ML20138P110
| ML20138P110 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 12/16/1985 |
| From: | ALABAMA POWER CO. |
| To: | |
| Shared Package | |
| ML20138P107 | List: |
| 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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.- 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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- 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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i 2FARLEY.- UNIT-1? 3/4 3-9 AMENDMENT NO.
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'- . N~ 1.
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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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71 1 .._ _
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*
-A _
'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
- w
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 -
]. .
- 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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x iCHANNELn . . MODES IN'WHICH- - -
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, ~. ' CHANNEL: "
. 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,
- g
- -
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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~
_ 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 . . .
4"~
- (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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