ML20138P110

Proposed Tech Spec Changes to Install Automatic Shunt Trip Breaker Mod Per Generic Ltr 85-09

ML20138P110
Person / Time
Site:	Farley
Issue date:	12/16/1985
From:	ALABAMA POWER CO.
To:
Shared Package
ML20138P107	List: ML20138P105 ML20138P110
References
GL-85-09, GL-85-9, TAC-55354, TAC-55355, NUDOCS 8512240301
Download: ML20138P110 (15)

Text

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] REACTOR TRIP = SYSTEM INSTD'JMENTATION-

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MINIMUM

- TGTAL N0. CHANNELS- CHANNELS ~ ^ APPLICABLE-FUNCTIONAL UNIT , OF CHANNELS- TO T.< IP l OPERABLE' MODES- ACTION

1. -

Manual Reactor Trip 2: 1 2 1, 2 12 3*, 4*, S* 13

2. Power. Range, Neutron Flux - '

A'. (Hi gh - 4- 2 -3 '1,:2 ,. _ 2#

B. . Low- 4: 2 3 2 2#

3. Power Range, Heutren Flux, 4- 2 3 1, 2 2#

High Positive Rate

4. Power l Range, Neutron Flux, 4 2 3- 1, 2 - 2#

-High Negativc' Rate . .

. s l [ 5. Intermediate Range, Neutron 2 1 2 '1, 2, and'* 3####

_4 Fl ux -

.. 6. Source Range, Neutron Flux A. Startup 2- 1 2 2##, and

• 4 B. Shutdown 2 -0 1### 3, 4 and 5 S-l 7. Overtemperature AT

Three Loop Operation 3 '2 2 1,'2- 7#

7 Two Loop Operation 3 1** 2 1, 2 9 1

8. Overpower. AT t  : Three Loop Operation . 3 2 2' .1, 2 7#

Two Loop Operation 3 1** 2 1, ' 2 9 ,

3 9. Pressurizer Pressure-Low 3 2 2 1- 7#

! 10. Pressurizer Pressure--High 3- 2 2 1,.2 7#

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TABLE 3.3-1 (Continued);.

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N 4" = REACTOR TRIP-SYSTEM INSTRUMENTATION M' i 2

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.- MINIMUMJ

--TOTAL NO.i , CHANNELS. -CHANNELSj APPLICABLE- *

' @' 'FUNCT10NAL UNIT- ,

0F - CHANNELS ': .TO TRIP -OPERABLE' ' -

MODES ACTION

~

21. ReactorfTrip' Breakers-2.1 ' l' 2/ 1, 2.! 1,f14'

. '3*, 4*','5* . 13,14

22. . ' Autom: tic' Trip Logic 2' 1: .2. 1~, .

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g h -ACTION 11,-LWith less than the Minimum Number of Channels OPERABLE,  ;

operation may continue. provided the inoperable channel is 4

placed:in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

t

- ACTION 12 - With the number of. 0PERABLE channels one less than required ,

,, - by the Minimum Channels. 0PERABLE requirement, restore.the. ,

ainoperable channel to OPERABLE status within .48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />,1-or:be in HOT' STANDBY within the next-6 hours.

$a. "; ACTION '13 - With the. number of 0PERABLE' channels one less than required -

by;the Minimum Channels ~ OPERABLE requirement, restore the'

@@ ' inoperable" channel .to OPERABLE: status-within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, or ty .open the: Reactor Trip System breakers within the next hour. -

~

JACTION 141 IWith one. of 'the diverse. trip Lfeatures (undervoltage .or - shunt-

~

' trip attachment) inoperable, the breaker'may be considered 0PERABLE provided that the diverse trip feature is. restored

" ^ to OPERABLE status withins.48' hours,'or declare the breaker

. inoperable and apply ACTION 1. The breaker shall not b'e bypassed while one of- the diverse-trip features is inoperable except. for the time required ~ for performing

-maintenance.to. restore'the breaker to OPERABLE status.  ;

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CHANNEL MODES'IN WHICH C ,_

.. . CHANNEL-  : CHANNEL: .

,FUNCTIONALD . SURVEILLANCE-g- _ FUNCTIONAL' UNIT- CHECK: CALIBRATION  : TEST -

REQUIRED-

1. Manual Reactor Trip 'N.A. -N.A.- .

R(11), ^

. ' 1 ', - 2 , -

S/U(1)(12) :3*,;4*, S*'

J

2. Power' Range, Neutron Flux ,

. A. Hi gh' iS D(2),.M(3)~ M 1, 2' ,

and Q(6)' -

B. Low- S O(2),'M(3) S/U(10)' 2 and Q(6)-

w. ~3. Power Range, Neutron Flux, High N.A. R(6) .M' 1, ' 2 -

3 Positive' Rate-

4. Power. Range, Neutron Flux, N.A. R(6) M 1, 2 High ;-Negativo Rate

5. Ir.termediate ' Range,-- Neutron S R(6) S/U(1) 1, 2'and-Flux *' /

.6. Source Range,- Neutron Flux ' 'S(7) R(G) M and S/U(1). 2, 3,'4, 5, and *

7. OvertemperaturehT S R -M ~ 1, 2

8. OverpowerkT -S R M 1, 2

9. Pressurizer Pressure-Low S R M 1 y -

10. Pressurizer Pressure--High .S R M 1, 2 h 11. Pressurizer Water Level--High S R M 11

'b 12. A.

Loss of Flow - Single Loop - .S R M 1 B. Loss of Flow - Two Loops S- R :N.A. 1

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,nR EACTOR. TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS ,- ,

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V 1 CHANNEL;.

MODES-IN WHICH'

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CHANNEL.. ..JCHANNEL'. LFUNCTIONAL' SURVEILLANCEc

~~' -CALIBRATION:

FUNCTIONAL': UNIT , CHECK: TEST REQUIRED

'13. : Steam GeneratorcWater Level--- ;S* .R'- .M' 1,.2~ -

Low-Low- 1

~

14. : Steam /Feedwater F1ow Mismatch - S' R: cM. 11,72.

and' Low. Steam Generator Water

Level .+. ,

15. Undervoltage - Reactor Coolant ' N. A. . R. ~' ;M ' ~ 1' Pumps y

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16. Underfrequency . Reactor. - N . A.' R. M M a Coolant-Pumps T 17. Turbine Trip - .

C- .A. Low- Auto Stop 011 Pressure: N. A. - N.A. :S/U(9)(1) :N.A.

B. Turbin'e Throttle- Valve ' N.A. N.A. .S/U(9)(1) .N.A.

Closure .

18. SafetyjInjection Input from ESF N.A.. N.A. M(4) .1', 2-

19. . Reactor Coolant. Pump' Breaker - N.A. N.A. R- 1-Position . Trip

20. Reactor Trip $ystem Interlocks

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N . A. :; .R' S/U(8) 1 2

21. Reactor Trip Breaker N.A. -

N.A. M(5)(14)(15), 1,:2,.

, j2 S/U(1)(14)(15) 3*,f4*, 5*

3 22. Automatic-Trip ~ Logic - N.A.  ? N. A.. -M(5) 1, 2, r- "2 - 3*,;4*, 5*

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'5-

23. Reactor Trip. Bypass-Breaker ~ N.A. N.A.. (13),R(11) :1, 2, l 1 3* , 4* , 5*

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d $)h TABLE NOTATION

'*J- n With the reactor trip system . breakers closed and the' control' rod drive system capable of rod withdrawal.

(1) -> If;not-performed:in previous 7: days'.

~

.(2)L : Heat balance only,;above 15% of RATED THERMAL! POWER. Adjust Q channel iff absolute difference is greater than 2 percent.

f(3) - Compare incore to excore axial flux. difference every'31 EFPb.

-Recalibrate if .the absolute difference 'is greater.than' or equal to 3 pe: cent.

-(4) - ~ Manual ESF functional. input check every 18 months.

.(5).- Each train or~ logic. channel shall' be tested at least every 62 days on a STAGGERED TEST BASIS..

(6) - :Nsutron detectors:may be excluded from CHANNEL CALIBRATION.

'(7)-- Below;the- P-6 -(Block of Source Range Reactor Trip) setpoint.

(8) = Logic only,iif not performed in previous 92 days.

~ (9) CHANNEL FUNCTIONAL TEST will-consist.of.. verifying that1each-

' channel indicates:a turbine trip prior to latching the-turbine and'indicatesino: turbine trip prior to P-9. -

(10)'-- -If not performed in the previous 31 days.

A

.(11) - Independently verify OPERABILITY. of the 'undervoltage and shunt .

trip circuitry for;the' Manual Reactor Trip Function..

2 (12): ~ . Verify reactor trip breaker. and reactor trip; bypass breaker?open upon ' actuation of each Main Control Board handswitch.

i(13)l ELocal! manual.- shunt trip prior to placing breaker in service.

Local . manual undervoitage trip prior to placing breaker in service.

(14) ;Under. voltage trip via Reactor Protection System.

~

l' (15) - ' Local manual. shunt trip..

? -FARLEY:- UNIT 1 3/4 3-14 AMENDMENT N0.

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REACTOR-TRIP-SYSTEM INSTRUMENTATION 3' MINIMUMi-4 :P TOTAL NO.- 'CHANNELSJ CHANNELS

.. APPLICABLE ro FUNCTIONAL-UNIT- . '0F CHANNELS -TO TRIP' OPERABLE. MODES- 1 ACTION'

1. Manual Reactor.. Trip.  :: 2 - 1 .2 1,E 2- .

. 12 2- 3*,-4*, 5* . 13

2. .- Power Range', Neutron - Flux ~ .

'A. Hi gh 4' 2 :3- .1, 2 12# .

B. : Low .

.4 2' ~3 2 -2#

-3. Power. Range, Neutron Flux, .4 2 3 1, 2 - 2#-

High Positive-Rate

4. Po'wer Range, Neutron Flux, 4 .2 3^ 1, 2 2#

High Negative Rate 'l ,

s

'[4 5. Intermediate Range, Neutron Flux 2 -1 -2 1,.2, and'*- ~ 3####-

6. Source Range, Nautron Flux -

A. Startup .

2 1 2 2##, and

• 4 B. Shutdown 2 0 1###- 3,.4 and 5  :- 5:

7. . Overtemperature. AT .

Three Loop Operation 3 2 2 1, 2 .7#

Two Loop Operation 3 1** 2 .1, 2 9-

8. Overpower AT-

' Three Loop Operation 3L 2 2 1, 2 7#'

Two Loop Operation 3 1** 2 1, 2 9 3 9. Pressurizer Pressure-Low 3 2 2 1 7#

10. Pressurizer Pressure--High 3 2 2 :1, 2 7#

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,'REACTORzTRIP-SYSTEM 1 INSTRUMENTATION

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MINIMUM-

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CHANNELSL ' LCHANNELS1 APPLICABLE: _ . ,

FUNCTIONAL UNIT ,

0F CHANNELS,. T -T0 TRIP. .

0PERABLE.,

y " MODES. -ACTIONS _ <

9 21..; Reactor Trip Breakers'- 12- l1! 2 --

1,12' :1,[141 -

• i 3* , 4* , 5* :13,14; - .

22. ' Automatic Trip. Logic a ' .2 1; ' 2- . . .1,12, and;*

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. - TABLE'3'.3-1 (Continued)

-" - ACTION 11.- With less-than the Minimum Number of Channels OPERABLE,

. operation may continue provided the inoperable channel is placed .in:the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

ACT10N :12 - With' the ~ number of 0PERABLE channels one less than required

- . by the Minimum Channels OPERABLE requirement, restore. the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, or be

'in H0T STANDBY'within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ACTION :13.- With the number of OPERABLE channels one less than required byJthe Minimum Channels OPERABLE requirement, restore the

inoperable ~ channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, or-

.open the. Reactor Trip System breakers within the next hour.

- ACTION'14 - With one-of the diverse trip features (undervoltage or shunt trip attachment). inoperable, tt.'e breaker may be considered OPERABLE provided that the diverse trip feature is restored to OPERABLE. status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, or declare the breaker inoperable ~ and apply ACTION 1. -The breaker shal; not be bypassed while-one of the diverse trip features is 11noperable except for the Ltime required for~ performing maintenance- to restore the breaker to OPERABLE status.

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TFARLEY - UNIT 2 3/4 3-9 AMENDMENT NO.

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'-4 CHANNEL- JCHANNEL! FUNCTIONAL- -SURVEILLANCE.- ' '

N. . FUNCTIONAL" UNIT ~ . - CHECK . CALIBRATION ^ TEST:- . ,

REQUIRED' -

1.? Manual [ Reactor.. Trip-

N.A. I N . A .m- R(11), '

-1,.2, . . . .

S/U(1)(12)L 13*,4*,..5*- -

2 2. - ' Power Range, Neutron Flux

. D(2),M(3).

~

"A. Hi gh . JS . M. -le 2__ ^.

h

. and Q(6)-

'B.. Low 'S :D(2),M(3)l S/U(10) * [2' L i

- and.Q(6).

m 3. Power Range,' Neutron Flux, High N.Ac ' R(6)' M 1, 2

) Positive Rate

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r 4, 4. . Power. Range, Neutron : Flux', N . A' .- R(6) M 1 1, .2 2 ; .

-N High Negative Rate W'

4

! 5. Intermediate Range, Neutron- ,

.S. R(6) S/U(1): 1, 2 and '

Fl ux *

6. . Source Range. Neutron Flux . S(7) - R(6). .M and S/U(1)'

2,.3, 4,1 5, and *-

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7. OvertemperaturehT S R M 1; 2 -

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8. Overpower h T S R M- 1, 2 4'

.g g 9. Pressurizer Pressure-Low S R M 'l .

E .

g 10. Pressurizer ~' Pressure--High S R 'M. 1, 2 -

i -4

$ 11. . Pressurizer Water: Level--High . S R M ,.1

12. -A. Loss of Flow - Single: Loop :S, R. M- 1- -'

B. Loss of Flow - Two Loops S~ R N.A. 1:

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. CHANNEL- *'.FUNCTIONAC '

' TSURVEILLANCE' * .'

' FUNCTIONAL UNITD CHECK iCALIBRATION , TESTJ- REQUIRED-

13. " Steam Generator Water. Level- ' S; :M-

R~ ' 1, : 2 ?

Low-Law s- '

14.- Steam /Feedwater Flow Mismatch. .

S' :R? iM- il, 2 and Low Steam Generator Water'- .-

Level' -

15. ;Undervoltage - Reactor Coolant.

'N.A.: .R: L. M . ,

'l

. Pumps

_16. 1Underfrequency -LReactor~ -N.A. .R- M 1-Coolant Pumps _

? 17. . Turbine Trip L T . A. .

Low AL uto Stop 011 Pressure - -N.A. N . A'. - S/U(9)(1)  :

N.A..

C ' B. - Turbine . Throttle Valve - N.A. N. A. : S/U(9)(1)- (N.A.

C1osure. .

.18. -Safety Injection Input.from ESF- 'NiA. 'N.A. M(4) 1, 2 l

19. Reactor Coolant Pump Breaker.

~

N.A. N.A. R 'l Position Trip .

20. Reactor Trip . System. Interlodks '.N. A. R S/U(8) 1

21. Reactor Trip Breaker.-  : N. A. . .N.A. . M(5)(14)(15), 1. 2,

.3 .S/U(1)(14)(15) . 3*,L4*, 5*

-E . .

Ei 22. Automatic Trip Logic -N.A.

'N.A. M(5) 1, 2,

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-

3*, 4*, 5*-

5 ' (13) ,R(11)'

2 23. . Reactor Trip Bypass' Breaker i N. A.. N.A.. 1, 2,1

~F 3* , 4* , 5* -

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$;w TABLE 4~.311 (Continued)

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_ ITABLE' NOTATION l

$gm ,  ; *1- qWith the . reactor trip' system breakers closed and the control rod .

lj):-  : drive system, capable of rod withdrawal'.

p _ l(1) .If7notfperformed in previousL7 days.

W . . .

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(2) - L Heat balance only~,-above 15% of RA.ED THERMAL-POWER. Adjust f channel ~if absolute difference lis greater than 2 percent.

(3)'- Compare-incore to'excore axial-flux difference every.31 EFPD.

'Recalibrate-if1the absolute. difference-is greater than ;or equal

~ to 3 percent.

$ (4) - ' Manual ESF functional input check every 18 months.

m ~

3 (5)l-u Each train.or logic channel shall be tested at least every-62 days.on:a STAGGERED' TEST BASIS.

I6)--.NeutrondetectorsmaybeexcludedfromCHANNELCALIBRATION.

n

(7) ~ - Below the P-6 (Block .of Source' Range Reactor Trip)'setpoint.

m (8);-' ; Logic only, if not performed in previous' 92 days.

-(9) -' CHANNEL 1 FUNCTIONAL TEST- will consist of. verifying that each

channel ~ indicates. a turbine trip prior to latching the turbine

-and indicates nol turbine trip prior to P-9.

.(10) f If.not performed in the-previous 31. days.

(11) ', Independently verify;0PERABILITY of the undervoltage and shunt

, trip circuitry:;for the Manual Reactor Trip Function.

~

4 Verify reactor trip breaker an'd reactor. trip bypass breaker open

~

(12) upo6 actuation of each Main Control Board handswitch.

(13)  : Local manual shunt trip prior to placing breaker in service.

- Local manual undervoltage trip prior to placing breaker in

~

service.: .

(14) - iUndervoltage trip via Reactor Protection' System.

(IS) - Local manual shunt trip.

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' FARLEY - UNIT 2 3/4 3-14 AMENDMENT N0.

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X ATTACHMENT 2 Significant- Hazards l Evaluation ' Pursuant' To 10CFR50.92

.for the Proposed Changes to the Reactor Trip and Bypass Breaker

~

_ Techni. cal Specifications in Response to Generic Letter 85 ,

Proposed Changes:

7 Add' Applicable Modes and Action Statements _ to-Table 3.3-1, Items 1 and

~ 21;' add applicable modes and table notation to Table '4.3-1, Items 1, 21 and .22; and add Item 23 to . Table 4.3-1 for the Reactor Trip Bypass.-

. Breakers.

Background:

Item 4.3 of Generic Letter 83-28, " Required Actions Based on Generic 1 Implications.of' Salem ATWS Events," established the requirement for the

' automatic actuation of'the shunt trip attachment for Westinghouse plants

and requested licensees to submit any needed Technical Specification changes.

' Westinghouse developed generic design modifications'for reactor trip system automatic actuation: using' shunt coil. trip attachments. The major aspects of tha Westinghouse generic design for Type 416 breakers are s  : applicable to the Farley Nuclear Plant breaker modifications.and have been implemented. Since some licensees had responded to Generic Letter 83-28 and had indicated that.no technical specification. changes were required, NRC _ issued Generic Letter 85-09 to request licensees to submit u

proposed technical' specification changes and'provided_ proposed changes to the Westinghouse Standard Technical Specifications (STS) and guidance for licensees to use in developing their Technical Specifications..

- Alabama Power' Company. therefore proposes changes to the reactor trip breakerispecifications and the addition of the reactor trip bypass

. breaker specifications for the Farley Nuclear Plant Unit 1 and 2 Technical Specifications in response to Generic Letter 85-09.

' Analysis:.

Alsbama' Power Company has reviewed the requirements of 10CFR50.92 as-they relate to the proposed ~ reactor trip.and-bypass breaker technical

specification changes and considers-the proposed changes not to involve

a significant hazards consideration. In support of this conclusion, the

~

following analysis is provided:

1)' The proposed change' s will not increase the probability or consequences of an accident previously evaluated because the 4 addition of reactor trip bypass breakers' ' surveillance requirements will provide!added assurance that diverse means of effecting a sreactor' trip exist ano will enhance the capability to shutdown the

reactor upon demand. Therefore, the probability or consequences of an! accident previously enluated have not been increased.

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' Attachment 2-

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' 0) The proposed l changes will not create the possibility of a new or different kind'of: accident previously evaluated because the proposed changes ensure that a common mode failure will not exist and the

-reactor will be;. shutdown when required. Since the ability to

. shutdown the reactor. has been previously analyzed, these proposed

changes do not create-the possibility of a new or different kind of

_ accident from any accident previously evaluated.

-3) The proposed changes will- not involv'e a significant reduction in a margin of- safety because the proposed changes enhance .the ability to

" shutdown the reactor'and therefore enhance.the margin of safety

~

associated with .a reactor shutdown. The proposed changes conform to -

'NRC accepted practice and'therefore doinot significantly reduce a margin of safety.

Conclusion Based upon the analysis provided herewith . Alabama Power Company-jconsiders the~ proposed changes to the technical specifications to not

"_ Linvolve a significant hazards consideration as defined by 10CFR50.92(c).

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