ML20154J541
| ML20154J541 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 05/18/1988 |
| From: | HOUSTON LIGHTING & POWER CO. |
| To: | |
| Shared Package | |
| ML20154J539 | List: |
| References | |
| NUDOCS 8805270037 | |
| Download: ML20154J541 (58) | |
Text
__ ______ _ .
1 ATTACHMENT 2 ST HL AE aSao PAGE I OF IB ATTACMIENT 2 PROPOSED REVISIONS TECHNICAL SPECIFICATIONS 8805270037 DR 880518 ADOCK 05000498 DCD
TABLE 3.3-3 Ys Si ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION x
M s"' MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION E
il 1. Safety Injection (Reactor
,. Trip, Feedwater Isolation, Control Room Emergency Ventilation, Start Standby Diesel Generators, Reactor Containment Fan Coolers, and Essential Cooling Water). ,
o a. Manual Initiation 2 1 2 1,2,3,4 19 s
- b. Automatic Actuation i' Logic 2 1 2 1,2,3,4 14 5
- c. Actuation Relays 3 2 3 1,2,3,4 14
- d. Containment 3 2 2 1,2,3,4 15 Pressure--High-1
- e. Pressurizer 4 2 3 1, 2, 3# 20 Pressure--Low g$3
-4
- f. Compensated Steam 3/ steam lin~e 2/ steam line 2/ steam line 1, 2, 3# 15 *C C)I'$i Line Pressure-Low any steam line in each steam P j jine c;h g;hi
-. c .m
- g. C;; pen ;ted TCOLD ' 00E 2I'"E i" 2 " I
- 2' #
""E I EP l Leu-L:= (interlecked :ny loop ::ch 100;
, with P-15) i
TABLE 3.3-3 (Continued) 8 5
1 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION M MINIMUM s
TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION E 4. Steam Line Isolation Z
~ a. Manual Initiation
- 1) Individual 2/ steam line 1/ steam line 2/ operating 1, 2, 3 24 steam line
- 2) System 2 1 2 1, 5, 3 23
- b. Automatic Actuation. 2 1 2 1,2,3 22
, y Logic and Actuation
- Relays
-Y m c. Steam Line Pressure -
Negative Rate--High 3/ steam 11ne 2/ steam line 2/ steam line 3### 15 any steam in each steam line line
- d. Containment Pressure - 3 2 2 1,2,3 15 o y3 >
- s High-2 1
- e. Compensated Steam Line 3/ steam line 2/ steam line 2/ steam line 1, 2, 3# 15 8*N Pressure - Low any steam in each steam WRI line line o)%
? $
%9
- f. C1_;n..;ud T COLD '* I' " ""y 2,9 % lu 10.J. , 2, X 15 *4 Lgl =hrhchzi 1m ;;;h i m I
1
~
3
~
- J vs TABLE 3.3-3 (Continued) o c-
, ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 1
4 4
g MINIMUM g TOTAL NO. CHANNELS CHANNELS APPLICABLE
, FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABtE MODES ACTION . ,
g S. Turbine Trip and Feedwater Isolation
~
-i a. Automatic Actuation 2 1 2 1,2,3 25
- Logic and Actuation Relays
- b. Steam Generator 4/stm. gen. 2/stm. gen. 3/stm. gen. 1,2,3 20 Water Level-- in any oper- in each
. High-High (P-14) ating stm. operating gen. sim. gen.
. betekm g c. Cc.pcasated Tg y Lew-finterlecked 3/locp 2Accp in 2/lcep in 1;;;;, 2, 3; 15 g with l'-15) any lecp cach loop
%\eAcet <
- d. Feeirater Fle" - High 3/st=. gen. 2-/+t=. gca. 2/sta. gca. Iffff, 2, 3 15 6 inter!Mked with P-15) in--any-st = . la coch stm.
soixident with either gen. gen.
-of the felle'iq in of 4 loops:
y,
> -i -4 Resetor Cocloat flew Oxy
' "' c o 4.ee 3Acep 2/ loop-in any ~2/lcep in
!cep each lecp 1, 2, 3 15 3[
Qug ee_-
j
- E*
eb T - L0u 1/lcop 1/lcep in eny 1/ loop 1, ^ , 3 15
-leeP
- e. Safety Injection See Item 1. for all Safety Injection initiating functions and requirements.
avg -L w coincident with
- f. T
- Reactor Trip (P-4)** 4 (1/ loop) 2 3 1,2,3 . 20
TABLE 3.3-3 (Continued) 8 y ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION M
g MINIMUM
- TOTAL NO. CHAN5tE8S CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION '
E q 8. Loss of Power -
~
Degraded Voltage o
Coincident with SI 4/ bus 2/ bus 3/ bus 1, 2, 3, 4 20
- c. 4.16 kV ESF Bus Under-y voltage - Sustained y Degraded Voltage 4/ bus 2/ bus 3/ bus 1, 2, 3, 4 20
- 9. Engineered Safety Features Actuation System Interlocks
, a. Pressurizer Pressure, 3 2 2 1,2,3 21 P-11 _
- b. Low-Low Tavg, P-12 4 2 3 1,2,3 21 2$D
- c. Reactor Trip, P-4 2 1 2 1,2,3 23 qv9 Or
- d. Prcr Sn;;; "strer. 4 2 3 1, 2, 3 21 - OEb Fl= Igut te E
- :::: !ve C = 1drr rietecticr., P-15 g(
) i- p e
I
ATTACHMENT A TA8LE 3.3-3 (Coatinued) ST HL AE. 7626 f PAGE 6 OF /8 TABLE NOTATIONS
- Feedwater Isolation only.
- Function is actuated by either actuation train A or actuatiun train B.
Actuation train C is not used for this function. .
- Trip function may be blocked in this MODE below the P-11 (Pressurizer Pressure Interlock) Setpoint.
- 0uring CORE ALTERATIONS or movement of irradiated fuel within containment.
- Trip function autcaatically blocked above P-11 and may be blocked below
- ' " t- #
c p. i4((k [h e."NobiIb "
- Trip function i: bicck:d in "00E 1 50v: the P-15 (Exce::ive Cocido.en
-Protection) cetpoint.
ACTION STATEMENTS ACTION 14 - With the number of OPERABLE channels one less than the Minimum Channels OPECABLE *equirement, be in at least HOT STANOBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />; however, one channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 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 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
ACTION 16 - (Not Used)
ACTION 17 - 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 set. One additional channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing per Specification 4.3.2.1.
ACTION 18 - With less than the Miniaue Channels OPERABLE requirement, operation may continue provided the containment purge supply and exhaust valves are maintained closed.
SOUTH TEXAS - UNIT 1 3/4 3-26
m TABLE 3.3-4 o
5x: ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS N TOTAL SENSOR ERROR TRIP SETPOINT ALLOWAL..
$ FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) c 1. Safety Injection (Reactor Trip, 3 Feedwater Isolation, Control Room Emergency
- Ventilation, Start Standby Diesel Generators, Reactor Containment Fan Coolers, and Essential Cooling Water)-
- a. Manual Initiation N.A. N.A. N.A. N.A. N.A.
- b. Autonatic Actuation Logic N.A. N.A. N.A. N.A. N.A.
R c. Actuation Relays N.A. N.A. N. A. N.A. N.A.
Y d. Containment Pressure--High 1 3.6 0.71 2.0 1 3.0 psig i 4.0 psig
($
- e. Pressurizer Pressure--Low 13.1 10.71 2.0 1 1850 psigff 1 1842 psigd
- f. Compensated Steam Line 13.6 10.71 2.0 1 735 psig 1714.7psid Pressure-Low e C m r.;
- ted TCOLO " 0- I0 1 2 2. '
- (fr.terhetid with F IS)-
- a. Manual Initiation N.A. .N.A. N.A. N.A. N.A.
! b. Automatic Actuation Logic N.A. N.A. N.A.
- N.A. N.A.
- c. Actuation Relays N.A. N.A. N.A. N.A. N.A.
l l' d. Containment Pressure--High-3 3.6 0.71 2.0 5 9.5 psig < 10.5 osig j gr 30 t 3st"d l
- 7t7 r -3Y1H-IS '
. g IN3WH3V11Y
(. . ._/
y.
TABLE 3.3-4 (Continued) 8 M ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMFf;iATION TRIP SETPOINTS h TOTAL SENSOR ERROR
$; FUNCTIONAL UNIT ALLOWANCE (TA) Z, (S) TRIP SETPOINT ALLOWABLE VA h
z
- 4. Steam Line Isolation i Z a. Manual Initiation N.A. N.A. N.A. N.A. N.A.
w
- b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.
and Actuation Relays I.
- c. Steam Line Pressure - 2.6 0.5 0 $ 100 psi i 126.3 psi *'
Negative Rate--High
- d. Containment Pressure - 3.6 0.71 2.0 5 3.0 psig 5 4.0 psig u High-2
- m .
w e. Compensated Steam Line 13.6 10.71 2.0 1 735 psig 1714.7 'psig' a
Pressure - Low .
e
- f. Compensated T COLD 4.5 0.5 1.0 1 532*F 1 528*F***
Low-Low (interlocked with P-15)
- 5. Turbine Trip and Feedwater M Isolation b a. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.
and Actuation Relays
- b. Steam Generator Water -4.5 2.35 2.0+0.2# < 87.5% o f < 88.9% of Level--High-High (P-14) -
iiarrow range liarrow range instrument instrument span. span.
httbi i c. Cc ;;r. sated Tg -Law 4.5 0.5 1.0 > 530"T : "t"T***
I (iiitei-locked wRh ."-15) -
[ 3)..jh;
. 't IN3*rIHOYLi'! .
- *w w =
vs TABLE 3.3-4 (Continued)
C
$ ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS TOTAL SENSOR ERROR ALLOWANCE (TA) (5) TRIP SETPOINT ALLOWABLE VAI FUNCTIONAL UNIT Z_
c 5. Turbine Trip and Feedwater ,
5 Isolation (Continued) .
bel&ct 2.70 4.0
- d. -Feedwater Flor!!igh 7.2 1 30.0% Flow 1 32.2%-H ow 4 inter 4ecked-wi4h-P45}-
-Coincident-With4ither-of-
-the-FoHowing in 2 of '
4:eepst-RES-A cw-Lcw 4.0 3.10 0-0 101.8% of 100.5 cf-loop design loop design flow **** flow ****
-oc-T -Low 4.5 1.30 0.0 1 574*r 1 571.1*F-
- e. Safety Injection See Item 1 above for all Safety Injection Trip Setpoints and Allowable Values.
- f. T -L w C incident with 4.5 1.36 0.8 1 574*F 1 571.1*F ,
avg f, Reactor Trip (P-4)
VJ (Feedwater Isolation Only)
D 6. Auxiliary Feedwater
- a. Manual Initiation N.A. N.A. N.A. N.A. N.A.
D b. Automatic Actuation Logic H.A. N.A. .4. A. N.A. N.A.
- c. Actuation Relays N.A. N.A. N.A. . H.A. H.A.
D d. Steam Generator Water 15.0 12.75 2.0+0.2# > 33.0% of > 31.5% of Level--Low-Low harrow range Harrow ranga
(/f instrument instrument I span. span.
- e. Safety Injection See Item 1. above for all Safety Injection ir'P 8: 30 b 30W !
Setpoints and Allowable Values. 9ew -3Y iH-1S I T IN3nNH3V11V
I
, TABLE 3.3-4 (Continued) l vs j g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS z
y TOTAL SENSOR ERROR i Q FUNCTIONAL UNIT ALLOWANCE (TA) Z (5) TRIP SETPOINT ALLOWARLE VA R
, 9. Engineered Safety Features c- Actuation System Interlocks 5
- a. Pressurizer Pressure, P-11 N.A. N.A. N.A. $ 1985 psig $ 1993 psig
]
- b. Low-Low T , P-12 N.A. N.A. N.A. > 563*F > 560.1*F
- c. Reactor Trip, P-4 N.A. N.A. N.A. N.A. M.A.
- d. P = r E ng: ":;;tra
. M.A. M.A. M.A. < 17. Sted < 12.3% h ie f!= !;rt *o -T t.c m 1 P r. : r - Tr.:=1 ru; i E cessive C M M _
, 7. et.ctic,r., ?-15 i S
,, 10. Control Room Ventilation
- % a. Manual Initiation N.A. M.A. N.A. N.A. N.A. *
- b. Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values.
- c. Autcoatic Actuation Logic M.A. N.A. N.A. M.A. N.A.
and Actuation Relays
- d. ControiRoomIntakeAir 3.7x10 5 2.2x10 5 1.6x10.s <6.1x10 5 <7.8x10 5 Radioactivity - High pCi/cc pCi/cc pCi/cc pC1/cc pC1/cc
- e. Loss of Power See Item 8. above for all Loss of Power Trip Setpoints and Allowable Values. .
- 11. FHB HVAC
- a. Manual Initiation N.A. N.A. N.A. ' IN.A. 91 30k_ a' 30Vd 1 er -3V-lH-1S r IN3WH3V11V
P ,
i
! TABLE 3.3-4 (Continued) ATTACHMENT A ST.HL.AE 4.26 TABLE NOTATIONS PAGE H OF # 8
- Time constants utilized in the lead-lag controller for Steam Line Pressure-Low are i t > 50 seconds and t < 5 seconds. CHANNEL CALIBRATION shall !
ensure Ihat these time constants are adjusted to these values.
- The time constant utilized in the rate-lag controller for Steam Line Pressurg-l Negative Rate-High is greater than or equal to 50 seconds. CHANNEL CALIBRATION '
shall ensure that this time constant is adjusted to this value.
N constant: "ti'f::d in tM 10:F1:; ::ntnller fer t;: ::t:d TCOLD are :, t 12 Oce:nd: :nd , 1 3 ::::nd:. C"f""5' "L!!P'.TIO" : hell
- n:er: th:t th::: t':: cen: tent: er: edj.;ted te these vs19es. .
j l
- Loop design flow = 95,400 gpa
- 2.0% span for Steam Generator Level; 0.2% span for Reference Leg RTDs 1
- Until resolution of the Veritrak transmitter uncertainty issue, the trip setpoint will be set at 1 1869 psig, with the allowable value at 1 1861 psig.
l l
l 1
SOUTH TEXAS - UNIT 1 3/4 3-36 i
i 6
- ---- - .- , - .----,-,------,,---------,-.m---------
ATTACHMENT 1 TABLE 3.3-5 (Continued) . ST ML.AE. %4 PAGE IZ.0F IB ENGINEERED SAFETY FEATURES RESPONSE TIMES I
INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS
- 3. Pressurizer Pressure--Low
- a. Safety Injection (ECCS) i 27(1)/12(5) ,
- 1) Reactor Trip < 2(3)
- 2) Feedwater Isolation 12(3)
- 3) Phase "A" Isolation [33(1)/23(2) ,
- 4) Containment Ventilation Isolation N.A. .
- 5) Auxiliary Feedwater < 60
- 6) Essential Cooling Water I62(1)/52(2) ,
- 7) Reactor Containment Fan Coolers 38(1)/28(2)
- 8) Control Room \entilation 72(1)/62(2)
- 9) Start Standby Diesel Generators 1 12
- 4. $ .$ bed T g -Lew-Lew
- a. Safety Injection (ECCS) "..^.
- 1) Resetcr Trip F.A.
,, -2) feenater I:olat4en u A.
Phu e "A" Isolatier
- 3) NA
- 4) Containment Vent 444th-holat-ica F.A.
-5) Auxiliary Fee sater N.A.
l
- 5) E::enti 1-Cecling V:ter N.^.
- 7) Reactor Containment Fan Cooler: N.A.
S) Control acca. Ventilatica N.A.
- 9) Start Die::1 C:ncrat+r: N.A.
- b. Ste= Lin: I:c14 tier u i.
- 5. Compensated Steam Line Pressure--Low
- a. Safety Injection (ECCS) 5 22(4}/12(5)
- 1) Reactor Trip -
1 2(3)
- 2) Feedwater Isolation i 12(3)
- 3) Phase "A" Isolation 1 33(1)/23(2)
- 4) Containment Ventilation Isolation N.A.
- 5) Auxiliary Feedwater < 60 1 6) Essential Cooling Water 62(1)/52(2)
- 7) Reactor Containment Fan, Coolers [38(1)/28(2)
SOUTH TEXAS - UNIT 1 3/4 3-38
. . 1 TABLE 3.3-5 (Continued) l ATTACHMENT 4 ENGINEERED SAFETY FEATURES RESPONSE TIMES ST HL.AE .2646 PAGE.83 Of'8 --
.) INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS
- 12. Loss of Power (Continued)
- 13. RCB Purge Radioactivity-High
- a. Containment Ventilation Isolation '
(48-inch lines) 1 73(2) -
- b. Containment Ventilation Isolation (18-inch lines) < 23(2)
- 14. 00mpen::ted Tg --Low
- . Turbin: Trip t'. A.
- b. Teedsater Isolstica ,
N.A.
- 15. Feedacter ficw - Nigh-Ccincident with 2
es of .. 4
_ Loop:
... ._ ,Having E.ither Reactor Coolant iivn wvn va s wvn
. Turbinc Trip - R::ctor Trip N. ;.
- b. feedwater Isclatica N.A.
-)
- 16. T,yg - Low Coincident with Reactor Trip Feedwater Isolation N.A.
- 17. Control Room Intake Air Radioactivity - High .
Control Room Ventilation ,
1 78(2)
- 18. Spent Fuel Pool Exhaust Radioactivity - High FHB HVAC Emergency Startup 5,42(2) ;
l
> 1 l
1 1
)
SOUTH TEXAS - UNIT 1 3I'43-40 SC7 h M I ee
m TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION y SURVEILLANCE REQUIREMENTS "A
g DIGITAL OR TRIP ANALOG ACTUATING MODES c CHANNEL DEVICE MASTER ' LAVE FOR MtICH 5 CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE
- FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED w
- 1. Safety Injection (Reactor Trip, Feedwater Isolation, Control Roon Emergency Ventilation, Start Staney Diesel Generators, Reactor .
Containment Fan Coolers, and Essential Cooling Water)
R
- a. Manual Initiatton N.A. M.A. N.A. R ,
M.A. N.A. N.A. 1, 2, 3, 4 Y
i g b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) N.A. N.A. 1,2,3,4 Logic
- c. Actuation Relays M.A. N.A. N.A. N.A. N.A. M(6) Q(4,5) 1, 2, 3, 4 l d. Containment Pressure- S R M N.A. N.A. N.A. N.A. 1, 2, 3, 4 High-1
- e. Pressurizer Pressure- S R M N.A. M.A. N.A. N.A. 1, 2, 3 Low
- f. Compensated Steam Line S R M N.A. N.A. N.A. M.A. 1,2,3 Pressure-Low i
- g. Cr;:.n:'d T ~
' A- ^- - ^- I' 2
- 3
, COLD t r le ( St w hchet me,6 n_ m 1 Gr.40 bl 30 W.
- oir*)r -3V 1H-iS i r 1N3WH3Y11Y ,
C .. J y, TABLE 4.3-2 (Continued) k I
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS a
- O DIGITAL OR TRIP 3; ANALOG ACTUATING MODES CHANNEL DEVICE MASTER SLAVE FOR WilCH e CilANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLAMI p
FUNCTIONAL UNIT CilECK CALIBRAT_]ON. TEST TEST LOGIC TEST TEST TEST IS REQUIREI g 4. Steam Line Isolation
- e. Compensated Steam Line S R H N.A. N.A. N.A. M.A. 1, 2, 3 Pressure-Low
- f. Compensated T -
S R M H.A. N.A. N.A. N.A. 1, 2, 3 COLD Low-Low (interlocked .
, with P-15) m S. Turbine Trip and Feedwater
). Isolation y a. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(6) Q(4) 1, 2, 3 g logic and Actuation Relays
- b. Steam Generator Water S 'R H N.A. N.A. N.A. h.A. 1, 2, 3 Level-High-High (P-14)
- c. he .-$tcdTg -tr.: S R M N.A. M.A. M.A. M.A. 1, 2, 3 fir.ter!rkM dth P-15)
- d. f *:
- e-F4ew-High S R M N.A. N . ^. . M.A. M.8 1, 2, 3 Hetedecked-with P-15)
Coincident with either S of the following in 2 of 4 loops: Reactor g( Coolant Flow-Low or -
g T,yg-Low
- e. Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.
5 R H N.A. N.A. N.A. N.A. 1, 2, 3 l f. T 3,g-Low Coincident ~
with Reactor Trip (P-4) gr 30 st 30Vri I (Feedwater Isolation . ec9e -3V9H-1S \
Only) t IN3WHOV11V l
g TABLE 4.3-2 (Continued)
C M ENGINEERED SAFETY FEATURES ACTUlm 3N SYSTEN INSTRUNENTATION
-4 SURVEILLANCE REQUIREMENTS O
M DIGITAL OR TRIP
. ANALOG ACTUATING N00ES c CHANNEL DEVICE MASTER SLAVE FOR %dHICN i 5 CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST 15 REQUIRED
]
- 8. Loss of Power (Continued)
- b. 4.16 kV ESF Bus M.A. R N.A. M N.A. N.A. N.A. 1, 2, 3, 4 Undervoltage (Tolerable Degraded Voltage .
Coincident with SI) ti
Degraded Voltage)
- 9. Engineered Safety Features Actuation -
System Interlocks
- a. Pressurizer N.A. R M N.A. N.A. N.A. N.A. 1, 2, 3 Pressure, P-11
- b. Low-Low T,,g, P-12 N.A. R M N.A. M.A. N.A. N.A. 1,2,3
- c. Reactor Trip, P-4 N.A. N.A N.A. R N.A. N.A. N.A. 1, 2, 3
- d. P=- *: ;;: "atm M.A. R(2) M(3) M.A. M.A. M.A. M.A. 1, 2, 3 F!= !; t 5 .
1 .c ---- w r_ w - .
- 9
- 30 9' 39Vd I
. Z. . .. . ,.' i. r_ . , . ' E. .
' Tor -3V-MIS
- 10. Control Hoom Ventilation c 1N3WH3Y11V 5
- j a. Manual Initiatton N.A. N.A. N.A. R N.A. N.A. N.A. All
~
j i
L .- . V g TABLE 4.3-2 (Continued)
S ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION
[ SURVEILLANCE REQUIREMENTS O DIGITAL OR TRIP
- E .
ANALOG ACTUATING MODES CilANNEL DEVICE . MASTER SLAVE FOR WHICH I c CHANNEL CilANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLAN 5
-4 FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRF
- 11. FHB HVAC (Continued)
- c. Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.
- d. Spent Fuel Pool S R M - N.A. N.A. N.A. N.A. With i Exhaust Radio- irradiated activity-High fuel in spent.fue{
, pool.
> ~
Y TABLE NOTATION '
> 0 (1) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.
(2) t detectors r.ey be excisded five C:l%GCL CALIBRATIGii.
(3) Nh. ted Theenal P;ucr ;;rcater-thar, or equal-to-the P 15 interleck set i ,sint, tM ."f". LOG CHANNEt 4)DERATIOM".L TEST-shah-censrist-of-verifying-that-the P-15 interlock-is-in-the required siste-by-
-observig tM p missive-annunicater-window.
(4) Except relays K807, K814, K829 (Train 8 only), K831 K845, K852 and K854 (Trains B and C only) which shall be tested at least once per 18 months during refueling and during each COLD SHUIDOWN exceeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless they have been tested within the previous 92 days.
D (S) Except relay K815 which shall be tested at indicated interval only when reactor coolant pressure is above 700 psig.
(6) Each actuation train shall be tested at least every 92 days on a STAGGERED TEST BASIS. Testing of
() each actuation train shall include master relay testing of both logic trains. If an ESFAS instru-mentation channel is inoperable due to failure of the Actuation Logic Test and/or Master Relay Test, i
increase the surveillance frequency such that each train is tested at least every 62 days on a STAGGERED TEST BASIS unless the failure can be determined by performance of an engineering evaluation to be a single random failure. -
91 30 1130 tci i
- During CORE ALTERATIONS or movement of irradiated fuel within containment. - Tyr -3V9H-1S ;
r IN3WH3Y11V
- 4
0 2 ATTACHMENT 1 l INSTRtMENTATION
- GT HL AE 2626 l .- PAG
..E18 OFlo . . . -
l REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued)
The EngineeNd Safety Features Actuation Systes interlocks perform the following functions:
P-4 Reactor tripped - Actuates Turbine trip via P-16, closes nata feed- -
water valves on T,,, below Setpoint, p nvents the opening of the
- main Steamfeedwater Generatorvalves whicherwere Water Level E"::::i": closed by a safet{ntetter :!gelwInjectio C^^1dr allows safety Injection block so that components can be reset or tripped :M u te t:: *-15. - . .
Reactor not tripped - prevents sanual block of Safety Injection.
P-11 On increasing pressurizer pressum, P-11 automatically reinstates -
Safety Injection "tetion on low pressurizer pressure org*cessrive-cooldown-signals, nis cates steaaline isolation on' excess 4ve-cJo4= 3.,4 down-signals, anc t*:* the accumulator discharge isolation valves.
On decreasing preen, P-11 allows the manual block of Safety Injec g I tion actuation on hw pressurizer pressure or' excessive-cooldown-sig-1 mais, allows the manual block of steaaline isolation on' excess 4ve-cooldown-sigcals3 and enables steaa line isolation on high negative steaa line pressure rate.
P-12 On .'ncreasing reactor coolant loop temperaum 0-12 automatically prc< ides an arming signal to the Steam Dump 5 . is. On decreasing l reactor coolant loop temperature, P-12 automatically removes the l arming signal from the Steam Dump System. I l
P-14 On increasing steam generator water level, P-14 .sutomatically trips i the turbine and the main feedwater pumps, and closes all feedwater isolation valves and feedwater control valves, Me '
- w. . u. u < - *e TiTC 'MEI7 U I t- Iessat cd (P-4) and er Wn MW-th F-ir Tinii Dist*
aller.-Safety-In(*etion-eetwation i
WiMHOUMtien en low-L~ T cold
- l I'
ksolatten :M te-bin tr49-#es-LOW CM OS58t*d celd nhWI +
water-Skw 3/4.3.3 MONITORING INSTRtMENTATION 3/4.3.3.1 RADIATION MONITORING FOR PLANT OPERATIONS The OPERABILITY of the radiation monitoring instrtmentation for plant operations ensures that: (1) the associated action will be initiated when the radiation level acnitored by each channel or combinatten thereof maches its Setpoint, (2) the specified coincidence logic is maintained, ara (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 a n 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 combination is cospleted, the system sends actuation signals to initiate alams or automatic isolation action and actuation of Energers.y Exhaust or Ventilation Systees. .
SOUTH TEXAS - UNIT 1 8 3/4 3-3
@) Low Compenseded Siendne bsswe sig%ls
ATTACHMENT 3 1 ST HL AE abh PAGE i OF 40 1
ATTACHMENT 3 PROPOSED REVISIONS FSAR l
l l
r 1
l 2
l TABLE 1.3-2 (Continued) 4
. SIGNIFICANT DESIGN CHANCES References I teen FSAR Description of Chang, a containment isolation for stema sections 7.3, 6.2.4 Containment isolation phase A signal changed to AFnt initiation signat generator blowdown and sample lines (Si or tow-tow SG water tevet).
Electrical penetration space Section 7.3 System is no longer actuated by control room emergency ventitation ventilation system actuation signat, only by si signal.
Actuated equipment lists Section 7.3 Various changes as regJired to support system design changes.
Radiation input to contalonent Section 7.3 Input via redJndant safety-grade RCs purge isolation monitors.
ventitation isolation I to.a colpea-Mad. Mcadme penwe sco u-c+a s. bl sf a Inc p<es.o,, r. r.
y Saesee' c: :::M: n F:::::'- Section 7.3 stock permissive for -c-?'- r" n m*-*'r changed from P-12
[ block permissive (Low-low T 45 avg
) to P-11 (pressurizer pressure). E Wide range RCS pressure Sections 7.4, 7.6 Addition of 2 RCS wide range pressure channets. Retocation of 3 outside containment. transaitters outside containment.
Post accident monitoring Section 7.5, Appendix 78 Upgrade instrumentation to address RG 1.97, Rev. 2. Instrumentation Qualified Display Processing Section 7.5 Addition of safety-related display processing system which provides System (QDPS) redJndant data acquisition and display. System provides Class 1E control of the SG power operated relief valves, head vent throttle (ps )>
i valves, AFU flow regulating valves, and ECW throttle valve for the N essential chitters. The system also performs SG reference leg temperature l '>
' tn compensation and RCS hot teg tesperature averaging calculations. N N,,,
O seu y
Emergency Response Facilities Data Acquisition and Display Section 7.5 Provides signal processing and display for Emergency Response Facit stles muz 4 and addresses SPDS requirement of NUREG0695.
- 2 gp Q.
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ATTACHM5NT 3 STP FSAR I ST.HL.AE AGA6 PAGE c5 0F YO 30 This transient is conservatively defined as an umbrella case to cover occur-rence of several events of the same general nature. These include:
- 1. Inadvertent opening of an FW control valve.
- 3. Small steam break with an open FW controlv'alve.
The excessive FW flow transient results frem inadvertent opening of an FW control valve when the plant is at hot shutdown and the SG is in the no load condition. The FV, Condensate, and Heater Drains Systems are in operation.
The stem of an FW control valve has been assumed to fail, with the valve imme-diately reaching the full open position. The FW flow to the affected loop is 30 assumed to step from essentially zero flow to the value determined by the system resistance and the developed head of all operating FW pumps, with no main feedwater flow to the other loops. Steam flow is assumed to remain at 30 zero, and the temperature of the FW entering the SG is conservatively assumed to be 32*F. F::d :ter flew i: i:0100:d ;n a reacter ceelant low Tceld sign =1; e cubecquent leu-1:e Tecid signal 20tuct:0 the S fety Injectica Syst:: (CIS).
Auxiliary Feedwater (AFV) flow, initiated by the SI signal, is assumed to continue, with all pumps discharging into the affected SG. It is also assumed, for conservatism in the secondary side analysis, that AFW flows to the SGs not affected by the malfunctioned valve, in the so called "unfailed 30 loops". Plant conditions stabili'ze at the values reached in 600 seconds, at which time AFW flow is terminated. The plant is then either taken to cold shutdown or returned to the no load condition at a normal heatup rate with the AFVS under manual control.
For design purposes, this transient has been assumed to occur 30 times during the 40 year life of the plant.
3.9.1.1.8 Emergency Conditions: The following primary system transients have been considered emergency conditions:
- 1. Small Loss-of Coolant Accident (LOCA)
- 2. Small steam line break 4I
- 3. Complete loss of flow 3.9.1.1.8.1 Small Loss of-Coolant Accident For design transient pur-poses, the small LOCA is defined as a break equivalent to the severance of a 1-in..inside diameter branch connection. (Breaks smaller than 0.375 in.
inside diameter can be handled by the normal makeup system and produce no significant fluid systems transients.) Breaks which are much larger than 1 in. will cause accumulator injection soon after the accident and are regarded as faulted conditions. For design purposes, it is assumed that this transient occurs five times during the life of the plant. It should be assumed that the ECCS is actuated immediately after the break occurs and subsequently delivers water at a minimum temperature of 32'F to the RCS.
3.9 11 Amendment 57
- _ _ , . - - - - .__ . , . _ . . - - . - _ , , ,- --m , w---
~
ATTACHMENT 3 STP FSAR St % AR. 3 26
_PAGEN M. W However, for those modes of operation when water solid operation may still be possible, procedures will further highlight precautions that minimize the severity of, or the potential for, an overpressurization transient. The l54 following precautions of measures are considered in developing operating procedures:
- a. Whenever the plant is water solid and the reactor coolant pressure is being maintained by the low pressure letdown control valve, letdown flow normally bypasses the normal letdown orifices.
- b. If all reactor coolant pumps have stopped for more than 5 minutes during plant heatup after filling and venting has been completed and the reactor coolant temperature is greater than the charging and seal injection water temperature, a steam bubble will be formed in the pressurizer prior to restarting a reactor coolant pump. This precaution minimizes the pres-sure transient when the pump is started and the cold water previously injected by the charging pumps is circulated through the warmer reactor coolant components. The steam bubble will accommodate the resultant expansion is the cold water rapidly warmed.
- c. If the reactor coolant pumps are stopped and the RCS is being cooled down by the residual heat exchangers, a nonuniform temperature distribution may occur in the reactor coolant loops. Prior to restarting a reactor coolant pump, a steam bubble will be formed in the pressurizer or an acceptable temperature profile will be demonstrated,
- d. During plant cooldown, all steam generators will normally be co;,nected to 53 the loops.steam header to assure a uniform cooldown of the reactor coolant ('
These special precautions back-up the normal operational mode of maximizing periods of steam bubble operation so that cold overpressure transient preven-tion is continued during periods of transitional operations. These precau-tions do not apply to reactor coolant system hydrostatic testing.
The specific plant configurations of emergency core cooling system testing and alignment cold will also highlight overpressurization procedural recommendations to prevent developing transients. During these limited periods of plant operation, procedures:
the following precautions / measures are considered in developing the a.
To preclude inadvertent emergency core cooling system actuation during heatup and cooldown, procedures required blocking the pressurizer pressure, and excessi"e the P-11 setpoint. t iov cmcoelde" protection signal actuation logic below peew.M tb h e pesme s'.r.
- b. During further cooldown, closure and power lockout of the accumulator !
isolation valves will be performed at 1,000 psig. When the RCS tempera- !
ture is reduced to or below 350'F, a maximum of one centrifugal charging pun.p and one HHSI pump is allowed operable by Technical Specifications.
The LHSI pump does not impact the COMS analysis because of the low ;
shutoff head (approximately 315 psi). 154 I
- c. The recommended procedure for periodic emergency core cooling system pum performance testing will be to test the pumps during normal power opera p tion or at hot shutdown conditions. This precludes any potential for developing a cold overpressurizatio6 transient.
5.2 4b Amendment 54
_ o
- ATTACHMENT 3 STP FSAR . ST.HL AE M %
PAGE 5 OF Yo 6.2.1.4 Mass and Energy Release Analysis for Postulated Secondary System Pipe Ruptures Inside the Containment. Following a postulated main steam line break or a main feedwater line break inside the Containment, the contents of one SG will be released to the Containment. Most of the contents of the other SGs will be isolated by the main steam isolation valves (MSIVs) and main feedwater isolation valves. Containment pressurization following a secondary side rupture depends on how much of the break fluid enters the Con-tainment atmosphere as steam. Main steam line break flows'can be pure steam or two-phase, while main feedwater line break flows are two phase. With a pure steam blowdown, all of the break flow enters the Containment vapor space atmosphere. With two-phase blowdown, part of the liquid in the break flow boils off in the Containment and is added to the vapor space atmosphere, while the remaining liquid falls to the sump and contributes nothing to Containment pressurization. For main steam line break cases with large break area, steam cannot escape fast enough from the two-phase region of the ruptured SG, and the two-phase level rises rapidly to the steam line nouzie. A two-phase blowdown results. The duration of this blowdown is short, therefore reducing primary-to-secondary heat transfer, and the break flow is largely liquid.
For main steam line break cases with small break areas, steam can escape fast enough from the two phase region of the SG with the ruptured line that the level swell does not reach the steam line nozzle, and a pure steam blowdown results. Because of the pressure reducing effects of active and passive Con-tainment heat sinks, the highest peak Containment pressure resulting from a main steam line break for a given set of initial SG conditions occurs for that case where the break area is the maximum at which a pure steam blowdown can occur. For conservatism, the main steam line break analysis assumed only pure steam blowdown for all break sizes and power levels.
Main steam line isolation is initiated on the following signals: high-2 Containment pressure, low steamline pressure or lev-low T - (above P 11 57 setpoint),highnegativesteamlinepressurerate(belowt%II'11setpoint), l and manual. Main feedwater line isolation is initiated by SG High-High water '
level, exeessiv; : eldown-pretection signal, reactor trip in conjunction with 2 l low T and SI. Both the MSIVs and the main feedwater isolation valves are fully *cfo, sed in 5 seconds.
The Auxiliary Feedwater System functions automatically following a secondary 49 system line break to assure that a heat sink is always available to the RCS by supplying cold feedwater to the SGs. For conservatism, it was assumed that i the Auxiliary Feedwater System attains full flow to the SG immediately follow- I ing feedwater isolation. In addition, the analysis includes the flashing of the volume of fluid located between the main Feedwater isolation valve and the 49 affected SC. This fluid then flows through the affected SG and into the Containment.
The feedwater enters the SG in the two phase region; therefore, main feedwater line break cases always result in two phase blowdowns through smaller size lines and do not produce peak Containment pressures as severe as main steam line break cases.
To permit a determination of the effect of main steam line break upon Contain- g4 ment pressure, a spectrum of break sizes was assumed to occur inside the Can-tain=ent, dovnstream from the integral steam line flow restrictors and up stream of the MSIVs. Unrestricted critical flow from the rupture was assumed.
6.2-21 Amendment 57 m
- . ~
- -~.
ATTACHMENT 3 STP FSAR . ST.HL.AE. M S PAGEto OF @
7.1.2.1.8 Diversity: Functional diversity has been designed into the system. Functional diversity is dir. cussed in Reference 7.1-1. The extent of the diverse system variables has been evaluated for a wide variety of pos-tulated accidents. Generally, two or more diverse protection functions would automatically terminate an accident before unacceptable consequences could occur. l 20 Regarding the ESFAS for a Loss of Coolant Accident (LOCA), a SI signal can be 43 obtained manually or by automatic initiation from two diverse parameter measurements: l2
- 1. Low pressurizer pressure
- 2. High containment pressure (HI 1)
For a steam line break accident, diversity of SI actuation is provided by: (3
-1. Lemlee comp - ated T l12 1, t . Low compensated steam line pressure
- 7. A Low pressurizer pressure 3.%. For a steam line break inside Containment, high Containment pressure l43 (HI-1) provides an additional parameter for generation of the signal.
All of the above sets of signals are redundant and physically separated and meet the requirements of IEEE 279-1971.
7.1.2.1.9 Bistable Trip Setpoints: Three values applicable to reactor trip and ESF actuations are specified: ,
- 1. Safety limit
- 2. Limiting value
- 3. Nominal setpoint The safety limit is the value assumed in the accident analysis and is the least conservative value.
The limiting value is the Technical Specification value and is obtained by subtracting a safety margin from the safety limit. The safety margin accounts for instrument error, process uncertainties such as flow stratification and transport factor effects, etc.
The nominal setpoi.c is the value set into the equipment and is obtained by subtracting allowances for instrument drift from the limiting value. The nominal setpoint allows for the normal expected instrument setpoint drifts so that the Technical Specification limits will not be exceeded under normal operation.
7.1 9 Amendment 57
ATTACHMENT 3 STP FSAR . ST HL AE MA6 PAGE 7 0F YO TABLE 7.1-2 (Continued)
PLANT COMPARISON
- DIFFERENCES FROM REACTOR TRIP SYSTEM (Continued) COMANCHE PEAK NUCLEAR STATION
- 8. Source Range Flux Detector 8. On Comanche Peak, each source Energization (Figure 7.2 3) range flux detector is ener-gized and de energized by logic output from a single train (the two detectors are on separate trains). On STP, to de energize each detector, outputs from both A and B actu-ation trains are used; to energize each detector, output from either actuation train (A or B) is used.
ENGINEERED SAFETY FEATURES 44 ACTUATION SYSTEMS Dele be C;;;nc}..:
- 1. Interi::h P-15 1. P: h d::: net provid (Figura ' 2-4)
- P-15 cign:1. Or STP, the P 15 g,),4,j eign:1 i: developed fr;; "-4 (seectuc l tripp:d) er 2/' p:ver range acte:t:rs rh0t'in; neutrOr flun 5:1;r ;otpuiui.
?-15 i: ;;;d :: :n interlock in
-exce :1.fe emelanun penr.r*<nn ingir (Figur 7.2 9).
- 2. Steam Generator High-High 2'. Four channels are used for each Water Level Signal SG (2/4 logic) on STP; three (Figure 7.2-7) channels are used for each SG l (2/3 logic) on Comanche Peak, i
- 3. Main Steam Line Isolation 3. Automatic actuation signals Initiation (Figure 7.2-8) for Comanche Peak are high l negative steam pressure rate, 57 low steamline pressure and Containment HI-2 pressure.
In addi+i:n-t; these same 44gnalegutomat4o-s teamHne- l57
-1solet-ion-signalt-for STP imelude--e-low-1;v T s i gnal-.
l57 l
l l
l 7.1 32 Amendment 57 W
. ST HL AE M-@
TABLE 7.1-2 (Continued) PAGE 8 0F YC l
PLANT COMPARISON
- ENGINEERED SAFETY FEATURES DIFFERENCES FROM ACTUATION SYSTEMS (Continued) COMANCHE PEAK NUCLEAR STATION
- 4. Safety Injection Initiation 4. Automatic SI actuation signals through Exterriv: C::ld:er used on both plants are Con-I tainment HI-l pressure and low Pre te: ~l er (Figures 7.2 8 and Pressurizer pressure. ,C;; nch:
7.2-9) low co.g abd -
stcachse pre nvre and P: h :1:: u::: low steamline pressure (2/3 in any loop). Sil'*
u::: th: ::::: ive cecideen cignal
_, ,_._ ,_.. ________ .a ,
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preceure (2/3 ir eneb leep) in any 1;;p.
- 5. Centainment Spray Actuation 5. On Comanche Peak, the spray .
(Figure 7.2 8) pumps are started by the SI l signal, while the containment I spray signal confirms pump I start and opens system valves. On STP, the SI signal does not actuate any containment spray equipment; only the con- l tainment spray signal actuates 44 Containment Spray System equipment.
- 6. Radiation Signal Inputs to 6. On Comanche Peak, the radiation Containment Ventilation inputs to the containment venti-Isolation (Figures 7.3 2A lation isolation signal are and 7.2 8) the three detectors (particulate, iodine, gas) of the containment air monitor. On STP, the radiation inputs are the two Class lE RCB purge isolation monitors (gas detectors) and the non Class lE Containment atmosphere monitor (particulate, iodine, gas detectors).
- 7. Control Room HVAC ESF 7. Both plants utilize the SI Actuation Signals (Figures signal for control room air 7.2 8 and 7.3 24) cleanup filtration. Comanche Peak has a common control room; each control room inlet radiation monitor actuates the corresponding control room HVAC train. Also each unit's plant vent stack vide range gas radiation monitor 7.1-33 Amendment 57 w- ,-
. ST.HL AE 2646
' TABLE 7.1 2 (Continued)_.P. AGE 9 0F 'f 6 PLANT COMPARISON
- ENGINEERED SAFETY FEATURES DIFFERENCES FROM ACTUATION SYSTEMS (Continued) COMANCHE PEAK NUCLEAR STATION actuates one control room HVAC train. STP has a separate control room for each unit.
Each control room has redundant air inlet radiation monitors, each actuating all three trains of control room HVAC.
44
- 8. Fuel Handling Building 8. STP uses SI signal or high Exhaust HVAC ESF Actuation radiation signal (from either Signals (Figure 7.3-27) of two redundant Class lE spent fuel pool exhaust monitors) to ,
initiate FHB exhaust filtration, j On Comanche Peak, fuel building l exhaust is always filtered; no actuation is required.
- 9. Deleted 61
- 9. Deleted Del fed I
- 10. Excercive Cecideen Feedenter 10. ..dd!tienforSTPefen::::ive
^
Isoletica Sign 1r (Figurer :::1deur. :Egnal: for feedveter
'.2-5, '.2 ', '.2-9 and i::1stien er:
' . 2 l'd # pa, [1:w primery leap flew er 10" I I 17 2/6 100F;
- Delet'd high $8Efle"
- P 15 ,
interie:P rign:1, or i
- i
- b. 1:= :::p:n::ted T P-15 inter 10:hci*gE1I. 44 l Cerench: P :h d::: n:t hree these
-cign:le.
- 11. Turbine Trip Signal From 11. Addition on STP of manual Feedwater Isolation Signals reset capability for the turbine (Figure 7.2-14) trip signal from the combined l signal of P-16 or any of the following signals: safety injection or P-14 signal 4HN
-e*:e e iv: 00:1dern feeducter--
4seletier. Comanche Peak does not provide this capability.
7.1 34 Amendment 61
. ST.HL AE J606 TABLE 7.1-2 (Continued) PAGE 10 OF Yo PLANT COMPARISON
- ENGINEERED SAFETY FEATURES DIFFERENCES FROM ACTUATION SYSTEMS (Continued) COMANCHE PEAK NUCLEAR STATION
- 12. P 4 Signal / Safety Injection 12. Comanche Peak: After P-14 or P 14 Signal er Ex::::1: - signal or SI signal is Ceelderr Feedwater Isolation received, feedwater isolation Interface (Figure 7.2-14) signal is sent. This signal is then sealed in through l coincidence with the P 4 reactor trip.
STP: The SI signal or P 14 signal er excer:17: ::aldown FW isolation signal sets a retentive memory for FW isolation. Absence of a P 4 reactor trip then allows reset of the memory.
- 13. P-4 Signal / Low T Signal 13. Comanche Peak: Presence of Feedwater IsolatioE Interface P 4 reactor trip and low T (Figure 7.2 14) signals sets a retentive ""E l memory (with actuation block). l Manual reset of this memory !
allows repositioning of all FW 44 control and bypass control valves (if closed by that signal). ,
1 l
STP: Presence of P 4 reactor trip and low T signals seals in the 1$w ST signal, sends a (non-resetla$le) closure signal to the FW control valves and sets a retentive memory (with actuation block), which can be manually reset to allow reposi-tioning of the FW bypass control valves. (Difference is that the STP FW control valves cannot be repositioned until the reactor trip signal is removed.)
- 14. Auxiliary Feedwater System 14 Comanche Peak: Two motor-Actuation (Figure 7.2 16) driven pumps are automatically actuated by SI signal or blackout (LOOP) signal or trip of both main feed pumps or low-low water level in any SG. One turbine driven pump 7.1-35 Amendment 57
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ATTACFDAENT 3 ST.HL AE 2W sEP FSAR PAGE 19 0F WD steam pressure rate), 7.2-8 (ESF actuation), 7.2 9 fs enere:1c; Los cocersde/ siem 6e pre tsureprefec hm
- 1devr.
pret:: tier.), 7.2-14 and 7.2-15 (feedwater control and isolation), and 7.2 16 (auxiliary feedwater).
To facilitate ESF actuation testing, six cabinets (two per train) are provided which enable operation, to the maximum practical extent, of safety features loads on a group by group basis until actuation of all devices has been checked. Final actuation testing is discussed in detail-in Section 7.3.1.2.
7.3.1.1.4 Final Actuation Circuitry: The Solid-State Protection System supplies the following signals: 43
- 1. Safety injection signal (Table 7.3 5 lists actuated equipment. Typical control logics for actuated equipment are shown on Figures 7.3-2 through 7.3 8.)
Containment spray signal (Table 7.3-6 lists actuated equipment. F3 2.
Typical control logics for actuated equipment are shown on Figures 7.3-9 and 7.6-14.)
- 3. Containment isolation Phase A signal (Table 7.3 7 lists actuated equip- 43 ment. Typical control logics are shown on Figures 7.3-11 through 7.3-13.)
- 4. Containment isolation Phase B signal (Table 7.3-8 lists actuated equipment. Typical control logics are shown on Figures 7.3-14 and 61 7.3-15.)
- 5. Containment ventilation isolation signal (Table 7.3 9 lists actuated equipment. Typical control logics are shown on Figures 7.3 16 and 7.3 17.)
- 6. Steam line isolation signal (Table 7.3-10 lists actuated equipment. !
Typical control logics are shown on Figures 7.3-18 and 7.3 18A.) !
43
- 7. Feedwater isolation signal (Table 7.3 11 lists actuated equipment.
Typical control logics are shown on Figures 7.3 19 and 7.3 20.) l l
- 8. Auxiliary Feedwater (AFV) initiation signal (Table 7.3-15 lists 45 QO32 actuated equipment. Typical control logics are shown on Figures *16 7.3-21, 7.3-21A and 7.3 21B.)
1 Loads are sequenced onto the three Class 1E ESF buses by the ESF load 43 sequencers, as described in Chapter 8. The design meets the requirements of l GDC 35, 7.3.1.1.5 Design Bases Information. The functional diagrams presented on Figures 7.2-5 through 7.2 9 and 7.2-14 through 7.2 16 provide a graphic outline of the functional logic associated with requirements for the ESFAS.
Requirements for the ESFAS are given in Chapter 15. Given below is the design bases information required in Institute of Electrical and Electronic Engineers (IEEE) 279-1971.
7.3 5 Amendment 61
. ST.HL.AE om PAGE Jo 0F Yo 7.3.1.1.5.1 cenerating Station Conditions - The following is a summary l43 of those generating station conditions requiring protective action:
- 1. Primary system:
- a. Rupture in small pipes or cracks in large pipes
- b. Rupture of a reactor coolant pipe or LOCA
- c. Rupture of a SG tube -
- 2. Secondary system:
- a. Minor secondary system pipe breaks resulting in steam release rates equivalent to a single dump, relief, or safety valve
- b. Rupture of a major secondary system pipe
- 3. Fuel handling accident inside Containment 7.3.1.1.5.2 Generating Station Variables: The accidents identified above are described in Chapter 15, including the ESFAS signals used to miti-gate the accident consequences. The variables listed below are monitored for the automatic initiation of ESF systems during these accidents. Post-accident monitoring requirements are discussed in Section 7.5.
- 1. Containment pressure
- 2. Pressurizer pressure '
- 3. Steam line pressure
+--- Re cter coetant ecid Icg temperatur: (T ) 43 4./. SG water level
- 5. Teedwetes flow Primary coolant flow-
- f. J. Normal and Supplementary Containment purgo exhaust radiation 61
(, . /
Containment atmosphere radiation (no credit taken for this parameter) belehf 7.3.1.1.5.3 bPaul$yh..
e / Ocrenden)t ":riabler : The caly riable meni-tored for deriving ESFAS cecident =itigating signals that could be wuoi-dered spatially dependent is the T sur W ich-ir ade at t he-sold 4eg-of-.-each leep domstreen ei de::e2cter r coolent pep. In this 43
-location turbulent mixing at the p=p vill climinate stratification.
7.3.1.1.5.4 Limits, Margins and Setpoints: Prudent operational limits, available margins, and setpoints before onset of unsafe conditions requiring protective action are discussed in Chapter 15 and the Technical 43 Specifications.
7.3 6 Amendment 61 g -- - - -- -
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ATTACHMENT 3
. ST HL AE 4634 STP FSAR pA(;F 21 0F Yo
- 2. Typical maximum allowable time delays in generating the actuation signal ;
for secondary system break protection, in addition to the above, are-l
- a. Steam line pressure (from which 0.6 seconds steam line pressure rate is also derived and to which add 0.5 see)
- b. Tcold (dir::t irrrreier ir ::1d 5.0 :::end: wi:h flav 4444- .i !. !. 7. .u l.~ I. .I " I ' . " ?. n.
1
- cend: :t : r: 21==.
l 43 b y. Actuation signals for auxiliary 2.0 seconds feedwater pumps (steam generator water level)
- d. Pslaary leep flee 1.0 ::: ends 1
- e. Feedeater fico 2.0 recends . I
- 3. The time delay in generating the Containment ventilation signal for a fuel handling accident inside Containment is the total of the time 43 delay in the radiation monitors and the time delay in the Solid State )
Protection System to generate the Containment ventilation isolation I signal. The maximum allowable time delay is 12.5 seconds for the design 53 58 basis release analyzed in Section 15.7.
7.3.1.1.5.6.2 System Accuracies - ,
l
- 1. Typical accuracies required for generating the reqaired actuation h3 {
signals for Reactor Coolant System break protection are:
- a. Pressurizer pressure (uncompensated) 114 psi
- b. Containment pressure 11.8 percent of 57 l full scale
- 2. Typical accuracies required in generating the required actuation signals for secondary system break protection, in addition to the I above, are: !
- a. Steam line pressure +2.5% of span 43 u , .n.-
' cold I' '
b g. Actuation signals for auxiliary feedwater pumps (steam generator 12.3 percent of span 43 water level)
- d. Pr4ea ry-loop-f1:e i2.75t or span 13 7.3 8 Amendment 58
ATTACHMENT 3 STP FSAR ST HL RE. Abas PAGE E2 OF 40
- . F::de:ter flee 15.0t i? psn 13 l
- 3. Typical accuracy in generating the required radiation actuation signals for the Containment ventilation isolation signal is 133 percent. 53 7.3.1.1.5.6.3 Ranges of Sensed Variables to be Accommodated Until Conclusions of Protective Action are Assured -
43
- 1. Typical ranges required in generating the actuation signals for Reactor Coolant System break protection are:
- a. Pressurizer pressure 1,700 to 2,500 psig
- b. Containment pressure -5 to 65 psig 53 '
- 2. Typical ranges required in gener: ting the eetuation signals for secondary system break protection, in addition to the above, are:
- a. Steam line pressure (from which steam line pressure rate is derived) O to 1,400 psig b T 510' t; ?!GiF
,a l b /.f Actuation signals for auxiliary + 6 ft from i feedwater pumps (steam generator nominal full-load '
water level) water level
- d. Pr'==ry lacp flee O te 120e AP
= Feedwatar flee O-te 100g AP* 1 l
- 3. The typical range required in generating the radiation actuation signals8 for the Containment ventilation isolation signal is 1 x 10,e pCi/cm to 0.1 pCi/cas 43 7.3.1.1.6 Final System Drawings. Functional block diagrams, electrical elementaries, and other drawings required to perform a safety review are listed in Section 1.7. I 7.3.1.2 Analysis 7.3.1.2.1 Failure Modes and Effects Analyses. Failure modes and effects 63 analyses have been performed generically on the ESTAS within the scope of Westinghouse and documented in Reference 7.3 4. The results verify that these 49 systems meet protection single failure criteria as required by IEEE 279 1971.
The STP ESFAS, although not identical, is designed to equivalent
- Corresponds-te O to 120 percent of rated ".' flev :.t design-ret 4ngr-7.3 9 Amendment 53 m
! ST HL.AE. A 6 M.
PAG _E Z3 0F 40 '
The output of each of the in'i tiation circuits consists of a master relay which 43 drives slave relays for contact multiplication as required. The master and slave relays are mounted in the ESFAS cabinets, designated Train A Train B, and Train C respectively, for the redundant counterparts. The master and slave relay circuits operate various pump and fan circuit breakers or starters, motor operated valve contactors, solenoid operated valves, standby 53 diesel generator starting equipment and other ESF actuation devices. l3 4
7.3.1.2.2.5.4.2 Analor Testing - Analog testing is identical to that 43 used for reactor trip circuitry as described in Section 7.2.2.2.3 and includes the following analog channels for other safety related circuits:
- 1. Containment pressure 43
- 2. Pressurizer pressure
- 3. "eeetor-coolant ::1d 1;; ..crr:r - r:ng: terper:tur: (exceeeive ceeldeea l43 pretectien) l l
^ Feeduster fleu (exceerive ee:1 deer pretection) l S. Prisery cealent flee (::::::iv: :::1dern prete:tien) 43
- 3. Jr. Steam line pressure An exception to this is Containment spray, which is energized to actuate 2/4 and reverts to 2/3 when one channel is in test.
7.3.1.2.2.5.4.3 Solid State Logic Testing - Except for Containment spray 43 channels, solid state logic testing is the same as that discussed in Section 53 7.2.2.2.3. During logic testing of one train, the other lo,;*c train can ini.
tiate the required ESF function (Ref. 7.3 2).
7.3.1.2.2.5.4.4 Actuation Testing - At this point, testing of the i'ni- 43 tiation circuits through operation of the master relay and its contacts to the coils of the slave relays has been accomplished. Slave relays do not operate because of reduced voltage.
The ESEAS final actuation device or actuated equipment testing is performed 43 from the Safeguards Test Cabinets. These cabinets are located adjacent to the ESFAS cabinets. There is one set of test cabinets provided for each of the three actuation trains, A, B, and c. Each set of cabinets contains individual test switches necessary to actuate the slave relays. To prevent accidental actuation, test switches are of the type that must be rotated and then depres-sed to operate the slave relays. Assignments of contacts of the slave relays for actuation of various final devices or actuators have been made so that groups of devices or actuated equipment can be operated individually during plant operation without causing plant upset or equipment damage. In the 2 unlikely event that an SI signal is initiated during the test of the final device that is actuated by this test, the device will already be in its safeguards position.
During this Inst procedure, close communication between the main control room operator and the operator at the test panel is required. Prior to the 7.3 13 Amendment 53
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- handle unexpected events which can be better dealt with by operator appraisal l
of changing conditions fol'.owing an accident.
It is most important to note that manual control of the spray system does not occur once actuation has begun by just resetting the associated logic devices alone. Components seal in (latch) so that removal of the actuation signal, in itself, neither cancels nor prevents completion of protective action nor 43 provides the operator with manual override of the automatic system by this single action. In order to take complete control of the system to interrupt its automatic performance, the operator must deliberately unlatch relays which have "sealed in" the initial actuation signals in the associated motor control l43 center, in addition to tripping the pump motor circuit breakers, if stopping the pumps is desirable or necessary.
The manual reset feature associated with Containment spray, therefore, does not perform a bypass function. It is merely the first of several manual oper-ations required to take control from the automatic system or interrupt its completion should such an action be considered necessary.
In the event that the operator anticipates system actuation and erroneously concludes that it is undesirable or unnecessary and imposes a standing reset condition in one train (by operating and holding the corresponding reset switch at the time the actuation signal is transmitted), the other trains 43 automatically carry the protective action to completion. In the event that the reset condition is imposed simultaneously in all three trains at the time 43 the actuation signals are generated, the automatic sequential completion of system action is interrupted and control has been taken by the operator.
~
Manual takeover is maintained, even though the reset switches are released, if 43 the original actuation signal exists. Should the actuation signal then clear and return again, automatic system actuation will repeat.
Any time delays imposed on the system action are applied after the initiating signals are latched. In this ,way, delays of actuation signals for fluid 43 system lineup, load sequencing, etc., do not provide the operator additional l time to interrupt automatic completion with manual reset alone, as would be l the case if a time delay were imposed prior to sealing of the initial actua-tion signal. ,
los ccesM sk.d.a 43 1
'Ihe nanual block controls of pressurizer pressure input and exeeerrive coel4own-P"ue* pret:: tion-input to the SI signal provide the operator with the m7tsns to block initiation of high SI during mafgjteam ine isolation on steamplant shutdown pressure negativeandrate startup and (eku W P allogj
- ":::1 km-$M,[Afon ec 53 block only). These block features meet the requirements of Paragraph 4.12 of IEEE 279 1971 in that automatic removal of the block occurs when plant conditions require the protection system to be functional.
7.3.1.2.2.7 Hanual Initiation of Protective Actions (RG 1.62): There are eight individual main steam isolation momentary control switches (two per 43 loop) mounted on the control board. Each switch, when actuated, isolates one of the main steam lines. In addition, there are two system level switches.
Operating either switch isolates all four steam lines at the system level.
No exception to the requirements of IEEE 279 1971 has been taken in the manual
. initiation circuit of safety injection. Although Paragraph 4.17 of IEEE 7.3 17 Amendment 53
- . ,e .
. ST.HL AE. 4626 TABLE 7.3 2 'FAGE2iOF 44 INSTRUMENTATION OPERATING CONDITION FOR VESTINGHOUSE ESFAS No. of No..of. . Channels No. Functional Unit Channels To Trip
- 1. Safety Injection Signal (See Figures 7.2-8 and 7.2 9)
- a. Muul 2 1 j 2 43
- b. HI 1 Containment pressure 3
- c. Low compensated steam line 12 (3/ steam 2/3 in any, steam pressure
- line) line (43 f
- d. Pressurizer low pressure
- 4 2 er Lee-lev compensated T
- I 00E) / 0"7 00E 57 (intericek:dwith?137Id (43
- 2. Containment Spray Signal 43 (See Figure 7.2 8)
- a. Manual ** 2 1
- b. Containment pressure 4 2 HI-3
- 3. Auxiliary Feedwater Initiation Signal '
43 (See Figure 7.2-16) Q32.16
- a. Safety Injection Signal See item 1 of th!s table (
- b. Steam generator low-low 16 (4/SG) 2/4 in any SG i water level l 1
- Permissible bypass if reactor coolant pressure is less than P-11 (nominally 43 1900 psig).
- Manual actuation of Containment spray is accomplished by actuating either of two sets (two switches per set). Both switches in a set must be actuated to obtain a manually initiated spray signal. The sets are wired to meet separation and single failure requirements of IEEE 279 1971. Simultaneous operation of two switches is desirable to prevent inadvertent spray actuation.
7.3 31 Amendment 57
- - - - - - - - - - - , - , . . ~ ~ __ .. ,_ . _ . , .
. ST4!L AE. aca6 F. AGE :u, OF 40 TABLE 7.3-3 INSTRUMENT OPERATING CONDITIONS FOR ISOLATION FUNCTIONS No. of No. of Channels No. Functional Unit Channels To Trip
- 1. Containment Isolation Phase A .
(See Figure 7.2-8)
- a. Safety Injection See item 1 (a through e) of Table 7.3-2
- b. Manual 2 l'
- 2. Steamline Isolation (See Figure 7.2-8) 43
- a. High steam 12 (3,'steamline) 2/3 in any pressure negative rate steamline (enabled by L cc::iv low cupo< del '
34" d" P'*TS 1deun Protection SI Block - see Figure 7.2-9)
~ '
- b. Low compensated steamline 12 (3/steamline) 2/3 in any pressure ** steamline
- c. "
Leu-leu ::=pencated T eold 1 (3/100F) 2/3 in On7 1**P (interlecked vith P-15) c . fr. Manual
- 2 1 A. p'. Containment Pressure HI-2 3 2
- 3. Feedvater Line Isolation (See Figures 7.2-8 and 7.2 14)
- b. Safety Injection See item 1 (a through e) of Table 7.3 2
- . Lev ::=peneeted 12 (3/leep) */1 1" eny ic"p h
-(interlecked with T-15)
- In addition to the two system level steamline isolation switches, each steam loop is provided with switches to effect steamline isolation in that loop.
-*+ "crxissible typess if reectec ceelout yi...uco 1. lo.. Lhou F 11 (nominally 1^00 psig). .
7.3 33 Acendment 57 O *,W
. ATTACHMENT 3 STP FSAR . ST HL AE ha8e PAGE 23,.6f.. y0 j TABl.E 7.3-3 (Continued)
INSTRUMENT OPERATING s.ONDITIONS FOR ISOIATION FUNCTIONS i l
1 No. of 1 l
Channels No. Functional Unit Channels To Trip
- d. !~u prissj leep S:: Tipr:: 7.2-5 l
,3 .o -. s._ , __2 ,e a
- ,,i ,
- :_ avg l 5'13.'5"'I5'Er.dP-15 C, g. 4 (1 per loop) 2 Low T,yg (interlocked 53 with P 4)
- 4. Containment Isolation Phase B
- a. Containment Spray See item 2 (a and b) of Table 7.3-2
- 5. Containment Ventilation Isolation
- a. Safety Injection See item 1 (a through e) of Table 7.3-2 43
- b. Manual Containment Spray See item 2a of Actuation Table 7.3 2
- c. Manual Containment Isolation See item lb of Phase A this table
- d. High radiation signal
- 2 1 1
l
- High radiation signal is derived from 1 of 3 radiation monitors:
two Class 1E RCB Purge Isolation monitors and one Containtnent atmosphere monitor (non Class 1E). High radiation signal is redundantly provided to logic trains R and S. These radiation monitors are discussed in Section 11.5.
7.3 34 Amendment 53
. ST.HL AE Mae TABLE 7.3 4 PAGE2s OF M r i INTER 1hCKS FOR ENGINEERED SAP 1!Ti FEATURES ACTUATION SYSTEM Function Designation Input Performed 1
P4 Reactor tripped Presence of P 4 signal l activates turbine trip
- l Presence of P 4 signal closes '
main W valves on T,yg below setpoint Presence of P-4 signal prevents opening of main W valves which are closed by SI or high SG water level. ee-
- iv: :::1d:= pr:tection Presence of P 4 signal allows manual reset / block of automatic safety injection 1 signal l43 Absence of P-4 signal defeats the manual reset / block for safety injection f.
-Inps; ;e T-15 P 11 2/3 pressurizer pressure Presence of P 11 allows below setpoint manual block of SI on low 43 pressurizer pressure Presence of P-11 allows manual block of exce**1v+
I w u~n.asab i T : ldr . prete :ic7, funeetens.
sico hae pn m ecSI):(see Figure 7.2-9)
Absence of F 11 opens all accumulator discharge isolation valves.
- P 4 is an input to P 16. The F-16 signal trips the turbine. The P-16 signal is present when either the P 4 signal is present (indicating the reactor trip circuit breaker (s) are open) or the reactor trip train oriented logic signal is present.
l r
7.3-35 Amendment 53 7
- e
~
- ST. HL AE MM TABLE 7.3 4 (Continu:d) PAGE M OF40 INTERIDCKS FOR ENGINEERED SAFETY FEATURES ACTUATION SYSTEM Function Descripe. ion Input Performed P 12 below low low Presence of P 12 blocks 43 2/4 T setpofM steam dump except for cooldown condenser dump valves Presence of P-12 allows l43 manual bypass of stamm dump block for the cooldown valves only P-14 2/4 SG water level above Presence of P-14 closes all setpoint on any SG W control and bypass valves Pxasence of P-14 trips all main W pumps and closes all W isolation and bypass 43 valves Presence of P-14 actuates turbine trip P-15 P.eseter--trip (P-6) Or 2/t Prese m f P 15 :,11;;; M neutrer fle: (;;r:r-reng:) cet"-tien end = ir ete r line b: lee eetpeint imeleti+n-en-lee-lee T g 57
- nd 11cus ".T iscletic.. ...d turbine trip from-low c;; pen::ted T 0 8 eold
-flow-7.3-36 Amend =ent 57 Ma g (4 M
ATTACHMENT 3 STP FSAR ST HL AE N6 ,
PAGE 30 OF Yo ..
10.3.2.4 Power Operated Relief Valves (PORVs). The PORVs, one for each MS line, are required for removal of heat from the Nuclear Steam Supply System l (NSSS) during periods when the condenser is not available as a heat sink or 39 when the MSIVs are closed. The valves are ASME Class 2 and are supplied with Class 1E power.
The design mass flowrate of each PORV (one per SG, four total) is 68,000 lb/hr l5456 2 saturated steam at 100 psia. The wide open condition does not exceed 1.05 x 10' lb/hr at 1,300 psia. The valve design is in accordance with ASME B&PV 56 Code,Section III, Su'Ssection NC and has a design pressure and temperature of 1,285 psig and 600'F, respectively. The operation of these valves is not required to protect against SG overpressure or to provide the necessary safety relief capacity.
The PORVs, which are equipped with electric hydraulic actuators and controlled through the Qualified Display Processing System (QDPS) discussed in Section 53 7.5.6, are set to open below the lowest SG safety valve setting to preclude the operation of safety valves during transients when the condenser is unavailable as a heat sink. The opening of the valves is automatic, based 39 upon steam line pressure. A remote pressure control station is provided for each PORV to permit setpoint adjustments of each valve over the entire pressure range up to the safety valve setting. Remote manual operation is provided for a safe shutdown at the control room and at the auxiliary shutdown panel. Local control is provided in case of complete loss of automatic control. Direct position indication is provided, with input also to the QDPS 39 co=puter.
10.3.2.5 Main Steam Isolation Valves. The MSIVs are located in each MS i line downstream of the PORV, and as close to the RCB as practical (see Figure 10.3 1). A small bypass valve at each isolation valve is provided for startup purposes.
Steam is conducted from each SG in a separate line through the RCB, each line being anchored at the Containment wall. Main steam line anchorage is covered in Section 3.8.1 and Containment isolation in Section 6.2.4 The lines have the capability to absorb thermal expansion. Testing of the MSIVs is discussed 39 in Section 14.2.3.1.
The MSIVs and bypass isolation valves are provided with remote manual con- 39 trols. Automatic signals which close the MSIVs and the small bypass isolation valves are the HI-2 Containment pressure, low steamline pressurecleu-Ice
-T and, the high negative steamline pressure rate signals. The MSIVs use 57 piik4on actuators and the bypass valves use diaphragm actuators. The valves are held open by instrument air pressure on the bottom of the actuator.
Spring pressure on the actuator acts as the driving force for valve closure.
The MSIV logic is shown in Figure 7.3 18. To assure safety function actua-tion, redundant actuation solenoid vent valves, powered from separate Class 1E 39 power sources, open to vent air from the bottom of the piston actuator through Q32.11 two separate vent lines. Remote valve position indications are provided in Q32.34 the control room. An annunciator located in the control room alarms on MSIV ,
closure.
10.3 4 Amendment 57
TABLE 15.0-6 PLANT SYSTEMS AND EQUIPMENT AVAttA8tE FOR TRANS!fMT AND ACCIDENT CONDITIONS 4 Incident Reactor Trip Functions ESF Actuation Functions Other Equipment ESF Eesipannt i
15.1 increase in Heat Removat by the l 60 secondary systems Feedwater system Power range high flux, Nigh-high steam Feedwater isolation Anatillery Feedwater System l60 g malfunctions causing overtemperature 4T, generator water levet- valves, steam generator an increase in feed- overpower aT, manual pro h ed feedwater safety valves, steam 60 water flow isolation and turbine generator PORVs trip
. Excessive increase in Power range high flux, -
Pressurizer safety valves, secondary steam flow overtaaperature af, pressurizer PORVs, main 60 y 7 overpower aT, marraat steam isolation valves. *ts O *=s I in j Inadvertent opening of Low pressurizer tr!r - . _W Feedwater footation valves, Auxiliary Feedwater Syster, g a steam generator re- pressure, safety injec- .L Low pressurizer main steam isolation valves safety Injection System tief or safety valve tion signal, overten- presbe,lowcompen-perature AT, overpower sated steam line pres-aT, power range high sure, manuel fIum, manual I 43l18 Steam system piping Low pressurizer L =- ! = . . , _ u:M Feedwater isolation valves, Auxiliary Feedwater System, l3 failure pressure, safety injec- T tow pressurizer main steam isolation valves safety Injection System tion signal, power ee4A t) '.n >
pressure, low compen- > -e range high flux, sated steam line pres- o I >4
- O overpower AT, manual sure, MI-1 and HI-2 m t/
'b r ,
Containment pressure, l manual O
't N
p bZ D .4 e H cs DD@ C El tu 3
rs CD O
. ST.HL AE AW PAGE 3L OF 40 TLo oui o( 6 <ee
- b. Ex::::1:: ::;1dern protectien '2/3 low compensated steamline 55 pressure,fr:: 2ny SC ::
'sanals 2/3 ' u leu :::p:n::ted T-celd in any loop)l 2
- 2. Reactor trip will occur from einn. a) high neutron flux, b) overpower AT, c) two out of four low 8 s
- er pressure signals, or d) receipt of 43 57 an SI signal.
- 3. Redundant isolation of the maxn feedwater lines: sust Qned high feed.
water flow would cause additional cooldown. Therefore W:n exce :ive 57
- 4. Closure of the fast acting main steam isolation valves (MSIVs) (designed to close in less than 5 seconds) from either a:
co m p uote)
- a. Lown steamline pressure :: 10 1;; T-celd signal (two out of three in any loop) above the P-11 setpoint, or 57
- b. High negative steamline pressure rate signal (two out of three in any loop) below the P-11 setpoint.
A block diagram summarizing various protection sequences for safety actions 2 required to mitigate the consequences of this event is provided in Figure Q211.6 15.0 9.
Systems and equipment which are available to mitigate the effects of the acci-dent are also discussed in Section 15.0.8 and listed in Table 15.0 6.
15.1.4.2 Analysis of Effects and Consequences.
Method of Analysis The following analyses of a secondary system steam release are performed for this section:
- 1. A full plant digital computer simulation using the LOFTRAN (Ref.15.1-1) code to determine RCS temperature and pressure during cooldown, and the effect of safety injection;
- 2. Analyses to determine that there is no consequential damage to the core or reactor coolant system. 2 3 The following conditions are assumed to exist at the time of a secondary sys-tem steam release:
2
- 1. End of-life shutdown margin at no-load, equilibrium xenon conditions, and with the most reactive RCCA stuck in its fully withdrawn position. Oper-ation of RCCA banks during core burnup is restricted in such a way that addition of positive reactivity in a secondary system steam release acci-dent will not lead to a more adverse condition than the case analyzed.
On ia /L tio n to the norml cootrel c4chu .Mel ell 4:e i k e m.,,, Ice.L/ccvalves (oil m a re=c for f rig 15.1-9 Amendment 57
- ATTACHMENT 3 ST HL AE up "
temperature coefficient, the cooldown results in an insertion of positive reactivity. If the most reactive RCCA is assumed stuck in its fully withdrawn position after reactor trip, there is an increased possibility that the core will become critical and return to power. The core is ultimately shut down by ,3 the boric acid delivered by the Safety Injection System, i The analysis of a main steam line rupture is performed to demonstrate that.the following criterion is satisfied:
Assuming a stuck RCCA, with or without offsite power, and assuming a 3 single failure in the SIS, the core remains in place and intact.
Although DNB and possible clad perforation following a steam pipe rupture are not necessarily unacceptable, the analysis, in fact, shows that no DNB occurs for any rupture assuming the most reactive assembly stuck in its fully with-drawn position.
A major steam line rupture is classified as an ANS Condition IV event (see section 15.0.1).
The major rupture of a steam line is the most limiting cooldown transient and, thus, is analyzed at zero power with no decay heat. Decay heat would retard the cooldown thereby reducing the return to power. A detailed analysis of this transient with the most limiting break size, a double-ended rupture, is presented here.
The following functions provide the necessary protection for a steam line rupture:
(-
- 1. Safety injection actuation from either: 2
- a. Two out of four low pressurizer pressure signals, or 57
- 6. Ex:: rive ecelde m protectie ( out of three low compensated steamline pressure frca any SC cr twc cue-of-three icw low-eespenseted-T- cid in any loop 7.
C. Lo o.t o f tLeee Ngh 1 c.o ntam e,ent pre rsv,e sepa l t .
- 2. The overpower reactor trips (neutron flux and AT) and the reactor trip 44 occurring in conjunction with receipt of the SI signal.
- 3. Redundant isolation of the main feedwater lines: sustained high feed-water flow would cause additional cooldown. Therefore, in addition to
@ ehe safety injection signal, an-exees:1 : eccide m protection-signal will 43 rapidly close all feedwater control valves and feedwater isolation valves, as well as trip the main feedwater pumps.
4 Closure of the fast acting Main Steam Isolation Valves (MSIVs) (designed to close in less than 5 seconds) from either a
- a. High 2 containment pressure signa 1 9 (ho ed d ihree) 57 Coe t e saie I
- b. Low steamline pressure er le -le.: T- cid signal (two out of three V
in any loop) above the P 11 setpoint, or
@ +Ac nor mal on trel achc Auk ull close f(e m,,
(c, b fer Alves fo %,q a r ea c tor tr f, a. '
15.1 12 Amendment 57 1
l
' ' ~
STP FSAR ATTACHMENT 3 ST4il AE 2626 P
Question 211.32 f{3 OF 90 Certain automatic safety injection signals are blocked to preclude unwanted actuation of these systems during normal shutdown and startup operations.
Describe the alarms available to alert the operator to a failure in the primary or secondary system during this phase of operation and the time frame available to mitigate the consequences of such an accident. Justify the time frame available.
Response
During the shutdown the following operator actions pertain to the isolation of Emergency Core Cooling System (ECCS) equipment and would effect a Loss of Coolant Accident (LOCA). (Start up is not addressed since shutdown is more limiting due to the high core decay heat generation).
(i) Below the P-11 setpoint, the operator is instructed to manually block the automatic safety injection (SI) actuation circuit. This action rii, c,, p e o ,J,1 disarms the SI signals from the pressurizer pressure transmitters and
,gg , e ::::::ive :::1d:= pr: tecti:n 1:gie. The containment high g ,,,y .g,,,, pressure signal remains armed and will actuate SI if the setpoint is exceeded. Manual SI actuation is also available. The circuit will automatically unblock if the Reactor Coolant System (RCS) pressure should increase above the P ll setpoint.
(ii) At 1000 psig or below, the operator closes and locks out the SI accumulator discharge isolation valves.
(iii) At approximately 350 psig and 350'F, the operator aligns the Residual 55 Heat Removal System (RHRS) for cooldown. )
The significance of these actions on the mitigation of a LOCA are:
(i) Below tha P 11 setpoint SI will be initiated by the HI-l containment pressure signal. For small LOCAs (<2 in, diameter break) manual SI initiation may be required. The results for this event are analyzed in the safety significance portion of this question.
(ii) Between 1000 psig and 350 psig, a portion of the ECCS may be actuated automatically on containment high pressure signal or manually by the operator. The equipment that can be energized are the Low Head and l High Head SI pumps. Three trains of SI are required to be operational in Modes 1, 2, and 3. In Mode 4 one train of SI plus one "
additional LHSI pump are required to be operational. The other MHSI pumps are locked out per Technical Specification requirements.
However, at least one of the locked out HHSI pumps can be restored to operable status within 30 minutes. The operator would reinstitute power in the main control room to the accumulator isolation valves.
(iii) Below 350 psig, the system is in the RRRS cooling mode. The operator would manually initiate SI and isolate the RHR system from the ROS.
Q&R 6.3 23 Amendment 55 b e * * . g Y
-,-_v- . ., - ,_
_ - _ . - + _ _ . - - -
ATTACHMENT 3 STP FSAR ST HL AE A626
. fAGE55 OF W Response (Continued)
- 3. The HHSI pumps and the acumulators are locked out when the break occurs. ,
However, operator action can be taken to unlock one of the HHSI pumps. l (This is a conservative asumption for South Texas because three trains of HHSI and MSI pumps are required operable in Mode 3.)
- 4. One MSI pump is available (a second pump is assumed to fail) from either manual SI actuation or automatic actuation by the contahaent HI-1 signal. l For breaks of 6 and 8 inches the calculations show that one low head SI pump turned on manually by the operator 10 minutes following the break gives sufficient flow to prevent the top of the core from being uncovered. For the 8 inch, break SI flow was initiated at 10 minutes plus 25 seconds (delay time between operator manually actuating safety injection and the beginning of flow). For the 6 inch break, although the SI signal was generated by the operator at 10 minutes, SI flow did not start until approximately 18 minutes following the break when the RCS pressure dropped below the GSI pump shutoff head of 700 f t.
The RCS pressure transient for a 4 inch break is so slow that the operator, in addition to manually activating the MSI pump at 10 minutes, is conservatively assumed to unlock one of the HHSI pumps at 30 minutes following the break.
With one GSI pump and one HHSI pump available at these times, the core remains covered.
Another facet which must be considered is the availability of alarms which would alert the operator to manually initiate SI for very small lhCAs (1 2 55 inch diameter) that do not pressurize the containment to containment HI l set pressure (5.5 psig) (which would automatically initiate safety injection).
The Class 1E indication available to the operator includes the narrow range water level sensors. In addition,.the alarms availaole would include the Reactor Coolant Pressure Boundary (RCPB) leak detection system alarms. Break flow from a 1 inch break is on the order of 500 gpm and a 2 inch break would have a flow of approximately 2000 gpm. Thus, these breaks would be expected to set off the RCPB leak detection alarms cuch sooner than an hour after the break occurs. Based on the Inadequate Core Cooling Study (WCAP 9753) for full operation, a 1 inch break would exhibit an extremely long transient prior to core uncovery from the initiation of break flow (approximately 2.5 hrs for a 4 loop plant). Other small break analyses with SI for similar 4 loop plants were reviewed and similar results were found. An even longer transient would be expected for a small break during shutdown. Thus, the operator would have ample time to diagnose the situation, initiate SI and prevent core uncovery.
For a 2 in. break, the RCPB leak detection alarms would sound within 30 minutes of initiation of the break. From McCuire low power test analyses (5 ,
percent power)4 for a 2 in, break no core uncovery occurs prior to 1,67 hours7.75463e-4 days <br />0.0186 hours <br />1.107804e-4 weeks <br />2.54935e-5 months <br />. l Thus, the operator again has ample time to initiate safety injection manually.
When RCS pressure is below the P 11 setpoint and SI is blocked on low pressurizer pressure and excessive-cooWewn-protection, a steamline rupture loa rowJek sfeml,a pr ue Q6R 6.3 23e Anendment 55
-e - __m. m e -
. ST.HL.AE. 2626
~PAGE 36 0F Yo Response (Continued) would be less severe from a core integrity standpoint than the steamline ruptures at hot zero power presented in Chapter 15. Technical Specification require shutdown margins such that the return to-power transient would be less 55 severe than the cases presented in Chapter 15.
The engineered safeguards functions desired during a steamline rupture are actuationofSIg^ndsteamlineisolation. When the low pressurizer pressure signals and the ner::1:: ::;1d;rr,pr;tecticr. signals are blocked, SI and l55 steamline isolation may be automatically initiated by the following signals:
- 1. HI-l Negative Steamline Pressure Rate Signal This signal is unblocked automatically when theG9exceerive cebldeer pretection signals are blocked.
29 (Actuates steamline isolation.)
- 2. Containment Pressure Signal 57 (Actuates SI (HI-1) & steamline isolation (HI-2).)
SI and steamline isolation may also be actuated manually by the operator.
During a steamline break, steamline pressure, pressurizer pressure, pressurizer level and steam generator water level will tend to decrease and l55 steam flow will increase. These partmeters are all displayed in the control room. The operator's attention may be drawn to them by the following alarms;
- a. Low pressurizer level deviation alarm
- b. Low pressurizer level alarm
- c. Steam flow /feedwater flow mismatch alarm
- d. Low steam generator level deviation alarm I
- e. Low steam generator level alarm i
b low co- pe n c b d c few lin . e presauce s, r, d&R6.323f Amendment 57
~
h , g* , ,
STP FSAR ATTACHMENT 3 ST.HL AE 2626 Response (Cont'd) PAGE 37 0F Yo System (RCS) for xenon decay. Replenishment of the AFST may be from one of two nonsafety systems (Demineralized Water or Secondary Make up Storage Tank) or from the Essential Cooling Pond (ECP) while boric acid is added to the RCS 55 via the safety grade Chemical and Volume Control System (CVCS).
For the following events involving breaks in the RCS or secondary system pip.
ing, additional requirements for operator action have been identified.
Main Steam Line Break (See Table Q211.52-1) 4g Following the hypothetical main steam line break (MSL3) incident, a main staam isolation signal will be generated, causing the main steam isolation valves to close within ten seconds of the break. If the break is downstream of the iso-lation valves, all of which subsequently close, the break will be isolated.
If the break is upstream of the isolation valves, or if one valve fails to close, three SGs will be isolated while one will continue to blow down. Only in the case in which one SG continues to blow down is operator action required.
In the analysis of a steam line break (Section 15.1.5), it has been assumed that the faulted steam generator is unisolable. The low steam line pressure signal automatically closes the MSIVs and initiates a SI signal which results 57 in KFIV closure in all loops, as well as SI flow. The analysis proceeds for 600 seconds (10 minutes). All applicable safety criteria are met for this $$
event without assuming operator action. It is implicitly assumed, however, that within a reasonable time period (30 minutes), the operator will begin corrective measures to orderly shutdown the plant, in accordance with the plant's Emergency Operating Procedures as discussed below, or m 4 tri s.nl m bid5 l
I The only source of water to the SG will be AFW since the sikna ver44r axcessive eeeldeer protectier :Egnal will cause main feedwater isolation to occur. Following main steamline isolation, steam pressure in the steam line for the unisolated SG will continue to fall rapidly, while pressure stabilizes in the remaining three main steam lines. The difference in steam line pressures, available seconds after steam line isolation, will provide the information necessary to identify the affected SG at which time the operator I will isolate the AFW to the SG. Manual controls for the AFW pumps and the APW regulating and isolation valves are provided in the control room. The required equipment for detecting the affected SG and isolating its AFV is safety grade. The operator's failure to isolate the APW to the SG or the 41 isolation of the wrong generator will result in the SG continuing to blowdown.
The second required operator action is to manually control repressurization of ,
the RCS. Fol'owing the automatic SI actuation and after the affected SG has '
been isolated, continued operation of the Safety Injection System (SIS) will increase the RCS pressure to the maximum SI pump shutoff head (-1600 psia).
(The RCS can be repressurized without isolating the affected SG; however the process will take longer). Above 1600 psia the operator must restore normal pressure and level control systems. The operator may terminate SI based on criteria established in the Emergency Operating Procedures. If the operator l55 fails to stop the SI pumps after the pressurizer level and pressure return to i .
Q6R 15.0 18a Amendment 57
-.,__,..,--,,n,_. - , . _ , ~ . , _ , - - - . , ,
ST.HL.AE na Question 440.01N In response to our previous question (211.85) regarding deletion of the emer-gency boration system (EBS) from the STP desi5 n. you have indicated that EBS deletion was justifiable, since, in the event of a main steam line break, the DNB design bases are met and the radiation release.s are within the limits set forth in 10 CP1t Part 100. We have reviewed the system aspects of the revised steam line break analysis in FSAR Section 15.1.5. Based on our review we have determined that the following additional information is required. If this in-formation has been included elsewhere in your FSAR, appropriate references in Section 15.1.5 will suffice. 1.ikewise, if the information has been provided in the form of other documentation (e.g., Westinghouse topical reports), ref-erence to such documentation (please be specific) is appropriate,
- a. Clarification of the methcdology for calculating reactivity feedback, including the effect of nonuniform core inlet temperatures from the reac-tor coolant loops; justification of the conservatism in the methodology with regard to the peak power obtained,
- b. Clarification of the methodology used in calculating DNBR and verification that the power distributions used for DNBR calculations reflect the effect of nonuniform core inlet temperatures from the reactor coolant loops.
- c. With respect to ESF actuatien functions for an SLB, describe and justify the differences between the protection functions at the STP and the actuation functions in NUREC 0452, "Standard Technical Specifications for Westinghouse PWR's". Describe the ' excessive cooldown protection" function, which, in accordance with the FSAR, provides safety injection in the event of an St.B. Identify the actuation set points.
Response
a) See revised Section 15.1.5.
b) See revised Section 15.1.5.
c) With respect to Engineered Safety Features (ESF) actuation functions for a SLB, the differences between the protection functions at STP and NUREG 0452 are as follows:
- 1. For safety injection (SI) actuation, the function [of J.a-)' low compensated steamline pgssure and (b) leu-lev :: penested T cel+ 44 e+ incident eith P-15 hn 'been added. The functions of M high differential pressure between steam lines and Ird)" high steam flow in two steam pressure lines have coincident been deleted. with low-low T""E or low steam line I
l
- l Vol. 3 Q&R 15.1 1N Amendment 53 j
~
. ST.HL AE. 2W PAGE 310F '/o
- 2. For steamline isolation, the functions of (a) high steam flow coincident with low steamline pressure of low low T , and (b) high high (HI 2) containment pressure have been del $TEd. In addition, the functions of (c) high steam pressure rate and (d) SI signal have been added.
This SI signal change was described to the NRC in December 1974 in Section 7.3 of RESAR 41. During CP review, the description of the ESF actuation functions referenced the RESAR-41 design and was subsequently approved by the NRC via the issuance of the STP SER, NUREG-75/075, August 1975.
S e ete rline imelatien eign=le $' sue beer changed circ- irrurce of the Sr" SER "SSt." 51 :h: :d :::=line iseleti n :n th: fell;uing : nditi ns:
-2. Cent:in::nt pre::ur: "I-2
- 5. Cent insent pr;;;ur:__,__,m "I 44
... . . . . u- , n. - -...
un 1-u --
- d. 1.cu ::: pen,yEted: ::: = lin press w.
^: indi
. ted above, iter: b, e, :nd d ::: ::d : generst: : SI sign:1, ul-th-leu pre: cur 4::: pre: ure th: enly :ther ::nditien generet4ng an SI sign:1.
Sin : th: ::tpeinte fer u;_1 3ng uy_7 cent 31 g ent pre..nv. vor, ek. ...., <e uas daciA*A .te dal.t, ek. uT-2 rignal and replace it by *ka SI sigul. n e-only actual difference is the additier f leu pre: cur 4ser-pr+seure. In M ate <nn, en preuta. *k. .nem.. tic imelatier ef the ste r liner uhen SI is bleeked (beleu P-11), the $'Q,h :te= pre::ure- ret: eign:1 i: used, The "eveessive eeelde" protectien" i: :: prie:d of the funct4en:-d::cribed in -
1 (a) =d (b) :,bev . f. :::: detailed de: r4pt4en felice:. Se exce:cive: (~
ooolde m-pretection logic i: che .- ir Figur: '.2-9 ^ctuatier of SI can-be
-cau: d by either (1) tue of *kree lev-leo cenpensa*.A T-enia eign=le in any Deacter Ceelant Syste'= leep coincident eith P-15 er (?) tue of three low l55
- p...... . ...,4,.
pr...n.. ,igne1. i_ my e e ,1, s te mune . g.u u caused by tue cf four4nst-rument+-indieet4eg : neut-r+n - flux below 10 pe rcent peuer er e reaeter trip.
~9 The actuation setpoints are given established in the Technical Specifications. 53 The etc,,sive cool b n p r e t e e.t, e , Ioye Aas bee., delete 4 hem f(e fo.dl i Texas Project p ro tection c ys tem. Tkes cys te% 6 <s + oles c ri bed in R ESA rt -4 1, a t not taba cr<hf for m tke fo d Ten s ?<oject plan f syect fee hce6 etny b us analyti*.
Vol. 3 Q&R 15.1-2N Amendment 55
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ATTACHMENT 3
. ST.HL.AE MA6 I Response PAGE vo 0F 40 . . . . l The Main Steamline Isolation Valve (MSIV) closure logic has been modified to l57 be consistent with that of other Westinghouse plants. The MSIV closure on manual and high steam pressure rate si5 nals will be maintained. The MSIV closure on a safety injection signal has been modified to MSIV closure only on l57 a HI-2 containment pressure signal and,-free the er.ce--ive coelda- pretection legie, on a low steamline pressure signal..:rd er e 1:e-1:e T eold 213 2*1' l i
l I
l
. l l
1 l
l Vol. 3 Q&R 15.6-17N Amendment 57 i
l 44
- O ( YM =, y yqe
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