer 4 2 3 1, 2, 3# 19 '

Pressure-Low

c. Stear Li c Precture 3/ste = linc 2/ tea line 2/ tes: line 1, 2, 3! 15-

++ -Lew- i 2 y stcar k= Tie hi 9

C. CATAWBA - UNIT 2 3/4 3-15 Amendment No. 142

, ; , ygay .n,-. , .- - - ~ .n,- m7~:- wgn-ce~~; mm 33 '-

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TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 1

1 FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE

1. Safety Injection (Reactor Trip, N' Phase "A" Isolation, Feedwater Isolation, Control Room Area ventilation Operation, Auxiliary Feedwater-Motor-Driven Pump, Purge & Exhaust Isolation, Annulus Ventilation Operation, Auxiliary Building Filtered Exhaust Operation, Emergency Diesel Generator Operation, Component Cooling Water,

?, Turbine Trip, and Nuclear

" ServiceWaterOperation)

- a. Manual Initiation N.A. N.A.

b. Automatic Actuation Logic N.A. N.A.

and Actuation Relays

c. Containment Pressure-High s 1.2 psig s 1.4 psig
d. Pressurizer Pressure-Low 2 1845 psig 2 1839 psig g e. Steer Lir.e oressure-Le:: 2 775 prig - 1 prig
2. Containment Spray
a. Manual Initiation N.A. N.A. -
b. Automatic Actuation Logic N.A. N.A.

and Actuation Relays

c. Containment Pressure-High-High s 3 psig s 3.2 psig CATAWBA - UNIT 2 3/4 3-29 Amendment No. 142
  • g .: ., -_. .._. _ .._ .. . ._ . ~ - _. _ . . . _ . _ . . _ _ _ %, m _ . . _._ c - .- 1.W %

TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES

INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONOS
4. Steam Line Pressure-Low
a. S=fety Injectier (ECCS) .:: 1 2(31/

i 1) actor Trip s2 -

2) Fee ater Isolation s I
3) Phase " Isolation (2) s 1 8(31/28(4)

. \

j 4) Purge and haust Isolation s6

5) Auxiliary Fee ater(5) s 60 I 6) Nuclear Service W er Operati s 6 5(33/76(4I i

l 7) Turbine Trip , N.A.

>l 8) Component Cooling Wa r s 6 5(31/ 7 6(43

9) Emergency Diesel enerator eration s 11 ,

l 10) Control Roo rea Ventilation N.A.

Operation

{ 11) Annul Ventilation Operation s 23 i

! 12) A iliary Building Filtered j xhaust Isolation N.A.

I M

1_, Centaia-ant S"i Dacirc"latier .

@ Steam Line Isolation s 10

5. Containment Pressure-High-High
a. Containment Spray s 45
b. Phase "B" Isolation s 6 5(33/ 7 6(43

" Nuclear Service Water Operation N.A.

< c. Steam Line Isolation s 10

d. Containment Air Return and Hydrogen s 600 Skimmer Operation

<c

6. Steam Line Pressure - Negative Rate-High Steam Line Isolation s 10 CATAWBA - UNIT 2 3/4 3-40 Amendment No. 142

. m w w a x m ws m ::-:.s

-  : . . .~ wwv1*rssts&: ' . , 15 TABLE 4.3-2 ~

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE0VIREMENTS TRIP ANALOG ACTUATING MODES CHANNEL DEVICE MASTER SLAVE FOR WHICH CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE' ,

FUNCTIONAL UNIT CHECK CALIBRATION TEST TES' LOGIC TEST TEST TEST IS REOUIRED 4

t*-

1. Safety Injection (Reactor Trip, (y' Phase "A" Isolation, Feedwater Isolation, Control Room Area  %@j Ventilation Operation, Auxiliary  !.y.

Feedwater-Motor-Driven Pump, frv Purge and Exhaust Isolation, $.#

Annulus Ventilation Operation, y Auxiliary Building Filtered Exhaust Operation, Emergency {b V

{

Diesel Generators Operation, Component Cooling Water, Turbine Trip, and Nuclear Service Water Operation) is

a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1,2,3,4 fa

$$1

b. Automatic Actua- N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4 .

tion Logic and Actuation Relays

c. Containment S R Q N.A. N.A. N.A. N.A. 1, 2, 3 Pressure-High
d. Pressurizer S R Q N.A. N.A. N.A. N.A. 1, 2, 3 Pressure-Low -
c. Ste= Line S P, Q N.A. M.A. M.A. N.A. 1, 2, 3 Pressurc '.cw CATAWBA - UNIT 2 3/4 3-44 Amendment No. 142 g

%- m

e

.t INSTRUMENTATION l>

A i BASES i

W i k REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM

! (I INSTRUMENTATION (Continued) 1 (i The Engineered Safety Features Actuation System interlocks perform the 3.! following functions:

7 4 P-4 Reactor tripped - Actuates Turbine trip, closes main feedwater valves l on Tavo below Setpoint, prevents the opening of the main feedwater

- valves which were closed by a Safety Injection or High Steam 4

Generator Water Level signal, allows safety injection block so that

, components can be reset or tripped.

P Reactor not tripped - prevents manual block of Safety Injection.

P-11 Defeats the manual block of Safety Injection actuation on low pres-surizer pressure and low stor 'he press"re and defeats steam line

isolation on negative steam line pressure rate. Defeats the manual l block of the motor-driven auxiliary feedwater pumps on trip of main '

]

feedwater pumps and low-low steam generator water level.

P-12 On decreasing reactor coolant loop temperature, P-12 automatically blocks steam dump and allows manual bypass of steam dump block for

) the cooldown valves only. On increasing reactor coolant loop temper-ature, P-12 automatically defeats the manual bypass of the steam dump block.

_ P-14 On increasing steam generator level, P-14 automatically trips all

feedwater isolation valves, pumps and turbine and inhibits feedwater j,- control valve modulation.

i b Surveillances for the Reactor Trip Bypass Breakers are included in

, ( response to the NRC's Generic Letter 85-09, dated May 23, 1985.

htV 3/4.3.3 MONITORING INSTRUMENTATION h; 3/4.3.3.1 RADIATION MONITORING FOR PLANT OPE 1ATIONS F The OPERABILITY of the radiation monitoring instrumentation for plant I operations ensures that: (1) the associated action will be initiated when the radiation level monitored by each channel or combination thereof reaches its Setpoint, (2) the specified coincidence logic is maintained, and (3) suffi-cient redundancy is maintained to permit a channel to be out-of-service for testing or maintenance. The radiation monitors for plant operations senses radiation levels in selected plant systems and locations 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 and abnormal conditions. Once the required logic combi-

] nation is completed, the system sends actuation signals to initiate alarms.or automatic isolation action and actuation of Emergency Exhaust or Ventilation Systems.

CATAWBA - UNIT 2 B 3/4 3-3 mummmmesww -, herM a s.m mas u m - m7FMam m umesnes

,i 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, Phase "A" Isolation, Feedwater Isolation, Control Room Area Ventilation Operation,

, Auxiliary Feedwater-Motor-Driven Pump, Purge & Exhaust Isolation, Annulus Ventilation Operation, Auxiliary Building Filtered Exhaust Operation, Emergency Diesel Generator Operation, Component Cooling Water, i Turbine Trip, and Nuclear Service Water Operation)

a. Manual Initiation 2 1 2 1,2,3,4 18
b. Automatic Actuation 2 1 2 1,2,3,4 14 Logic and Actuation Relays
c. Containment 3 2 2 1,2,3 -

15 Pressure-High

d. Pressurizer 4 2 3 1, 2, 3f 19 Pressure-Low l

i 4

CATAWBA - UNIT 1 3/4 3-15 Amendment No.

~

TABLE 3.3-4 .

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE .

1. Safety Injection (Reactor Trip, Phase "A" Isolation, Feedwater Isolation, Control Room Area Ventilation Operation, Auxiliary Feedwater-Motor-Driven Pump, Purge & Exhaust Isolation, Annulus Ventilation Operation, Auxiliary Building Filtered Exhaust Operation, Emergency Diesel Generator Operation, Component Cooling Water, Turbine Trip, and Nuclear Service Water Operation)
a. Manual Initiation N.A, N.A.
b. Automatic Actuation Logic N.A. N.A.

and Actuation Relays

c. Containment Pressure-High s 1.2 psig s 1.4 psig
d. Pressurizer Pressure-Low 2: 1845 psig 2: 1839 psig i
2. Containment Spray
a. Manual Initiation N.A. N.A.
b. Automatic Actuation Logic N.A. N.A.

and Actuation Relays

c. Containment Pressure-High-High s 3 psig s 3.2 psig CATAWBA - UNIT 1 3/4 3-29 Amendment No.

. . s TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATION SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

4. Steam Line Pressure-Low l Steam Line Isolation s 10
5. Containment Pressure-High-High I
a. Containment Spray s 45 l
b. Phase "B" Isolation s 65W/76W Nuclear Service Water Operation N.A.
c. Steam Line Isolation s 10
d. Containment Air Return and Hydrogen s 600 Skimmer Operation
6. Steam Line Pressure - Negative Rate-High i Steam Line Isolation s 10 i i

I CATAWBA - UNIT 1 3/4 3-40 Amendment No.

TABLE 4.3-2 .

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION ,

SURVEILLANCE REQUIREMENTS TRIP .

ANALOG ACTUATING MODES CHANNEL DEVICE MASTER SLAVE FOR WHICH CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED

1. Safety Injection (Reactor Trip, Phase "A" Isolation, Feedwater Isolation, Control Room Area Ventilation Operation, Auxiliary-Feedwater-Motor-Driven Pump, Purge and Exhaust Isolation, Annulus Ventilation Operation, Auxiliary Building Filtered Exhaust Operation, Emergency Diesel Generators Operation, Component Cooling Water, Turbine Trip, and Nuclear Service Water Operation)
a. Manual Initiation N.A. N.A. N.A. R N.A. N.A N.A. 1, 2, 3, 4
b. Automatic Actua- N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4 tion Logic and Actuation Relays
c. Containment S R Q N.A. N.A. N.A N.A. 1,2,3 Pressure-High
d. Pressurizer S R Q N.A. N.A. N.A N.A. 1, 2, 3 Pressure-Low CATAWBA - UNIT 1 3/4 3-44 Amendment No.

INSTRUMENTATION

)

l l

BASES l

REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued) l l

The Engineered Safety Features Actuation System interlocks perform the following functions:

P-4 Reactor tripped - Actuates Turbine trip, closes main feedwater valves on T,yg below Setpoint, prevents the opening uf the main feedwater valves which were closed by a Safety Injection or High Steam Generator Water Level signal, allows safety injection block so that components can be reset or tripped.

Reactor not tripped - prevents manual block of Safety Injection.

1 P-11 Defeats the manual block of Safety Injection actuation on low pres- l surizer pressure and defeats steam line isolation on negative steam l line pressure rate. Defeats the manual block of the motor-driven auxiliary feedwater pumps on trip of main feedwater pumps and low-low steam generator water level.

P-12 On decreasing reactor coolant loop temperature, P-12 automatically l blocks steam dump and allows manual bypass of steam dump block for the cooldown valves only. On increasing reactor coolant loop temper- ,

ature, P-12 automatically defeats the manual bypass of the steam dump I block.

P-14 On increasing steam generator level, P-14 automatically trips all feedwater isolation valves, pumps and turbine and inhibits feedwater control valve modulation.

Surveillances for the Reactor Trip Bypass Brukers are included in response to the NRC's Generic Letter 85-09, dated May 23, 1985.

3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING FOR PLANT OPERATIONS The OPERABILITY of the radiation monitoring instrumentation for plant operations ensures that: (1) the associated action will be initiated when the radiation level monitored by each channel or combination thereof reaches its Setpoint, (2) the specified coincidence logic is maintained, and (3) suffi-cient redundancy is maintained to permit a channel to be out-of-service for testing or maintenance. The radiation monitcrs for plant operations sense radiation levels in selected plant systems and locations and determine whether or not predetermined limits are being exceeded. The radiation monitors send actuation signals to initiate alarms or automatic isolation action and actuation of Emergency Exhaust or Ventilation Systems. Some of the final actuations are dependent on plant condition in addition to the actuation signals from the radiation monitors.

CATAWBA - UNIT 1 B 3/4 3-3

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, Phase "A" Isolation, Feedwater Isolation, Control Room Area Ventilation Operation, Auxiliary Feedwater-Motor-Driven Pump, Purge & Exhaust Isolation, Annulus Ventilation Operation, Auxiliary Building Filtered Exhaust Operation, Emergency Diesel Generator Operation, Component Cooling Water, Turbine Trip, and Nuclear Service Water Operation)
a. Manual Initiation 2 1 2 1,2,3,4 18
b. Automatic Actuation 2 1 2 1,2,3,4 14 Logic and Actuation Relays
c. Containment 3 2 2 1,2,3 15 Pressure-High
d. Pressurizer 4 2 3 1, 2, 37 19 Pressure-Low CATAWBA - UNIT 2 3/4 3-15 Amendment No.

TABLE 3.3-4 .

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE .

1. Safety Injection (Reactor Trip, Phase "A" Isolation, Feedwater Isolation, Control Room Area Ventilation Operation, Auxiliary Feedwater-Motor-Driven Pump, Purge & Exhaust Isolation, Annulus Ventilation Operation, i

Auxiliary Building Filtered Exhaust Operation, Emergency Diesel Generator Operation, Component Cooling Water, Turbine Trip, and Nuclear Service Water Operation) -

a. Manual Initiation N.A. N.A.
b. Automatic Actuation Logic N.A. N.A.

and Actuation Relays

c. Contair, ment Pressure-High s 1.2 psig s 1.4 psig
d. Pressurizer Pressure-Low 2: 1845 psig a 1839 psig l
2. Containment Spray
a. Manual Initiation N.A. N.A.
b. Automatic Actuation Logic N.A. N.A.

and Actuation Relays

c. Containment Pressure-High-High s 3 psig s 3.2 psig i

CATAWBA - UNIT 2 3/4 3-29 Amendment No.

. e TABLE 3.3-5 (Continued)

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

4. Steam Line Pressure-Low Steam Line Isolation s 10
5. Containment Pressure-High-High
a. Containment Spray s 45
b. Phase "B" Isolation s 65 l3)/76DU Nuclear Service Water Operation N.A.
c. Steam Line Isolation s 10
d. Containment Air Return and Hydrogen s 600 Skimmer Operation
6. Steam Line Pressure - Negative Rate-High Steam Line Isolation s 10 l

1 i

CATAWBA - UNIT 2 3/4 3-40 Amendment No. ,

1

~

TABLE 4.3-2 .

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION ,

SURVEILLANCE REQUIREMENTS-TRIP .

ANALOG ACTUATING MODES CHANNEL DEVICE MASTER SLAVE FOR WHICH .

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

1. Safety Injection (Reactor Trip, Phase "A" Isolation, Feedwater Isolation, Control Room Area Ventilation Operation, Auxiliary Feedwater-Motor-Driven Pump, Purge and Exhaust Isolation, Annulus Ventilation Operation, Auxiliary Building Filtered Exhaust Operation, Emergency Diesel Generators Operation, Component Cooling Water, Turbine Trip, and Nuclear Service Water Operation)
a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3, 4
b. Automatic Actua- N.A. N.A. N.A. N.A. M(1) .M(1) Q 1,2,3,4 tion Logic and Actuation Relays
c. Containment S R Q N.A. N.A. N.A. N.A. 1, 2, 3 Pressure-High
d. Pressurizer S R Q N.A. N.A. N.A. N.A. 1, 2, 3 Pressure-Low I

CATAWBA - UNIT 2 3/4 3-44 Amendment No.

I.

INSTRUMENTATION-BASES REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued)

The Engineered Safety Features Actuation System interlocks perform the following functions:

P-4 Reactor tripped - Actuates Turbine trip, closes main feedwater valves on T,yg below Setpoint, prevents the opening of the main feedwater valves which were closed by a Safety Injection or High Steam Generator Water Level signal, allows safety injection block so that components can be reset or tripped.

Reactor not tripped - prevents manual block of Safety Injection.

P-11 Defeats the manual block of_ Safety Injection actuation on low pres-surizer pressure and defeats steam line isolation on negative steam l, line pressure rate. Defeats the manual block of the motor-driven auxiliary feedwater pumps on trip of main feedwater pumps and low-low steam generator water level .

P-12 On decreasing reactor coolant loop temperature, P-12 automatically blocks steam dump and allows manual bypass of steam dump block for the cooldown valves only. On increasing reactor coolant loop temper-ature, P-12 automatically defeats the manual bypass of the steam dump block.

P-14 On increasing steam generator level, P-14 automatically trips all feedwater isolation valves, pumps and turbine and inhibits feedwater control valve modulation.

Surveillances for the Reactor Trip Bypass Breakers are included in response to the NRC's Generic Letter 85-09, dated May 23, 1985.

3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING FOR PLANT OPERATIONS The OPERABILITY of the radiation monitoring instrumentation for plant operations ensures that: (1) the associated action will be initiated when the radiation level monitored by each channel or combination thereof reaches its Setpoint, (2) the specified coincidence logic is maintained, and (3) suffi-cient redundancy is maintained to permit a channel to be out-of-service for testing or maintenance. The radiation monitors for plant operations sense radiation levels in selected plant systems and locations and determine whether or not predetermined limits are being exceeded. The radiation monitors send actuation signals to initiate alarus or automatic isolation action and I actuation of Emergency Exhaust or Ventilation Systems. Some of the final actuations are dependent on plant condition in addition to the actuation signals from the radiation monitors.

CATAWBA - UNIT 2 B 3/4 3-3

Attachment II Technical Justification l

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l Technical Justification for Removal of SI on Low Steam Line Pressure The functional design and safety system performance basis  !

for the Safety Injection (SI) function comes from the safety analyses for the plant. The Safety Injection System is designed to provide borated makeup water during loss-of-coolant accidents as well as cooldown accidents such as steam line breaks. Introduction of cold SI water into the primary coolant system could result in a thermal transient and pressurization of the reactor coolant system (RCS),

introducing unnecessary stresses to the plant. It is also

desirable to prevent unnecessary actuation of the SI System  !

because the injection of highly borated water into the core j j

creates unnecessary challenges to plant safety equipment. l In addition, repeated SI System actuations can lead to an 4

increased number of design transients which decrease the available operating life and cause unneeded actuation and wear of various safety system components.

2 In order to minimize unnecessary actuations of the SI System, the actuation of SI on low steam line pressure will be eliminated. (The low steam line pressure signal will

, continue to initiate main steam isolation valve closure.)

i The results of all licensing basis transients (evaluated using NRC-approved methodology) remain within their respective acceptance criteria with the elimination of SI on low steam line pressure. A discussion of the effects of 1 removing SI on low steam line pressure is provided for each

licensing basis transient which involves a significant decrease in steam line pressure. The remaining transients i listed in Table 1 are unaffected by the removal of SI actuation on low steam line pressure. For most transients, SI actuation does not occur due to the lack of significant primary or secondary side depressurization. For transients j where automatic SI actuation occurs, the initiating signal is either low pressurizer pressure or high containment i pressure.

The steam line break transient (UFSAR 15.1.5) has been analyzed to demonstrate short-term core cooling capability.

A spectrum of break sizes was examined to determine the limiting break size. For smaller breaks (including the limiting break size), SI actuation on low pressurizer pressure occurs prior to reaching the setpoint for SI actuation on low steam line pressure. However, for larger breaks, the setpoint for SI actuation on low steam line pressure is reached first. Thus, the larger breaks have been reanalyzed with the SI actuation on low steam line pressure removed. With the removal of SI on low steam line

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l pressure, SI actuation is slightly delayed for these break sizes until the setpoint is reached for SI on low pressurizer pressure. For these larger break sizes, the minimum DNBRs remain above those calculated for the limiting break size. Thus, the acceptance criteria for the steam line break transient continue to be met with the removal of SI actuation on low steam line pressure. For the scenarios considered in these analyses, there is no requirement (or need) to provide diverse actuation signals (e.g., low steam line pressure as well as low pressurizer pressure).

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The steam line break mass and energy release analysis (UFSAR 6.2.1.4) was evaluated to demonstrate that the conditions inside containment during a steam line break do not violate the existing environmental qualification envelope.

Regardless of break size, SI actuation occurs on high containment pressure prior to reaching the setpoint for SI actuation on low steam line pressure. For steam line breaks outside containment (e.g., in the doghouse), SI actuation occurs on low pressurizer pressure prior to reaching the setpoint for SI actuation on low steam line pressure. ,

Therefore, the steam line break mass and energy releases are i not adversely impacted by the elimination of SI actuation on low steam line pressure.

The loss of AC power transient (UFSAR 15.2. 6) was analyzed to demonstrate long-term core cooling capability. For the worst-case loss of AC power transient analyzed in the UFSAR, there is no primary or secondary system depressurization, and thus SI actuation does not occur. However, under less limiting conditions during a loss of AC power, it is possible for primary and secondary side depressurization to occur. This is due to excessive auxiliary feedwater being delivered to the steam generators as well as extraction steam loads and potentially open steam line drains. If such overcooling does occur, SI actuation on low pressurizer pressure is still available, and thus SI actuaticn on low steam line pressure is not necessary. If prompt operator action prevents overcooling, no SI actuation is required.

The feedwater line break transient (UFSAR 15.2.8) was also analyzed to demonstrate long-term core cooling capability.

During a feedwater line break, the secondary system will depressurize only if the break occurs between the check valve and the steam generator. Breaks are required to be postulated only at the terminal ends of the feedwater piping (i.e., at the steam generator or at the main feedwater pumps). For a feedwater line break at the main feedwater pumps, the check valve will prevent depressurization of the steam generator. For a feedwater line break at the steam

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generator, SI actuation occurs on high containment pressure. '

Therefore, the elimination of the SI actuation on low steam line pressure does not adversely impact the feedwater line i break transient.

l In summary, only those transients which involve a secondary i system depressurization have the potential to be affected by the removal of the SI actuation on low steam line pressure.

For all other transients in which SI actuation occurs, the .

initiating signel is low pressurizer pressure or high  !

containment pressure. The steam line break core response i was reanalyzed with the removal of the SI actuation on low steam line pressure, and all acceptance criteria continue to be met. For all licensing basis transients, all acceptance criteria are met with the elimination of the SI actuation on low steam line pressure.

UFSAR Revisions Required The low steam line pressure SI signal is discussed in the i following sections in the Catawba UFSAR: I 3.1, 6.2.4.1, 6.3.3, 7.1.2.1.6, 7.3.1.2.6, 7.3.2.4.2, Tables 7-5, -6, and -7, 15.1.4.1, 15.1.5.1, 15.1.5.2, 15.2.8.1.

These sections will be revised during the first applicable revision following the modification for the affected unit (s).

4 TABLE 1

SUMMARY

OF SI ACTUATION FOR ALL LICENSING BASIS TRANSIENTS

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UFSAR SI Actuating Signal Transient Section Actuation (ifapplicable) l Increase in Feedwater Flow 15.1.2 No i Increase in Steam Flow 15.1.3 No j Steam Line Break 15.1.5 Yes Low pressurizer pressure

Turbine Trip 15.2.3 No i Loss of AC Power 15.2.6 No Feedwater Line Break 15.2.8 Yes High containment pressure Partial Loss of Forced Reactor Coolant 15.3.1 No j Flow Complete Loss of Forced Reactor 15.3.2 No
Coolant Flow 1 Locked Rotor 15.3.3 No l Uncontrolled Bank Withdrawal from 15.4.1 No j Subcritical i Uncontrolled Bank Withdrawal at Power 15.4.2 No Dropped Rod 15.4.3 a No Single Uncentrolled Rod Withdrawal 15.4.3d No j Startup of an Inactive Reactor Coolant 15.4.4 No l Pump at an Incorrect Temperature

! Boron Dilution 15.4.6 Yes Boration provided by BDMS or I

manual operator action l Inadvertent Loading and Operation of a 15.4.7 No j Fuel Assembly in an Improper Position j Rod Ejection 15.4.8 No I

Inadvertent ECCS Operation 15.5.1 Yes Initiating event is inadvertent SI

actuation

! Inadvertent Opening of a Pressurizer 15.6.1 Yo Low pressurizer pressure l Safety or Relief Valve j Instrument Line Break 15.6.2 No l Steam Generator Tube Rupture 15.6.3 Yes None for overfill and short-term core cooling evaluations; Manual operator action for dose evaluation

! Loss of Coolant Accident - Peak Clad 15.6.5 Yes Low pressurizer pressure or high l Temperature containment pressure Loss of Coolant Accident - Mass and 6.2.1.3 Yes Low pressurizer pressure or high Energy Release containment pressure i Steam Line Break Inside Containment - 6.2.1.4 Yes High containment pressure 1 Mass and Energy Release

! Steam Line Break Outside Containment - 3.11 Yes Low pressurizer pressure j Mass and Energy Release

i Attachment III No Significant Hazards Analysis The following analysis is presented, pursuant to 10 CFR 50.91, to  !

demonstrate that the proposed change will not create a Significant Hazard Consideration.

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1. The proposed change will not ir.tolve a significant increase in  !

the probability or consequences of an accident previously evaluated.

The proposed change, to delete the SI signal on low steam line pressure, will only prevent an unnecessary SI actuation as an event occurs which involves secondary system depressurization.

No consequences will significantly increase, because for each l

event previously analyzed it has been shown that either SI on low steam pressure is not demanded, or that another SI signal (e.g.,

low pressurizer pressure) is generated in sufficient time to meet i applicable acceptance criteria. The probability of an accident will not increase.
2. The proposed change will not create the possibility of any new i accident not previously evaluated.

The initiation of SI on a low steam line pressure signal may occur during events which involve a depressurization of the secondary side, including excessive auxiliary feedwater addition. 3 There are other SI initiation signals which will accomplish this l same function if needed. Removing this actuation signal will not )

create any new failure modes or necessitate any new hardware j configurations (other than the deletion of the signal itself) . I No new accident scenarios are created.

3. There is no significant reduction in a margin of safety.

Analysis has shown that for any transient for which SI would have occurred on low steam line pressure, transient response is maintained within acceptable limits. Steam line break mass and energy releases inside containment do not violate the existing environmental qualification envelope. Steam line breaks outside

! containment are not adversely affected by this change.

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4 l Based on the foregoing analysis, it is concluded that the 1 proposed amendment will not create a significant hazards i consideration.

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4 lt Statement of Environmental Impact >

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This proposed deletion of the SI signal on low steam pressure 1

. will not create any new effluents, will not significantly ,

! increase any previously identified effluents, and will not result j in an increase in either personal or cumulative radiation 4

exposure. Therefore, this change is considered to have a-  ;

l negligible effect on the environment. ,

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