ML20154M528
| ML20154M528 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 10/15/1998 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20154M512 | List: |
| References | |
| NUDOCS 9810200323 | |
| Download: ML20154M528 (850) | |
Text
{{#Wiki_filter:. .- . . . . - - . . _ - - . - RCS Pressure. Temperature. and Flow DNB Limits 3.4.1 l. p p e 13,4-' REACTOR C00LANTz SYSTEM {(RCS) A' F s' , 13.411c RCS' Pressure. Temperature' and Flow Departure from Nucleate Boiling. (DNB) Limitsi l l L. LCO: 3.4.11 ' RCS-DNB parameters for pressurizer pressure. RCS average-l ' temperature ' and RCS total flow rate shall be within the limitsispecified below:
- - a. Pressurizer pressure a 2219 psig
c; - b. .RCSaveragetemperaturel(T,f)s591.2*F: and
- 5
- c. ~ RCS total- flow rate a 371.400 gpm.
c e o.
- 1. .
NOTE Al -: Pressurizer pressure;1imit does _not apply during:
+O a '. THERMAL POWER ramp > 5% RTP per mintde; or
- b. . THERMAL POWER step > 10% RTP'.
Ih. 9 k v; { APPLICABILITY:: MODE 1.
'ACTIONSI ,
CONDITION' REQUIRED ACTION COMPLETION TIME
;A. 'One or more RCS-DNB A.1' Restore RCS DNB 2 hours
. parameters not within parameter (s).to
' l. imits . within limit.
B.(RequiredAction'and B.1 Be in MODE 2. 6 hours
- associated Completion
; Time not met. ~
4 L li L
- .1 l 7 3:
p%f LBYROND- UNITS--1 &L2 3.4.1 - 1 8/21/98 Revision K 9810200323 981015 PDR ADOCK 05000454 P PDR w v
. - . . . . . - -.~ .._.--- ..- - . .- . . . - . , .... . . . - . . . _ . ~ .- -
3 # # RCS; Pressure.. Temperature,andFlowDNBLimits 3;4.-1
, +.
SURVEILL'ANCE REQUIREMENTS D]#
\
SURVEILLANCE = FREQUENCY?
+ g lSR13,4.1.1)
~
Verify; pressurizer pressure-is 2.2219 psig. - 12-hours- ,
-1
-I 4
7 ' SRM3.4.1.26 Verify RCS' average' temperature (T,fis' 12 hours
, s 591.2*F. -
l SR f 3.' 4 ; 1'.'3 ' Verify' RCS1 total flow ' rate is - .12 hours-
-a 371,400-gpm.:
mg
- g1 SR:J3.4.1;4' ..
. NOTE _
o -Not required to.be performed until 7' days i =after 2.90% RTP.
.;. .x.: : ----
1
.. Q' . ;, ' Verify by precision heat balance that RCS 18' months. l
( 4lt *, r l '-
;>,J total. flow rate-is- = 371,400 gpm.- 1
, i 1
1 Y l
- j j [
. BYRON ?-UNITS'1 &'2 3,4.1 - 2 8/21/98 Revision K l
j g +, g. g ..1- g -ys* T W ut 9 yr- e N 1 '"
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.RCS~ Minimum Temperature for Criticality.
3.4.2 jl (
/~i 3 M REACTOR' COOLANT. SYSTEM!(RCS). '
l
'J '
\
3.4.2- RCS Minimum Temperatu'e r for Criticality 1
.u l
f LC0v3 4.2= . . Each RCS. loop _ average' temperature (T.,,) shall.be = 550*F. o r , ;.:T lLAPPLICABILITY: : MODE 1. . i
- MODE 2:with k,,,= 1 0.
JACTI0N'S. i 1 3, ,. iCONDITION : REQUIRED ACTION- COMPLETION TIMEi
~
A. T;,:in one or more RCS A .1. Be in. MODE 2 with ' 30-minutes loop's .: not; withini k,,, < 1.0. limit. ,D Y]
' SURVEILLANCE REQUIREMENTS J SURVEILLANCE _ FREQUENCY
?:SR -3.4.2;1f Verify RCS T.,,iin each loop = 550 F. 12 hours f-I f
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1 j A [y y; L fBYRON UNITS 11 & 2: 3.4.2 - 1 8/21/98 Revision K X i .. I 'i 'i. 4 y w e .. 4 - . , -.v. - _ , , . .
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.RCS P/T. Limits 3.4'3
' f i J3.4' REACTOR. COOLANT SYSTEM (RCS)-
.' U 3.4'.3 RCS Pressure-and. Temperature (P/T)' Limits LC0 -3;4.3 .RCS pressure. RCS temperature..and.RCS heatup and cooldown rates'shall be maintained within the limits specified in the '
PTLR. 1 i
. 1 APPLICABILITY: At'all times. 1
; .' ACTIONS-
~ CONDITION REQUIRED ACTION COMPLETION TIME.
' A. ---NOTE .
A.1 . Restore parameter (s) 30 minutes l Required Action A,2 to'within limits. shall be completed .
-whenever this AND l Condition is entered.
A.2 Determine RCS is 72 hours acceptable for l
;[ . Requirements of LC0 continued operation. 1
'N not inet' in MODE 1. 2. !
- 3. or-4. j l
B; Required Action and~ B.1. Be in MODE 3. 6 hours
. associated Completion
' Time of Condition A AND
. not met.
B2 Be in MODE 5. 36 hours. 1 (continued). H l l
- '< A l
' A,f BYRON.- UNITS l'&'2 3.4.3 - 1 -8/21/98 Revision A l L l I
I RCS P/T Limits-3.4.3 l j, ACTIONS -(continued) ! CONDITION REQUIRED ACTION COMPLETION TIME L
)
C, NOTE C.1 Initiate action to Immediately Required Action C.2 restore parameter (s) < , , shall be completed to within limits, l whenever this Condition.is entered. AND !
)
, C.2 Determine RCS is . Prior to Requirements.of LC0 acceptable.for entering MODE 4 .. not met any tiale other continued operation. l than in MODE 1.-2, 3. . or 4. .) l
. l
~ SURVEILLANCE REQUIREMENTS SURVEILLANCE, FREQUENCY ]
y SR 3.4.3.1 --- NOTE- - Only required to be performed during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing Verify RCS pressure. RCS temperature, and 30 minutes RCS heatup and cooldown rates are within the limits specified in the PTLR. 1 l [
. BYRONL- UNITS.1 & 2 3.4.3 - 2 8/21/98 Revision A l
I
l RCS Loops-MODES 1 and 2 3.4.4 i /~') 3.41 REACTOR COOLANT SYSTEM'(RCS) LJ L 3.4.4 RCS Loops-MODES =1 and 2 - l LC0 3.4.4- Four RCS loops shall be OPERABLE'and in operation. l. L APPLICABILITY: . MODES 1 and 2. l
. ACTIONS CONDITION REQUIRED ACTION- COMPLETION TIME i
h .' L Requirements of A.1 Be in MODE 3. 6 hours A. LCO not met,
,- s . SURVEILLANCE REQUIREMENTS, i A-- SURVEILLANCE FREQUENCY :
SR 3.4.4.1' Verify each RCS loop is in operation. 12 hours > s :. I' l 7g- .
- (
(..
. BYRON - UNITS 1 & 2 3.4.4 - 1 8/21/98 Revision A i
h RCS Loops-MODE 3 - l 3.4.5 l hf~} Ku 13.4 : REACTOR COOLANT SYSTEM'(RCS) ', 3.4.5 RCS Loops-MODE 3:
- l. iLC0~ 3.4.5 'Tw6'RCS loops _shall be OPERABLE.~and either:
- a. -Two OPERABLE RCS'. loops shall be in operation when the i .
Rod Control System .is capable.of r6d withdrawal; .or. , 1
- b. One OPERABLE RCS loop shall be in operation when the Rod' l Control System is not. capable of rod withdrawal. '
1'
---. -NOTE All' reactor coolant pumps may.be removed from operation for..'
p :s 1 hour per 8 hour; period provided:
'No'o)erations are permitted that would cause' reduction
.- a-.
7 .of tie RCS' boron zoncentration: and 1
.Hta b. Core; outlet temperature-is maintained 2 10*F below W 1
- saturationLtemperature.
o .%p - __ __
.n
'l APPLICABILITY: . MODE 31 A _.)V.
. ACTIONS l
. CONDITION! REQUIRED ACTION COMPLETION TIME
. A ione required RCS loo r A'.1 Place.the Rod' Control- 1 hour notiin operation witi System in a condition 1 Rod: Control System incapable of rod ,
capable of rod- withdrawal withdrawal. (continued) l yj 'V
# l BYRON ~.UNITSL1&2- 3.4.5 - 1 10/8/98 Revision K u _ _ . . _ _- . _ -- . .
RCS Loops-MODE 3 ; 3.4.5
}
ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B. No required RCS loop B.1 Suspend all Immediately in operation with Rod operations involving Control System not a reduction of RCS capable of rod boron concentration. withdrawal. AND B.2 Initiate action to Immediately restore one RCS loop to operation. C. Two required RCS loo)s C.1 Initiate action to Immediately not in operation wit 1 place the Rod Control Rod Control System System in a condition capable of rod incapable of rod withdrawal. withdrawal. E AND
,x L') ! Required Action and associated Completion C.2 Suspend all operations involving Immediately Time of Condition A a reduction of RCS not met. boron concentration.
AND C.3 Initiate action to Immediately restore RCS loop (s) to operation. D. One required RCS loop D.1 Restore required RCS 72 hoprs inoperable. loop to OPERABLE status. (continued) L l ,,
~ BYRON - UNITS 1 & 2 3.4. 5 - 2 8/21/98 Revision A
RCS Loops -MODE 3 3.4.d , - Q-
\
ACTIONS (continuedt 3 CONDITION. REQUIRED ACTION COMPLETION TIME
~
E. Required Action and . E .1- Be in MODE 4. 12 hours b associated Completion Time of Condition 0- ; not met.
'F. two required RCS loops F,1- Initiate action to Immediately
- = inoperable. place._the-Rod Control l System in a condition :
- , incapable of rod withdrawal.
AND j J F.2 Suspend all Immediately operations involving a reduction of RCS j boron concentration. < p)' (' AND 1 Initiate action to F.3 Immediately i restore one RCS loop ! to OPERABLE status. ! c ) SURVEILLANCE REQUIREMENTS
-SURVEILLANCE FREQUENCY
- SR 3.4.5.1- . Verify each required RCS loop is in 12 hours o peration.
. (continued)
(+
.t i l.
BYRON - UNITS 1.&'21 3.4.5 - 3 8/21/98 Revision A d se,u e u -- g --
J 1
. RCS Loops -MODE : 3 3.4.5 l' '
c L : /Y'
-SURVEILLANCE RE0VIREMENTS (continued)
SURVEILLANCE-FRE0VENCY L @j? . l- ?T ; SR.3.l4.5.2 Verify steam generator secondary side 12 hours j] narrow range water level is.218% for each g- required RCS loop, p
' SR -3;4.5.3- Verify correct. breaker alignment and 7 days indicated power are available~to each required pump thatfis not.in operation.
3 ,. : f-~ - a\ ,_ - I ,m o - g L . l-
- - j-
.8YRON-UNITS)1&2 3.4. 5 - 4 8/24/98 Revision K i:
o [: l
- .. .. = .
1 RCS' Loops-MODE 4 3.4.6 fi. '3.4 REACTOR COOLANT S[ STEM (RCS) s^ ji 3.4.'6 RCS Loops-MODE 4.-
-LC0. 3;4.6. Two loops consisting of any combination of RCS loops and Residual Heat Removal (RHR). loops shall be OPERABLE. and one
. LOPERABLE loop .shall be in operation.
NOTES .
- 1. Al.1-Reactor Coolant Pumps (RCPs) and RHR pumps may be removed from operation for s I hour per 8 hour p'eriod provided:
- '" a. No' operations are permitted that would cause l
P j reduction of the RCS boron concentration: and M b. Core outlet temperature is maintained a 10 F below
% 2.
1 saturation temperature.
.No RCP shall be started with any BCS cold leg 1 temperature s 350*F unless the secondary side water temaerature of each Steam Generator-(SG) is <.50*F above
-eac1 of-the RCS cold leg temperatures.
j% U,. . 1 APPLICABILITY: MODE 4. tACTIONS e 1 CONDITION' REQUIRED ACTION COMPLETION TIME LA. .No required. loop in- 'A.1 Suspend all _ Immediately operations involving
. operation.
a reduction in RCS-boron concentration. AND A.2 Initiate action to Immediately restore one loop to operation. (continued) i-ro ,n.) ,
%s. >
l , ' BYRON --UNITS 1 & 2 3.4.6 - 1 10/8/98 Revision K
.e
, . . , .- . . . - ;.-~~.. .. .- .. .- . . - ..
1 L L , 1RCS Loops-MODE 4 ! 3.4,6 L . 1 YN '
- ACTIONS -(contiinued)
! CONDITION REQUIRED ACTION- COMPLETION TIME' u
i -q .
- fl
- LBl. 'Orie required sloopf B.1~
. Initiate action'to.- Imediately'- 1 inoperable.
A restore a second loop ! to'OPERABLEistatus. [, AND B.2! -- NOTE Only. required if RHR
- t. loop is 0PERABLE.
1 Be in MODE 5. 24 hours i
'CL Two' required loops C.1 Suspend'all . Imediately . >
. operations involving
~
- inoperable. I
-a-reduction of RCS i boron concentration. j 1
,[ 7: d' i Initiate action to- Imediately C.2 restore one loop:to OPERABLE status. ,
g ~ SURVEILLANCE REQUIREMENTS
-SURVEILLANCE FREQUENCY SR;l3 4._6.1 Verify required RHR or RCS loop is in 12 hours gg
. operation.
g
- w. '
s
. eiL . . . .
4 .:SR: ..43.4.6 2' . Verify SG secondary side narrow range water - 12 hours
.p.. flevelfis = 18% for each required RCS loop.
g . . I (continued)
- f. }
'g
_k. BYRON : . UNITS 11 &. 2 : 3.4.6 - 2 8/21/98 Revision K 1
< }NJ
..(..,
,, ,[ .RCS - Loops - MODE--4 H'< . <
.3.4~.6
? --
-! SURVEILLANCE ~REOUIREMEN S (continuedi
- o' ' ~ ;
SURVEILLANCE FREQUENCY'
.1 t
- SR t- 3.4. 6,3 ': . Verify correctibreaker al< gnment i and 7 days ~!
1 indicated power are available-to,each !
.- 1 required pumn that is notfin operation. !
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.py ' X 4.
!BRON-lVNITSli&'2' 3.4.6 - 3 . 8/21/98 Revision K
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. -. . _ .- ... . . - _..__._.~.m -.
?RCS Loops--MODE 5. Loops Filled 3.4.7 o
hy 3.4! REACTOR COOLANT SYSTEM-(RCS) v '
~
..3.4,7. RCS Loops-MODE 5. Loops Filled l
cv 1 o . o LC0H3;4.7 . L0ne Residual Heat Removal (RHR) loop shall be OPERABLE and. J N in operation; and either: x . .
% a. One additional _RHR11oop shall be OPERABLE: or-
"' 'b. The secondary side water level of at least two Steam-Y Generators (SGs) shall be = 18%.
.J [h NOTES
, g 1. 'The RHR pump may be' removed from-' operation for s-1_ hour
'per 8 hour period provided:
-a. No operations are permitted that would cause' x . reduction of the RCS boron concentration; and y
MM .b. Core outlet. temperature is maintained a'10 F below
$$S '
. saturation _ temperature.
- 2. .One required RHR loop may be inoperable for s 2 hours-for surveillance testing provided that the other RHR
. loop _is OPERABLE and -in operation.
'f
\j 3. No reactor coolant _ pump shall:be' started with any RCS cold. leg temperature's 350*F unless the secondary side water temperature of each SG is < 50 F above each of the RCS cold leg temperatures. ,
.4. All RHR loops may.be removed from operation during planned heatup to MODE 4 when at least one RCS loop is
_in operation. l APPLICABILITY: LMODE 5 with RCS loops filled.
,+
i: k,,. . p Q '
~
-BIRON -UNITS 1:& 2 3.4.7 - 1 10/8/98 Revision K 1 .
RCS . Loops-MODE 5. Loops Filled 1
'.4.7 I n i
- e. .
1 ACTIONS r/~)" ,K~L H i" , CONDITION! ' _ REQUIRED ACTION- COMPLETION TIME. I m I
-A; :-No required RHR. loop'. A'.1 '.Susp~ e nd ' all Imediately in cperation r.r , . operations. involving L a. s ,
fa reduction in RCS l boron concentration' . (, AND
- l. [ 'A . 2 - Initiate. action,to. ImediatelyL restore one RHR loop-q' to operation.
r.
.O-
;x_ o r +
B. 'One ' required RHRr icop B .1 ' Initiate action.to Immediately inoperable. restore required RHR loop to OPERABLE
,.;l 3 status.
j 1
..4
' 7(^$$f v 1.EOnejor.'bothfrequired C
, SG secondary side : ,
C.1: : Initiate action to restore required SG Immediately N #'ia.,- J
-Twater; level (s)inot' secondary side water-i EN .
.within limits. level (s) to within
- qq 't ' ~ ~ -
limits. (continued) 4 I
- (,-
s I/ t p f I, Ig-r- . . '. Jj ;g l Q + N' * 'BYRONE-UNITS!11&'2l 3.4.7 - 2 -10/5/98 Revision K _._/ __',c' _I h :Y-l
, , .' - n . . .- .. - - . ,
-. A
. 4
' RCS Loops-MODE 5. Loops Filled =
g 3. 4 : 7-
'p i .ACTIONSi (contir.ued) '
.a h M -
' CONDITION .- -REQUIRED ACTION: . COMPLETION TIME' T D. nTwo required.RHR-loops .D .1 Suspend a'il.- Immediately..
. inoperable, operations' involving a reduction of RCS
-@L boron concentration.
. Required RHR: loop. AND-
'ino3erable and one or.
bota. required ~SG- D.2.1 Initiate. action to.. Immediately
- secondary side water restore one RHR loop
- level (s) not within' to'0PERABLE status,
- e. ;-limits.
2 D.2.2' Initiate: action to Immediately restore required SG secondary side water
-level (s) to within limits.
SURVEILLANCE REQUI'REMENTS-
' SURVEILLANCE FREQUENCY g SR .3.4.7;1_
' Verify required RHR loop is in operation.
L,. 12 hours
+
p, M,'( af.
, N. .SR: 3.4.7.2.
Verify SG secondary side narrow range water 12 hours level is E 18% in required SGs'. . J- n. ISR'3.47.3' ' Verify correct breaker alignment and 7 days
. indicated power are available to each required RHR, pump that is not in operation. '
O L BYRON.- UNITS 11 & 2. 3.4. 7 - 3 10/6/98 Revision K , 1 R
i RCS Loops-MODE 5. ' Loops Not Filled 3.4'8-a ,. j3.4LREACTOR:COOLANTSYSTEM(RCS) $f%[A_ . . . l' - 3.4' 8 : LRCS ? Loops -MODE ,5
. Loopsi Not- Filled : i
.LC0?3.4'.8 .Two Residual-Heat Removal (RHR) loops.shall be OPERABLE and one 0PERABLE'RHR loop'shall be.in operation- . ;
NOTES
- 1. All/ RHR pumps may be. removed from operation.for s 1: hour provided: '
y a. No operations are permitted that would cause a ,
.o: reduction of the RCS boron concentration:
- N% b, lThe core.o'utlet temperature is maintained a~10 F.
L([ below saturation temperaturei:and c, No draining operations are permitted that sould 1 further-reduce the RCS water volume.' j c 2. ;0ne RHR loop may be inoperable for s 2 hours fors
- surveillance . testing provided that the,other. RHR' loop is OPERABLE and in operation.
Mj >-u
- < APPOCABILITY
- MODE 5 with RCS loops not filled.
9 i ACTIO'NS
- ' CONDITION' REQUIRED ACTION- COMPLETION TIME I. A. [No required RHR loop' A.1 ' Suspend all- Immediately
-in. operation, operations involving a reduction in RCS boron concentration.
AND
'A.2 Initiate action to Immediately
-restore one RHR . loop
*' to operation'.
b r-I (continued) L A Q,) . - L : BYRON:- UNITS'1;&f2 34.8-1 10/8/98 Revision K - l:
,'a .
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- - - . - .- - -. ~
L- RCS Loops-MODE 5. Loops Not Filled . 3.4 8 i -ACTIONS' (continued)
- . CONDITION ' REQUIRED ACTION COMPLETION TIME r
1 1 ! ' B. :One required RHR loop: B1 Initiate action to Immediately
. inoperable restore RHR loop to OPERABLE status.
I 1 I
.C, Two required RHR' loops C,1 -Suspend all'- .Immediately. !
operations involving inoperable,- reduction in RCS boron concentration. AND C.2 Initiate action ~to Immediately restore one RHR loop l to OPERABLE status, i
- p.
SURVEILLANCE REQUIREMENTS 1 1 SURVEILLANCE FREQUENCY l
.SR -3,4.8.1- Verify required RHR loop is in operation. 12 hours SR 3.4.8,2. Verify correct breaker alignment and 7 days indicated power are available to each required RHR pump that is not in operation.
- (,)
-BYRON - UNITS 1 & 2 3.4.8 - 2 8/21/98 Revision A
m , y Pressurizer
'e -3.4.9
'3.4 ' REACTOR COOLANT SYSTEM (RCS)
.3.4.9:. Pressurizer.
4.y: t.
, L.LC0. 3.4!9 The_ pressurizer shall;be OPERABLE with:
g a. -Pressurizer water level!s 92%: and'
- o. .
; f. . .b. Two groups-of pressurizer heaters.-OPERABLE-with the
+ capacity of each group = 150.kW and capable of being 4- powered from redundant-Engineered Safety Features (ESF)
- u power supplied buses.
N. APPLICABILITY: MODES'1. 2. and 3. o 4- ACTIONS ' CONDITION RE0VIRED' ACTION COMPLETION TIME A~ -Pressurizer water
. A.1- .Be in MODE 3, 6 hours a1 level: not within-
.' . limit; AND' A.2 Fully insert all 6 hours rods.
AND A.3 Place Rod Control 6 hours System in a condition incapable of rod withdrawal. 6H0 A.4 Be in MODE 4. 12 hours B. -One or more' required B.1 Restore required 72 hours
' groups:of. pressurizer groups of pressurizer
' heaters inoperable.- heaters to OPERABLE status.
.. g- (continued)
, LJ i BYRON - UNITSL1 &'2- 3.4.9 - 1 8/21/98 Revision K
Pressurizer 3.4.9 - ' PY IACTIONS' (continued)
^s'/
~ CONDITION- COMPLETION TIME REQUIRED ACTION C ._ Required ~ Action and C.1 Be in MODE 3. 6 hours associated Completion Time of. Condition B. AND h not met. .
C. 2 - Be in MODE 4. 12 hours I'
' SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
'l, SR 3.4.9.1 Verify pressurizer water level is s 92%. 12 hours l
SR 3.4.9.2 Verify-capacity of each required group of- 18 months ,;y( )s pressurizer heaters'is a 150 kW. l- SR. 3.4.9.3 Verify required pressurizer heaters are 18 months-l - capable of being powered from an ESF power supply. I L BYRON-UNITS 11&2 3.4.9 - 2 8/21/98 Revision A
Pressurizer Safety Valves 3.4.10
~
lf','y- ,g - 3.4 ~ REACTOR COOLANT SYSTEM (RCS) l 3.4.10' Pressurizer-Safety Valves . -)
- LCO: 3.4.10 Three-pressurizer safety valves shall be OPERABLE with lift
.. settings a 2460 psig and 5 2510 psig.
.L i
o - NOTE a -The lift settings are not required to be within the
@ LCO limits during MODE 3 for the purpose of setting the 4 l -' pressurizer safety valves under ambient (hot) conditions.
- This exception is allowed for 54 hours following entry into El MODE 3 provided a preliminary cold setting was made prior to G' heatup.
-APPLICABILITY: MODES 1, 2, and 3.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME (N One' pressurizer safety
-A. A.'l- Restore valve to ~15 minutes valve inoperable. OPERABLE status.
. B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time not' met. AND.
08 B.2 Be in MODE 4. 12 hours Two or more
. pressurizer safety valves inoperable.
i i p. w BYRON.- UNITS 1 & 2 3.4.10 - 1 8/21/98 Revision K
s, . _ . . _. . . _ _ . _ _ _ . _ - - _ . _ _ . _ . . _ . . . . . _ _ _ . _ . . .~.
- y. >
Pressurizer Safety Valves _ 3.4.10 ' 1 SURVEILLANCE REQUIREMENTS f: 'r$ '
; SURVEILLANCE-
. FREQUENCY ~
7, e . f? ;
-y"?
e 1
' SR L: 3.4i10/1' ; Verify each pressurizer safety _ valve is'- In accordance-Q:- ' OPERABLE in accordance with the' Inservice with the M c. Testing' Program. :Following testing, lift - Inservice, p
- q.
1 settings'shall.be within~'- 1%, Testing Program tw ! i i
.(
l i
)
i
/
4 h '
! ~
, BYR'NO UNITS 1 &.21 3.4.10 - 2 8/21/98 Revision K )
i
g-Pressurizer PORVs'
- A 3.4.11
- i. a,,-
i 1 1 . ThNp "3;4'. REACTOR COOLANT SYSTEM (RCS) M L 3.'4.112 Pressurizer $ Power Operated ~ Relief Valves'(PORVs) LL'C033.4.11 4 Each PORVIand' associated block valve'shall be OPERABLE. [ APPLICABILITY: MODES 1~.2. and 3. l' ACTIONS
. -. . NOTES --
- 1.- ? Separate ~ Condition entry .is allowed -for each PORV and each block valve.
- 2. LCO 3.0.4 isinot applicable.
CONDITION REQUIRED-ACTION. COMPLETION TIME 1 lA. One-or more PORVs- . A.I. Close'and maintain 1 hour. s M- ino)erable.and capable - aower to associated
- 4'- Jof Deing manually and lock valve.
~
a i)!j T- automatically cycled. cu. (continued)
- b. . ,. -
BYRON - UNITS 1.&'2~ 3.4.11 - 1 10/5/98 Revision K
'Et-
-----_._ m_-_.. C""LN
.. m -. , ,- s _ _ ~- . _ . - _ . _ , . . . . . .
Q ' :, .
- ( . ,
l l Pressurizer PORVs'- I
'1 13.4.11-bs - - -
~.
l
$,N < iACT'IONSii:(contintied)?
N)T ' LCONDITIONi , REQUIRED l ACTION: LCOMPLETION TIME-1 1 i p r :81[One1PORViinoperable: B.12 ~ Clos'e~ associated 1 hour Land not' capable of , block valve.
~
J ,
- being zmanually cycled
~
7 '
.@ j
~
;gg~ .;
p .
~
B;2: ' NOTE.~ . ---
.i
'7 W)? One PORV(inoperable'- Not required i f - 1 W Land'not:capa31e of- , associated PORV- 1 k -:bei:.g automatically;- remains capable-of" l f
~
cyceed.- being manually.
.H Lcycled. q al
.. k.
.y l
i w . Remove-power'from 1 hour. > associated block 1 valve.' -i g .- .; E tu
?, /
^
, 7C. One block-valve- '
C.1- Place-associated PORV- 1-hour i
. inoperable. .in-manual control.
M a C 2' Restore block valve -72 hours ; to OPERABLE: status. o 9 .D. Required Action and D.1' Be in MODE 3.- 6 hours
~ * . .T associated Completion.
.M Time:of Condition A. @-
,B,Jor C not met.
I.H D.2 - Be in MODE 4. 12 hours
- (continued).
y ,y
- . a 4
o)xOM M UNITS 1 & 2 . 3.4.11 - 2 8/21/98 Revision K
,w r ,
~,{:
, , . :+.. . . - . , . - , - . ,, - ,. .
L
? Pressurizer PORVs W . :3. 4.11. .
.c. . }- "E '
} ACTIONS ~(continued)-
iCONDITION- REQUIRED ACTION- COMPLETION TIME L q_ -_ n coV .Ei ::Two PORVs inoperable- E111: Be in MODE 3. 6 hours 1.3.0) andinotcapableof
'l Q.
being' manually? cycled; fNQ ;
'n h~ .
?QB- E.2? :Be :in. MODE 4 12-hours' .
LTwo PORV's inoperable-dy ;3'.
- and no.tfcapable of; M:
n
~
, , ibeing; icycled. automatically' r llilF;lTwoblockEvalves.: F .1. Resto're.one block 2 hours ninoperable.
valve to OPERABLE-status _s=
'o; T
Required Action andL G.I. LBe in MODE 3-- . 6 hours- > /NLl W G.associated. Completion: ()g ,4 l Time of; Condition F;-
- not met.
A_NQ ' 4 g[ G.2 Be in MODE'4. 12 hours i 1 LSURVEILLANCE REQUIREMENTSL 0-T: SURVEILLANCE. FREQUENCY-i w-3- lSR;;3.4.11.1: . NOTE .
,nP
-Not' required to be met with block valve
- N l closed in 'accordance with the Required
$ ll : Action of Condition B.
,5 " ,. ~ c s . Perform a. complete cycle' of- each block' 92 days -
- valve.
p (lf-g (continued)
- Q',
S ' BYRON;-- UNITS 1-& 23 3.4. 11 - 3 10/13/98 Revision K g .
.i_
- . . . .. . . . - .. . . .~ -. . - - , - . - . . _.- -.
1 Pressurizer PORVs 3.4.11 ; N ? SURVEILLANCE REQUIREMENTS '(continued).
?' f
\
. SURVEILLANCE FREQUENCY j SR:;3.4 ell.2
, . ' NOTE
'_ Only~ required to be. performed.in MODES 1-and.2
~
Perform a' complete cycle'of each PORV. 18 months :
- SRT-3.4.11.3 Perform a. complete cycle of each' solenoid- 18 months
- air controlivalve and' check' valve on the
..y air accumulators in PORV control systems.
1 i
'4; p q -
) "y SR 3.4.11.4. Perform CHANNEL CALIBRATION of PORV '18 months-
-t actuation instrumentation, v.-
y~t
;\ )[
b L; {:'. , ~ {3; ,H BYRON - UNITS 1 & 2 3.4.11 - 4 10/5/98 Revision K L. t. p 4 < - , - , - -..
~LTOP System ~
3.4.12 [I I34 REACTOR ' COOLANT SYSTEM '(R' CST - 1 J g ; W 3.4:12: Low Temperature.0verpressure Protection-(LTOP) System H 0-4:
~-
, ].l LCO .3.4;12.' -
~ Ari LTOP System shall be OPERABLE with: 1 b ta. A maximum of;one charging' pump -(centrifugal) capable of -
injecting-into the RCS,-
'b. LNo. Safety Injection'(SI) pumps capable of injecting into the RCS '
- c. Each SI accumulator. isolated whose pressure is greater than or equal to the maximum RCS pressure for the i existing RCS cold leg temperature allowed by the P/T j limit curves-provided'in the PTLR. and !
.d. One of the following pressure relief capabilities:
H > 1. Two Power Operated Relief Valves (PORVs) with: lift settings within the' limits specified in the PTLR.
- 2. ~Two Residual Heat Removal-(RHR) suction relief valves with setpoints s 450 psig.
- (A U
E
-3. ^ 0ne-PORV with a lift setting within the limits .
specified in the PTLR and one RHR' suction-relief '
. valve with a.setpoint's 450 psig. or
~4. .The RCS depressurized and an RCS vent of )
= 2.0 square inches.
_5 y
- ' ~
NOTE i . Operation in MODE 4 with alt SI pumps and charging pumas capable of injecting into the RCS is allowed when all TCS yl-w cold legs exceed 330*F. .
-MODES 4 and 5.
'kAPPLICABILITY:
zu . 1 MODE 6 when the reactor vessel head is on.
" M.n:: I j
1 L 3 L N" l
- )3 y
fBYRON'-UNITS-1&2. '3.4.12 - 1 10/5/98 Revision K
3
~
ls p, 4 4LTOP System-L3 4.12 1 ?? :.. 7 Y.;: ACTIONS-f . N 7 --A s)/ t p.L--- . . NOTE . 1 5
. p$l f LCOl3;0.411s:not? applicable tolthe. RCS. pressure 1 relief. capab
~
CONDITION REQUIRED' ACTION' COMPLETION TIME' $p , g Ai(Twocharging' pumps =A 1 -
. --NOTE l W :(centrifugal) capable. ;Two charging pumps, , s: d:-
foflnjectingintothe RCSi may be capable of-injecting into the-
'..'. RCS during pump swap-3* LOR; ; operation for EW .5 15 minutes.
L$ {Onechargingpump:
?(positivel.. i ,. .
sdisplacement) capable: Initiate. action to Immediately j
- of 1njecting:into'the1
~ '
verify.a maximum of ;
, RCS. one charging pump
- (centrifugal).1s- ']
' capable of> injecting-
'into:the RCS. ,
3
' i t-B. L0ne'or more SI pumps B .1 ' ; Initiate' action to: Immediately capable Lof- injecting verify no SI . pumps-linto the~RCS. are capable of.s ; )
injecting'into the 1 RCS. i
. TC: .'An accumulator not'-
C.1 Isolate affected I hour iisolated when-the' accumulator. ,
; accumulator 3ressure is greater <tlan or equal to the' maximum RCS pressure fore e pe ed in 1
.ithe PTLR. ' 1 I
-(continued) k4 L ;
"f gf-
- =. , .
.-BYRON. . UNITS'1-&L2 l 3.4.12 - 2 8/21/98 Revision K im):
?" .
. m.
. - - . . - =.
LTOP System 3.4.12 A ACTIONS (continued) V CONDITION. REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Depressurize affected 12 hours associated Completion. accumulator to less Time of Condition C than the maximum RCS
.not met pressure for existing cold leg temperature allowed in the PTLR.
E. One required RCS E.1 Restore reautred RCS 7 days relief vc've relief valve to inoperable-in MODE 4. OPERABLE status. F. One required RCS. F.1 Restore required RCS 24 hours relief valve relief valve to inoperable in MODE 5 OPERABLE status. or MODE 6 when the reactor vessel head is on. !(] G. Two required RCS- G.1 Depressurize RCS and 8 hours relief valves establish RCS vent of
' inoperable. 2 2.0 square inches.
2 Required Action and associated Completion L Time of Condition D. E. or F not met. 2 LTOP System inoperable for any reason other than Condition A. B. .. C. D.'E. or F. s"
.G BYRON' UNITS'1 & 2 3.4.12 - 3 8/21/98 Revision A i
' I q; ,
LLTOP System .! y:# - 3.4.12. .l JSORVEIL'LANCEREQUIREMENTS-C"' ')
^
' SURVEILLANCE ' FREQUENCY:- i
- SRE314;12.1':- Verify.no SI pumpLis: capable'of:inje'cting. 12l hours into the'RCS.
4:
W q -
1 ., ; l .
~
p '~ LSRn3.4.12.2 Verify a maximum of 'one charging pump :12' hours i
' 4' Tc /
-(centrifugal) .is capable. of: injecting into
< ~ the RCS.
~
.j l
1 SR 13.4;12.-3
~~
, NOTE-- . .
j
'Only required..to be met for~ accumulator-1."
- whose' pressure is. greater than or: equal to. 9 1the. maximum RCS pressure for the existing j u , ,, RCS~ cold leg temperature allowed by the P/T L , limit curves 1provided in.the'PTLR. ,
p Verify;each-accumulator is isolated. 'x)y@ 12 hours
.# 1 "jm
\
y
- w. .
.SRn314.12.4 ; Verify required.RCS ' vent a 2.0_ square -12 hours for j il' '
;4 inches open, unlocked open ,
vent. valve (s)- y n 31 days for locked open vent valve (s) i i y[- SRl3.4.12.5 l Verify 'RHR suction valves are open for each 72 hours j g required RHR. suction relief valve. i
+
,;4
) L o
[LSRL3;4.12'6 . . Verify PORV block valve is open for each 72 hours required PORVi l M (continued)- Qf J BYRON - UNITS :1. & 2
- 3.4.12 - 4 10/5/98 Revision K
~ :e y -
g- ~ e - ,
s' j'm s.; ,
'l LTOP System
%3 ,' 3,4,12 m . . . 7NJe .8
? SURVEILLANCE RE0VIREMENTS- (continued) -
1
)
M' '
^
- (SURVEILLANCE ~ ~ FREQUENCY l
.. 1 r ~ , sSR.)3L4.12!7 .- -
. NOTE------ .
l Not recuired to be' performed until 12 hours I
/after.= cecreasing RCS.' cold. leg temperature j s
ito;s 350 F . Perform'a-COT on'each required PORV, 31 days. excluding actuation. 1
, -SR :3.4.12.8L -Perform CHANNELLCALIBRATION for each 18 months required.PORV actuation channel.
"'.u
\"
.i
, l ,
-' L
.p[
i
- 'l
'i (
. ' ,)
- -y e f
(" r . . . . , . e Q ?^ / BYRON' UNITS l'& 2 3.4.12 - 5 10/5/98 Revision K
..] y
- P V I ) y e, . ~ 4 % e -
.v.4
m , 1 RCS Operational LEAKAG'E- i
.3.4.13-e p ",d- ?3.4jREhCTORCOOLANTSYSTEM(RCS)
' Q^ f 3.4.13' RCS Operational LEAKAGE-
- -RCS operational LEAKAGE'shall-be limited to:
LC0!-3.4.13
-ac No pressure boundary LEAKAGE:
- b. 1 gpm unidentified LEAKAGE: l L 'c. .10 gpm identified LE KAGE: 1
'd , 600 gallons per day total primary to secondary LEAKAGE through all Steam Generators (SGs): and L 'e. 150 gallons per day primary to secondary LEAXAGE through
- any one SG, i
~APPLICABILITYF . MODES:1. 2; 3.'and 4.
L' .
~ ACTIONS.
l CONDITION REQUIRED' ACTION 1 COMPLETION TIME '%f)
~
A. IRCS LEAKAGE not within A.1 . Reduce LEAKAGE to. .4 hours
' limits'for reasons within limits.
other than pressure boundary LEAKAGE. iB, Required Action and B.1. Be in MODE 3. 6 hours associated Completion
. Time of Condition A AND
.not met.
B.2 Be in MODE 5, 36 hours
.2 Pressure boundary LEAKAGE exists; L
(N
. BYRON-UNITS l1&2' 3.4,13 - 1 8/21/98 Revision A t
%_ v <
- aM..
l .t , 1 % RCS Operational: LEAKAGE
- 4. '
3.4;13 ; 1 \ (
- ; . ,...t . - / , . . . . , . .
Y'S SURVEILLANCE REQUIREMENTS ~ 4 3,J t-
," n g. , -
! SURVEILLANCE FREQUENCYi !
4:-~.
~
g.: iSR1":; 3.4'.13.1:- - _ l NOTE--- ' q
.n ?Not required:to'be performed untiljl2 hours ;
x- T , :after, establishment of steady state ' sfs operation, f [M[,. .
. %U , ' .
.. . . l
' 4
/ Verify RCS-operational LEAKAGE is within-. 72 hours 1
' limits by.>erformance.of RCS water-
>: , i
-inventory Jalance. .
-i c[.y. ' '
- p V:"
?SR d 4.13.2: :Verifyisteam generat'or . tub' lintegrity is-in
~
e In'accordance
, W ,
.accordance with the' Steam Generator Tube- with:the Steam Surveillance Program. ' '
Generator. Tube. Surveillance--
. Program
.)
1 > b L i I ,- i ; 4 n;_ R, I < t -. ; T {l 6 py 3:. , [ .',' M t r;-. m :p; e 1
.f.
y; ;n , a(/ o.
- '. 1 } ,
S BYRON-:UNITSL1&2- 3 4.13 - 2 9/18/98 Revision K
.g
') ,.
',; ;. ,.y.
4 s
. . , - . -. ~ . - _ ..~. .
e
'o i R' ,av
_,w w.. RCS PIV-Leakage
' 1 3.4;14 d,
.L3.4L REACTORTC00LANT SYSTEM (RCS)-
ffj ? L .3 - .. 1RCS'
~ ' '
. .SPressure' j @ 1Isolation 4.14 Valve..(PIV) Leakage w
-s' .
j l '; c wY yy x ! a .$lyLC003.4.14; Leakage from each RCS'PIV shall.be within' limits. z e;
* ; APPLICABILITY: 'MODESL1 b 3..an'd 4 h
, :I
< i=' J 3
iACTIONS
--a ,
A ---- --- NOTES - - -
$? ;1.HSeparat'e Condition entry is . allowed .for each flow path, m ,
Enter:aap'licable C'onditions and Required. Actions for_ systems made ~ g[lb2; . inoperaale by.anjinoperable' PIV.
- q j
30NDITIO'NI
- REQUIRED ACTION. COMPLETION. TIME 1
fA.10ne or more flow paths; --NOTE Jb,l. s with leakage from'one- Each valve used to satisfy
?
Ad , or-more RCS PIVstnot- Required Action A,1 and
- within limit.- Required ActionLA'2 must have .
been verified to meet? ' SR.3 4.14.1 and be in the. reactor-coolant 3ressure , boundary. or' the ligh' pressure portion of the system. A.1 Isolate the high . 4 hours pressure portion of the affected system from the low pressure portion by'use of one closed manual,
... .de-energized power operated.
de-activated
. automatic, or check a ,
. valve. #-
-s i ,
; MD
. . _ . . (continued)
A M-n .
.~
; BYRON UN.ITS 1 & 2 3.4.14 - 1 8/21/98 Revision K m,
x
'I g k d I I y-eg -- e+
n' .RCS PIV-Leakage'
-i 3.4'.14 I
^
JACTIONS' f.. M LCONDITION' . REQUIRED, ACTION = COMPLETION TIME I L :A, :(continued)- A.2 : Isolate the high 72 hours
. pressure portion of the affected system from the low pressure
' c$ portion by use of a !
'3'
.second-closed manual. -l
.7 . de-energized power 1 3M operated.
'de-activated
'@ M automatic, or check
- valve.
- j B. Residual Heat Removal B.1 Isolate the affected 4 hours
-(RHR): System suction flow path'by use of
. isolation valve- one.de-energized interlock function _ power operated valve.
N inoperable. x, e .i ! f- ; ,M i.
' C. I Required Action and C.1 Be in MODE 3. 6 hours l associated.. Completion
~
, Time not inet. ' $@
C.2 Be in MODE.5. 36 hours l 6 1 p .
+:
- p 3 .o <
BYRONl ' UNITS-:1 & 2 3.4.14 - 2 10/1/98 Revision K [ ..
- b. - J t
+
c 4 s n - we- -- - V-
,?
~ '
-s,
~,
' ' ' -' ~ ' ~ - ' ~ ' ' -
8 < RCS PIV. Leakage-
- m. <
3.4.14 [ N [ SURVEILLANC'E RE0VIREMENh5 xj"a-
.' SURVEILLANCE
, FRE0VENCY. .
JSRD3.L4i14.1 . .
---NOTES---
- 1. Only required;to be performed in'.
-MODES;1 and-2; i
2 .- .RCS PIVs actuated during the j performance of this Surveillance:are - '
.not. required to be test'ed more than
- once if = a' repetitive testing loopi -j
-cannotibe. avoided. i o.
q
'3. Not-required to be performed for
.RH8701A and B and RH8702A and B on the
> . Frequency required following valve actuation or flow through .the valve.
c
$T
.4 i: : Verify Lleakage from each RCS ~PIV is. In accordance g *i
- equivalent to s 0.5 gpm'per nominal inch of with the 4 ii valve size up.to a maximum of 5 gpm' at an Inservice. l
.g H- RCS pressure = 2215 psig and s 2255 psig. Testing i w) g i Program.'and i
( 18' months v AND Prior to entering MODE'2
.whenever the unit has been ;
in~ MODE 5 for' a 7 days , .i f
. leakage testing has not been performed once within the
*- previous
- j. 9 months
~
T[ , AND y (continued) 4 i .
- l. ,
tyx .Y (BYRON - UNITS 1 & 2 3.4.14 - 3 10/5/98 Revision K
~
' r ;.
f' 1 .
-RCS PIV.. Leakage 3.4.14 ng .
i [)n'
.\
"^ -
K SURVEILLANCE REQUIREMENTS ':(continuedT
,fSURVEILLANCE, . FREQUENCY
; 9.c
.y
($ ,[SRF3114.1..;(continued) Within 24 hours
- S following valve-
. Y ' '* actuation due--
to~ automatic or.
- , .t manual action- .)
=t or flow through- 1 the valve i
. r c . .
'I
' SR 3.4.14.2, . Verify RHR System suction isolation valve 18 months interlock.arevents the valves from being '
, , opened wit 1 a simulated or actual RCS pressure. signal =l360 psig.
1
-l a, '
l r , 1
-1 q,g . . }
A4 ,
}
, l q
l 4 .j l L l :- 1 I" .- l !u n:a. r r .. ..
' BYRON -LUNITS 1 &-2:. 3.4.14 - 4 8/21/98 Revision K l
w 1
,RCS' Leakage Detection Instrumentation-
%, ' 1 3.4.15
, :-i.;) , .
, s7
~y
' 76[J: s . 13.4f15. 3.4L RCS REACTOR COOLANT SYSTEM (RCS) Leakage DetectionLInstrumentation. '
~
L LLC 0"3.4.151 .TheLfollowinq RCS; leakage detection. instrumentation shall be. a OPERABLE - )[.
.a. 10ne containment sump monitori and; 1 v b. ,0ne containment. atmosphere radioactivity monitor--
~
(gaseous or-particulate). 1 l i f'N v! APPLICABILITY:- MODES 1.2l3,'and4. I ACTIONS
~ CONDITION. REQUIRED ACTION COMPLETION TIME
-A. -~ Required containment- .
.. NOTE-- ;
' sump monitor LC0'3.0.4_is.not applicable. !
v;.n;g : inoperable.- d,.[ o A.1 .. ---NOTE Not required to be.- y : performed-until
* ~12 hours'after-t establishment'of JV. - steady state
.l Loperation.
Perform SR 3.4.13.1. Once per 24 hours AND A.? Restore required 30 days containment sump monitor to OPERABLE status. .
,, , (continued) iq>%.
gj [ .. - BYRON'- UNITS 1-&'2 3.4.15 - 1 10/1/98 Revision K L I t -.
n Lt.' 4' i.- i
- RCS Leakage. Detection Instrumentation n
;:6 ' l^ '
3.4.15-n 9' , 0 ACTIONS (continuedF fb . l CONDITION! ~ REQUIRED' ACTION' COMPLETION TIME-
- B):: Required' containment- ----NOTE-
~
atmosphe.re? LCO O 0.4 is'not applicable. ) eicaicactivity: monitor -- j
-inoperable.
1 B .1~. l . Anal ze grab samples Once,per.
, of;t e containment- 24 hours F atmosphere. .
L , g> 'g
, :4: ,
, sn B .1. 2 - . NOTE 3 Not:. required to be I n.', ;M "
performed-until- ,
% .12 hours ~after 1
- establishment of.
.' { L steady state.
1.. operation.- Perform SR'3.4.13.1. Once per ) 24 hours , ng[: a gg e
.B.2. Restore required 30 days ,
icontainment atmosphere-radioactivity monitor
-to' OPERABLE status; g 'i (C; Required Action and C.1 Be in MODE 3. 6 hours
' associated: Completion
> Time not met. AND C.2- Be in MODE 5. 36 hours
- f '. -
. 9ED. lAll regijired monitors D.1 Enter LC0 3.0.3. Immediately.
M a inoperable. IW id F LBYRON -: UNITS l'&-2? 3.4.15 - 2 10/1/98 Revision K 5
- 4
- -.2._- __ _ . . _. . . . - _ - _ . . .
.. - - . . . . . . ~ . . .
P L RCS Leakage Detection Instrumentation 3.4.15 l I- . 'fl- SURVEILLANCE REOUIREMENTS u ' SURVEILLANCE FREQUENCY 7
-SR ~3 4.15.1' Perform CHANNEL CHECK of the required 12 hours containment atmosphere radioactivity monitor.
l I SR 3.4.15.2- Perform COT of the required containment 92 days atmosphere radioactivity monitor. L t l ,SR 3.4.15.3' Perform CHANNEL- CALIBRATION of the required 18 months
. containment sump monitor.
.SR 3.~ 4.15 ; 4 ~ Perform CHANNEL CALIBRATION of the required 18 months
. containment atmosphere radioactivity monitor.
,7d.. i t i L t lb 1 q,, L-
' f) -.
BYRON - UNITS 1 & 2 3.4.15 - 3 8/21/98 Revision A
1(f;
^ RCS Specific' Acti.vity -
3.4.16' L . k(/N.g :3?4# REACTOR:C00LANTSYSTEM-(RCS)? l
'm '1
.. ~ . . , .
3.4.16 L RCS: Specific" Activity ~
. , .)
b TLC 0.33i4:16( EThe specific activity.ofithe reactor coolant shall' be within
-the.following limits:
{ a' .- ? Dose Equivalent.1-131 specific activity s.l.0:vCi/gm:
,$..p and? ~ ~ ' i
' ;' ' Ny L. ,~ 1 b .' ; Gross specific activity s 100/l pCi/gm. g: , . - u
&
- MODES'1 and 2. .
Q l.kAPPLICABILITY:
. MODE 3 with ~RCS average temperature:(Tm)ia 500 F. ;
istt ;
' ACTIONS'
.1
' CONDITION .
. REQUIRED' ACTION- COMPLETION TIME- q
- J4 ;+
?Ai iDOSE EQUIVALENT I-131:
- specific: activity NOTE
.LCO 3 0.4 is not applicable.
- y(
- >51.0'pCi/gm.
[' X A.1L- l Verify ' DOSE EQUIVALENT 61-131; Once per 4 hours
- specific' activity
.within the c racceptable' region of-
. Figure:3.4.16-1. ,
AND - K A.2l Restore DOSE 48 hours-EQUIVALENT .-I-131-specific' activity to Tl < within limit. " i '(continued) O L : i 7 4
- w)l i& ' ' ' '
' LBYRONL. UNITS:1L&2} 3.4.16 - 1 8/21/98 Revision K jy : wi ,
l l.g RCS Specific Activity F 3.4.16 l ! ';f^f [ ACTIONS- (continued)
CONDITION- REQUIRED ACTION . COMPLETION TIME l :
I
'B, : Required Actionand L. '
B.1 Be in MODE.3 with 6 hours L associated Completion- T,,, < 500*F. Time of Condition A i not met. , l 03 DOSE. EQUIVALENT I-131 specific activity-in
;the. l unacceptable region of !
. Figure 3.4.16-1. !
l l l- l l Gross specific. C. C.1 Be in MODE 3 with 6 hours
.-activity not within T,,, < 500 F.
limi t .- ! l 77
.N.} i
-SURVEILLANCE REQUIREMENTS- .
l l SURVEILLANCE' FREQUENCY SR 3.4.16.1 . Verify reactor coolant gross specific 7 days I activity 5 100/E #Ci/gm. (continued) l f-~3
- l%.) ;
BYRON'- UNITS 1 & 2 3.4.16 - 2 8/21/98 Revision A
-y
-f ~
f
^$ d y-pys', '
RCS Specific' Activity. R 3;4.16 [y; LSURVEILLANCE REQUIREMENTS -(continuedt ,u +
--SURVEILLANCE . FREQUENCY--
;SRP3.4;16.2
~ ~ ' '
. . - . - NOTE o
*k g j Onlyirequired to be performed in MODEL1'. y
@'7 -Vehify: reactor coolant DOSE EQUIVALENTJ .14 days
% 11-131 spectfic activity 51.0 pCi/gm.
A AND
'M , .} < jp Betweeni2 and 6 hoursLafter.a-THERMAL l .
POWER change of m'15% RTP within a l' hour period ..
, MSR=-3.4.16.3 .. . . . . NOTE- .-
- 1
^-
Not required toibe performed until.31-days [- <after a minimum of.2 effective full sower days and=20 days of MODE l'~ operation- %' M'f 1 ave elapsed since the reactor was:last
'J
;subcritical for. = 48 hours.
d 3
- Determine'~f from'a reactor coolant-sample-taken-in MODE l after a minimum of 184 days 3M _ 2' effective ful1.>ower days and 20 days of
.: MODE'1 operation lave elapsed since the a% -d reactor was .last~ subcritical- for a 48 hours.
E J i; , i,_ & f.
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<4 i 97 a / h/ v., R-
' BYRON - UNITS 1 & 2 3.4.16 - 3 8/21/98 Revision K m ,
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. Reactor Coolant DOSE EQUIVALENT I-131 Specific Activity )
Limit Versus Percent of RATED THERMAL POWER i u i i I ,O' 1
- e i
!BVRON UNITS 11 &'2 3 A.16 - 4 9/18/98 Revision K
- .. = . .- . - . . - - . . . .. . . - . . . ~ . - . . - - . .
RCS Loop Isolation Valves
'3.4.17 L
i ' f j. 3.4 REACTOR COOLANT SYSTEM (RCS) v 3.4.17 'RCS Loop Isolation Valves LCO 3.4.17 Each RCS hot and cold leg-loop isolation valve shall be open with ' power removed from each isolation valve operator. i APPLICABILITY:- MODES 1, 2. 3. and 4-. ACTIONS
. NOTE i Separate Condition entry.is allowed for each RCS loop isolation valve. ,
l 1 CONDITION REQUIRED ACTION COMPLETION TIME A. Power available to one A.1 Remove power from 30 minutes or more loop isolation loop isolation valve valve operators. operators, d B. NOTE - B.1 Maintain valve (s) Immediately All Required Actions closed. shall. be completed whenever this AND
' Condition is entered.
B.2. Be in MODE 3. 6 hours One or more RCS loop AND isolation valves i closed. B.3 Be-in MODE 5. 36 hours i f jr 3.v . C ' BYRON.- UNITS.1 &'2 3.4.17 - 1 8/21/98 Revision A
-l L RCS Loop Isolation. Valves 3.4.17:
( ' SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY i
- SR 3.4.17.1 Verify each RCS loop isolation valve is 31 days:
open and power is removed from each loop isolation valve operator. ; i I l l l 1 f i l 1 1 4 (- . l BYRON - UNITS 1 & 2 3.4.17 - 2 8/21/98 Revision A
. . ~ . . . - - _ _ ._. _ __ .
u RCS' Loops-Isolated u, 3.4.18 !!/ N S3.4'~ REACTOR: COOLANT. SYSTEM (RCS) u
- 3.4'.-18 'RCS! Loops-Isolated l
~
l iLC013.4T18- Each RCS. isolated loop shall remain isolated with:
- a. The~ hot:and cold leg loop stop isolation' valves closed' l -
if boron concentration of the isolated loop is less than , the. required SDM boron concentration of the unisolated fe? . . .
-portion of the RCS: and- j
$ b. The-cold leg loop stop i. solation valve closed if the
. cold leg temperature of the' isolated loop is > 20 F i
9 at I 'below the highest-cold leg temperature of the unisolated portion of the'RCS. .J APPLNABILITY: MODES 5 and 6. i ACTIONS-
-CONDITION- REQUIRED ACTION COMPLETION TIME g- -
r f .A. ': Isolated. loop hot or A.1 Close hot =and cold Immediately 1 cola' leg isolation leg isolation valves. .!
- valve.open with boron i concentration ..
requirement not met.
. 1 B-. Isolated loop cold leg. B.1 -Close cold leg Immediately
. isolation. valve open. isolation valve.
with temperature frequirement.not met. b 3 I. 1 e f;W N BYRON -' UNITS;1:& 2 3.4.18 - 1 10/5/98 Revision K w
.e . . . . -- . - - . - ... - . . .-. . . .
RCS Loops-Isolated: i t
-3.- 4.'18 'i L ,
-~
- .SURVEILLA'NdE pie 0VIREMENTS
' r~~f
-r 1:
e . . . .
^~'y . ' ,
LSURVEILLANCE! . FREQUENCY p g - T. J4
- SRE3.'4.18.1, VerifyLcold leg .. temperature of. isolated : Within j
A.
- loopLis s' 20 F below the' highest cold leg 30. minutes temperature of the unisolated portion of prior to- . .1 the RCS. . opening the- !
m cold -leg lt isolation valve in the isolated-r , loop
,*: -l
-L i 4 l:- SR4 3.'4018.2-t ~ Verify boron concentration of isolated loop' Within.4. hours i
,W is greater than or equal-to the required prior .to ;
;Li j- SDM boron concentration of the unisolated opening the hot !
'1 -porti_on of the RCS. or cold leg 1 isolation valve l in the isolated .i loop.
k . R l i
-I o
5' i G ( p
- fs.
- t.: g) 1 LBIRON UNITS 1.&.2. 3.4.18 - 2 10/5/98 Revision K
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'RCS. Pressure. Temperature, and Flow DNB Limits B 3.4.1 (N BL3.4- REACTOR COOLANT SYSTEM-(RCS)
B 3.4.1L RCS Pressure. Temperature. and Flow Departure from Nucleate Boiling.
.(DNB) Limits BASES..
BACKGROUND These Bases address requirements for maintaining RCS pressure.~ temperature, and flow rate within limits assumed
'in the safety analyses. The safety analyses (Ref. 1) of-normal operating 1 conditions and anticipated operational l occurrences assume initial conditions within the normal i L . steady state envelope. The limits ] laced on RCS pressure. I temperature, and flow. rate ensure tlat the departure from
_ll nucleate boiling (DNB) will be met for each of the transients-analyzed. . j L The RCS pressure limit is consistent w-ith operation within the nominal operational envelope. Pressurizer pressure ! indications-are averaged to come up with a value for ; comparison to the limit. A lower pressure will cause the 1 reactor core-to approach DNB limits. i s,% The RCS_ coolant average temperature (To limit is consistent with full power operation wit)in the nominal
' M'
! h operational- envelope. -Indications of temperature are averaged to determine a value for comparison to the limit.
. A higher average temperature will cause the core to approach DNB limits.
The RCS flow rate normally remains constant during an operational fuel cycle with all pumps running. The minimum RCS flow limit corresponds to that assumed for DNB analyses. Flow rate indications are averaged to come up with a value for comparison to the limit. A lower RCS flow will cause the core to approach DNB limits. Operation for significant periods of time outside these DNB limits increases the likelihood of a fuel cladding failure in a DNB limited event. l t- . U
~
BYRON - UNITS 1 & 2 B 3.4.1 - 1 10/13/98 Revision K e s'
RCS Pressure. Temperature, and Flow DNB Limits B 3.4.1 [h - X.f
' BASES
-APPLI'ABLE C The requirements of this LC0 represent the initial- ,
-SAFETY ANALYSES. conditions for DNB limited transients analyzed in the plant 1 i;4 safety analyses (Ref. 1). The safety analyses have shown j
~0- that transients initiated from the limits of this LCO will !
' result Tin meeting the DNBR criterion of. = 1.4. This is the ;
W^ acceptance limit for the RCS DNB parameters. Changes to the l Ef N unit that could impact these parameters-must be assessed.for their. impact on the DNB criteria. The transients analyzed 4 ~l for. include loss of' coolant flow events and dropped or stuck
-l l
rod events.. A key assumption for the analysis of these
. events is that the core power distribution is within the -l
-limits of LC0 3.1.6. " Control Bank' Insertion Limits:" 1 LC0 3.2.3. " AXIAL FLUX DIFFERENCE (AFD):" and LCO 3.2.4 a "OUADRANT POWER TILT-RATIO (0PTR)." i o
, j. Safety Analyses assumed a value of 2207 psia (2192.3 psig).
-W This value is bounded by the LC0 value of 2219 psig assuming g[q a measurement accuracy of less than 26.7 psi. ;
Safety Analyses assumed a value of 588.4 F for the vessel average temperature. In addition, the analyses assumed the -l calculated error-(including the 4 F dead band for the rod control-system) for the temperature is 8.74 F (20 random j -) V error of 7.6 F plus the 1.14*F bias error). The value assumed in the non-Revised Thermal Design Procedure (non-RTDP) analyses is -8 F. +9.5 F. For the_RTDP analyses. a value of 7.6 F with a bias of +1.5'F is assumed. Safety Analyses assumed a total RCS flow rate of 358.800 gpm. This_value is bounded by the LC0 value of 371.400 gpm assuming a flow measurement uncertainty of 3.5%. This 3.5% flow measurement uncertainty assumed in the analyses included errors from all sources including fouling
- in the venturi. The use of 3.5% flow error is acceptable if actual uncertainty is unknown. At the time analyses were performed, the flow accuracy was unavailable. Subsequent calculations determined the error to be less than 3.5%.
Any fouling that might bias the flow rate measurement
' greater than 0.1% can be detected by monitoring and trending various plant performance parameters. If detected, either the effect of-the fouling shall be quantified and compensated for in the RCS flow rate measurement or the venturi shall be cleaned to eliminate the fouling.
l l X BYRON - UNITS-1 & 2 B 3.4.1 - 2 10/13/98 Revision K a
'(
RCS Pressure. Temperature, and Flow DNB Limits B 3.4.1 BASES
- APPLICABLE SAFETY ANALYSES (continued)
The RCS DNB parameters satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). , i LCO This LC0 specifies limits on the monitored process variables-pressurizer pressure. RCS average 7 temperature (Tg) and RCS total flow rate-to ensure the
- core operates within the limits assumed in the safety analyses. Operating within these limits will result in 3 meeting the DNB design criterion in the event of a DNB -
f limited transient. A Note has been added to indicate the limit on pressurizer 4 1s not applicable during short term operational transients
.$ such as a THERMAL POWER ramp increase > 5% RTP per minute or a THERMAL POWER step increase > 10% RTP. These conditions represent short term perturbations where actions to control pressure variations might be counterproductive. Also, since they typically represent transients initiated from power levels < 100% RTP. an increased Departure from Nucleate m Boiling Ratio (DNBR) margin exists to offset the temporary pressure variations.
}
Another set of limits on DNB related parameters is provided in SL 2.1.1. " Reactor Core SLs." LC0 3.4.1 re) resents the initial conditions of the safety analysis whic1 are far more restrictive than the Safety Limit (SL). Should a violation of this LCO occur, the operator must check whether or not an SL may have been exceeded. APPLICABILITY In MODE 1. the_ limits on pressurizer pressure. RCS coolant average temperature, and RCS total flow rate must be maintained during steady state operation in order to ensure DNB design criteria will be met in the event of an unplanned loss of forced coolant flow or other DNB limited transient. In all other MODES the power level is low enough that DNB is not a concern. ( BYRON - UNITS 1 & 2 B 3.4.1 - 3 8/22/98 Revision K
RCS Pressure. Temperature. and Flow DNB Limits B 3.4.1
~ '
BASES J
-ACTIONS Al RCS pressure and RCS average temperature are controllable
,= l - and measurable parameters. With one or both of these parameters not within LCO limits, action must be taken-to restore parameter (s).
1 RCS total flow rate is not a controllable parameter and is not expected to vary during steady state operation. If the
; indicated RCS total flow rate is below the LC0 limit. power r must be reduced. as required by Required Action B.l. to 2 restore DNB margin and eliminate the potential for violation
.i of the accident analysis bounds.
kl The 2 hour Completion Time for restoration of the parameters t provides sufficient time to adjust unit parameters, to determine the cause for the off normal condition, and to restore the readings within limits, and is based on plant i operating experience. l U i n s If Required Action A.1 is not met within the associated V ) Completion Time. the unit must be brought to a MODE in which v the LC0 does not apply. To achieve this status. the unit must be brought to at least MODE 2 within 6 hours. In MODE 2.'the reduced power condition eliminates the potential for violation of the accident analysis bounds. The Completion Time of 6 hours is reasonable to reach the required unit conditions in an orderly manner. SURVEILLANCE- SR 3.4.1.1 REQUIREMENTS Since Required Action A.1 allows a Completion Time of 2 hours to restore parameters that are not within limits. the 12 hour Surveillance Frequency for pressurizer pressure is sufficient to ensure the pressure can be restored to a normal operation steady state condition following load changes and other expected transient operations. The 12 hour interval has been shown by operating practice to be sufficient to regularly assess for potential degradation and to verify operation is wi' thin safety analysis assumptions. t_
- l BYRON - UNITS 1 & 2 83.4.1-4 8/22/98 Revision K
. -. ___ _ . _.4._ . _ _ _ . . _ . _ _ _ . _ . _ . _ _ _ . . . _ _ _ _ _ _ . - . _ . _ . _
RCS' Pressure. Temperature. and Flow DNB Limits B 3.4.1 i BASES.
- SURVEILLANCE REQUIREMENTS (continued)
. SR 3.4.1,2 Since' Required Action A.1 allows a Completion: Time of 2 hours to restore-parameters that are not within limits, the 12 hour Surveillance Frequency for RCS average l temperature (T can be' restore'.) d'to is sufficient a normal to ensure operation. the temperature
. steady state
- condition following. load changes and other expected l transient operations. The 12 hour ~ interval has been shown
! by operating practice lto be sufficient to. regularly assess - for potential degradation and.to verify operation is within
~
l: safety analysis assumptions. I ! SR 3L4.1.3 t The 12 hour Surveillance Frequency for RCS total flow rate
- is performed using the installed flow instrumentation. For
L
. Unit 1. the required minimum RCS flow rate is met with a 95% l indicated ' flow rate. For Unit 2. the required minimum RCS l flow rate is met with a 92% indicated flow rate. The
- 12 hour interval has been shown.by operating practice to be i sufficient to regularly assess potential degradation.and to I
. verify ' operation within safety analysis assumptions.
L l e '[ BYRON - UNITS 1 &'2 B 3.4.1 - 5 8/21/98 Revision A y' ,. , -
-94, y ..* c -. m y- y 9 -i py 5 e- enw
RCS Pressure. Temperature, and Flow DNB Limits B 3.4.1-f'e BASES G' ~ L SURVEILLANCE REQUIREMENTS (continued 1 j SR 3.4.1.4 s 3 Measurement of RCS total flow rate by performance of a iy precision calorimetric heat balance once every .18 months. wg allows the installed RCS flow instrumentation to be: . y calibrated and verifies the actual RCS flow rate is greater g .than or equal-to the minimum required RCS flow rate. j ' The Frequency 'of 18 months reflects the importance of
~H- verifying flow after'a refueling outage when the' core has b c been altered, which may have caused an alteration of flow
" resistance.
This SR is modified by a Note that allows entry into MODE 1.
-without having performed the SR and placement of the unit in:the best condition for performing the SR. The Note
-states that the SR is not required to be performed until
.7 days after A 90% RTP. This exception is appropriate since the heat balance' requires the unit to be at a minimum of 90% RTP to'obtain the stated RCS flow accuracies. The
~
Surveillance shall be performed within 7 days after reaching 7 .g - 90% RTP. U , REFERENCES 1, UFSAR. Chapter.15. I hy 'N}
'BYRONL-' UNITS-1 & 2-' B 3.4.1 - 6 8/22/98 Revision K
l i' RCS Minimum Temperature for Criticality B 3.4.2 f G) i 3.4 . REACTOR COOLANT SYSTEM (RCS)
- B 3.4.2 RCS Minimum Temperature for Criticality BASES
' BACKGROUND This'LCO is based upon meeting several major considerations before the-reactor can be maoe critical ano wnile the reactor is critical.
The first consideration is Moderator Temperature Coefficient (MTC). LCO 3.1.3. " Moderator Temperature Coefficient (MTC)." In the transient and accident analyses. the MTC is assumed to be in a range from slightly positive to negative and the operating temperature is assumed to be within the nominal operating envelope while the reactor is critical. The LC0 on minimum temperature for criticality helps ensure the unit is operated consistent with these assumptions. The second consideration is the protective instrumentation. Because certain protective instrumentation (e.g. excore neutron detectors) can be affected by moderator temperature. a temperature value within the nominal operating envelope is O' chosen to ensure proper indication and response while-the reactor is critical. The third consideration is the pressurizer operating characteristics. The transient and accident analyses assume that the pressurizer is within its normal startup and I operating range (i.e. saturated conditions and steam bubble present). It.is also assumed that the RCS temperature is
~w ithin its normal expected range for startup and power I operation. Since the density of the water, and hence the. l response of the pressurizer to transients, depends upon the 1 initial temperature of the moderator a minimum value for ,
moderator temperature within the nominal operating envelope 1 is chosen. The fourth' consideration is that the reactor vessel is above its minimum nil ductility Reference temperature when the reactor is critical. This parameter is also assured through compliance with LC0 3.4.3. "RCS Pressure and Temperature (P/T) Limits." l l l! p)y (_ o l BYRON - UNITS 1 & 2 B 3.4.2 - 1 8/21/98 Revision A
I l j RCS Minimum Temperature for Criticality B 3.4.2 BASES-l- (vl APPLICABLE Although the RCS minimum temperature for criticality is not l SAFETY ANALYSES itself an initial condition assumed in Design Basis Accidents (DBAs), the closely aligned temperature for Hot Zero Power (HZP) is a process variable that is an initial condition of DBAs. such as the Rod Cluster Control Assembly l (RCCA) withdrawal. RCCA ejection, and main stean line breal , accidents performed at zero power that either assumes the l failure of, or presents a challenge to, the integrity of a ! fission product barrier. I All low power safety analyses assume, initial RCS loop temperatures greater than or equal to the HZP temperature of 557 F (Ref. 1). This minimum temperature for criticality I limitation provides a small band. 7'F. for critical I operation below HZP. This band allows critical operation below HZP during unit startup and does not adversely affect any safety analyses since the MTC is not sigrificantly affected by the small temperature difference Letween HZP and , the minimum temperature for criticality. l The RCS minimum temperature for criticality satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). G 1 LCO Compliance with the LC0 ensures that the reactor will not be l made or maintained critical (k r, a 1.0) at a temperature less than a small band below tfie HZP temperature, which is l assumed in the safety analysis. Failure to meet the I requirements of this LCO may produce initial conditions inconsistent with the initial conditions assumed in the safety analysis. APPLICABILITY In MODE 1 and MODE 2 with k 2 1.0. LCO 3.4.2 is applicable sincethereactorcanonlybere critical (k,fr a 1.0) in these MODES. l l t i l l l [ b/ BYRON - UNITS 1.& 2 B 3.4.2 - 2 8/21/98 Revision A l
.l
RCS Minimum Temperature for Criticality B 3.4.2 - T BASES I APPLICABILITY (continued) The special test exception of LCO 3.1.8. "MCDE 2 PHYSICS TESTS Exceptions." permits PHYSICS TESTS to be performed at s 5% RTP with RCS loop average temperatures slightly lower
$, than normally allowed so that fundamental nuclear characteristics of the core can be verified. In order for N nuclear characteristics to be accurately measured. it may be T necessary to operate outside the normal restrictions of this M LCO. For example, to measure the MTC at beginning of cycle. ;
4 it is necessary to allow RCS loop average temperatures to i
# fall below Tno % .which may cause RCS loop average temperatures to fall below the temperature limit of this LCO.
ACTIONS 6.l If the parameters that are outside the limit cannot be restored, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to MODE 2 with k er < 1.0 within 30 minutes. Ra)idreactorshutdowncanbereadilyandpractically /_') ' aciieved within a 30 minute period.
~.-
The Completion Time is reasonable, based on operating experience, to reach MODE 2 with kg, < 1.0 in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.4.2.1 REQUIREMENTS RCS loop average temperature is required to be verified 2: 550'F once every 12 hours. The SR to verify RCS loop average temperatures every 12 hours is frequent enough to prevent the inadvertent violation of the LC0 and takes into account indications 6nd alarms that are continuously available to the operator in the control room. REFERENCES 1. 'UFSAR. Section 15.0.3.
- s. s
)
BYRON - UNITS 1 & 2- B 3.4.2 - 3 8/22/98 Revision K
f RCS P/T Limits B 3.4.3 l .B 3.4 _. REACTOR COOLANT SYSTEM (RCS) lB 3.4.3. RCS Pressure and Temperature (P/T) Limits BASES -l L ' BACKGROUND All'comoonents of the RCS are designed to withstand effects- 1 l of cyclic loads due to system pressure and temperature ; l- changes. These loads are. introduced by startup (heatup) and ! L shutdown (cooldown) operations, power transients, and-L reactor trips. This LC0-limits the pressure ano temperature
- changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.
The PTLR contains P/T limit curves for heatup. cooldown. l Inservice Leak and Hydrostatic (ISLH) testing and data for f the maximum rate of change of reactor coolant temper'ature (Ref. 1). Each P/T limit' curve defines an acceptable region for normal operation. The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region.
- The"LC0 establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the Reactor Coolant Pressure Boundary (RCPB). The vessel is the component most subject to brittle failure. and the LCO limits apply to the entire RCS (except the pressurizer).
The limits' do not apply to the pressurizer, which has different design characteristics and operating functions. l 10 CFR 50. Appendix G (Ref. 2), requires the establishment of P/T limits for specific material fracture toughness
, requirements of the RCPB materials. Reference 2 requires an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests. It mandates the use of the American Society of Mechanical Engineers (ASME) Code. Section III. Appendix G (Ref. 3).
L l t O i- BYRON - UNITS 1 & 2 B 3.4.3 - 1 8/21/98 Revision A p-
- e. v - ,e e v s-+-. wr -
--.--n e------.-m e-~-- .
ve < - .
RCS P/T Limits B 3.4.3 BASES
'~'
BACKGROUND (continued) The neutron embrittlement effect on the material touahness is reflected by increasing the Nil Ductility Reference Temperature (RT,) as exposure to neutron fluence increases. The actual shift in the RT e of the vessel material will be established periodically by removing and evaluating the irradiated reactor vessel material specimens. In accordance with ASTM E 185 (Ref. 4) and Appendix H of 10 CFR 50 (Ref. 5). The operating P/T limit curves will be adjusted. as necessary, based on the evaluation findings and the recommendations of Regulatory Guide 1.99 (Ref. 6). The P/T limit curves are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive. At any specific pressure, temperature, and temperature rate of change, one location within the reactor vessel will dictate the most restrictive limit. Across the span of the P/T limit curves, different locations are more restrictive, and. thus the curves are composites of the most restrictive regions.
.[ The heatup curve represents a different set of restrictions G than the cooldown curve because the directions of the thermal gradients through the vessel wall are reversed. The thermal gradient reversal alters the location of the tensile stress between the outer and inner walls during heatup and cooldown. respectively.
The criticality limit curve includes the Reference 2 requirement that it be a 40 F above the heatup curve or the cooldown curve. and not less than the minimum permissible temperature for ISLH testing. However, the criticality curve is not operationally limiting: a more restrictive limit exists in LCO 3.4.2. "RCS Minimum Temperature for Criticality." n
'N BYRON - UNITS 1 & 2 B 3.4.3 - 2 8/21/98 Revision A l
l
, .. . - ~ --- . ..
I l RCS P/T Limits , B 3.4.3 BASES , BACKGROUND (continued) The consequence of violating the LCO limits is tnat the RCS l has been operated under conditions that can result in brittle failure of the RCPB. possibly leading to a ' nonisolable leak or loss of coolant accident. In the event i these limits are exceeded, an evaluation must be performed ; to determine the effect on tne structural integrity of the l RCPB components. The ASME Code. Section XI. Appendix E (Ref. 7), provides a recommended methodology for evaluating l an operating event that causes an excursion outside the I l limits. l APPLICABLE The P/T limits are not derived from Design Basis Accident : SAFETY ANALYSES (DBA) analyses. They are prescribed during normal operation l to avoid encountering pressure, temperature. and temperature ' rate of change conditions that might cause undetected flaws l to propagate and cause nonductile failure of the RCPB. an unanalyzed condition. Reference 1 establishes the methodology for determining the P/T limits. Although the P/T limits are not derived from any DBA. the P/T limits are acceptance limits since they preclude operation in an (n) wJ unanalyzed condition. RCS P/T limits satisfy Criterion 2 of I 10 CFR 50.36(c)(2)(i1). LCO The two elements of this LCO are:
- a. The limit curves for heatup.'cooldown and ISLH testing; and
- b. Limits on the rate of change of temperature.
The LC0 limits apply to all components of the RCS. except the pressurizer. These limits define allowable operating regions and permit a large number of operating cycles while providing a wide margin to nonductile failure. I b v BYRON - UNITS 1 & 2 B 3.4.3 - 3 8/21/98 Revision A
RCS P/T Limits B 3.4.3 r
- O BASES V
l LCO (continued) The limits for the rate of change of temperature control the thermal gradient through the vessel wall and are used as inputs for calculating the heatup. Cooldown, and ISLH testing P/T limit curves. Thus the LCO for the rate of change of temperature restricts stresses caused by thermal gradients and also ensures the validity of the P/T limit curves. Violating the LC0 limits places the reactor vessel outside of the bounds of the stress analyses and can increase stresses in other RCPB components. The consequences depend on several factors, as follow:
- a. The severity of the departure from the allowable operating P/T regime or the severity of the rate of change of temperature;
- b. The length of time the limits were violated (longer violations allow the temperature gradient in the thick vessel walls to become more pronounced); and
- c. The existences. sizes, and orientations of flaws in p)
L the vessel material. v APPLICABILITY The RCS P/T limits LCO provides a definition of acceptable operation for prevention of nonductile failure in , accordance with 10 CFR 50. Appendix G (Ref. 2). Although j the P/T limits were developed to provide guidance for operation during heatup or cooldown (MODES 3. 4 and 5) or ISLH testing their Applicability is at all times in keeping with the concern for nonductile failure. The Applicability includes MODE 6 and conditions with no fuel in the reactor vessel. This arovides continued prevention of nonductile failure even w1ile the reactor is "defueled" so that the RCS is acceptable for operation when fuel is returned to the reactor vessel. The limits do not apply to the pressurizer. l l l l l^ !n L.) l BYRON - UNITS 1 & 2 B 3.4.3 - 4 8/21/98 Revision A
- ~ . - - .- - - __. -.-.---- -._ _..- - - -
RCS P/T Limits B 3.4.3
- BASES l APPLICABILITY (continued).
! During MODES 1 and 2. Other Technical Specifications provide L Timits for operation'that can be more restrictive than or L .can supplement these P/T limits. LCO 3.4.1. "RCS Pressure. L Temperature. and Flow Departure from Nucleate Boiling (DNB) h Limits:" LCO 3.4.2. "RCS Minimum Temperature for Criticality;" and Safety Limit 2.1. " Safety Limits:" also ' provide operational restrictions for pressure, temperature and maximum pressure. Furthermore. MODES 1 and 2 are above the temperature range of concern for nonductile failure, and stress analyses have been performed for normal maneuvering profiles. such as power ascension or descent. . A.1 and A.2 ACTIONS 0peration outside the P/T limits during MODE 1. 2. 3. or 4 must be corrected so that the RCPB is returned to a condition that has been verified by stress analyses. The 1 30 minute Completion Time reflects the urgency of restoring the parameters to within the analyzed range. Most t . violations will not be severe, and the activity can be l: accomplished in this time in a controlled manner. Besides restoring operation within limits, an engineering ! evaluation is required to determine if RCS operation can continue. The evaluation must verify the RCPB integrity remains acceptable and must be completed before-continuing l operation. Several methods may be used. including i
-comparison with pre-analyzed transients in the stress ,
analyses, new analyses, or inspection of the components. l l
.For the vessel beltline only. ASME Code. Section XI.
Appendix E (Ref 7), may be used to support the evaluation. The 72 hour Completion Time is reasonable to accomplish the evaluation. The evaluation for a mild violation is possible within this time, but more severe violations may require special, event specific stress analyses or inspections. A L favorable evaluation must be completed before continuing to operate. 1 y
/"]
)
-BYRON - UNITS 1 & 2 B 3.4.3 - 5 8/21/98 Revision A l
L RCS P/T Limits B 3.4.3 9 H(7 ' BASES i - %) L' ' ACTIONS (continued) l l Condition A is modified by a Note reauiring Required Action A.2 to be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion'outside the '
- allowable limits. Restoration alone per Recuired Action A.1
.is insufficient because higher than analyzec . stresses may have occurred and may' have affected the RCPB.1ntegrity.
L B 1 and B.2 1 l If.a Required Action and associated Completion Time of
. Condition A are not met, the unit must be placed in a lower MODE because either the RCS remained in an unacceptable P/T
. region for an extended period of increased stress or a l i~ sufficiently severe event caused entry into an
- unacceptable region. .Either possibility indicates a_need for more careful examination of the event. best accomplished j with the RCS at reduced pressure and temperature. In l reduced pressure and temperature conditions, the possibility i of propagation with undetected flaws is decreased, '
, ,3 ' If the required restoration activity of Required Action A.l. -f cannot be accom]lished within 30 minutes. Required-Action B.1 and Required Action B.2 must be implemented to reduce pressure and temperature.
If the required evaluation for continued operation cannot be accomplished within 72 hours or the results are indeterminate or unfavorable. action must proceed to reduce y pressure and temperature as specified in Required Action B.1 and Required Action B.2. A favorable engineering evaluation must be completed and-documented before returning to l operating pressure and temperature condi_tions. j Pressure and tem]erature are reduced by bringing the unit to I MODE 3 within 6 1ours and to MODE 5 within 36 hours. i The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging plant systems. l-lL BYRON - UNITS 1 & 2 B 3.4.3 - 6 8/21/98 Revision A' t -
- i. ~ . . . , . . _ _. -. _ - _ . . - --
RCS P/T Limits B 3.4.3-BASES ( ACTIONS (continued) C.1 and C.2 Actions must be initiated immediately to correct operation l outside of the P/T limits at times other than wnen in MODE 1. 2. 3. or 4. so that the RCPB is returned to a condition that.has been verified by stress analysis. The immediate Completion Time reflects the urgency of initiating action to restore the parameters to within the analyzed range. Most violations will not be severe, and the activity I can be accomplished in this time in a controlled manner. Besides restoring operation within limits, an evaluation is required to determine if RCS operation can continue. The i evaluation must verify that the RCPB integrity remains , acceptable and must be completed prior to entry int 6 MODE 4. l Several methods may be used, including comparison with pre-analyzed transients in the stress analyses, or inspection of the components. For the vessel beltline only. ASME Code. Section XI. ! Appendix E (Ref. 7) may be used to support the evaluation. ; [) v Condition C is modified by a Note requiring Required Action C.2 to be completed whenever the Condition is ! entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits. Restoration alone per Recuired Action C.1 is insufficient because higher than analyzec stresses may have occurred and may have affected the RCPB integrity. l 1 l l l 1 i t T'T I V 4 BYRON - UNITS 1 & 2 B 3.4.3 - 7 8/21/98 Revision A 1
1 RCS P/T Limits B 3.4.3 BASES f'g]- A SURVEILLANCE SR 3.4.3?1
' REQUIREMENTS ~
Verification that operation is within the PTLR limits is.
. required every 30 minutes when RCS 3ressure and temperature conditions are undergoing planned c1anges. This' Frequency ,
-is considered reasonable in view of the control room- :
indication-available to monitor RCS status. Also, since l M temperature rate of change . limits are specified in hourly I 9L increments. 30 minutes permits assessment and correction for W minor deviations within a reasonable time-4 4 Surveillance for'heatup. cooldown, or ISLH testing may be I
- discontinued when.the definition given in the-relevant plant
$ procedure for ending the activity is satisfied.
This SR is modified by a Note that only requires th'is SR to
- be performed during system heatup. cooldown, and ISLH testing. This SR is not required during ' critical operations because the combination of LC0 3.4.2 establishing a lower - :
bound and the Safety Limits establishing an upper bound will l provide adequate controls to prevent a change in excess of ; 100*r prior to entry into the performance condition of i
- 74. heatup and cooldown operations.
REFERENCES 1. WCAP-14040 " Methodology used to Develop Cold 0verpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves." June 1994.
- 2. 10 CFR 50 Appendix G.
- 3. ASME Boiler and Pressure Vessel Code. Section III.
Appendix G.
- 4. ASTM E 185-82. July 1982.
- 5. 10 CFR-50. Appendix H.
- 6. Regulatory Guide 1.99. Revision 2. May 1988.
- 7. ASME. Boiler and Pressure Vessel Code. Section XI.
. Appendix E.
p {. u: '
~ BYRON - UNITS 1 &'2 B 3.4.3 - 8 8/22/98. Revision K
_ _. _ _ _ - . _ _ . _ . _ . _ . _ . - ~ . _ . . - . - - _ - _ _ . - . - - RCS Loops-MODES 1 and 2 B 3.4.4 IG' B 3.4- REACTOR COOLANT. SYSTEM (RCS) U ' BL3.4.4 RCS Loops-MODES 1 and 2-L BASES l l BACKGROUND The primary function of the RCS is removal of the heat
- . generated in the fuel due to the fission process. and transfer.of this heat via the Steam Generators (SGs). to the secondary plant.
'Th'e secondary functions of the.RCS. include:
- a. Moderating the neutron energy level to the thermal-state, to increase the probability of-fission:
} b .' Improving the_ neutron economy by acting as a !. . reflector: !- c. Carrying-the soluble neutron poison, boric acid: e d. Providing a second barrier against fission product L release to the environment; and l e. Removing the heat generated in the _ fuel due to fission l- product decay following a unit shutdown. i- The reactor coolant is' circulated through four loops connected in parallel to the reactor ~ vessel, each containing an SG. a Reactor Coolant Pump (RCP).. and appropriate flow and temperature instrumentation for.both control and protection. The_ reactor vessel contains the clad fuel. The !~ SGs provide the heat sink to the isolated secondary coolant. The RCPs circulate the coolant through the reactor vessel and SGs at a ~ sufficient rate to ensure proper heat transfer and prevent fuel damage. This forced circulation of the 1 reactor coolant ensures mixing of the coolant-for proper l boration and chemistry control. -i i i t l' l !!( BYRON -l UNITS 1 & 2_ B 3.4.4 - 1 8/21/98 Revision A
l RCS Loops-MODES 1 and 2 B 3.4.4 l/3 BASES lV l APPLICABLE Safety analyses contain various assumptions for the design
- SAFETY ANALYSES bases accident initial conditions including RCS pressure.
RCS temperature, reactor power level, core parameters. and safety system setpoints. The important aspect for this LCO is the reactor coolant forced flow rate, which 15 represented by the number of RCS loops in service. !' Both transient and steady state analyses have been performed to establish the effect of flow on the Departure from Nucleate Boiling (DNB). _The transient and accident analyses for the plant have been performed assuming four RCS loops are in operation. The majority of the plant safety analyses are based on initial conditions at high core power or zero power. The accident analyses that are most important to RCP operation are the four pump coastdown, single pump locked rotor. single pump (broken shaft or coastdown), and rod withdrawal events (Ref. 1). The Uncontrolled Rod Cluster Control Assembly Bank Withdrawal from a Subcritical or Low Power Startup Condition and the spectrum of Rod Cluster Control Assembly Ejection events were analyzed assuming only two of four RCPs in operation. This conservatively bounds operation in the t T lower modes. Analyzing these transients with only two RCPs O in operation will result in a lower Departure from Nucleate Boiling Ratio (DNBR). thus producing more ,imiting results. Steady state DNB analysis has been performed for the four RCS loop operation. For four RCS loo) operation, the steady state DNB analysis, which generates t1e pressure and temperature Safety Limit (SL) (i.e.. the DNBR limit) assumes a maximum power level of 118% RTP. This is the design overpower condition for four RCS loop o)eration. The value for the accident analysis setpoint of t1e nuclear overpower (high flux) trip is-109% RTP and is based on an analysis assumption that bounds possible instrumentation errors. The DNBR limit defines a locus of pressure and tem]erature points that result in a minimum DNBR greater tlan or equal to the critical heat flux correlation limit.
.A
'a\j s BYRON - UNITS 1 & 2 B 3.4.4 - 2 8/21/98 Revision A
1 RCS Loops-MODES 1 and 2 I B 3.4.4 l L t fD. v
-BASES ~ ,
1
~ APPLICABLE' SAFETY ANALYSES (continued) I
.The. unit is designed to 03erate with all PCS loops in
! operation-to maintain DNBR above the SL. during all normal .
. operations and anticipated ~ transients. By ensuring _ heat !
transfer in the nucleate boiling region. adequate heat-- L i- transfer is provided between the fuel cladding and the reactor coolant. RCS Loops-MODES 1 and 2 satisfy Criterion 2 of - 10 CFR 50.36(c)(2)(ii). LCO' -The purpose of this LC0 is to require an adequate forced; flow rate for_ core heat removal. Flow is represented by the
- number of RCPs in operation for removal of heat by the SGs.
To meet safety analysis acceptance criteria for DNB. four j .
-pumps are required at rated power.
2 An OPERABLE RCS.. loop consists of an OPERABLE RCP in y operation providing forced flow for heat transport and an
't OPERABLE SG in accordance with the~ Steam Generator Tube d
j : Surveillance. Program.. K_)[ Ih Q
-BYRON - UNITS 1 & 2. B 3.4.4 - 3 8/22/98 Revision K f
I. - - , ,
l l RCS Loops-MODES 1 and 2 i B 3.4.4 L Y'T - BASES V l APPLICABILITY In MODES 1 and 2. the reactor is critical and thus nas the l potential to produce maximum THERMAL POWER. Thus, to ensure
- that the assumptions of the accident analyses remain valid. l all RCS loops are required to be OPERABLE and in operation 1 in tnese MODES to prevent DNB and core damage. l The decay heat production rate is much lower than the full power heat rate. As such, the forced circulation flow and heat sink requirements are reduced for lower. noncritical MODES as indicated by the LCOs for MODES 3. 4. and 5.
Operation in other MODES is covered by: LCO 3.4.5. "RCS Loops -MODE 3": LCO 3.4.6. "RCS Lcops -MODE 4": LCO 3.4.7. "RCS Loops -MODE 5. Loops Filled" . LCO 3.4.8. "RCS Loops -MODE 5. Loops Not Filled": LC0 3.9.5. " Residual Heat Removal (RHR) and Coolant Circulation-High Water Level" (MODE 6); and LCO 3.9.6. " Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level" (MODE 6), in ACTIONS A.1 If the requirements of the LCO are not met, the Required Action is to reduce power and bring the unit to MODE 3. This lowers power level and thus reduces the core heat removal needs and minimizes the possibility of violating DNB limits. The Completion Time of 6 hours is reasonable, based on operating experience to reach MODE 3 from full power conditions in an orderly manner and without challenging safety systems. l i i,m 0 BYRON - UNITS 1 & 2 B 3.4.4 - 4 8/21/98 Revision A
RCS Loops-MODES 1 and 2 8 3.4,4
/~N BASES G)'
SURVEILLANCE SR 3 4 4.1 REQUIREMENTS This SR requires verification every 12 hours that each RCS loop is'in operation, Verification may include flow rate. temperature..or pump status monitoring, which helps ensure that forced flow is providing heat removal while maintaining the margin to DNB. The Frequency of 12 hours is sufficient considering other indications and alarms available to the operator in the control room to monitor RCS loop performance. REFERENCES 1. UFSAR, Chapter 15. n - ! I u i (' )
'(/ l 3
t BYRON -UNITS 1 & 2 B 3.4.4 - 5 8/21/98 Revision A , i
I i RCS Loops -MODE 3 L B 3.4.5 l l J
' B - 3. '4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.5 .RCS Loops -MODE 3 l l
- l BASES l l .-
1 l BACKGROUND .In MODE 3. the primary function of the reactor coolant is
. removal of decay heat and transfer of this heat. via the L Steam Generator (SG). to the secondary plant fluid. A secondary function of.the reactor coolant is to act as a carrier for soluble neutron poison, boric acid. !
The reactor coolant is circulated through four RCS locos, , connected in parallel to the reactor vessel, each containing l an-SG, a Reactor Coolant' Pump (RCP), and appropriate flow. ! pressure. level, and temperature instrumentation for control . protection, and indication. The reactor vessel contains the clad fuel. The SGs provide the heat sink. The RCPs circulate the water through the reactor' vessel and SGs at a sufficient rate to ensure proper heat transfer and L prevent fuel damage. In MODE 3 RCPs are used to provide forced circulation for l heat removal during heatup and cooldown. The MODE 3 decay
- e heat removal' requirements are low enough that a single RCS o loop with one RCP running is sufficient to remove core decay heat. However, two RCS loops are required to be OPERABLE to ensure redundant capability for decay heat removal.
APPLICABLE. Whenever the Rod Control System is capable of rod ,
-SAFETY: ANALYSES withdrawal (i.e., the Reactor Trip Breakers (RTBs) are in ;
the closed position and the Control Rod Drive Mechanisms 1 (CRDMs) are energized) an inadvertent rod withdrawal from l subtritical. resulting in a power excursion, is possible l
-(Ref. 1).. Such a: transient could be caused by a malfunction of the rod control system. In addition; the possibility of i a power excursion due to the ejection of an inserted control i rod is possible with the breakers closed or open. Such a l transient could be caused by the mechanical failure of a l CRDM.
t l(
-BYRON - UNITS 1.& 2 B 3.4.5 - 1 8/21/98 Revision A I 1
~, , - - - . . . , , , , , , , , . , . . - - - . - - - .-,,,w, , - - - . , -
RCS Loops -MODE 3 B 3.4.5
/~N BASES APPLICABLE SAFETY ANALYSES (continued)
Therefore, in MODE 3 with the Rod Control System capable of rod withdrawal, accidental control rod withdrawal from subcritical is postulated and requires at least two RCS loops to be OPERABLE and in operation to ensure that the accident analyses limits are met. For those conditions when the Rod Control System is not capable of rod withdrawal. two 4 loops are required to be OPERABLE. but only one RCS loop is required to be in operation to be consistent with MODE 3 accident analyses. Failure to provide decay heat removal may result in challenges to a fission product barrier. The RCS loops are part of the primary success path that functions or actuates-to prevent or mitigate a Design Basis Accident or transient that either assumes the failure of. or presents a challenge to, the integrity of a fission product barrier. RCS Loops-MODE 3 satidv Criterion 3 of 10 CFR 50.36(c)(2)(ii). [)
~ LC0 The purpose of this LC0 is to require that at least two RCS loops be OPERABLE. In MODE 3 with the Rod Control System capable of rod withdrawal. two RCS loops are required to be in operation due.to the postulation of a power excursion because of an inadvertent control rod withdrawal. The required number of RCS loops in operation ensures that the Safety Limit criteria will be met for all of the postulated accidents.
When the Rod Control System is not capable of rod withdrawal only one RCS loop in operation is necessary to ensure removal of decay heat from the core and homogenous boron concentration throughout the RCS. An additional RCS loop is required to be OPERABLE to provide backup forced flow capability. l l r 4 p Li !- BYRON - UNITS 1 & 2 B 3.4.5 - 2 8/21/98 Revision A l l
1 1 L RCS Loops -MODE' 3- i g o
-B 3.4.5-t .
i tBASES I M) '* i LCOL(contiinued) - i The Note permits all'RCPs to be removed from' operation-f(i.e; not-in operation) for s 1 hour per 8 hour period. 4The purpose:of the Note is to perform tests that are
- designed to validate-various accident ' analyses values. One of these'. tests is validation'of the pump coastdown curve ,
.used as input to a' number of accident analyses including a - -)
aloss of flow accident. This-test is' generally performed in ! MODE 3 during the. initial startup testing program, and as such-should only be performed once. If, however, changes are.made to the RCS that would cause a change to the flow characteristics of the RCS. the input values of:the coastdown curve must be revalidated by conducting the test
'again. Another test performed during the startup testing
, program is the ' validation of rod drop times during cold -
' conditions, both'with and without' flow.
- The no' flow .te'st may be )erformed in MODE 3, 4. 'or 5 and requires that the pumps. Je stopped for a short period:of time. The Note permitsithe stopping of the pumps in order to perform this test and validate the assumed' analysis
-values. As with the validation of the pump coastdown curve, this test should be aerformed only once unless the flow
_f_Y characteristics' of. t1e RCS are changed. The 1 hour time ' L7 period specified Lis adequate to perform the desired tests, and operating experience has shown that boron stratification _, is'not a problem during this short period with no . forced 1 flow. -! Utilization of the Note .is permitted 3rovid'ed the following conditions are met. along with any otler conditions imposed by procedures:
, a. No operations are permitted that would dilute the RCS boron concentration, thereby maintaining the margin to y criticality. Boron reduction is prohibited because a 1 o -- unifon concentration distribution throughout the RCS 43: cannot be-ensured when in natural circulation: and 4 b. -Core outlet temperature is maintained at least 10 F i
1 N~ below saturation temperature, so that no vapor bubble
~
may form and possibly cause a natural circulation flow j L obstruction. p 4 1.' , M l
' BYR0'N'-' UNITS 1 & 2- B 3.4.5 - 3 10/5/98 Revision K
RCS Loops-MODE 3 l B.3.4.5
~
. i ! l ' .?N. - R,fy(BASES' 1. 0 LCO-(continued) g
?J An OPERABLE RCS loop consists of one OPERABLE RCP and one m OPERABLE SG in accordance with the Steam Generator Tube 4
Surveillance Program which has the minimum water level
--specified:in SR 3J4.5.2.' An RCP is OPERABLE if it is ca,>able of being powered and is able to provide forced flow if required.~ ^
l
. APPLICABILITY- In MODE 3. this-LCO ensures forced circulation of the 1 reactor coolant to remove decay heat from the core and to provide proper boron mixing. The most stringent condition ;
of the LCO. that is. two RCS loops OPERABLE and two RCS q loops in operation applies to MODE 3 with' the Rod Control ' System capable of rod withdrawal. The least stringent condition, that is, two RCS loops OPERABLE and one RCS' loop in. operation, applies to MODE 3 with the Rod Control System
. not capable of rod withdrawal.
l
- Operation in other MODES is covered by:
LC0 3.4.4. "RCS Loops-MODES 1 'and 2": , ( -
. LC0 3.'4.6 "RCS Loops-MODE _4":
I N>
' LCO 3.4.7. "RCS Loops--MODE 5. Loops Filled": '
LCO 3.4.'8. "RCS Loops-MODE 5. Loops Not Filled": LCO 3.9.5. " Residual Heat Removal (RHR) and Coolant I Circulation-High Water Level" (MODE 6): and ; LC0 3.9.6. " Residual Heat Removal-(RHR) and Coolant 1 Circulation-Low Water Level" (MODE 6). I l
/
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.A l l
' u);
. . l BYRON = UNITS.1 & 2 ~ B 3.4.5 - 4 8/22/98 Revision K ;
m
~ . .
RCS Loops-MODE 3 B 3.4.5 /~'T BASES O ACTIONS A.1 If the required RCS loop is not in operation, and the Rod Control System is capable of rod withdrawal. the Requirec Action is to place the Rod Control System in a concition incapable of rod withdrawal (e.g., disable all CRDMs by opening the RTBs or de-energizing the motor generator (MG) sets). When the Rod Control System is capable of rod withdrawal,-it is postulated that a power excursion could occur in the event of an inadvertent control rod withdrawal . This mandates having the heat transfer capacity of two RCS loops in operation. If only one loop is in operation. the Rod Control System must be rendered incapable of rod withdrawal. The Completion Time of 1 hour to defeat the Rod Control System is adequate to perform these operations in an orderly manner without exposing the unit to risk for an undue time period. B.1 and 8.2 If no RCS loop is in operation with the Rod. Control System not capable of rod withdrawal except as permitted by the f3 Note in the LCO section all operations involving a r i reduction of RCS boron concentration must be suspended, and V action to restore one RCS loop to operation must be i immediately initiated. Boron dilution requires forced ; circulation for proper mixing. l The immediate Completion Time reflects the importance of l maintaining operation for heat removal. The action to restore must be continued until one loop is restored to o[. ration. 1 ('O BYRON - UNITS 1 & 2 B 3.4.5 - 5 8/21/98 Revision A
RCS Loops -MODE 3 B 3.4.5
/7 BASES l T '^
l-ACTIONS (continued) C.1. C.2. and C.3 l If no RCS loop is in operation with the Rod Control System I capable of rod withdrawal, except as permitted by the Note in the LCO section or if the Required Action and associated Completion Time of Condition A are not met, action must be initiated to place the Rod Control System in a condition incapable of rod withdrawal (e.g.. disable all CRDMs by I opening the RTBs or de-energizing the MG sets). Additionally, all operations involving a reduction of RCS l boron concentration must be suspended, and action to restore i one of the RCS loops to operation must be immediately ' initiated. Boron dilution requires forced circulation for proper mixing. and disabling the CRDMs removes the possibility of an inadvertent rod withdrawal. The immediate Completion Time reflects the importance of maintaining operation for heat removal. The action to restore must be continued until one loop is restored to l operation. 1 p ., U l V If one required RCS loop is inoperable, redundancy for heat removal is lost. The Required Action is restoration of the required RCS loop to OPERABLE status within the Completion Time of 72 hours. This time allowance is a justified period to be without the redundant, nonoperating loop because a single loop in operation has a heat transfer capability greater than that needed to remove the decay heat produced in the reactor core and because of the low probability of a failure in the remaining loop occurring during this period. El If the Required Action and associated Completion Time of Condition D are not met, the unit must be brought to MODE 4. In MODE 4, the unit may be placed on the Residual Heat Removal System. The additional Completion Time of 12 hours is compatible with required operations to achieve cooldown and depressurization from the existing unit conditions in an orderly manner and without challenging plant systems. l O
\ /
BYRON - UNITS 1 & 2 B 3.4.5 - 6 8/21/98 Revision A
~1 RCS Loops-MODE 3 B 3.4.5
' h) A" . BASES' ACTIONS (continuedt , FL1 Fi2.'and F 3' cIf two re' quired RCS loops are inoperable action must be .
';nitiated to place the Rod Control System in a condition Lincapable of rod withdrawal (e.g., disable all'CRDMs-by ,
-opening.the RTBs or de-energizing the MG sets). All
, _ operations involving a reduction of RCS boron concentration
-must be. suspended, and action to restore one of the RCS '
loops to 0PERABLE status must be initiated. ' Boron dilution ! _ requires forced circulation for proper mixing, and disabling
.the CRDMs removes.the. possibility of an inadvertent rod withdrawal.. The-immediate Completion H me reflects the importance'of maintaining the capabi1ity'for heat removal.
The action to restore must'be continued until one' loop is' 1 restored to OPERABLE: status. i, ., I SURVEILLANCE- SR 3.4.5.1-
, REQUIREMENTS
'This SR requires verification every 12 hours that the.
c . required operating loops are in operation. Verification may 'v) I ,' include flow rate.- temperature, and pump status monitoring, which hel
- removal. ps Theensure'that Frequency' forced flow isisproviding of 12 hours sufficient-heat considering other indications and alarms available to the operator in the control room to monitor RCS loop performance.
SR= 3.4'.5.2. w.
.SR 3.4'.5.2 requires verification'of required SG OPERABILITY.
- 9 SG OPERABILITY is verified by ensuring that the secondary side narrow range water level is a 18% for each required RCS
}#ll loop. If the SG secondary side narrow. range water level is o U < 18%, the tubes may become uncovered and the associated dl- ' loop may not be. capable of providing the heat sink for
- removal-of the decay heat. The 12. hour Frequency is l considered adequate
- in view of other indications available L in the control room to alert the operator to a loss of- SG level ~.
I 1 L h ~Qf BYRON - UNITS 1 & 2- B 3.4.5 - 7 8/22/98 Revision K t... .
~ y 4 - =
. -, .. .. - . - _ . >- . ~ .- . _- . . - . .- - . , - _ _ . .
e- N RCS Loops -MODE :3 ,
.B 3.4.5 1
[' -BASES
% /.
~
g- SURVEILLANCE REQUIREMENTS (continued)
, s- ,
fg < .SR 3L4.5.3-Verification that the required RCPs are OPERABLE ensures-(1 that-safety analyses' limits are. met. -The requirement also ensures that an additional:RCP can be placed in operation. 7 *t ~ cif needed.oto maintain decay heat removal-and reactor coolant circulation.- Verification-is performed by verifying 1 proper breaker-alignment and power availability to the
. required RCP. The Frequency of 7 days is considered 1 reasonable .in view-of other administrative controls l available and'has been shown to be'_ acceptable by operating
- . experience.
1
REFERENCES:
UFSAR,.Section 15.4.1. J M .. < {} j i 1 l,y
.(_)/
. BYRON-l. UNITS 1&2 B 3.4.5 - 8 10/5/98 Revision K' l i
RCS Loops-MODE 4 B 3.4.6 (d B 3.4 REACTOR'C00LANT SYSTEM (RCS) B 3.4.6 RCS Loops - MODE ' 4 BASES BACKGROUND In MODE 4. the primary function of the reactor coolant is the removal of decay heat and the transfer of this heat to either the Steam Generator (SG) secondary side coolant or the component cooling water via the Residual Heat Removal (RHR) heat exchangers. The secondary function of the reactor coolant is to act as a carrier for soluble neutron poison, boric acid, t In MODE 4. the reactor coolant is circulated through at least two of the four RCS loops connected in parallel to the reactor vessel, each loop containing an SG. a Reactor Coolant Pump (RCP), and appropriate flow, pressure. level. and temperature instrumentation for control, protection, and. indication. The RCPs circulate the coolant through the reactor vessel and SGs at a sufficient rate to ensure proper heat transfer and to prevent boric acid stratification. n In MODE 4. RHR loops can be used in lieu of RCS loops to f, j provide forced circulation. The intent of this LCO is to A - provide forced flow from at least one RCP or one RHR loop for decay heat removal and transport. The flow provided by : one RCP loop or RHR loop is adequate for decay heat removal. ! The other intent of this LCO is to require that two paths be available to provide redundancy for decay heat removal. I APPLICABLE In MODE 4. circulation of the reactor coolant increases the SAFETY ANALYSES time available for mitigation of the accidental boron dilution event. The RCS and RHR loops provide this circulation. RCS Loops-MODE 4 satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii). t l /7 LJ BYRON - UNITS 1 & 2 B 3.4.6 - 1 8/21/98 Revision A
I RCS Loops-MODE 4 B 3.4.6
'~ j BASES . /
LC0 The purpose of this LCO is to require that at least two loops be OPERABLE in MODE 4 and that one of these loops be in' operation. The LC0 allows the two loops that are required to be OPERABLE to consist of any combination of RCS loops and RHR loops. Any one loop in operation provides enough flow to remove the decay heat from the core with ; forced circulation. An additional loop-is required to be 1 OPERABLE to provide redundancy for-heat removal. Note 1 permits all RCPs and RHR pumps to be removed from o)eration for s 1 hour per 8 hour period. The purpose of tie Note is to permit tests that are designed to validate various accident analyses values. One of the tests performed durir.g the startup testing program is the-validation of rou drop times during cold conditions, both with and without flow. The no flow test may be performed in MODE 3. 4 or 5.and requires that the pum]s be stopped for a short period of time. The Note permits tie stopping of the pumps in order to perform this test and validate the assumed J analysis values. If necessary, this test may also be conducted after the initial startup testing program. The 1 hour time period is adequate to perform the test, and
-x operating experience has shown that boron stratification is
). not a problem during this short period with no forced flow.
Utilization of Note 1 is permitted provided the following conditions are met along with any other conditions imposed by procedures:
- a. No operations are permitted that wculd dilute the RCS boron concentration, therefore maintaining the margin eJ to criticality. Boron reduction is prohibited because
? a uniform concentration distribution throughout the.
y' RCS cannot be ensured when in natural circalation; and n W b. Core outlet temperature is maintained at least 10 F
@ below saturation. temperature, so that no vapor bubble may form and possibly cause a natural circulation flow obstruction.
. ,,,3 BYRON-- UNITS 1 & 2 B 3.4.6 - 2 10/5/98 Revision K
RCS Loops -MODE 4 l B 3.4.6 l '(] BASES QJ LC0 (continued) Note 2 requires that the secondary side water temperature of each SG be < 50 F above each of the RCS cold leg temperatures before the start of an RCP with any RCS cold leg temperature s 350'r. This restraint is to prevent a 104 l temperature overpressurc event due to a thermal transient ! when an RCP is started. 1 An OPERABLE RCS loop comprises an OPERABLE RCP and an l OPERABLE SG which has the minimum water level specified in i SR 3.4.6.2. Similarly for the RHR System, an OPERABLE RHR loop is i comprised of an OPERABLE RHR pump capable of providing i forced flow to an OPERABLE RHR heat exchanger. RCPs and RHR , pumps are OPERABLE if they are capable of being powered and are able to provide forced flow if required. 1 APPLICABILITY In MODE 4. this LCO ensures forced circulation of the reactor coolant to remove decay heat from the core and to m - provide proper boron mixing. One loo) of either RCS or RHR provides sufficient circulation for t1ese purposes. (G) However two loops consisting of any combination of RCS and RHR loops are required to be OPERABLE to provide adequate redundancy for decay heat removal. Operation in other MODES is covered by: LC0 3.4.4. "RCS Loops-MODES 1 and 2": LCO 3.4.5. "RCS Loops -MODE 3": LC0 3.4.7. "RCS Loops-MODE 5. Loops Filled": LC0 3.4.8. "RCS Loops-M3DE 5. Loops Not Filled". LCO 3.9.5 " Residual Heat Removal (RHR) and Coolant Circulation-High Water Level" (MODE 6); and LCO 3.9.6. " Residual Heat Removal (RHR) and Coolant Circulation-Low Water Leval" (MODE 6). l
/~)
LJ BYRON - UNITS 1 & 2 B 3.4.6- 3 8/21/98 Revision A
- RCS ' Loops - MODE 4 B 3.4.6 BASES.
, ACTIONS A.1 and-AL2 l
If no loop is in operation, except during conditions 1 permitted by the' Note in the LCO section, all operations involving a reduction:of.RCS boron concentration must be , i suspended and action to restore one .RCS or RHR~ loop to operation must be immediately initiated. Baron dilution
' requires forced circulation to provide proper mixing and E preserve the margin to criticality. The immediate L Completion Times reflect the importance of maintaining L operation for decay heat removal.
.B.1 and B.2 If.one required RCS or RHR loop is inoperable and o'nly one required loop remains OPERABLE. the intended redundancy for heat removal is lost. Action must be initiated to restore a
!- second RCS or RHR loop to OPERABLE status. The immediate l- Completion Time reflects the importance of maintaining the availability of two paths for heat removal.
- If the one required OPERABLE loop is an RHR loop and if the
.. required loop is not restored to OPERABLE status, the unit L must be brought to MODE 5'within 24 hours. Bringing the l W unit to MODE 5 is a conservative action with regard to decay
- heat. removal. With only one RHR loop OPERABLE. the intended redundancy for! decay heat removal is lost and, in the event of a loss of the remaining RHR loop, it would be safer to
-1nitiate that loss from MODE 5 (s 200 F) rather than MODE 4 . (200 to 350 F). The Completion Time of 24 hours is a reasonable time. based on operating experience to reach MODE 5 from MODE 4 in an orderly manner and without L challenging plant systems.
C.1 and C.2 If no loop is OPERABLE. all operations involving a reduction of RCS boron concentration must be suspended and action to restore one RCS or RHR loop to OPERABLE status must be l initiated. Boron dilution requires forced circulation to
- provide proper mixing and preserve the margin to
! criticality. The immediate Completion Times reflect the importance of maintaining the capability for decay heat removal. . L l BYRON'- UNITS 1 & 2 B 3.4.6 - 4 8/21/98 Revision A ; I
. - . _ __ _ - ., - - . _ _ . _ . . ~ .. _ - _ __
n, - 1 RCS Loops.-MODE 4- l B 3.4.6
< l "d BASES-
- M L
1
,. SURVEILLANCE. 'SR 3.4.6.1 REQUIREMENTS .
I
- This SR requires-verif1 cation every 12 hours that the
_ required operating RCS or RHR loop is'in operation, j Verification.may . include flow rate. tem)erature, or pump ; status monitoring which helps ensure' t1at forced flow is providing heat removal. The Frequency of 12 hours is
- sufficient consicering other indications and alarms available to the operator in the control room to monitor RCS-and RHR loop performance.
g SR '3.4.6.2 kT' - SG SR OPERABILIT 3.4.6.2 rehuires verification is verified of required by ensuring thatSG theOPERABILITY secondary l
*l 4
side narrow range water-level is a 18% for each required RCS loop. If'the SG secondary side: narrow range water level is j
- }l - < 18%, the tubes may become uncovered and the associated !
loop may not be. capable of providing the heat sink necessary i for removal 1of. decay heat. The 12 hour Frequency is I
. considered adequate in view'of other indications available !
. in the control room to alert the operator to the' loss of SG !
level. l 'X) ' SR 3.4.6.3 Verification that the required pum) is 0PERABLE ensures that
' an' additional RCS or RHR pump can 3e placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. . Verification is performed by verifying proper
, breaker alignment and power available to-the required pump.
The Frecuency of 7 days is considered reasonable in view of other acministrative controls available and has been shown to be acceptable by operating experience. REFERENCES
- None.
V yLE BYRON - UNITS 1 & 2 B 3.4.6- 5 8/22/98 Revision K t + - - *e - - t -
L RCS~ Loops-MODE 5. Loops Filled B 3.4.7 1
- B 3.4 -REACTOR COOLANT SYSTEM (RCS)
;B'3.4.7: RCS Loops-MODE 5, Loops Filled 1
BASES l L l i BACKGROUND -In MODE _5 with the RCS loops filled, the primary function of ) l the reactor coolant is-the removal of decay heat and transfer this heat either to the Steam Generator (SG) secondary side coolant via natural circulation (Ref.1) or l- -the component cooling water via the Residual Heat Removal F (RHR) heat exchangers. While the principal means for decay heat removal is via the RHR System, the SGs via natural [ circulation are specified as a backup means for redundancy. L Even though the SGs cannot produce steam in this condition. , they are capable'of being a heat sink due to' their large l contained volume'of. secondary water As long as the SG-secondary side water is at a -lower temperature than the 4 L reactor coolant, heat transfer will occur. The rate of heat I
- transfer is directly proportional to the temperature l
l difference. The secondary function'of the reactor coolant
'is to act as a carrier for soluble neutron pcison boric aci0.
I In MODE 5 with the RCS loops filled, the reactor coolant is L , circulated by means of two RHR loops connected to the RCS. each loop containing an RHR heat exchanger, an RHR pump, ana appropriate flow and temperature instrumentation for control, protection, and indication. One RHR pump ! circulates the water through the RCS at a sufficient rate to prevent boric acid stratification. The number of loops in operation can vary to suit the l operational needs. The intent of this LCO is to provide L fvced flow from at least one RHR loop for decay heat removal and transport. The flow adequate for decay heat removal. Theprovided by one other intent RHR loop is of this LCO is to require that a second path be available to provide H ' redundancy for heat removal. L The second path can be another OPERABLE RHR loop or two l' OPCRABLE SGs to provide a' alternate method for decay heat removal via natural circu, sion. L L 1 e ~) BYRON UNITS 1 & 2 B 3.4.7 - 1 8/21/98 Revision A 1 i
RCS Loops-MODE 5. Loops Filled B 3.4.7
"') BASES O
APPLICABLE In MODE 5. RCS circulation increases the time available for SAFETY ANALYSIS mitigation of an accidental boron dilution event. The RHR loops provide this circulation and have been identified as important contributors to risk reduction. RCS Loops-MODE 5. Loops Filled. . satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii). t) Y N h LCO The purpose of this LC0 is to require that at least one of
-@ Nv the RHR locas be OPERABLE and in operation with an additional RHR loop OPERABLE or two SGs with secondary side t water level a 18%. One RHR loop provides sufficient forced f circulation to )erform the safety functions of the reactor 46 coolant under taese conditions. An additional RHR loop is k6 required to be OPERABLE to meet ~ single failure w considerations. However, if the standby RHR loop is not y? cPERABLE. an acceptable alternate method is two SGs with Nl -0 their secondary side water levels a 18%. Should the operating RHR loop fail, the SGs via natural circulation could be used to remove the decay heat.
f) v Note 1 permits all RHR pumps to be removed from operation
.l s 1 hour per 8 hour period. The purpose of the Note is to permit tests designed to validate various accident analyses values. One of the tests performed during the startup ;
testing program is the validation of rod drop times during l cold conditions, both with and without flow. The no flow l test may be performed in MODE 3. 4. or 5 and requires that i the pumps be stopped for a short period of time. The Note l permits stopping of the pumps in order to perform this test and validate the assumed analysis values. If changes are made to the RCS that would cause a change to the flow characteristics of the RCS. the input values must be I revalidated by conducting the test'again. The 1 hour time period is adequate to Jerform the test, and operating i experience has shown tlat boron stratification is not likely l during this short period with no forced flow. l l _r LJ j BYRON - UNITS 1 & 2 B 3.4.7 - 2 10/13/98 Revision K l
, .. - - . . . - ,. . . ~ .. . ..
l RCS- Loops-MODE 5. Loops Filled B 3.4.7 N: -BASES: Q '
.LCO (continued)=
1
' Utilization of Note 1 is ' permitted'provided the following. _ i conditions.are' met. along with any other conditions imposed. ;
by. procedures:
- a. No operations.are permitted that would dilute the RCS j boron concentration. therefore maintaining the margin to criticality. Boron. reduction is prohibited because
- M- l a uniform. concentration distribution throughout the ;
l rti .- RCS cannot be ensured when in natural circulation; and
,C .
l
'm
- b. Core outlet temperature-is maintained at least 10*F j below saturation-temperature, so that no vapor bubble imay form and-possibly cause a natural circulation flow obstruction.
Note.2 allows one RHR loop to be inoperable for a period of s 2 hours, provided that the other RHR loop is OPERABLE and ~ in operation. This permits periodic surveillance tests to be performed on the inoperable loop when such testing is safe-and possible.
,s Note 3 requires that the secondary side water temperature of 1 l each SG be < 50 F above each of the RCS cold leg V( temperatures before the start of a Reactor Coolant Pump )
l (RCP) with an'RCS cold leg temperature s 350 F. This ! restriction'is to prevent a low temperature overpressure event due to a thermal transient when an RCP is started, i Note 4 provides for an orderly' transition from MODE 5 to MODE- 4 during a planned heatup by permitting removal of RHR' ' loops from operation.when at least one RCS loop is in .
.o)eration. This Note provides for the transition to MODE 4 :
w1ere an RCS loop is permitted to be in operation and replaces the RCS circulation function provided by the RHR L loops. An OPERABLE RHR loop is comprised of an OPERABLE RHR pump capable of providing forced flow to an OPERABLE RHR-heat exchanger. RHR pumps are OPERABLE if they are capable of being Jowered and are able to provide flow if required. An i OPERAB_E SG via natural circulation has' greater than or l , equal to _the minimum water level .specified in SR 3.4.7.2 and is otherwise capable of providing the necessary heat sink via natural circulation, q k J_ . BYRON - UNITS 1 & 2 B 3.4.7 - 3 10/5/98 Revision K i l y+r,.e.+
R I RCS ' Loops-MODE 5. Loops Filled B 3.4.7-
. m .
> gi 1 BASES l ' lwf I LAPPLICABILITY- In MODE 5 with RCS loops-filled;:this LCO requires forced 1
-circulation'of the reactor coolant to remove decay heat from the core and to provide proper boron mixing. One. loop of.
RHR provides sufficient circulation for these purposes. .; lHowever, one additional. RHR loop is required to be OPERABLE- l or the secondary side water level of at least two SGs is ; required;to be a.18%. I l 0peration in other MODES is covered by: LC0 3.4.4, "RCS Loops-MODES 1 and 2":
. ' LCO 3.4.5. ~"RCS Loops -MODE 3":
LC0 3.4.6, :"RCS Loops -MODE 4": ;
-LCO 3.4.8. "RCS Loops -MODE 5. Loops Not- Filled":
LCO 3.9.5 " Residual Heat Removal (RHR) and Coolant j Circulation-High Water Level"-(MODE 6): and LC0.3.9.6 " Residual Heat Removal-(RHR) and Coolant Circulation-Low Water Level" (MODE 6). ACTIONS L1 and' A. 2 [A{k
'L If.no required MR loop is in operation, except'during-conditions permitted by . Note 1. all operations-involving .a
. reduction of RCS boron concentration must be suspended and action toJrestore'one RHR loop to' operation must be immediately :1nitiated. Boron dilution requires forced circulation to provide proper mixing and preserve the margin
.to criticality. The immediate Completion Times reflect the
( importance'of maintaining operation for decay heat removal. (( B.1 and C.1-
?
h-
'If the required RHR loop is inoperable or the required SG(s) have secondary side water levels < 18%. redundancy for heat
, p' 3 removal is' lost. Action must be initiated immediately to ~
? restore either the required RHR loop to OPERABLE status or k:
-to restore the required SG secondary side water level (s).
~'
The Required Actions will restore an ave'lable alternate heat removal path. The immediate Completion Times reflect
.the importance of. maintaining the availability of two paths for heat removal.
l ) i (: . w. IBYRON - UNITS 1 &.2 B 3.4.7 - 4 10/5/98 Revision K b
- RCS Loops-MODE 5. Loops Filled B 3;4.7
~
< /~V LBASES' V 1 ACTIONS (continued) D.E D.2L1.-and D.2 1 If two required RHR loops are inoperable or the required RHR. . loo) and one or both SG secondary side water levels ~are not o witain limit (s), all operations involving a reduction of RCS. H A -boron concentration must-be sus) ended and action to restore l 4 one' RHR loop to. operation must se immediately initiated.or 1 p; . initiate action.to restore required SG secondary side water ! p levelato within limits. Boron dilution requires forced. g circulation to. provide proper mixing and preserve the. margin
- to' criticality. The immediate Completion Times reflect the ;
.importance of maintaining operation.for decay heat' removal.
i
' SURVEILLANCE- SR 3.4.7.1 REQUIREMENTS
, This SR requires verification every 12 hours that- the.
required operating ~ RHR loop.is in operation. Verification may include flow rate, temperature, or pump' status monitoring, which helps ensure that forced flow is providing heat: removal. The Frequency of 12 hours is sufficient considering 'other indications and alarms available to the
.~)'
[ L >
- - operator in the control room to monitor RHR loop
. performance.
Ien { SR 3.4.7.2 y Verifying that at least~ two SGs are OPERABLE by ensuring 4 their secondary side narrow range water 1evels are a 18% 4 ensures an alternate decay heat removal method via natural circulation in the event that the second RHR loop is not . OPERABLE. If both RHR loops are OPERABLE. this surveillance is not needed. The 12 hour Frequency is. considered adequate ca : in view of other indications available in the control room
. ,1 to alert the operator to the less of SG 1evel.
.+ .
Il 1 I i. L.. .
- v '
L BYRON - UNITS l'& 2. B 3.4.7 - 5 10/5/98 Revision K
1
' ~
' RCS ~ Loops-MODE 5. Loops Filled B 3.4.7 I' , . l
'2N o
- BASES Lh ' '
' SURVEILLANCE REQUIREMENTS:(continued) ..
SR' 3.4.7.3
,Verificationcthat a second RHR pump is OPERABLE, when required, ensures that an additional:pum) can be placed in zoperation, if needed, to maintain decay 1eatfremoval and
" reactor coolant circulation. Verification is performed by 3 verifying proper' breaker. alignment and power available to the.RHR pumpt If: secondary side. water level is a 18% in at i least1two SGs..this surveillance is not needed. The-Frequency of:7; days is. considered reasonable-in' view of. i other administrative controls.available and has been shown i toL be-acceptable by operating experience. ;
l REFERENCES '1. 'NRC Information Notice 95-35, " Degraded Ability of Steam Generators to Remove. Decay Heat by Natural Circulation," August 28._1995. 4 a a 1 r4 d 1A p k/ f i BYRON. -UNITS 1 & i?. B 3.4.7 - 6 10/5/98 Revision K
.\
j:
?- -
RCS Loops-MODE 5. Loops Not Filled B 3.4.8 (~T- B 3.4 REACTOR COOLANT SYSTEM (RCS) C/ B 3.4.8 RCS Loops-MODE 5. Loops Not Filled i BASES BACKGROUND In MODE 5 with the RCS loops not filled, the primary function of the reactor coolant is the removal of decay heat generated in the fuel, and the transfer of this heat to the component cooling water via the Residual Heat Removal (RHR) heat exchangers. The Steam Generators (SGs) are not available as a heat sink when the loops are not filled. The secondary function of the reactor coolant is to act as a carrier for the soluble neutron poison, boric acid. In MODE 5 with loops not filled. only RHR pumps can be used for coolant circulation. The number of pumps in operation can vary to suit the operational needs. The intent of this LCO is to provide forced flow from at least one RHR pump for decay heat removal and transport, and to require that two paths be available to provide redundancy for heat removal. [] APPLICABLE In MODE 5. RCS circulation increases the time available for
- SAFETY ANALYSES mitigation of an accidental boron dilution event. The RHR loops provide this circulation and have been identified as important contributors to risk reduction. The flow provided by one RHR loop is adequate for heat removal and for boron mixing.
RCS loops in MODE 5. Loops Not Filled, satisfies Criterion 4 of 10 CFR 50.36.(c)(2)(ii). 1 i O] BYRON - UNITS 1 & 2 B 3.4.8 - 1 8/21/98 Revision A
.-, - , . - . .. . - . - ~ - ..
- I RCS Loops-MODE 5. Loops Not Filled B 3.4'8 b BASES u
i LC0 iThe purpose of this LC0 is to require that at least two RHR-loops be OPERABLE and one of these loops be in operation. l
.An-OPERABLE loop is one that has the capability of
.g o .- transferring heat _from the reactor coolant at a controlled
' rate.. Heat cannot be removed via the RHR System unless 10 forced flow is used. A minimum of one running RHR pump
# meets the LCO requirement for one loop in operation; An i
% additional RHR loop is required to be OPERABLE to meet !
91 single failure considerations. t-
- j Note 1 permits all RHR pumps to be removed from cperation for s'1 hour. Utilization of Note'l is permitted provided R
.the'following conditions are met, along with any other 1 conditions imposed by procedures: i i
- a. No operations.are permitted that would dilute the RCS I y i-boron concentration, therefore maintaining the margin-to' criticality. Boron reduction is prohibited because 1
t7 a uniform concentration distribution throughout the 1 V- RCS cannot be ensured when in natural circulation: 6 y b. Core outlet temperature is-maintained at least 10 F below saturation temperature, so that no va or bubble- [uNx b/ may form and possibly cause a natural circu ation flow obstruction; and b-o
- c. No draining operations are-permitted that would ;
further reduce the RCS water volume. !
@ l x? Note 2 allows one RHR loop' to l'e inoperable for a 3eriod of I
% s 2 hours. -provided that the other loop is OPERABLE and in )
operation. This permits periodic' surveillance tests to be performed on the inoperable loop'when these tests are safe and possible. An OPERABLE RHR loop is comprised of an OPERABLE RHR pump , capable of providing forced flow to an OPERABLE RHR heat exchanger. RHR pumps-are OPERABLE if-they are capable of _
. being powered and are able to provide flow if required.
I
+.
e BYRON -' UNITS 1 & 2' B 3.4.8- 2 10/8/98 Revision K r
~
RCS Loops-MODE 5. Loops Not Filled. B-3.4,8- } Ba.SES APPLICABILITY In MODE- 5 with loops ~ not filled, this LCO requires core heat removal and coolant circulation by the RHR System; Operation in other MODES is covered by: n LCO 3.4.4. "RCS Loops -MODES 1 and 2": LCO 3.4.5. -"RCS Loops-MODE 3": LCO' 3.4.6. "RCS Loops - MODE 4" . LC0 3.4.7. "RCS Loops-MODE 5. Loops Filled"; LCO 3.9.5. " Residual Heat Removal.(RHR) and Coolant Circulation-High' Water Level" (MODE 6): and : LCO 3.9.6. " Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level" (MODE 6). L
- ACTIONS A.1 and A.2 If no RHR loop is in operation, except during conditions ~
permitted by Note-1. all operations involving a reduction of RCS boron concentration must-be suspended and action to restore one RHR loop to operation must be immediately initiated. Boron dilution requires forced circulation to O provide proper mixing and preserve the margin-to criticality. The immediate Completion Times reflect the importance of maintaining operation-for decay heat removal. 1.1 If only one RHR loop is OPERABLE. except during conditions permitted.by Note 2.-redundancy for decay heat-removal is lost and' action must be initiated immediately to restore a , I second loop to OPERABLE status. The immediate Completion Time reflects the importance of maintaining the availability , of two paths for heat removal. l-b i l:
- O o
-BYRON.- UNITS 1 & 2 B 3.4.8 - 3 8/21/98 Revision A 1-l
RCS Loops-MODE 5. Loops Not Filled
- B 3.4.8 l
/]-
w BASES ACTIONS (continued) C 1 and C.2 If no required RHR loops are OPERABLE. all operations involving a reduction of RCS boron concentration must be suspended and action must be initiated immediately to restore an RHR loop to OPERABLE status. Boron dilution requires forced circulation to provide proper mixing and preserve the margin to criticality. The immediate Completion Times reflect the importance of maintaining the capability for heat removal. SURVEILLANCE SR 3.4.8.1 REQUIREMENTS This SR requires verification every 12 hours that the required operating RHR loop is in operation. Verification may include flow rate, temperature, or pump status monitoring, which helps ensure that forced flow is providing heat removal. The Frequency of 12 hours is sufficient considering other indications and alarms available to the es operator in the control room to monitor RHR loop
/
u) performance. SR 3.4.8.2 Verification that a second RHR pump is OPERABLE ensures that an additional pump can be placed in operation if needed. to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to the required pumps. The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience. REFERENCES None. l l
'v BYRON - UNITS 1 & 2 B 3.4.8 - 4 8/21/98 Revision A
rr $ Pressurizer. l B 3.4.9
.B 3.4 REACTOR COOLANT SYSTEM (RCS) b(~4 B.3.4.9 Pressurizer
! BASES BACKGROUND :The pressurizer provides a point in the RCS where liquid and vapor are maintained in equilibrium under saturated conditions for pressure control purposes to prevent bulk-boiling in the remainder of the RCS. Key. functions include maintaining required primary system pressure 'during steady state operation, and limiting the pressure changes caused by reactor coolant thermal expansion and contraction during normal load transients. The pressure control components addressed by this LC0 include the pressurizer water level. the required heaters, and their controls and Engineered Safety Features (ESF) power. supplies. Pressurizer safety valves and pressurizer. power operated relief valves are addressed by LCO 3.4.10. " Pressurizer Safety Valves." and LC0 3.4.11.
" Pressurizer Power Operated Relief Valves (PORVs)."
respectively. YD ' The intent of the LC0 is to ensure that a steam bubble V exists.in the pressurizer dur.ing. MODES 1. 2. and 3 to , minimize the consequences of potential overpressure transients. The presence of a steam bubble is consistent with analytical assumptions. Relatively small amounts of noncondensible gases can inhibit the condensation heat transfer between the pressurizer spray and the steam, and
. diminish the spray effectiveness for pressure control.
BYRON-- UNITS 1 & 2 B 3.4.9 - 1 8/21/98 Revision A
Pressurizer B 3.4.9
'~T BASES BACKGROUND (continued)
Electrical immersion heaters, located in the lower section of the pressurizer vessel, keep the water in the pressurizer at saturation temperature and maintain a constant operating pressure. A minimum required available capacity of pressurizer heaters ensure.s that the RCS pressure can be maintained. The capability to maintain and control sptem pressure is important for maintaining subcooled conditiom in the RCS and ensuring the capability to remove core decay heat by either forced or natural circulation of redctor coolant. Unless adequate heater capacity is available. the hot, high pressure condition cannot be maintained indefinitely and still provide the required subcooling margin in the primary system. Inability to control the system pressure and maintain subcooling under conditions of natural circulation flow'in the primary system could lead to a loss of single phase natural circulation and decreased capability to remove core decay heat. The pressurizer heaters are powered from the non-Class 1E
~
buses. The pressurizer heaters are non-safety related.
? Plant design includes a total heater capacity of 1800 kW 4
+ that is divided into four groups, with separate controls for 1 m
(1 the~ proportional and backup grou]s. The non-Class 1E ESF w/ i buses servicing the pressurizer 1 eaters can be powered from g the Unit Auxiliary Transformer, the System Auxiliary e Transformer (SAT). or the emergency diesel generator by td, closing the ESF to non-ESF crosstie breaker. APPLICABLE- In MODES 1, 2. and 3. the LC0 requirement for a stoam bubble SAFETY ANALYSES is reflected implicitly in the accident analyses. Safety analyses performed for lower MODES are not liraiting. All analyses performed from a critical reactor condition assume the existence of a steam bubble and saturated conditions in the pressurizer. In making this assumption, the analyses neglect the small fraction of noncondensible gases normally present. Safety analyses presented in the UFSAR (Ref. 1) do not take credit for pressurizer heater operation: however, an implicit initial condition assumption of the safety analyses is that the RCS is operating at normal pressure. I BYRON - UNITS 1 & 2. B 3.4.9- 2 9/18/98 Revision K
l Pressurizer B 3.4.9 VN - BASES' O- APPLICABLE SAFETY ANALYSES'(continued).
.The-maximum pressurizer water level limit which ensures that a' steam-bubble exists in the pressurizer satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). Although the heaters are not specifically used in accident analysis..they provide the capability.to maintain subcooling in the long term-during-loss of. offsite power, as indicated in NUREG-0737
-(Ref; 2), and thus, satisfy Criterion 4 of
.10 CFR 50.36(c)(2)(ii).
LCO The LCO requirement for the pressurizer to be OPERABLE with
'a water volume 5 1656 cubic feet, which is equivalent to .1 s 92%; ensures that a steam bubble exists. Limiting the 1 LCO maximum operating water level preserves the steam space i for pressure-control. The LCO has been established to l ensure the capability to' establish and maintain pressure control for steady state operation and to' minimize the consequences of potential overpressure transients.
T
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Requiring the presence of a steam bubble is also consistent with analytical assumptions. -
.m i f i y1 The LCO requires two groups of OPERABLE pressurizer heaters.
L' g each with a capacity a 150 kW capable of being powered from gel redundant ESF power supplied buses. Since the only safety function for pressurizer heaters ~is in a loss of offsite power condition, normal power is not required for OPERABILITY. The minimum heater capacity required is
-sufficient to maintain the RCS near normal operating pressure when accounting for heat losses through the pressurizer insulation. By maintaining the pressure near the operating conditions. a wide margin to subcooling can be
'obtained in the loops. The value of 150 kW is derived from generic evaluation of Westinghouse pressurizer heat loss calculations (Ref. 3).
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-BYRON -' UNITS 1 & 2 B 3.4.9 - 3 8/22/98 Revision K L _
.. ., _ _ _ _ _ _ _ . . _ . . - _ . _ - . _ . _ _ _ _ _ . ~ . . _
Pressurizer B 3.4.9
' BASES' L APPLICABILITY The need for pressure. control is most pertinent when core heat-can cause the greatest effect on RCS temperature, resulting in the greatest effect on pressurizer level and E RCS' pressure control. Thus. applicability has been designated for MODES 1 and 2. The applicability'is also provided for MODE 3. The purpose is to prevent solid water
.RCS operation during heatup and cooldown to avoid rapid pressure rises caused by normal operational perturbations.
such as reactor coolant pump startup. In MODES 1. 2. and 3. there is need to maintain the availability of pressurizer heaters, ca)able of being powered from an ESF power supply. In t1e event of a loss of offsite power the~ initial conditions of these MODES give the greatest demand for maintaining the RCS in a hot pressurized condition with loop subcooling for an extended _ period. For MODE 4, 5. or 6. it is not necessary to control pressure (by heaters) to ensure loop subcooling for heat transfer when the Residual Heat Removal (RHR) System is in service. and therefore. the LCO is not applicable. ACTIONS- .A.L1. A.2. A.'3.'and A.4 Pressurizer water level control malfunctions or other plant evolutions may result in a pressurizer water level above the nominal upper 'imit, even with the unit at steady state conditions. In MODE 1 at > 10% RTP (P-7), the unit will trip since the' upper limit of this LCO-is the same as 'the :
. Pressurizer Water Level-High Trip. J If.the pressurizer water level is not within the limit.
action must be taken to bring the plant to a MODE in which the LC0 does not apply. To achieve this status, within 6 hours the unit must be brought to MODE 3, with all rods fully inserted'and incapable of withdrawal. Additionally, the unit must be brought to MODE 4 within 12 hours. This takes the unit out of the applicable MODES. The allowed Completion Times are reasonable. based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging plant systems. 4 f t l i !~
' BYRON - UNITS 1 & 2 B 3.4.9 - 4 8/21/98 Revision A t
t Pressurizer i B 3.4.9 f} ? BASES; l ACTIONS (continued): u
.If the required groups of. press' izer heaters are inoperable, restoration is required within 72 hours. The I c9.. Completion Time. of.72 hours is reasonable considering the j
't -
antici)ation that a demand caused by loss of offsite power T would 3e unlikely-in this period. Pressure control may be ! maintained durin] this time using the remaining pressurizer 4( heater. capability. l C.1'and C.2
~
If Required Action B.1 and its associated Completion Ti p are not met, the' unit must be brought to a MODE.in which the-LC0 does not a) ply. To achieve this status, the unit must- > be brought to iODE'3 within 6 hours and to MODE 4 within 12 hours. The allowed Completion Times are reasonable.. ! H based on operating experience, to reach the recuired unit- ; conditions from full power conditions in an orcerly manner i and without challenging plant systems.
- p. . .. .
1
. SURVEILLANCE .SR 3.4.9.1~
c,jii REQUIREMENTS This SR requires that during steady state operation y pressurizer-level .is maintained below the nominal upper
'-Q limit to provide a minimum space for a steam bubble. The Surveillance is performed by observing the indicated level.
The Frequency of 12' hours corresponds to verifying the
)arameter each ' shift. The 12 hour interval has been shown ay operating' practice to be sufficient to regularly assess level for any deviation and verify that operation is consistent with the safety analyses-assumption of ensuring that a steam bubble exists in the pressurizer. Alarms are also available for early detection of abnormal level indications. >
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( . BYRON - UNITS 1 & 2 B 3.4.9 - 5 8/22/98 Revision K L a
[ A 1 Pressurizer-B 3.4.9 ((h BASES ' Af 1 SURVEILLANCE REQUIREMENTS (continued) LSR 3.4 9.2-The SR is satisfied when the power supplies are demonstrated of .to be capable of producing the minimum power and the 1 associated pressurizer heaters are verified to be* 150 kW.
- i. This is performed by energizing the heaters and measuring -
M< circuit current; The Frequency of 18 months is considered
. Hl adequate to detect' heater degradation and has been shown by j operating experience to be acceptable.
SR 3.4.9.3 This Surveillance demonstrates ~that the heaters-can be manually transferred from the normal non-ESF power supply to the ESF power supply and energized. .The Frequency of 18 months is ' based on a typical fuel cycle and is consistent (with similar verifications of-ESF- power supplies. I REFERENCES' 1. UFSAR. Chapter 15.- NOREG-0737, " Clarification of TMI Action Plan
~
T'i 2. b 'r Requirements." November 1980.
- 3. Westinghouse Ownert Group Study. " Emergency ~ Power il Supply- Requirements for the Pressurizer Heaters."
.1 transmitted.via B.'L.. King to C. Reed. TMI-0G-83.
- .;, -September 26. 1979.
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BYRONL-LUNITS 1 & 2 B 3.4.9 - 6 9/18/96 Revision K s
Pressurizer Safety Valves B 3.4.10 Q. B 3.4 RE.CTOR COOLANT S'/ STEM (RCS) V B 3.4.10 Pressurizer Safety Valves BASES BACKGROUND The pressurizer safety valves provide. in conjunction with ) the Reactor Protection System. overpressure protection for i the RCS. The pressurizer safety valves are totally enclosed ! pop type, spring loaded, self actuated valves with
~ backpressure compensation. The safety valves are designed I to prevent the system pressure from exceeding the system !
Safety Limit (SL). 2735 psig. which is 110% of the design pressure. j Because the safety valves are totally enclosed and self l actuating. they are considered independent components. The relief capacity for each valve. 420.000 lb/hr. is based on postulated overpressure transient conditions resulting from a complete loss of steam flow to the turbine. This event results in the maximum . surge rate into the pressurizer. which specitles the minimum relief capacity for the safety valves. The discharge flow from the pressurizer safety i valves is directed to the pressurizer relief tank. This ! r O cT discharge flow is inoicated by an increase in temperature ! downstream of.the pressurizer safety valves or increase in the pressurizer relief tank temperature or level. Overpressure protection is recuired in MODES 1. 2. 3, 4 and 5: however, in MODES 4 anc 5. and in MODE 6 with the : reactor vessel head on. overpressure protection is provided by operating procedures and by meeting the requirements of LCO 3.4.12. " Low Temperature Overpressure Protection (LTOP) System." The upper and lower pressure limits are based on the 1% tolerance requirement (Ref.1) for lifting pressures above 1000 psig. The lift setting is for the ambient conditions associated with MODES 1. 2. and 3. This requires either that the valves be set hot or that a correlation between hot and cold settings be established. (O-BYRON - UNITS 1 & 2 B 3.4.10 - 1 8/21/98 Revision A
r j L Pressurizer Safety Valves B 3.4.10 BASES
. BACKGROUND (continued)
, The pressurizer safety valves are part of the primary !.4 success.. path and mitigate the effects of postulated 1 accidents. OPERABILITY of the safety valves ensures that the RCS pressure will be limited to 110% of design pressure. The consequences of exceeding the American Society of Mechanical Engineers (ASME) pressure limit (Ref. 1) could include damage to RCS components.' increased leakage. or a requirement to perform additional stress analyses prior to resumption of reactor operation. 1
' APPLICABLE All accident and safety analyses in the UFSAR (Ref. 2) that SAFETY ANALYSES require safety valve actuation assume operation of three pressurizer safety valves to limit increases in RCS pressure. The overpressure )rotection analysis (Ref. 3) is i also based on operation of t1ree safety valves. Accidents that could result in overpressurization if not properly terminated include:
- a. Uncontrolled rod withdrawal from full power:
- b. Loss of reactor coolant flow:
- c. - Loss of external electrical load;
- d. Loss of normal feedwater:
- e. Loss of. all AC power to station auxiliaries:
- f. Locked rotor:'and g .- Feedwater line break.
Detailed analyses of. the above transients are contained in Reference 2. Safety valve actuation is required in events c.- d. and e (above) to limit the pressure increase.
-Compliance with this LCO is consistent with the design bases and accident analyses assumptions.
Pressurizer safety valves satisfy Criterion 3 of 1
'10 CFR 50.36(c)(2)(ii).
i' 1 LO l BYRON - UNITS'1 &'2 B 3.4.10 - 2 8/21/98 Revision A
Pressurizer Safety Valves B 3.4.10 ! p l : BASES l LCO The three pressurizer safety valves are set to open at the ll RCS design pressure (2500 psia). and within the ASME. ; I specified tolerance. to avoid exceeding the maximum design '
. pressure SL. to. maintain accident analyses assumptions, and to comply with ASME-requirements. The upper and. lower
. pressure tolerance limits are based on the 1% tolerance- ,
requirements (Ref.1):for lifting pressures above 1000 psig. i
.The limit protected by this Specification is the Reactor Coolant Pressure Boundary (RCPB) SL of 110%_of design
. pressure. Inoperability of one or more valves could result in exceeding the. SL 'if a transient were to occur. The
.consecuences of exceeding the ASME pressure. limit could incluce damage to one.or more RCS components, increased leakage. or additional stress analysis being required prior to resumption of reactor operation.
The Note allows entry into MODE 3 with the lift settings outside.the LC0 limits. This permits testing and examination of the safety valves at high pressure and temperature near their normal operating range, but only after the valves have had a preliminary cold setting. The cold setting.gives assurance that the valves are 0PERABLE near their design condition. Only one valve at a time will be. removed from service for testing. The 54 hour exception lsA) is based on 18 hour outage time for each of the three valves. T_he 18 hour period is derived from operating experience that hot testing can be performed in this time frame. l LAPPLICABILITY- In MODES 1; 2. and 3. OPERABILITY of three valves is i required because the combined capacity is required to keep ' reactor-coolant pressure below 110% of its design value during certain accidents. MODE 3 is conservatively included, although the listed accidents may not require the ! safety valves for protection. l The LCO is not applicable in MODES 4 and 5. and in MODE 6 with the reactor vessel head on. because Low Temperature Overpressure Protection-(LTOP) is provided. Overpressure
]rotection is not required in MODE 6 with reactor vessel lead detensioned.
l BYRON'- UNITS-1 & 2 8 3.4.10 - 3 8/21/98 Revision A
l Pressurizer Safety Valves B 3.4.10
/7 BASES
[ LJ ACTIONS' A1 With one pressurizer safety valve inoperable, restoration must take place within 15 minutes. The Completion Time of 15 minutes reflects the importance of maintaining the RCS Overpressure Protection System. An inoperable safety vahe coincident with an RCS overpressure event could challenge the integrity of the pressure boundary. B.1 and 8.2 If Required Action A.1 and its associated Completion Time are not met or if two or more pressurizer safety valves are inoperable. the unit must be brought to a MODE in which the requirement does not apply. To achieve this status. the unit must be brought to at least MODE 3 within 6 hours and to MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience. to reach the required unit conditions from full power conditions in an orderly manner and without challenging plant systems. In MODE 4. overpressure protection is provided by the LTOP System. The change from MODE 1. 2. or 3 to MODE 4 reduces
,, the RCS energy (core power and pressure). lowers the potential for large pressurizer insurges, and thereby 'v) t.
removes the need for overpressure protection by three ; pressurizer safety valves. SURVEILLANCE SR 3.4.10.1 REQUIREMENTS SRs are specified in the Inservice Testing Program. Pressurizer safety valves are to be tested in accordance with the requirements of Section XI of the ASME Code (Ref. 4), which provides the activities and Frequencies necessary to satisfy the SRs. No additional requirements are specified. l The pressurizer safety valve setpoint is 1% of a nominal 2485 psig. l l () !v BYRON - UNITS 1 & 2 B 3.4.10 - 4 8/21/98 Revision A
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i Pressurizer Safety Valves B 3.4.10 i [f'Y {- V BASES REFERENCES 1. ASME. Boiler and Pressure Vessel' Code, Section III.
- 2. UFSAR. Chapter 15.
- 3. WCAP-7769. Rev. 1. June 1972.
! 4 .. ASME. Boiler and Pressure Vessel. Code, Section XI. l Vl3 !O l'
~J 8/21/98 Revision A BYRON - UNITS 1 & 2 B 3.4.10 - 5
l Pressurizer PORVs B 3.4.11 , 1 o (N i-B 3.4 REACTOR COOLANT SYSTEM (RCS) B'3.4.11' Pressurizer l Power Operated Relief Valves-(PORVs) l l l
. BASES '
BACKGROUND The pressurizer-is equipped with two types of devices for i pressure relief: pressurizer safety valves and PORVs. The PORVs are' air. operated valves that.are controlled to open at a specific set pressure when the pressurizer pressure
-increases and close when the pressurizer pressure decreases.
The PORVs~may also be manually operated from the control l room. ! Block valves, which are normally open, are located between I
~the pressurizer and the PORVs. The block' valves are used to
)
. isolate the.PORVs in case of excessive leakage or a stuck l open PORV. Block valve closure is accomplished manually using controls in the control-room. A stuck open PORV is, in effect a small break Loss Of Coolant Accident (LOCA).
As such, block valve closure terminates the RCS depressurization and coolant inventory loss.
-The PORVs and their associated block valves may be used by tO plant operators to depressurize the RCS to recover from d certain transients if normal pressurizer spray is not available. Additionally, the series arrangement of the PORVs and'their block valves permit performance of-surveillances on the valves during power operation.
The PORVs may also be used for feed and bleed core cooling in the case of multiple equipment failure events that are not within the design basis, such as a total loss of _feedwater.
.L'8
. The PORVs. their block valves, and their controls are T powered from the vital buses that normally receive power
- from offsite power sources, but are also capable of being 4 powered from emergency power _ sources in the event of a loss
- h. of offsite power. Two PORVs and their associated block valves are powered from two separate safety trains (Ref. 1).
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~ BYRON - UNITS 1~& 2- B 3.4.11 - 1 8/22/98 Revision K
Pressurizer PORVs B 3.4.11 f'j - BASES BACKGROUND (continued) l The unit has two PORVs. each having a relief capacity of 210.000 lb/hr at 2350 psia. The functional design of the PORVs is based on maintaining pressure below the Pressurizer Pressure-High reactor-trip setpoint following a step , reduction of 50% of full load with steam dump. In addition. l the PORVs minimize challenges to the pressurizer safety ' valves and also may be used for Low Temperature Overpressure l Protection (LTOP). .See LCO 3.4.12. " Low Temperature l Overpressure Protection (LTOP) System." APPLICABLE Plant operators employ the PORVs to depressurize the RCS in I SAFETY ANALYSES response to certain unit transients if normal pressurizer spray is not available. For the Steam Generator Tube Rupture (SGTR) event, the safety analysis assumes that l manual operator actions are required to mitigate the event. 1 If a loss of offsite power is assumed to accompany the event, normal pressurizer spray is unavailable to reduce RCS pressure. The PORVs are assumed to be used.for RCS depressurization, which is one of the steps performed to _.x equalize the primary and secondary aressures in order to (
) terminate the primary to secondary areak flow and the
- radioactive releases from the affected steam generator. PORVs are also credited for automatic pressure control during recovery from an inadvertent safety injection (SI).
}A While the automatic pressure control is assumed to function for this event, the PORV(s) can initially be in " manual" if 7 available for the operator to place in " auto" upon 7 recognition of the inadvertent SI (Ref. 5). Automatic
[ operation of the PORVs to control reactor coolant system c pressure reduces challengos to the pressurizer safety valves S during an inadvertent S1 at power.
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BYRON - UNITS 1 & 2 B 3.4.11 - 2 8/22/98 Revision K
Pressurizer PORVs y B 3.4.11
- W BASES Lt i !
APPLICABLE: SAFETY ANALYSES (continued)
'The-PORVs are also modeled in safety analyses for events that result in increasing RCS pressure for which Departure.
from Nucleate Boiling Ratio (DNBR) criteria are critical 1 (Ref. 2). .=By assuming PORV actuation, the primary pressure-remains below the high pressurizer pressure trip setpoint: ! thusL the DNBR calculation is more conservative. As such, this actuation is not required to mitigate.these. events, and i this PORV automatic operation is, therefore, not required to .; r- support an assumed safety function. ! i Pressurizer.PORVs satisfy Criterion 3 of '
- 10 CcR 50.36(c)(2)(ii).. .
'LC0 The LC0 requires both manual and automatic operation of the PORVs and manual operation of their associated block valyes to be OPERABLE to mitigate the effects associated with an SGTR and inadvertent SI. However placing the PORV(s) in l 6 " manual" is acceptable..and the PORV may continue to be ~
-OPERABLE if. the automatic circuitry remains OPERABLE and the
- x. operator remains capable of returning the PORV(s) to " auto."
(Vli ' L By-maintaining two PORVs and their associated block valves
.i OPERABLE, the single failure criterion is satisfied. - An
? OPERABLE block valve may be either open, or closed and
- - energized with the capability to be osened, since the
> required safety function is accomplis 1ed by manual
.m o)eration. Although typically open to allow PORV operation, u- tie block valves may be OPERABLE when closed to isolate the
- j. flow path of an inoperable PORV that is capable of being xl manually.and' automatically cycled (e.g. as in the case of excessive PORV leakage). Similarly, isolation of an OPERABLE PORV does not render that PORV or block valve ino)erable provided the relief function remains available
-l = wit 1 manual and automatic action.
An OPERABLE PORV is required to be capable of manually and automatically opening and closing, and not experiencing
. excessive seat leakage. Excessive seat leakage although not associated with a specific acceptance criteria, exists when conditions,di' tate closure of the block valve to limit
! leakage. n A.f LBYRON - UNITS.1L& 2- B 3.4.11 - 3 8/22/98 Revision K
Pressurizer PORVs B 3.4.11
< S BASES LC0 (continued)
Satisfying the LCO helps minimize challenges to fission product barriers. APPLICABILITY ~In MODES 1. 2. and 3. the PORV and its block valve are required to be OPERABLE to limit the potential for a small break LOCA through the flow path. The most likely cause for a PORV small break LOCA is a result of a pressure increase transient that causes the PORV to automatically open. Imbalances in the energy output of the core and heat removal by the secondary system can cause the RCS pressure to r- increase to the PORV opening setpoint. The most rapid
? increases will occur at the higher operating power and
- pressure conditions of MODES 1 and 2. The PORVs are also f required to be OPERABLE in MODES 1, 2. and 3 for manual m actuation to mitigate a steam generator tube rupture event y and for automatic actuation to mitigate an inadvertent SI
% event.
Pressure increases are less prominent in MODE 3 because the
-core input energy is reduced, but the RCS pressure is high.
, j Therefore, the LC0 is applicable in MODES 1. 2 and 3. The w/ LC0 is not applicable in MODE 4. 5. or 6. when both pressure and core ~ energy are decreased and the pressure surges become much less significant. LC0 3.4.12 addresses the PORV requirements in MODES 4 and 5. and in MODE 6 with the reactor vessel head in place. (~ BYRON - UNITS 1 & 2 B 3.4.11 - 4 8/22/98 Revision K
l Pressurizer PORVs > B 3.4.11 1 O. BASES
- (d ACTIONS L
An ACTION Note 1 has been added to clarify that all . pressurizer PORVs and block valves are treated as separate
- w- entities, each with separate. Completion Times (i.e., the t Completion Time is on a. component basis). -The exception for >
LC0 3.0.4. Note 2.. permits entry into MODES 1. 2. and 3 to
'd perform cycling of the PORVs or block valves to verify their .,
Ql_ OPERABLE status. Testing is not performed in lower MODES.
; y b.1 I PORVs may be inoperable but capable of being manually and M
H< automatically cycled (e.g. excessive seat leakage). In this condition, either the PORVs must be restored or the flow path isolated within 1 hour. The associated block valve is required to be closed but power must be maintained
-to the associated block valve, since removal of power would render the block valve inocerable. This permits operation of the unit until the next refueling. outage (MODE 6) so that I maintenance can be' performed on the PORVs to eliminate the j problem condition.
B.1. B.2 and B.3 If one PORV is inoperable and not capable of being manually
?(mAJ cycled, it must be either restored, or isolated by closing the associated block valve and removing the power to the associated block valve. The Completion Times of 1 hour are reasonable.. based on challenges to the PORVs during-this
' time period, and provide the operator adequate time to correct the situation. If the-inoperable valve cannot be restored to OPERABLE status. it must be isolated within the specified time. Because there is at least one PORV that remains OPERABLE. 72 hours'is provided to restore the inoperable PORV to OPERABLE status. If the PORV cannot be
. restored within this time, the unit must be brought to a MODE in which the LC0 does not apply, as required by Condition D.
$ Required Action B.2 is modified by a Note stating that 7 removing power from the block valve is not required unless T* the associated PORV is inoperable due to being incapable of being manually cycled. In the event the PORV-is inoperable H solely due to inoperable automatic cyclin capability, it is I desiredforpowertoberetainedtotheb$ockvalvetoallow more readily accessible manual relief capabilities.
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- BYRON - UNITS 1:& 2 B 3.4.11 - 5 8/22/98 Revision K i
Pressurizer PORVs : B 3.4.11 l l-g/ y BASES TACTIONS.(continued). C.1. and C.2 4 If one block valve is inoperable, then it is necessary to either restore the block valve.to OPERABLE status within the
' Completion Time of 1 hour or place the associated PORV in U ,
- manual control. The prime importance for the capability to close the block valve is to isolate a stuck open PORV. 4 Therefore, if the block valve cannot be restored to OPERABLE - l status within 1 hour, the Required Action is to place'the ,
- PORV in manual control (i.e.. closed) to preclude its )
a automatic opening for an overpressure event and to avoid the '
> potential for'a stuck o)en PORV at a time that the' block i valve is ino)erable. T1e Completion Time of 1 hour is !
reasonable. Jased on the small potential for challenges to the system during this. time period, and provides the j 6 operator time to correct the situation.
. Because at least one PORV remains OPERABLE. the operator is '
. permitted a Com31etion Time of 72 hours to restore the inoperable blocc valve to OPERABLE status. The time allowed to restore-the block valve is based upon the Completion Time
.for' restoring an inoperable PORV in Condition B. since the PORVs may not be.ca able of mitigating an event if the
>[s}b inoperable block va ve'is not full-open. If the block valve '
- is restored within the Completion Time of 72 hours. the
. power will be restored, and the PORV restored to OPERABLE status. If it cannot be restored within this additional time the unit must be brought to a MODE in which the LCO does not apply.'as-requi. red by Condition D.
~
D.1 and D.2
~# If the Required Action'of Condition A. B. or C is.not met.
T then the unit must be brought to a MODE in which the LCO does not apply. -To achieve this status, the unit must be j brought to at least MODE 3 within 6 hours and to MODE 4 w within 12 hours. The allowed Completion Times are reasonable? based on operating experience, to reach the recuired plant conditions from rull power conditions in an orcerly_ manner and without challenging plant systems. In l ' MODE 4. 5. and 6 with the reactor vessel head on, automatic p PORV OPERABILITY may be required. See LCO 3.4.12. X C
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BYRON - UNITS 1 & 2 B 3.4.11 - 6 9/18/98 Revision K T
E Pressurizer PORVs B 3.4.11 , ~BASES-I ' (\.~} J ' v-
. ACTIONS (continud) ;
I 7l -E.'1'and E.2 , J . .
. l MO LIf two PORVs are inoperable and not capable of being e --
manually-and automatically cycled. Condition B and its 1 C 7' associated Required Actions would already be entered. .The. i Required Actions.would either-restore at least one valve-within the Completion Time of 1 hour or isolate the flow l 3ath by closing and removing the power to the associated , alock valvo . The Completion Time of 1 hour is reasonable. based on the smalll potential for challenges to the system during this time and provides the operator time to' correct ,
*- the situation. If no PORVs are restored within the .i
~ 'J
- Completion Time, then the unit must be brought to a MODE in.
which the:LCO does not apply; To achieve this' status. the
=,
.M unit must be brought'to at least MODE 3 within 6 hours and to_ MODE 4 within 12 hours. The allowed Completion Times are 'l I
" { **I reasonable, based on operating experience, to reach the I recuired unit conditions.from full power conditions in an orcerly manner and without challenging plant systems ~. In
_gl MODE 4. 5. and 6 with the reactor vessel head on.-automatic o PORV OPERABILITY may.be required. See LC0 3.4.12. l 5:, f~y , g D- y t If two block valves are-inoperable, it is necessary to-N restore at least one block valve within 2 hours. The
-@~ Completion Time is reasonable, based on the small potential for challenges to the system during this time and provide
- the: operator time to-correct the situation.
G.1 and G.2.
" If the Required Actions'of Condition F are not met, the unit
.y : -must be brought to a MODE in which the LC0 does not apply.
g' "i To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating 4 experience, to reach the required unit conditions from full
. power conditions in an _ orderly manner and without b.2 challenging plant systems .In MODE 4. 5. and 6 with the reactor vessel head on. automatic PORV OPERABILITY may be required. See LC0 3.4.12.
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', BYRON'- UNITS-1 &.2 B 3.4.11 - 7 10/13/98 Revision K
p Pressurizer PORVs B 3.4.11' fy . BASES. l;Q-L l SURVEILLANCE SR 3.4.11.1 REQUIREMENTS L
] Block valve cycling verifies ~ that the valve (s) can be opened -
and closed if needed. 'The basis for the Frequency of l- l 4;; - 92 days is the ASME Code. Section-XI (Ref. 3). ; ~ m 1 L
'"lW The Note modifies this SR by stating that it is not required j
-to be met'with the block valve closed in accordance with the i
!l Required Actions of this LCO. If-the block valve is. closed I
- i. to isolate an inoperable PORV that is incapable of being
'jl manually and automatically cycled the maximum Completion I s Time to restore the PORV and open the block valve-is i L 72 hours, which is well within the allowable limits (25%) to extend the block valve Frequency of 92 days. Furthermore; these test requirements would be completed by the reopening of a recently closed block valve upon restoration-of.the PORV to OPERABLE status (i.e. completion of the Required Actions fulfills the SR).
z SR -3.4.11.2
=. l SR 3.4.11.2 requires a complete c cle of each PORV. !
3,. -
- Operating a PORV through one comp ete cycle ensures that the Txi lV 4 PORV can be manually and automatically actuated for mitigation of an SGTR and inadvertent SI. The Frecuency of g
-18 months is based on a typical refueling cycle anc industry accepted practice.
The Note modifies the SR to allow entry into and operation ! in MODE 3 prior to performing the SR. This allows the test to be performed in MODE 3 under operating temperature and pressure conditions prior to entering MODE 1.or 2. In accordance with Reference 4. this test should be performed in MODE.3 or 4 to adequately simulate operating temperature and pressure effects on PORV operation. ! SR 3.4.11.3 Operating the solenoid air control valves and check valves on the air accumulators ensures the PORV control system actuates properly when called upon. The Frequency of 18-months is based on a typical refueling cycle and the Frequency of the other Surveillances used to demonstrate PORV OPERABILITY. fy V BYRON - UNITS 1 & 2 B 3.4.11 - 8 9/18/98 Revision K r
- f'
- . . . _ _ _ _ - __ . _ ~ _ _ _ _ . -- .
L Pressurizer PORVs B 3.4.11 'q . BASES'
.V '~
SURVEILLANCE REQUIREMENTS (continued) l , SR 3.4.11.4 SR 3.4.11.4 is the performance of-a CHANNEL CALIBRATION. A CHANNEL CALIBRATION is performed every 18 months. or approximately at every refueling. . CHANNEL CALIBRATION.is a
-complete check of the instrument loop, including the sensor.
The: test verifies that the channel responds to measured parameter within the necessary range and' accuracy.
'T 2 CHANNEL CALIBRATIONS must be performed consistent with the
;; assumptions of the plant specific setpoint methodology. The
,4 . difference between the current "as found" values and.the previous test "as-left" values must be consistent with the
_{ drift allowance used in the setpoint methodology.
' The Frequency of 18 months is based on the'. assumption of.an 18 month calibration interval in the determination of the magnitude.of equipment drift in the setpoint methodology.
. ~ REFERENCES- 1. Regulatory Guide 1.32. February 1977,
- 2. UFSAR. Section 15.2.
-pl- 3. . ASME. Boiler and Pressure Vessel Code. Section XI.
}
~
- 4. Generic Letter. 90-06. " Resolution of Generic Issue 70.
- . " Power Operated. Relief Valve and Block-Valve v Reliability." and Generic Issue 94. " Additional ^ Low
]li -Temperature Overpressure Protection for Light Water g Reactors." pursuant.to 10 CFR 50.54(f). June 25, 1990.
kl'. 5. UFSAR3 Section 15.5.1. '7x , k,I BYRON--' UNITS-1.& 2.
. B 3.4.11 - 9 9/18/98 Revision K e -f -
_ . _. ._ = - - _ _ - . -_ LTOP System B 3.4.12
- l N B 3.4 REACTOR COOLANT SYSTEM (RCS) i (d B 3.4.12 Low Temperature Overpressure Protection (LTOP) System i
j BASES 1 BACKGROUND The LTOP System controls RCS pressure at low temperatures so l the integrity of the Reactor Coolant Pressure Boundary (RCPB) is not compromised by violating the pressure and temperature (P/T) limits of 10 CFR 50. Appendix G (Ref.1). The reactor vessel is the limiting RCPB component for demonstrating such protection. The PTLR provides the ) maximum allowable actuation logic setpoints for the ; pressurizer Power Operated Relief Valves (PORVs) and the maximum RCS pressure for the existing RCS cold leg 1 temperature during cooldown, shutdown, and heatup to meet the Reference 1 requirements during the MODES in which LTOP is necessary. The reactor vessel material is less ductile at low temperatures than at normal operating temperature. As the vessel neutron exposure accumulates, the material toughness decreases and becomes less resistant to pressure stress at low temperatures (Ref. 2). RCS pressure, therefore, is , i ]c maintained low at low temperatures and is increased only 1 ( within the limits specified in the PTLR. The potential for vessel overpressurization is most acute when the RCS is water solid, occurring only while shutdown: ! a pressure fluctuation can occur more quickly than an ' operator can react to relieve the condition. Exceeding the RCS P/T limits by a significant amount could cause brittle cracking of the reactor vessel. LCO 3.4.3. "RCS Pressure and Temperature (P/T) Limits " requires administrative , control of RCS pressure and tem 3erature during heatup and ' cooldown to prevent exceeding t7e PTLR limits. l <n BYRON - UNITS 1 & 2 B 3.4.12 - 1 8/21/98 Revision A l
l LTOP System B 3.4.12 r" BASES- l N) - BACKGROUND (continued) 1 This LCO provides RCS overpressure protection by having a I minimum coolant input capability and having adequate i pressure relief capacity. Limiting coolant input capability requires all Safety Injection (SI) pumps and all but or,e charging pump (a centrifugal charging pump) incapable of 1 injection into the RCS and isolation of the 51 accumulators. The pressure relief capacity requires either two redundant I RCS relief valves or a depressurized RCS and an RCS vent of sufficient size. One RCS relief valve or the open RCS vent is the overpressure protection device that acts to terminate l an increasing pressure event. With minimum coolant input capability. the ability to provide core coolant addition is restricted. The LCO does not require the makeup control system deactivated or the SI actuation circuits blocked. Due to the lower pressures in the LTOP MODES and the expected core decay heat levels, the makeup system can provide adequate flow via the makeup control valve. If conditions require the use of more than I one centrifugal charging pump for makeup in the event of loss of inventory, then pumps can be made available through manual actions. ,{} < v'- The LTOP System for pressure relief consists of two PORVs with reduced lift settings, or two Residual Heat Removal (RHR) suction relief valves, or one PORV and one RHR suction relief valve, or a depressurized RCS and an RCS vent of sufficient size. Two RCS relief valves are required for redundancy. One RCS relief valve has adequate relieving capability to prevent overpressurization for the required coolant input capability. l l l A U ' l BYRON - UNITS 1 & 2 B 3.4.12 - 2 8/21/98 Revision A I p
( LTOP System ! B 3.4 12 l (~}
%j BASES BACKGROUND (continued)
PORV Reauirements As designed for the LTOP System. each PORV is signaled to open if the RCS pressure approaches a limit cetermined by the LTOP actuation logic. The LTOP actuation logic monitors both RCS temperature and RCS aressure and determines when a condition not acceptable in t1e PTLR limits is approached. The wide range RCS temperature indications are auctioneered to select the lowest temperature signal. ' The lowest temperature signal is processed through a function generator that calculates a pressure limit for that temperature. The calculated pressure limit is then compared with the indicated RCS pressure from a wide range pressure channel. If the indicated pressure meets or exceeds the calculated value, a PORV is signaled to open. The PTLR presents the PORV setpoints for LTOP. The setpoints are normally staggered so only one valve opens during a low temperature overpressure transient. Having the setpoints of both valves within the limits in the PTLR ' ensures that the Reference 1 limits will not be exceeded in i u) ( any analyzed event. When a PORV is opened in an increasing pressure transient. the release of coolant will cause the pressure increase to slow and reverse. As the PORV releases coolant, the RCS pressure decreases until a reset pressure is reached and the valve is signaled to close. The pressure continues to decrease below the reset pressure as the valve closes l l O lL) BYRON - UNITS 1 & 2 B 3.4.12 - 3 8/21/98 Revision A
LTOP. System B 3.4.12
./^V . BASES
- BACKGROUND 2(continued)
RHR Suction Relief-Valve Reauirements , -During LTOP MODES.. the RHR. System is operated for decay. heat removal.and low pressure letdown control. Therefore the: L RHR suction isolation valves are open in the. piping from the
. RCS hot legs to the inlets of the RHR. pumps. While these:
-< valves are open. the RHR suction relief valves'are exposed to the RCS and are able to relieve pressure transients in the RCS.
.The RHR suction isolation valves must ue open to make the-RHR suction relief valves OPERABLE for'RCS overpressure mitigation. .The RHR suction relief valves are spring .
loaded.. bellows type water relief valves with pressure tolerances and accumulation limits established by [ -Section III of the American Society of Mechanicai Engineers (ASME) Code (Ref. 3) for Class 2 relief valves. RCS Vent Reauirements-Once the RCS is dearessurized, a vent exposed to the
- containment atmosplere will maintain the RCS at containment
; V( \< ambient pressure in an RCS overpressure transient, if the relieving requirements of the transient'do not exceed the
-capabilities of'the vent. Thus, the vent path must be-capable of relieving the flow resulting from the limiting.
LTOP mass or heat input transient. and maintaining pressure below the P/T limits. The required vent capacity may be ' provided by one'or more vent. paths. For an RCS vent to meet the flow capacity requirement._-it-requires removing' a pressurizer safety valve removing a
~ PORV's internals, and disabling its block valve in the open
' position, or simi_larly establishing any comparable vent.
The vent path (s) must be above the level of reactor coolant,. so as not-to drain the RCS when open. x e s f( L1: ~
. BYRON-lUNIl>;&'2 B 3.4.12 - 4 8/22/98 Revision K J
b *
- e exv , , , , -
LTOP System B 3.4.12 ('T BASES
'% )
APPLICABLE Safety analyses (Ref. 4) demonstrate that the reactor vessel SAFETY ANALYSES is adequately protected against exceeding the Reference 1 P/T limits. In MODES 1. 2. and 3. the pressurizer safety valves.will prevent 1CS pressure from exceeding tne Reference 1 limits. In MODE 4 and below, overpressure prevention falls to two OPERABLE RCS relief valves or to a depressurized RCS and a sufficient sized RCS vent. Each of these means has a limited overpressure relief capability. Tne actual temperature at which the pressure in the P/T limit curve falls below the pressurizer safety valve setpoint increases as the reactor vessel material toughness decreases due to neutron embrittlement. Each time the PTLR curves are revised, the LTOP System must be re-evaluated to ensure its ftoictional requirements can still be met using the RCS relief valve method or the depressurized and vented RCS condition. The PTLR contains the acceptance limits that define the LTOP requirements. Any change to the RCS must be evaluated against the Reference 4 analyses to determine the impact of the change on the LTOP acceptance limits.
'(] Transients that are capable of overpressurizing the RCS are As categorized as either mass or heat input transients.
examples of which follow: Mass-Inout Tyne Transients
- a. Inadvertent safety injection; or
- b. Charging / letdown flow mismatch.
Heat Inout Tvoe Transients
- a. Inadvertent actuation of pressurizer heaters:
- b. Loss of RHR cooling: or
- c. Reactor Coolant Pump (RCP) startup with temperature asymmetry within the RCS or between the RCS and steam !
generators. ! l l l O l BYRON - UNITS 1 & 2 B 3.4.12 - 5 8/21/98 Revision A i 1 L
!- l l l LTOP System B 3.4.12 BASES iPPL.ICABLESAFETYANALYSES(continued) The following are required during the LTOP MODES to ensure i that mass and heat input transients do not occur, which l either of the LTOP overpressure protection means cannot handle:
- a. Rendering all SI pumps and all charging pumps but'one centrifugal charging pump incapable of injection:
l b. Deactivating the accumulator discharge. isolation / ,
. valves in their. closed. positions: and I
- c. . Disallowing start of an RCP-if secondary temperature I
.is more than 50*F above primary temperature in any one ,
loop. LC0 3.4.6. "RCS Loo >s-MODE 4." and LCO 3.4.7 l "RCS Loops-MODE 5.' Loops illed," provide this ; protection. j The Reference 4 analyses demonstrate that either one RCS l relief valve or the dearessurized RCS and RCS vent can ! maintain RCS pressure 3elow limits when only one centrifugal '
. charging pump is. actuated. Thus, the LCO allows only one. 3 centrifugal charging pump OPERABLE during the LTOP MODES.
'O O Since none of the overpressure protection methods can handle the pressure transient need from accumulator injection, when j RCS temperature is low,- the LC0 also requires the - accumulators isolation when accumulator pressure is: greater than or equal to the maximum RCS pressure for the existing RCS cold . leg temperature allowed in the PTLR, The isolated accumulators must have their discharge valves closed and the valve power supply breakers in their.open positions. I l i 1* ; i o . BYRON - UNITS _1 & 2 - B 3.4.12 - 6 8/21/98 Revision A
LTOP SysteT E 3 4 ':
*g\ BME:.
V ADPLICABLE SAFETY ANALYSE 5 (continusa) DOD t. Der #~ mance Tre fracture mechanics analyses snow tnat tne vessel is protected when the PORVs are set t: coen at er Deicr. :re limit snown in the PTLR. Tne setpoints are cerise cy analyses Inat model the performance Cf tne LIOD System assuming the limiting mass addition transient of one centrifugal charging pump injecting into a water soild RCS or the limiting heat input transient of the startup of an idle RCP with the secondary water in tne steam generator s 50 F above the RCS cold leg temperatures. These analyses consider pressure overshoot and underchcot beyond the PORV opening and closing, resulting from signal processing and valve stroke times. The PORV setpoints at or below the derived limit ensures the Reference 1 P/T limits will be met. The PORV setpoints in the PTLR will be updated. as necessary, when the P/T limits are revised. The P/T limits are periodically modified as the reactor vessel material toughness decreases due to neutron embrittlement caused by
. neutron irradiation. Revised limits are determined using neutron fluence projections and the results of examinations (n) " of the reactor vessel material irradiation surveillance specimens. The Bases for LCO 3.4.3 discuss these examinations.
The PORVs are considered active components. Thus, the failure of one PORV is assumed to represent the worst case. single active failure. .A () l BYRON - UNITS 1 & 2 B 3.4.12 - 7 8/21/98 Revision A
LTOP System B 3.4.12 r~' BASES
-k APPLICABLE SAFETY ANALYSES (continued) 1 RHR Suction Relief Valve Performance The RHR suction relief valves do not have variable pressure l and temperature lift setpoints like the PORVs. Analyses l must show that one RHR suction relief valve with a setpoint j s 450 psig will pass flow greater than that required for the limiting LTOP transient while maintaining RCS pressure less than the P/T limit curve. Assuming all relief flow requirements during the limiting LTOP event, an RHR suction relief valve will maintain RCS pressure to within the valve l
rated lift setpoint, plus an accumulation s 10% of the rated ! lift setpoint. As the RCS P/T limits are decreased to reflect the loss of toughness in the reactor vessel materials due to neutron embrittlement the RHR suction relief valves must be analyzed to still accommodate the design basis transients i for LTOP. l The RHR suction relief valves are considered active components. Thus, the failure of one valve is assumed to i
, represent the worst case single active failure. /s\
' () RCS Vent Performance With the RCS depressurized. analyses show a vent size of 2.0 square inches is capable of mitigating the allowed LTOP overpressure transient. The capacity of a vent this size is j greater than the flow of the limiting transients for the LTOP configuration, maintaining RCS pressure less than the l maximum pressure on the P/T limit curve. The RCS vent size will be re-evaluated for compliance each time the P/T limit curves are revised based on the results . of the vessel material surveillance. l The RCS vent is passive and is not subject to active ! failure. l The LTOP System satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). 1 A "k_) BYRON - UNITS 1 & 2 B 3.4.12 - 8 8/21/98 Revision A 1 l }.
LTOP System B 3.4.12 , 1
~
, , BASES
.J g 9 LC0 This LCO requires that the LTOP System is OPERABLE. The I 4 LTOP System is OPERABLE when the minimum coolant input and s pressure relief capabilities are OPERABLE. Violation of Q this LC0 could lead to the loss of low temperature overpressure mitigation capability and violation of the ,
Reference 1 limits as a result _of an operational transient. . To limit the coolant input capability, the LCO requires no SI pumps and a maximum of one charging pump (centrifugal) be capable of injecting into the RCS. and all accumulator discharge isolatien valves be closed and de-energized (when-accumulator pressure is greater than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed in the PTLR). ; Q The LC0 is modified by a note that permits the 03eration in l
? MODE 4 with all SI pumps and charging pumps capa)le of RCS N injection whenever all RCS cold legs exceed 330 F. This is y necessary to allow transition between MODES 3 and 4.
m Q The elements of the LC0 that provide low temperature V overpressure mitigation through pressure relief are': v
' ') a. Two OPERABLE PORVs:
- b. Two OPERABLE RHR suction relief valves;
- c. One OPERABLE PORV and one OPERABLE RHR suction relief.
valve; or
- d. A depressurized RCS and an OPERABLE RCS vent.
A PORV is OPERABLE for LTOP when its block valve is open. its lift setpoint is set to the limit required by the PTLR and testing proves its ability to open at this setpoint. and motive power is available to the two valves and their control circuits. An RHR suction relief valve is OPERABLE for LTOP when its RHR suction isolation valves are open. its setpoint is s 450 psig, and testing has proven its ability to open at this setpoint. r BYRON - UNITS 1 & 2 B 3.4.12 - 9 10/1/98 Revision K
6-LTOP System B 3.4.12 /'; BASES V LCO (continued) An RCS vent is OPERABLE wnen open with an area of a 2,0 square inches. Each of these methods of overpressure prevention is capaDie of mitigating the limiting LTOP transient. APPLICABILITY This LCO is applicable in MODES 4 and 5. and in MODE 6 when the reactor vessel head is on. The pressurizer safety valves provide overpressure protection that meets the Reference 1 P/T limits above 350*F. When the reactor vessel head is off, overpressurization cannot occur. LCO 3.4.3 provides the opert urial P/T limits for all MODES. LCO 3.4.10. " Pressurizer Safety Valves." requires the OPERABILITY of the pressurizer safety valves that provide overpressure protection during MODES 1. 2. and 3. Low temperature overpressure prevention is most critical during shutdown when the RCS is water solid and a mass or heat input transient can cause a very rapid increase in RCS !o 'd 3 pressure resulting in little or no time available to allow operator action to mitigate the event. i l l l l y
*+
BYRON - UNITS 1 & 2 B 3.4.12 - 10 8/21/98 Revision A
LTOP System p , L , B 3.4.12 j J TV ; BASES' .-
~,
I U - 3 -ACTIONS LThe Actions'are modified by a Note that indicates that the ! provisions of LC0 3.0.4 are not applicable to the RCS. j'j pressure relief capabilities:(PORVs and RHR suction relief
. valves or vent of = 2.0 square inches with the RCS l
$ depressurized). As a result. MODE' changes are allowed when !
g .one or more of these capabilities are inoperable. This. L allowance is-provided because the Required Actions have been i G ' determined to provide ~an acceptable level of safety. ; 1 A.1 and B.1 j With two centrifugal charging pumps capable of injecting. i into the RCS, or one positive displacement charging pump 'l-
; capable of injecting into the RCS, or any SI pump capable of j injecting into the RCS, RCS overpressurization is possible. 1
- - The requirement to immediately initiate action (except i O during charging pump swap operation) to restore restricted k:- Lcoolant in)ut capability to the RCS-reflects the urgency of removing'tle RCS from this condition.
3 M - Required Action A.1 is modified by a Note that permits two
-h '
, charging pumps capable of RCS injection for s 15 minutes to l
.Q allow for pump swaps. i 7)
'v - C.1 and D.1 An unisolated accumulator requires isolation within 1 hour.
-This is only required when the accumulator pressure is at or more.than the-maximum RCS pressure for the existing
-temperature allowed by:the P/T limit curves. l If the Required Action and associated Completion Time of l Condition C are not met. Required Action D.1 must be- I performed in the next 12 hours. Depressurizing the .
I accumulators below the LTOP limit from the PTLR' prevents an accumulator pressure from exceeding the LTOP-limits if the 1 accumulators are fully injected. r . The Completion Times are based on operating experience that these activities can be accomplished in these time periods and on engineering. evaluations indicating that an event T, requiring LTOP is not likely in the allowed times. 4
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%)
BYRON - UNITS 1-& 2 B 3.4.12 - 11 3/18/98 Revision K 1 v-
LTOP System B 3.4.12 l f] BASES V ACTIONS (continued) L.1 In MODE 4, with one required RCS elief valve inoperable. the RCS relief valve must be restored to OPERABLE status-within a. Completion Time of 7 days. Two RCS relief valves in any combination of the PORVS and the RHR suction relief valves are required to provide low temperature overpressure mitigation while withstanding a single failure of an active component. The Completion Time considers that only one of the RCS relief valves is required to mitigate an overpressure transient and that the likelihood of an active failure of the remaining valve path during this time period is very low. F.1 The consequences of operational events that will overpressurize the RCS are more severe at lower temperature (Ref. 5). Thus with one of the two RCS relief valves inoperable in MODE 5 or in MODE 6 with the head on the
/T Comaletion Time to restore two valves to OPERABLE status is V 24 1ours.
The Completion Time represer.ts a reasonable time to investigate and repair several types of relief valve failures without exposure to a lengthy period with only one OPERABLE RCS relief valve to protect against overpressure events. ! l
\
\ ! V BYRON - UNITS 1 & 2 B 3.4.12 - 12 8/21/98 Revision A l
l l LTOP System I i B 3.4.12 )
- (~}
BASES ACTIONS (continued) G.1 l The RCS must be depressurized and a vent must be estaoiished within 8 hours when: , Both required RCS relief valves are inoperable; or a.
- b. The Required Action and associated Completion Time of Condition D. E. or F is not met; or
- c. The LTOP System is inoperable for any reason other than Condition A. B. C. D. E. or F.
The vent must be sized a 2.0 square inches to ensure that the flow capacity is greater than that required for the worst case mass input transient reasonable during the applicable MODES. This action is needed to protect the RCPB i frtn a low temperature overpressure event and a possible I brittle failure of the reactor vessel. ; The Completion Time considers the time required to place the
, unit in this Condition and the relatively low probability of l an overpressure event during this time period due to (V) increased operator awareness of administrative control l
requirements. I l t I( BYRON - UNITS 1 & 2 B 3.4.12 - 13 8/21/98 Revision A
1 LTOP System B 3.4.12
'N BASES
' (O SURVEILLANCE SR 3. 4.12 1. SR 3.4.12.2 and SR 3.4.12.3 REQUIREMENTS To minimize the potential for a low temperature overpressure l event by limiting the mass input capability. all $1 oumps and all cnarging pumps but one centrifugal charging Dump are verified incapable of injecting into the RCS. and tne I accumulator discharge isolation valves are verified closed i and de-energized. ! , The SI pumps and charging pumps are rendered incapable of l injecting into the RCS through removing the power from thc 4 pumps by racking the breakers out under administrative control. An alternate method of LTOP control may be i employed using at least two independent means to prevent a i mass addition event such that a single failure or single l action will not result in an injection into the RCS. This 1 may be accomplished through the pump control switch being placed in pull to lock and at least one valve in the discharge flow path being closed. This latter method is l appropriate when the SI pump needs to be available for i mitigation of the effects of a loss of decay heat removal event (Ref. 6). Another alternate method of LTOP control i may be utilized when a pump must be energized for testing or '()- U for filling accumulators to assure positive control of the capability for injection by the pump. This may be accomplished by closing the isolation valve and removing power from the valve operator, or by securing a manual . isolation valve in the closed position. These methods are l acceptable provided that an OPERABLE flow path exists from ! the RWST to the RCS. The Frequency of 12 hours is sufficient. considering other j indications and alarms available to the operator in the ' control room, to verify the required status of the equipment. SR 3.4.12.3 is modified by a Note stating that accumulator isolation is only required to be met for an accumulator if ! its pressure is greater than or equal to the maximum RCS l 3ressure for the existing RCS cold leg temperature allowed ' 3y the P/T limit curves provided in the PTLR. i
.(~)
v BYRON - UNITS 1 & 2 B 3.4.12 - 14 8/21/98 Revision A f
b , _ M , LTOP System 1 p B 3.4,12 i q o fN ' cBASESi )
' J
. SURVEILLANCE REQUIREMENTS (continued)-
]
SR '3I4.12.4 The RCS vent of = 2.0. square inches is proven OPERABLE by
- verifying its open condition either- R 1
- a. Once every 12. hours for'a valve-that'cannot~be locked. )
l o b. Once every'31 days for a valve that is loc'ked, sealed, )
#4 or secured in position' A removed pressurizer safety i
Q. valve fits this category. j LThe passive' vent arran ement must only be open to be: -l
-@ OPERABLE. This-Survei 1ance is required to be performed if -l Lthe vent _is being used to satisfy;the pressure relief l requirements of LC0 3.4.12.d.4. l l
SR 3.4.12.5 1 1 Each required RHR suction relief valve shall be demonstrated -l
~J OPERABLE'by verifying its RHR suction isolation valves are
_=% ;o)en. - This Surveillance is only required to be performed if w . yg tie-RHR suction relief valve is being used to satisfy this-LCO.
< O, The RHR suction isolation valves. RH8701A and:RH87018 for relief valve RH8708A. and RH8702A and RH8702B for. relief valve RH8708B. are verified to be opened every 72 hours.
.The Frequency is considered adequate in view of-other adainistrative controls such as valve status indications available to the operator in the control room that verify the-RHR' suction valves remain open.
I The ASME Code. Section XI (Ref. 7). test per Inservice 2 Testing Program verifies OPERABILITY by proving proper i relief valve mechanical motion and by measuring and, if required, adjusting the lift setpoint. QN); Q ' i ' BYRON'- UNITS 1 & 2 B 3.4.12 - 15 10/5/98 Revision K I F 9 - e
LTOP System B 3.4.12 L > L:?%
-BASES:
's SURVEILLANCE REQUIREMENTS-(continued)
.SR 3 '. 4.12. 6 :
u .
, , :The PORV block. valve must be verified open every 72 hours.to .
provide the flow path for each required PORV. to perform its function when actuated. 'The valve.must be remotely verified open inlthe main control room. The block valve is a remotely controlled, motor operated ; valve. The power to the valve operator is not required
. removed, and the manual operator is not required locked in. 1 V the inactive position; Thus the block valve can be closed intheeventthePORVdevelohsexcessiveleakage'ordoesnot- i close.(sticks.open) after relieving an overpressure j situation.
~
1 2 The 72, hour Frequency.is. considered adequate-in! view of 4 other administrative controls available to the operator in
'g the control room, such as valve position-indication, that g verify that.the PORV block valve remains open.
.H .
i
--$ SR 3.4.12.7 .
OAJ"
. Performance of a COT is required within -12' hours after decreasing _RCS temperature to s 350*F and every 31 days on l each required PORV to verify and. as necessary, adjust its '
s- lift setpoint. The COT will verify the setpoint is within T 'the allowed maximum limits in the PTLR. PORV actuation
~d could depressurize_the RCS and is not required.
T g The 12 hour Frequency considers the unlikelihood of a low q temperature overpressure event during this time. A Note indicates that this SR is not: required to be performed until-12 hours after decreasing RCS cold leg temperature to s 350*F. The COT cannot be performed until
~
Lin the LTOP MODES when the PORV lift setpoint can' be reduced :
.to the LTOP setting. The test must be performed within 12 hours after' entering'the'LTOP MODES.
+
x
'3 .
o
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, BYRON - UNITS 1 & 2: B 3.4' ,12 -'16 10/5/98 Revision K L ,
N:
LTOP System. B 3.4.12.
'{ BASES
' SURVEILLANCE REQUIREMENTS (continued)
SR 3.4.12.8 Performance of a CHANNEL CALIBRATION on each required PORV
- actuation ' channel is required every 18 months to adjust the whole channel so that it responds and the valve opens within the required range and accuracy to known input.
REFERENCES 1. 10.CFR 50 Appendix G.
- 2. Generic Letter 88-11.
l 3. ASME, Boiler and Pressure Vessel Code, Section III.
- 4. UFSAR, Chapter 15.
5- Generic Letter 90-06.
$. 6. . Safety Evaluation Report, dated August 31, 1990.
g 73 M .l 7. ASME, Boiler and Pressure Vessel Code. Section XI. _ '_ ' k p"*% J. BYRON - UNITS 1 & 2 B . 3. 4.12 - 17 10/5/98 Revision K
RCS Operational LEAKAGE B 3.4.13 A B 3.4 REACTOR COOLANT SYSTEM (RCS) U- B 3.4.13 -RCS Operational LEAKAGE BASES BACKGROUND Components that contain or transport the coolant to or from the reactor core make up the RCS. Component joints are made by welding, bolting, rolling, or pressure loading. Valves isolate connecting systems from the RCS. During plant life, the joint and valve interfaces can produce varying amounts of reactor coolant LEAKAGE through either normal operational wear or mechanical deterioration. The purpose of the RCS Operational LEAKAGE LCO is to limit system operation in the presence'of LEAKAGE from these sources to amounts that do not compromise safety. This LC0 specifies the . types and amounts of LEAKAGE. 10 CFR 50. Appendix A. GDC 30 (Ref. 1). requires means for detecting and to the extent practical, identifying the source of reactor coolant LEAKAGE. Regulatory Guide 1.45 (Ref. 2) describes acceptable methods for selecting leakage
- detection systems. The leakage detection instrumentation is
'/9
%.J discussed in Section 3.4.15.
The safety significance of RCS LEAKAGE varies widely depending on its source, rate, and duration. Therefore, detecting and monitoring reactor coolant LEAKAGE into the containment area is necessary. Quickly separating the identified LEAKAGE from the unidentified LEAKAGE is
- necessary to provide quantitative information to the operators, allowing them to take corrective action should a leak occur that is detrimental to the safety of the facility and the public.
f A limited amount of leakage inside containment is expected from systems that cannot be made 100% leaktight. Leakage from these systems should be detected, located (identified), and isolated in such a manner, if possible, to not interfere with detection of unidentified RCS leakage. L (~T () BYRON - UNITS 1 & 2 B 3.4.13 - 1 8/21/98 Revision A
RCS. Operational LEAKAGE- ., B 3.4.13 l W BASES:
' BACKGROUND (continued). l This LCO deals with protection of the Reactor Coolant Pressure Boundary (RCPB) from degradation and the core from .
inadequate cooling. in addition to preventing the accident- 1 analyses radiation release assumptions from being exceeded. The consequences 'of-violating this LC0 include the
' possibility of a Loss Of Coolant-Accident (LOCA). However.
the ability to monitor leakage provides advance warning to- ) Jermit unit shutdown before a LOCA occi ' 'his' advantage las been shown by " leak before break" a ..as, i
. APPLICABLE Except for primary to secondary LEAKAGE, the safety analyses )
SAFETY ANALYSIS do not address operational LEAKAGE. However, other operational LEAKAGE is related to the safety analyses for j LOCA: the amount of leakage can affect the probability of l such an event. The safety analysis for an event resulting in steam discharge to the atmosphere assumes 1 gpm primary to secondary LEAKAGE as the initial condition. ; Primary.to secondary LEAKAGE is a factor in the dose releases outside containment resulting from a Steam Line . N Break (SLB) accident because such leakage contaminates the 1 i V
-l secondary fluid. Other accidents.or transients involve 1 secondary steam release to the atmosphere, such as a Steam l Generator Tube Rupture (SGTR). The SGTR is more limiting than the'SLB for site radiation releases.
The UFSAR (Ref. 3) analysis for SGTR assumes the i contaminated secondary fluid is released for a limited time j via the steam generator PORV. .After a tube rupture occurs, reactor coolant immediately begins flowing from the primary ; system into the secondary side of the ruptured steam generator causing the RCS pressure to decrease until a reactor trip occurs on low pressurizer. pressure. The analysis assumes a Loss of Offsite Power occurs coincident with the reactor trip causing the Reactor Coolant Pumps to !
+
trip and the main condenser to become unavailable when the circulating water pumps are lost I J [ l. /sj: BYRON - UNITS 1 & 2 B 3.4.13 - 2 8/22/98 Revision K 1
RCS Operational LEAKAGE I B 3.4.13
/7 BASES V' i l APPLICABLE SAFETY ANALYSES (continued) ;
After the ' reactor trips, the core power quickly decreases to decay. heat levels. The steam dump system cannot be used to dissipate the core decay heat.due to the unavailable condenser. Therefore. the secondary pressure increases in the Steam Generators (SGs) until the steam generator PORVs ., open at which time the ruptured steam generator PORV is ! assumed ~to fail in the open position. .The ruptured SG i failed PORV is' isolated when the block valve is manually l closed twenty minutes after the PORV first opened.- The I
, 1--gpm pMmary to secondary LEAKAGE is relatively inconsegeential to the results of this analysis. ,
The dcse consequences resulting from the SLB accident are well withii the limits defined in 10 CFR 100. .; l To support-the use of sleeving techniques for steam i generator tube repair, the Unit 1 primary to secondary J leakage limits are conservatively reduced from 500 gpd for any single steam generator and 1 gpm. total to 150 gpd for any single steam generator and 600 gpd total (Ref. 4). The RCS operational LEAKAGE satisfies Criterion 2 of [ ') 10 CFR 50.36(c)(2)(ii).
- 9) .
LCO RCS operational LEAKAGE shall be limited to:
- a. Pressure Boundary LEAKAGE No pressure boundary LEAKAGE is allowed, being indicative of material deterioration. LEAKAGE of this type is unacceptable as the leak itself could cauce further deterioration, resulting in higher LEAKAGE.
Violation of this LCO could result in continued degradatior. of the RCPB. LEAKAGE past seals valve seats, and gaskets is not pressure boundary LEAKAGE. 'V BYRON - UNITS 1 & 2 B 3.4.13 - 3 8/22/98 Revision K l
,. ~. . . . . . . - . . . - - . . ~ . . - . - . - - . - ~ - - . - - - . _ . - . . - . - . . - RCS Operational LEAKAGE ', B 3.4.13
--BASES ,
LCO (contin'ued)L r
- b. Unidentified LEAKAGE One gallon per minute (gpm) ~of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount tnat
-the containment air monitoring and containment sump
' discharge flow monitoring equipment can detect within a reasonable time period. Violation of this LCO could result-in continued degradation of the RCPB. if the LEAKAGE is from the pressure boundary.
s
- c. Identified LEAKAGE
'Up to 10 gam of identified LEAKAGE is considered-allowable.3ecause LEAKAGE is from known sources that u do not interfere with detection of, unidentified ~ '
LEAKAGE and is well.within the' capability of_the RCS-Makeup' System. Identified LEAKAGE' includes LEAKAGE to > the containment from specifically'known and . located
. sources, but does not include pressure boundary LEAKAGE or controlled Reactor Coolant Pump (RCP) seal ;
leakoff (a normal function not considered LEAKAGE). Violation of this LC0 could result in continued-degradation of a component or system. .
- d. Primary to Secondarv LEAKAGE throuah All Steam.
Generators (SGs) ! Total primary to secondary LEAKAGE amounting to
.600 gallons per day through all SGs not isolated from the RCS produces acceptable offsite doses in the SLB ,
accident analysis. Violation of this LCO could exceed the offsite' dose limits for this accident. Primary to secondary LEAKAGE must be included in the total p allowable limit for identified LEAKAGE.
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RCS Operational LEAKAGE , B 3.4.13
/N BASES-O LCO (continued)
- e. Primarv to Secondarv LEAKAGE throuan Any One SG The 150 gallons per day limit on one SG is based on the assumption that a single crack leaking this amount would not propagate to a SGTR under the stress conditions of a LOCA or a main steam line rupture. If leaked through many cracks then the cracks are very small. and the above assumption is conservative.
LCO 3.4.14. "RCS Pressure Isohtion Valve (PIV) Leakace." measures leakage through each mdividual Pressure Isolation Valve (PIV) and can impact this LCO. Of the two PIVs in series in each isolated line, leakage measured through one PIV does not-result in RCS LEAKAGE when the other is leak tight. If both valves leak and result in a loss of' mass from the RCS. the loss must be included as identified LEAKAGE. APPLICABILITY In MODES 1, 2, 3. and 4. the potential for RCPB LEAKAGE is g greater due to RCS pressure. V) i In MODES 5 and 6. LEAKAGE limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses and reduced potentials for LEAKAGE. ACTIONS A_l Unidentified LEAKAGE. identified LEAKAGE. or primary to secondary LEAKAGE in excess of,the LCO limits must be reduced to within limits within 4 hours. This Completion i Time allows time to verify leakage rates and either identify ' unidentified LEAKAGE or reduce LEAKAGE to within limits before the reactor must be shut down. This Required Action is necessary to prevent further deterioration of the RCPB. l O V BYRON - UNITS 1 & 2 B 3.4.13 - 5 S. 21/98 Revision A 1
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'RCS:0perational LEAKAGE ~
B 3.4.13
=
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# i. BASES-
' O' ' ACTIONS?(continued): p .
- B;1 and'B.21
'If any' pressure boundary LEAKAGE exists . or if unidentified-
' F
-f LEAKAGE. identified LEAKAGE or/ primary to secondary LEAKAGE Y. cannot.be reducedLto within limits within 4 hours, the- I 1 -f - reactor must.'be; brought to'. lower pressure conditions to 4 ; reduce' the: severity of the LEAKAGE and its-potent 1al u u consequences. .It'should be vloted that. LEAKAGE past seals 8 - and gaskets is not pressure boundary LEAKAGE. The unit must.
' d be brought:to MODE 3 within 6 hours.and MODE 5 within 36 hours. : This . action reduces -the. LEAKAGE ~ and- also' reduces
.the factors that tend to degrade the pressure boundary. -j The allowed Completion Times are reasonable. based on operating experience. to reach the required unit conditions 1 from fullLpower conditions in an orderly manner and without- a challenging plant systems. In MODE 5. the )ressure stresses i acting on the RCPB are much lower, end furtier deterioration
.is much less likely.
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RCS-Operational LEAKAGE B 3.4.13 BASES [ W,n Q - 1
' SURVEILLANCE SR 3L4.13.1 REQUIREMENTS Verifying RCS LEAKAGE to be within the LC0 limits ensures
' the integrity of the RCPB is maintained.: Pressure boundary
.. LEAKAriE would 'at first appear as unidentified LEAKAGE and L can only be positively. identified by inspection. . It should ? , be noted that LEAKAGE past seals, valve.setts..and gaskets. is not pressure boundary: LEAKAGE.- Unidentified LEAKAGE and identified LEAKAGE are determined by perfornunce of an RCS water inventory balance. Primary-to secondary leakage is a component of the gross -leakage as determined by the performance of an RCS inventory balance. Primary ~to secondary leakage:1s quantified by analysis of ~the
, radionuclides present in secondary.feedwater, steam, or Tcondensate. ,or the'noncondensible gaseous effluent; The RCS water inventory balance must be performed with the reactor at steadyfstate operating conditions ~and:near o)erating pressure._ Therefore, a Note-is added allowing tlat this.SR is not required to be performed until-12 hours
+ after. establishing steady state operation. ' Thel 12 hour.
allowance provides sufficient time to. collect and. process all necessary data after. stable plant conditions are k,3 established. O P Steady, state operation is required.t'o perform a proper inventory balance since calculations'during: maneuvering are s- not useful. For RCS operational LEAKAGE determination by water inventory balanca, steady state .is defined as
' ; . stable RCS pressure 1(i 2150 psig). tem)erature, power level.
1ressurizer and' makeup rank levels, ma(eup and letdown, and { y. . k'CP seal injection and return flows. m o m "c 4 .An early warning of-pressure boundary LEAKAGE or unidentified LEAKAGE is provided by the systems that monitor 1' i:s the containment atmosphere radioactivity and the. containment sum? level. It should be noted that LEAKAGE past seals and y . gas (ets is not pressure boundary LEAKAGE. These leakage ( j
, detection systems are specified in LCO 3.4.15. "RCS Leakage Detection Instrumentation."
The 72 hour Frequency during ste.ady state operation is a
.? reasonable interval to trend LEAFAGE and recognizes the f importance of early leakage detect. ion in the prevention of 1
", accidents.
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1 i
- RCS Operational LEAKAGE l B 3.4.13 c7" , LBASES l Os s~
^ I
, SURVEILLANCE REQUIREMENTS-(continued) '
.SR 3'4.13.2
.This SR provides the meansLnecessary'to determine SG H 0PERABILITY. LThe requirement-to demonstrate SG. tube 1
- integrity in accordance with the' Steam Generator Tube. 1 Surveillance; Program emphasizes.the importance~of SG tube -l
' integrity, even though this Surveillance cannot be performed:
at normal operating conditions; A REFERENCES 1. :10.CFR 50. Appendix-A GDC 30.
-2. -Regulatory Guide 1.45, May 1973.
- 3. 'UFSAR,; Chapter 15.
[ 4. Safety Evaluation Report, dated May 7, 1994. 1 03bj: 1 0
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RCS PlV Leakage B 3.4.14 B 3.4' REACTOR COOLANT SYSTEM (RCS) 8.3.4.14 ;RCS-Pressure Isolation Valve (PIV) Leakage BASESL BACKGROUND .10 CFR 50.2. 10 CFR 50.55a(c). and GDC 55 of 10 CFR 50. Appendix A (Refs._1. 2. and 3), define RCS PIVs as any two normally closed valves in series within the Reactor Coolant Pressure Boundary (RCPB), which separate the high pressure RCS from an attached low pressure system. During their lives. these valves can produce varying amounts of reactor coolant leakage through either normal operational wear or . mechanical deterioration. The RCS PIV Leakage LCO allows l RCS high pressure operation when leakage through these j valves exists in amounts that do not compromise safety. J The PlV leakage limit applies to each individual valve. l Leakage.through both series PlVs in a line must be included I as part of the identified LEAKAGE. governed by LCO 3.4.13.
"RCS Operational LEAKAGE." This is true during operation only when the loss of RCS mass through two series valves is determined by a water inventory balance (SR 3.4.13.1). A
- l. known component of the identified LEAKAGE before operation begins is the least of the two individual leak rates determined for leaking series PlVs during the required surveillance testing: leakage measured through one PlV in a line -is not RCS operational LEAKAGE if the other is L leaktight.
l L Although this specification provides a limit on allowable l ! PlV leakage rate. its main purpose is to prevent overpressure-failure of the low pressure portions of connecting systems. The leakage limit is an indication that i the PIVs between the RCS and the connecting systems are degraded or. degrading. PIV leakage could lead to I overpressurization of the low pressure piping or components. Failure consequences could be a Loss Of Coolant Accident (LOCA) outside of containment, an unanalyzed accident, that could degrade the ability for low p "ssure injection. 1: 1-L i LO 8/21/98 Revision A . BYRON-- UNITS 1 & 2 B 3.4.14 - 1
RCS PIV Leakage-B 3.4.14
- BASES
. BACKGROUND:(continued)
The basis for this LCO is the 1975 NRC " Reactor Safety
-Study"'(Ref. 4):that identified potential intersystem LOCAs as a significant contributor to.the risk of core melt. A subsequent study (Ref. 5) evaluateo various PlV.
configurations to determine the probability of intersystem - LOCAs. -PlVs are provided to isolate the Rr5 from the following connected systems:
- a. Residual; Heat' Removal (RHR) System:
- b. Safety Injection (SI) System; and
- c. Chemical and Volume Control' System.
2 Violation'._of this LC0 could result in continued degradation of a PIV. which could lead to overpressurization of a low pressure system and the loss of the integrity of a fission product barrier. 1 TAPPLICABLE- Reference '4 identified potential intersystem LOCAs as a '- . SAFETY ANALYSES- significant contributor to the risk of core melt. The dominant accident ~ sequence in the intersystem LOCA category i is-the" failure of the low pressure portion of the RHR System - outside of containment. The accident is the result of a ! postulated failure of the PIVs. which are part of the RCPB. i and the subsequent pressurization of the RHR. System I downstream of the PIVs from the RCS. Bec'ause'the low pressure portion of the RHR System is designed for 600 psig. l overpressurization failure of the RHR low pressure line- l could result in a LOCA outside containment and subsequent I l [ risk of core melt. ] Reference 5 evaluated various P1V configurations. leakage
. testing of the valves and operational changes to determine the effect on the probability'of intersystem LOCAs. This study concluded that periodic leakage testing of the PIVs can substantially reduce the probability'of an intersystem L- LOCA.
RCS PIV leakage satisfies Criterion.2 of 10.CFR 50.36(c)(2)(ii). l
-BYRON - UNITS 1 & 2 B 3.4.14 - 2 8/21/98 Revision A 1 i-
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RCS PIV Leakage B 3.4,14 r'~N BASES b LCO RCS PIV OPERABILITY protects the low pressure systems attached to the RCS from potential failure due to overpressurization. This protection (i.e., RCS PIV OPERABILITY) is provided by both the leak tight PIVs and the RHR System suction isolation valve interlocks. RCS PIV leakage is identified LEAKAGE into closed systems connected to the RCS. Isolation valve leakage is usually on the order of drops per minute. Leakage that increases significantly suggests that something is operationally wrong and corrective action must be taken. Tne LCO PIV leakage limit is 0.5 gpm per nominal inch of valve size with a maximum limit of 5 gpm. The previous criterion of I gpm for all valve sizes imposed an unjustified penalty on the larger valves without providing information on potential valve degradation and resulted in higher perscnnel radiation exposures. A study concluded a leakage rate limit based on valve size was superior to a single allowable value (Ref. 6). Reference 7 permits leakage testing at a lower pressure differential than between the specified maximum RCS pressure
./
and the normal pressure of the connected system during RCS
' C] operation (the maximum 3ressure differential) in those types of valves in which the ligher service pressure will tend to diminish the overall leakage channel opening. In such cases, the observed rate may be adjusted to the maximum pressure differential by assuming leakage is directly proportional to the pressure differential to the one half L power.
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.c BYRON - UNITS I & 2 B 3.4.14 - 3 8/21/98 Revision A
m
.RCS PIV' Leakage'
~B 3.4;14.
s M I:
. BASES .;
V' { LCO_(continued)-
$ LThe(following valves are RCS PIVs:
Valve' Number Function SI8900A..B. C. D Charging /SI check valve. . ! SI8815- LCharging/SI backup check! ivalve
~
o . SI8948A. B.-C. D -Accumulator check' valve SI8956A. B. C..D' . Accumulator _ backup check-valve
.SI8818A B. C. D' RHR cold leg check valve-S18819A.'.B. C.' D. SI. cold leg check valve' SI8949A, B..C..D Sl. hot. leg check valve. ,
SI8905A. B. C. D SI hot-leg backur check valve l SI8841A.!B RHR hot leg checc valve.. 1 RH8701A.: B . RHR~ suction Motor Operated ValveJ(MOV) 1RH8702A.-B -RHR suction MOV
" APPLICABILITY. In.M' ODES 1, 2. 3. and 4. this-LC0 ap) lies because the PIV I Q.. .-- leakage potential is greatest.when tie RCS is pressurized. -j x ;. -
In MODES 5 and 6. leakage limits are not provided because the' lower-reactor ccolant pressure results in a reduced
; potential .for -leakage and for'a LOCA outside the - H containment.
r1
~
h2 ACTIONS The Actions are modified.by two. Notes. Note 1 provides i clarification that separate entry into a Condition'is ' 3- ~ allowed for each flow path. This is-allowed based upon the 1 M . functional independence of the flow path. Note 2 requires
-@l an evaluation of affected systems if a-PIV is inoperable.
te The leakage may have affected system operability, or isolation of a leaking flow path with an alternate valve may
' have degraded the ability of the interconnected system to
. perform its safety. function. j 0
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y -
-,s - ,- -
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- RCS PIV Leakage B'3.4.14
' BASES
- ' G( ,L
+ .l
. ACTIONS (continued)
A.1 and A.2 l The flow path must be isolated by two valves. Required l Actions A'.1 and A.2 are modified by a Note that the valve ; used for isolation must meet the same leakage requirements ! as the PIVs-and must be within the RCPB or the high pressure ! portion of the system. 1 Required Action A.1 recuires' that the isolation with one I valve must be performec within 4 hours. Four hours provides i time ~to reduce leakage in excess of the allowable limit and l to isolate the affected system if leakage cannot be reduced. ! M The 4 hour Completion Time allows the actions.and restricts ! 9 the operation with leaking isolation valves. ; y
~
j' Required Action A.2 specifies that the double isolation , g barrier. of two valves be restored by . closing some other ! g valve qualified -for isolation or restoring one leaking PIV.
~
g The 72. hour' Completion Time after exceeding the limit-W considers the time required to complete this Action and the , low probability of a second valve failing during this a g period.
- 8) y The inoperability of the RHR System suction' isolation valve j interlock could allow inadvertent opening of the valves at '
RCS 3ressures in excess of the RHR Systems design aressure. , If tie RHR System suction isolation valve interloc( is !
, inoperable, operation may continue as long as the affected RHR suction penetration is closed by at least one de-energized power operated valve within 4 hours. This Action accomplishes the purpose of the interlock function.
7
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,(c
,r
(
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1 BYRON'- UNITS 1 & 2- B 3.4.14 - 5 8/22/98 Revision K
RCS-PIV Leakage B 3.4.14'
. i
;% BASES A* /' .
' ACTIONS (continued)~
C:1 and Cf2 If the Required Actions and. associated ' Completion Times of' i
; Conditions A~and.B are not. met. the unit must be brought-to
~
a a MODE in:which the requirement does.not apply. To achieve this status. the unit must be brought to. MODE 3 within .
~
6 hours and MODE 5 within 36 hours. This. Action may reduce the leakage and also reduces the potential for a LOCA outside the containment. The allowed Completion Times are reasonable based on operating' experience, to reach the. recuired unit- conditions. from; full power conditions in an-orcerly manner and without challenging plant systems. SURVEILLANCE' SR 3. 4 L14 '.1 REQUIREMENTS
- Performance of leakage testing on each RCS PIV or isolation valve used to satisfy Required Action A.1 and Required ,
-Action 'A',2 is required to' verify' that' leakage. is below the -
specified limit and to. identify each leaking valve.- The s . . leakage limit of 0'5 gpm per inch of' nominal valve diameter.
.f/,3 -
up to 5 gpm maximum applies:to each valve.. Leakage. testing V requires a stable pressure' condition. For two PIVs in series the-leakage requirement applies to each-valve individually and not to the combined leakage across both valves. If'the PIVs are not individually leakage tested, one valve may have . failed completely and not-be detected if the other valve in. series meets the leakage recuirement. 'In this situation, the protection provided by
- a. recundant valves would be lost.
g Testing is to be performed every 18 months, a typical refueling cycle, if the plant does not go into MODE 5 for at T least 7 days. The 18 month Frequency.is consistent with i 10 CFR 50.55a(g) (Ref. 8) as contained in the Inservice Testing Program, is within the frequency allowed by the
.r# . American Society of Mechanical Engineers (ASME) Code.
"l- Section XI: (Ref. 7).
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BYRON.- UNITS-1 & 2 B 3.4.14 - 6 10/5/98 Revision K o e v v . tl .- -
RCS PIV Leakage B 3.4.14
~
f~y .. BASES ~ SURVEILLANCE REQUIREMENTS (continued)-
't
- l:
.. Testing must.also be performed prior to entering MODE 2- ;
-whenever the unit has- been in MODE 5' for a 7 days -if leakage :
testingLhas not been' performed once within the previous 9 months. The-leakage: testing is typically performed at the RCS.
. pressure associated with MODES 1 and 2. .This permits >
3 leakage testing at high differential pressures with- , o- stable conditions. However. : test pressures less than - ' Lt '2235 psig but greater than 350 psig are allowed. When
.g measured at these reduced pressures, observed leakage must.
A be adjusted for the actual test pressure up to 2235 psig_ W assuming the leakage to be.directly' 6 differential to the one half power. proportional.to pressure This SR is. modified by three Notes. Note 1-allows entry. into MODES 3.and 4 to. establish-the necessary differential pressures and stable conditions to allow for performance of-this-Surveillance. Note 1 is applicable to all Frequencies
'of this Surveillance.
In addition. -testing must be performed once after the valve b, -has been opened by' flow or exercised to ensure tight U reseating. PIVs-disturbed in the performance'of this Surveillance should also be tested unless it has been
.~ established (per Note 2) that an infinite testing-loop cannot practically be avoided. ' Testing must be performed
.within~24 hours after the valve has been reseated if in MODE 1 or 2,1 or. prior to entry into MODE 2 if not in MODE -1 or 2 at the end of the.24 hour period. Within 24 hours is-a reasonable and practical time limit for performing this test after opening or reseating a valve.
Note 3 exemats the RHR suction isolation valves (RH8701A and B and Ri8702A and B) from the specified Frequency of this testing since these MOVs are not subject to the same failure characteristics as a check' valve that has actuated due to flow. k M g BYRON . UNITS 1 & 2- B 3.4.14 - 7 10/5/98 Revision K am
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, i RCS PIV Leakage ,
B.3.4.14 t z .. f'y R 1
.. BASES :
? ~^ :-SURVEILLANCE. REQUIREMENTS-(continued)
'SR 3.4.14'.2 l
The interlock setpoint'that prevents the'RHR System suction isolation valves from being opened is set-so the actual.RCS.
,. pressure must be < 360 psig to open the valves. This setpoint-ensures the RHR design pressure will not be exceeded and the RHR relief valves will not lift. The i 18 month Frequency is. based on the need to perform the.
-Surveillance under conditions that apply during a unit. ;
outage. The 18 month: Frequency'is also acceptable based:on 1 consideration of the design reliability (and confirming
~ operating experience).of the equipment.
REFERENCES- 1. 10'.CFR 50.2. !
~ 2. 10 CFR.50.55a(c).
- 3. 10 CFR 50.-Appendix A, Section V. GDC'55 l
j 4. WASH-1400(NUREG-75/014).AppendixV. October 1975. '(U) - , 5. .NUREG-0677. May 1980. T 6. EG&G Report. EGG-NTAP-6175. .-i-i; 7. ASME Boiler and Pressure Vessel Code. Section XI. Ij! e
- 8. 10 CFR 50.55a(g).
i i.9 K., ,l BYRON - UNITS 1 & 2.. B 3.4.14 - 8 8/22/98 Revision K t. p. tr
RCS Leakage Detection Instrumentation B 3.4.15
/~~N u B 3.4 REACTOR COOLANT SYSTEM (RCS) 8 3.4.15 RCS Leakage Detection Instrumentation BASES BACKGROUND GDC 30 of Appendix A to 10 CFR 50 (Ref. 1) requires means for detecting and. to the extent practical, identifying the location of the source of RCS LEAKAGE. Regulatory Guide 1.45 (Ref. 2) describes acceptable methods for selecting leakage detection systems.
Leakage detection systems must have the capability to detect significant Reactor Coolant Pressure Boundary (RCPB) degradation as soon after occurrence as practical to minimize the potential for propagation to a gross failure. Thus, an early indication or warning signal is necessary to permit proper evaluation of all unidentified LEAKAGE. Industry practice has shown that water flow changes of 0.5 to 1.0 gpm can be readily detected in contained volumes by monitoring changes in water level. in flow rate. or in the operating frequency of a pump. The containment sump. used
,s -
to collect unidentified LEAKAGE. is instrumented to alarm ( 3 for leakages of 1.0 gpm. This sensitivity is acceptable for U detecting increases in unidentified LEAKAGE. The reactor. coolant contains radioactivity that, when released to the containment, can be detected by radiation l monitoring instrumentation. Instrument sensitivities of 1044 pCi/cc radioactivity for particulate monitoring and of 10 pC1/cc radioactivity for gaseous monitoring are practical for these leakage detection systems. Radioactivity detection systems are included for monitoring both particulate and gaseous activities because of their sensitivities and rapid responses to RCS LEAKAGE. An increase in humidity of the containment atmosphere would i indicate release of water vapor to the containment. Dew I Joint temperature measurements can thus be used to monitor l lumidity levels of the containment atmosphere as an ; indicator of potential RCS LEAKAGE.
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RCS Leakage Detection Instrumentation B 3 A 15
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BASES BACKGROUND (continued) Since the humidity _ level is influencec by several factors. a quantitative evaluation of an indicateo leakage rate by this means may be questionable and should be compared to observed increases in liquid flow into or from the containment sump. Humidity level monitoring 15 considered most useful as an indirect alarm or indication to alert the operator to a potential problem. Humidity monitors are not required by this LCO. Air temperature and pressure monitoring methods may also be used to infer unidentified LEAKAGE to the containment. Containment temperature and pressure fluctuate slightly during unit operation, but a rise above the normally indicated range of values may indicate RCS leakage into the containment. The relevance of temperature and pressure measurements are affected by containment free volume and, for temperature, detector location. Alarm signals from these instruments can be valuable in recognizing rapid and sizable leakage to the containment. Temperature and pressure monitors are not required by this LCO. (v APPLICABLE The need to evaluate the severity of an alarm or an SAFETY ANALYSES indication is important to the operators, and the ability to compare and verify with indications from other systems is necessary. The system response times and sensitivities are described in the UFSAR (Ref. 3). The safety significance of RCS LEAKAGE varies widely depending on its source, rate and duration. Therefore. detecting and monitoring RCS LEAKAGE into the containment area is necessary. Quickly separating the identified , LEAKAGE from the unidentified LEAKAGE provides quantitative l information to the operators. allowing them to take l corrective action should a leak occur detrimental to the ! safety of the plant and the public. RCS leakage detection instrumentation satisfies Criterion 1 of 10 CFR 50.36(c)(2)(ii). l l ! A l ) v BYRON - UNITS 1 & 2 B 3.4.15 - 2 8/21/98 Revision A l
RCS Leakage Detection Instrumentation ; B 3.4.15 ; BASES
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.LC0 One method of protecting against large RCS leakage derives from the ability of instruments to rapidly detect extremely
.small leaks. This LC0 requires instruments of diverse monitoring principles to be OPERABLE to provide a high degree of confidence that extremely small leaks are detected in time to allow actions to place the unit in a' safe- l condition, when RCS LEAKAGE indicates possible RCPB j degradation. ,
The LC0 is satisfied when monitors of diverse measurement means are available. Thus, the containment' sump monitor, in ! combination with a gaseous or particulate radioactivity monitor provides an acceptable minimum. The containment floor drain sump flow monitor (RF008) and the reactor cavity
' sump flow monitor (RF010) are utilized to fulfill the containment sump monitor requirement.
For the containment atmosphere radioactivity monitor, the PR011A (particulate) or PR011B (gaseous) monitor satisfies the LC0 requirement.
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APPLICABILITY- Because of elevated RCS temperature and pr sure in MODES 1.
- 2. 3 and 4. RCS leakage detection instr' .ntation is required to be OPERABLE.
In MODE 5 or 6, the temperature is to be s 200*F and pressure is maintained low or at atmospheric pressure. Since the temaeratures and pressures are far lower than thcse for MODES 1. 2.'3. and 4. the likelihood of leakage and track propagation are much smaller. Therefore. the recuirements of this LC0 are not applicable in MODES 5 anc 6. y#~^ u] BYRON - UNITS 1 & 2 B 3.4.15 - 3 8/22/98 Revision K
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- J _RCS Leakage DetectionLInstrumentation
'B 3.4.15 j PN BASES M -
ACTIONS A.1 and A.2 3
~ With the required containment sump monitor inoperable, no .i
- other form of sampling can provide the equivalent l information:;however, the containment atmosphere- .
l radioactivity monjtor will provide indications.of changes in-
' leakage. 'Together with the atmosphere monitor. the periodic. ;
surveillance for RCS water inventory balance..SR-3.4.13.1. l 2 must-be performed at an increased frequency of 24 hours to
.arovide information that is adequate to detect leakage. A .
iote is added allowing that SR 3.4.13.1 is not required to y be performed until 12 hours after-establishing steady state-V . operation (stable RCS pressure, temperature, power level,
$ pressurizer and makeup tank levels, makeup and letdown, and
+- < .RCP seal injection and return flows). ~The 12 hour allowance-
" provides sufficient time to collect and process all necessary data after stable plant conditions are established..
Restoration of the required sump monitor.to'0PERABLE status within a Completion Time of 30 days is re
=the function after the monitor's failure. quired This to regain time is a_ acceptable' considering the Frequency and adequacy.of the f1 Af
'RCS. water inventory balance required by Required Action A.1.
- Recuired Action 'A.1 and Required Action A,2 are modified by al ote that indicates that the provisions of LC0 3.0c4 are not applicable As a result, a MODE change is allowed when the containment sump monitor is inoperable. This allowance is provided be'cause other instrumentation is available to monitor RCS' leakage. I 1
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-BYRON - UNITS :1 & 2 B 3.4.15 - 4 10/5/98 Revision K L
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RCS Leakage Detection Instrumentation-B 3.4.15 FN BASES:
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ACTIONS-(continued)- B 1.1. B:1.2. and 8.2 M With both gaseous and particulate containment atmosphere
-radioactivity monitoring instrumentation channels "r ,
inoperable, alternative action is required. Either grab 4 ' samples.of the containment atmosphere must be taken and-g- analyzed for gaseous and particulate radioactivity or water y- - inventory balances. in accordance with SR 3.4.13.1. must be performed to provide alternate periodic information. With a sample obtained and analyzed or water inventory balance performed every 24 hours, the reactor may be operated for up to 30 days to allow' restoration of the required containment atmosphere radioactivity monitors. g The 24 hour interval provides periodic information that is
, adequate to. detect leakage. A Note.is added allowing that
- SR 3.4.13.1 is not required to be performed until 12 hours ,
iT. after establishing steady state operation (stable RCS pressure. temperature. power level ~. pressurizer and makeup
-i tank: levels, makeup and letdown, and RCP seal injection and H - return flows).. The 12 hour allowance provides sufficient
/-, F $ .. time to collect and process all necessary data after V ' stable plant conditions are established. The 30 day Com)letion Time recognizes at least one other form of
'leacage detection is available.
Recuired Action B.1 and Required Action B.2 are modified by a bote that indicates that the provisions of LC0~3.0.4 are not applicable. As a result, a MODE change is allowed when the gaseous and particulate' containment atmosphere radioactivity channel is inoperable. This allowance is provided because other instrumentation is available to
. monitor RCS-leakage.
5 L 'q. . $,,( . ' I BYRON UNITS 1 & 2 B 3.4.15 - 5 8/22/98 Revision K
l l RCS Leakage Detection Instrumentation l B 3.4.15 'O \J BASES ACTIONS (continued) C.1 and C.? If a Required Action and associated Completion Time of Condition A or B is not met. the unit must be brougnt to a MODE in which the requirement does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging plant systems. i D_.1 With all required monitors inoperable no means of monitoring leakage are available. and immediate actions. in I accordance with LCO 3.0.3, are required. l SURVEILLANCE SR 3.4.15.1 ; fm REQUIREMENTS l SR 3.4.15.1 requires the performance of a CHANNEL CHECK of I (\ M' the required containment atmosphere radioactivity monitor. The check gives reasonable confidence that the channel is : operating properly. The Frequency of 12 hours is based on instrument reliability and is reasonable for detecting off ' normal conditions. SR 3.4.15.2 i SR 3.4.15.2 requires the performance of a COT on the required containment atmosphere radioactivity monitor. The test ensures that the monitor can perform its function in the desired manner. The test consists of exercising the digital computer hardware using data base manipulation and injecting simulated process data to verify OPERABILITY of alarm and trip functions. The test verifies the alarm setpoint and relative accuracy of the instrument string. The Frequency of 92 days considers instrument reliability. and operating experience has shown that it is proper for detecting degradation. l [ L p LJ BYRON - UNITS 1 & 2 B 3.4.15 - 6 8/21/98 Revision A
RCS Leakage Detection Instrumentation B 3.4.15
~3 BASES (V - . SURVEILLANCE REQUIREMENTS (continued)
SR 3.4.15.3 and SR 3.4.15.4 These SRs require the performance of a CHANNEL CALIBRATION for each of the required RCS leakage detection instrumentation channels. The calibration verifles the accuracy of the instrument string, including the instruments
. located inside containment. The Frequency of 18 months is a typical refueling cycle and considers channel reliability.
Again operating experience has proven that this Frequency is acceptable. REFERENCES 1. _10 CFR 50. Appendix A. Section IV. GDC 30,
- 2. Regulatory Guide 1.45.
- 3. UFSAR Section 5.2.5.
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-BYRON - UNITS 1 & 2 B 3.4.15 - 7 8/21/98 Revision A
RCS Specific Activity
. B 3.4.16 M B 3.4 REACTOR-COOLANT SYSTEM (RCS) d B 3.4.16 RCS Specific Activity
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BASES BACKGROUND The maximum dose to the whole body and the thyroid that an ' individual at the site boundary can receive for 2 hours during an accident is specified in 10 CFR 100 (Ref.1). 'The limits on specific ~ activity ensure that the doses are held to a.small fraction of the 10 CFR'100 limits during analyzed transients and accidents. The RCS specific activity LCO limits the allowable ' concentration level of radionuclides in the reactor coolant. "
- The LCO limits are established to minimize the offsite radioactivity dose consequences in the event of a Steam Generator Tube Rupture (SGTR) accident.
The-LC0 contains specific activity limits for both DOSE EQUIVALENT I-131 and gross specific activity. The allowable levels are intended to limit the 2 hour dose at.the site boundary to a.small fraction of the 10 CFR 100 dose w.
~-
. guideline limits. The limits in the LCO are standardized.
i b based on parametric evaluations of offsite radioactivity , AJ - dose consequences -for typical site locations, i l The parametric evaluations showed the potential offsite dose levels for a SGTR accident'were an appropriately small l fraction of the 10 CFR 100 dose guideline limits. - Each ;
~' evaluation assumes a broad range of site applicable l atmospheric dispersion factors in a parametric evaluation.
I 4 APPLICABLE The LC0 limits on the specific activity of the reactor SAFETY ANALYS5 coolant ensures that the resulting 2 hour doses at the site i boundary will not exceed a small fraction of the 10 CFR 100 ) 3 dose guideline limits following a SGTR accident. The SGTR !
-e, safety analysis (Ref. 2) assumes the specific activity of l
.s the reactor coolant at the LC0 limit and an existing reactor 1 3 coolant Steam Generator (SG) tube leakage rate of 1 gpm. j The safety analysis assumes the specific activity of the g
A secondary coolant at its limit of 0.1 gCi/gm DOSE EQUIVALENT L 5 I-131. from LC0 3.7.3. " Secondary Specific Activity." u -
-BYRON - UNITS 1 & 2 B 3.4.16 - 1 10/1/98 Revision K
4 i RCS Specific Activity B 3.4116
' BASES APPLICABLE SAFETY ANALYSES (continued)
The analysis for the SGTR accident establishes the acceptance limits for RCS specific activity. Reference to this analysis is used to assess changes to the unit tnat could affect RCS specific activity, as they relate to the acceptance limits. The analysis is for two cases of reactor coolant' specific activity. 'One cc:s assumes specific activity at 1.0 pCi/gm-
' DOSE EQUIVALENT-I-131 with a concurrent large iodine spike that. increases the I-131 iodine release rate from the~ fuel
-to the co lant to a value 500 times greater than the release rate corresponding to the initial primary system iodine concentration. 'The second case assumes the initial reactor coolant iodine activity at 60.0 pCi/gm DOSE EQUIVALENT I-131 due to a pre-accident iodine spike caused by an RCS transient. In both cases, the noble gas activity in the reactor coolant assumes 1% failed fuel, which closely equals the LCO -limit of.100/E pCi/gm.for' gross specific activity.
An.SGTR event causes a reduction'in reactor _ coolant I inventory The reduction initiates a reactor. trip from a l
. low pressurizer pressure signal or an RCS Overtemperature !
AT signal. [. If a coincident loss of offsite power occurs. the steam dump valves close to protect the condenser. The rise in pressure in the ruptured SG discharges radioactively contaminated steam to the atmosphere through the SG power operated relief valves and the main steam safety valves. The unaffected SGs remove core decay heat by venting steam to.the atmosphere until the cooldown ends. The safety analysis shows the radiological consequertes of an SGTR accident are within a small fraction of the Reference 1 dose guideline limits. Operation with iodine specific activity levels greater than the LCO limit is permissible, if.the activity levels do not exceed the limits ; shown in Figure 3.4.16-1. in the applicable s)ecification, for more'than 48 hours. The safety analysis las concurrent and pre-accident iodine spiking levels up to 60.0 pCi/gm DOSE. EQUIVALENT l-131. .y f.J ; BYRON - UNITS'1 & 2. B 3.4.16 - 2 8/21/98 Revision A L L L. _ _ _
RCS Specific Activity i B 3.4.16 (7 BASES L/ -
' APPLICABLE SAFETY-ANALYSES (contined). l The remainder of the above limit permissible iodine levels
'shown in Figure 3.4.16-1 are acceptable because of the low probability of a SGTR accident occurring during the H established 48 hour time hmit. The occurrence of an SGTR l
# accident at these .ermissible levels could increase the site !
boundary dose.leve s. but'stil' be' within 10 CFR 100 dose guideline limits. I T M The limits on RCS-specific activity are also used for
-@ establishing standardization-in radiation shielding and. ;
y plant personnel radiation protection practices. RCS srcific activity satisfies Criterion 2 of i 10 CFR 50;36(c)(2)(ii). LC0 The specific' iodine activity is limited to 1.0 pCi/gm DOSE EQUIVALENT I-131. The gross specific activity'in the ; reactor coolant is limited to the number of pC1/gm equal to j 100' divided'by C (average disintegration energy of the sum 1 e.' of the average beta- and gamma energies of the coolant l 1 .nuclides). The limit on DOSE EQUIVALENT I-131 ensures the ' A]o 2 hour thyroid dose' to an individual at the site boundary during the Design Basis Accident (DBA) will be a small fraction of the allowed thyroid dose. The limit on gross specific activity. ensures the 2 hour whole body-dose to an ) individual at the site boundary during the DBA will.be a i small. fraction of the allowed whole body dose. I The'SGTR accident analysis (Ref. 2) shows that the 2 hour site boundary dose levels are within acceptable limits.
~ Violation of the LC0 may. result in reactor coolant radioactivity levels that could, in the event of an SGTR.
lead to site boundary doses that exceed the 10 CFR 100 dose < guideline limits. n ' 1 O' BYRON - UNITS 1 & 2 B 3.4.16 - 3 8/22/98 Revision K l'
l RCS Specific Activity ! B 3.4.16 () BASES 1 V APPLICABILITY In MODES 1 and 2. and in MODE 3 with RCS average temperature a 500 F. operation within the LCO limits for DOSE EOUlVALENT I-131 and gross specific activity are necessary to contain the potential consequences of an SGTR to within the acceptable site boundary dose values. For operation in MODE 3 with RCS average temperature
< 500 F. and in MODES 4 and 5. the release of radioactivity l in the event of a SGTR is unlikely since the saturation '
pressure of the reactor coolant is below the lift pressure settings of the main steam safety valves. l l l ACTIONS A.1 and A.2 With the DOSE EQUIVALENT I-131 specific activity greater l than the LCO limit. samples at intervals of 4 hours must be l taken to demonstrate that the limits of Figure 3.4.16-1 are not exceeded. The Completion Time of 4 hours provides l sufficient time to obtain and analyze a sample. Sampling is ! done to continue to provide a trend.
- 1
( The DOSE EQUIVALENT I-131 s]ecific activity must be restored l to within limits within 48 1ours. The Completion Time of l 48 hours is required, if the limit violation resulted from ! normal iodine spiking. A Note to the Required Actions excludes the MODE change restriction of LC0 3.0.4. This exception allows entry into the applicable MODE (S) while relying on the ACTIONS even though the ACTIONS may eventually require unit shutdown. This exception is acceptable due to the significant conservatism incorporated into the s)ecific activity limit, the low probability of an event whic1 is limiting due to exceeding this limit. and the ability to restore transient specific activity excursions while the unit remains at, or proceeds to power operation. lg d BYRON - UNITS 1 & 2 B 3.4.16 - 4 8/21/98 Revision A
l RCS Specific Activity B 3.4.16 ' 'r3 BASES l i ACTIONS (continued) B_1 If the Required Action and associated Completion Time of Condition A is not met or if the DOSE EQUIVALENT I-131 l specific activity is in the unacceptable region of l Figure 3.4.16-1. the reactor must be brought to MODE 3 with RCS average temperature < 500 F within 6 hours. The Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 3 below 500 F from full power , conditions in an orderly manner and without challenging I plant systems. l L.1 With the gross specific activity in excess of the allowed limit, the unit must be placed in MODE 3 with RCS average l temperature < 500 F. This action lowers the saturation pressure of- the reactor coolant below the setpoints of the main steam safety valves and prevents venting the SG to the environment in an SGTR event. The Completion Time of 6 hours is reasonable, based on operating experience, to n . reach MODE 3 below 500 F from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.4.16.1 REQUIREMENTS SR 3.4.16.1 requires performing a gamma isoto]ic analysis as a measure of the gross specific activity of t1e reactor , coolant at least once every 7 days. A gross radioactivity analysis consists of the quantitative measurement of the total specific activity of the reactor coolant except for radionuclides with half lives < 10 minutes and all radiciodines. The total specific activity is the sum of the . degassed beta-gamma activity and the total of all identified gaseous activities in the sample within 2 hours after the l sample was taken. Determination of the contributors to the gross specific activity are based upon those energy peaks identifiable with a 95% confidence level. The latest available data may be used for pure beta emitting radionuclides. This Surveillance provides an indication of any increase in gross specific activity. l l l p r ( BYRON - UNITS 1 & 2 B 3.4.16 - 5 8/21/98 Revision A
' RCS Specific Activity B 3.4.16 pMT BASES Q, )
. SURVEILLANCE REQUIREMENTS (continued)' i l
Trending the results of this Surveillance allows proper i remedial action to be taken before reaching the LCO limit' . under normal operating conditions. The Surveillance is ;j applicable in MODES 1 and;2. and in MODE 3 with RCS average - !
. temperature a 500*F. The 7 day Frequency considers the.
unlikelihood of a gross fuel failure during the time. SR 3 A 16.2 i This Surveillance is performed in MODE 1 only to ensure j iodine remains within limit..during normal operatio'n and L.. following fast power changes when fuel failure is more apt i to occur. The 14 day Frequency is adequate to trend changes Q in the iodine activity level, considering gross activity is jl monitored every 7 days. The Frequency, between 2 and
'g' 6. hours after a power change = 15% RTP within a 1 hour period is established because the iodine levels peak during Y: this~ time following fuel failure: samples at'other times wl. would provide inaccurate results.
SR- 3.4.16.3 ,b ~ NM-
-A radiochemical analysis for [ determination.is required
'every 184 days (6 months) with the unit operating in MODE 1 equilibriurconditions. The E determination directly relates to the LC0 and is required to verify unit operation
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within the specified' gross activity LC0 limit. The analysis for E-is a measurement of the average energies per
-disintegration for isotopes with half lives longer than The Frequency of 184 days
.10 minutes,[excludingiodines recognizes does not change rapidly.
This SR has been modified by a Note that- indicates sampling is required to be performed within 31 days after a minimum of.2 effective full power days and 20 ddys of MODE 1 operation have elapsed since the reactor was last
. subtritical for. at least 48 hours. This ensures that the radioactive materials are.at equilibrium so the analysis for
& is representative and not skewed by a crud burst or other similar abnormal event.
U. BYRON'- UNITS 1 & 2- B 3.4.16 - 6 8/22/98 Revision K
RCS Specific Activity ! !- B 3.4.16 ) BASES. ('}
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l REFERENCES 1. 10 CFR 100.11. 1973. l
- 2. UFSAR, Section-15.6.3.
- 3. Safety Evaluation Report. dated May 7. 1994. 1 I
- 4. ' Safety Evaluation Report. dated August 18. 1994. ^
Safety Evaluation Report. dated November 9. 1995. I 5. i l l l i i 1,r'N l
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I 1 1. l -. , LJ - BYRON - UNITS 1 & 2 B 3.4.16 - 7 8/21/98 Revision A L
_ _ _ . . . _ . _ - . . . . ~ . . - _ . _-_m ...-_m . 1 L RCS Loop Isolation Valves B 3.4.17 7 , y]' i
;B 3.4 REACTOR: COOLANT SYSTEM (RCS)-
f . l B 3.'4.17: RCS Loop Isolation' Valves
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!- i N BASES . l n-. \ BACKGROUND. TheRCSlmaybeloperatedwith;1oopsisolated~in'orderte i perform maintenance. While' operating with a loop isolateJi there is potential;for. inadvertently opening the isolation ) valves in the' isolated loop. -In this event. the coolant in- .l
, the isolated' loop would. suddenly begin to mix with the l ' - [' -coolant in the unisolated portion of the RCS. This -i situation'has the potential of causing a positive reactivity addition with~a corresponding reduction of SDM if:
- a. TheLtemperature-in the isolated loop is lower than the !
r[ . temperature in the unisolated portion of the RCS-(cold 1- water. incident);.or 1: [ .b. The boron concentration ~in the isolated loop is' lower i t .than the boron concentration required in the RCS to '
; meet SDM (boron dilution incident). .
! As discussed in the UFSAR (Ref. 1), the startup~of an. :
Ll
?("'F 1 isolated : loop:1s performed-in a controlled manner that V ! virtually eliminates any sudden' positive. reactivity. addition-from cold water or boron dilution because:
~_. Ja; LC0 3.4.18. "RCS Isolated Loop Startup_" and plant -l o . operating procedures require that the boron. '
r concentration in the isolated loop be maintained y (higher than the required-SDM boron' concentration of
-4 the unisolated portion of the RCS.. thus eliminating -
% the potential for introducing coolant from the 4 isolated loop that could dilute the boron 1 concentration in the unisolated portion of the RCS to less than the required SDM boron concentration:
l
/ b-. The' cold leg. loop isolation valve cannot be opened ;
unless the temperatures of both the hot and cold legs of the isolated loop are within 20 F of the temperatures of the hot and cold legs of +ha unisolated portion of the RCS (compliance is. ensurd
.. by operating procedures and automatic interlocks); and L s ;
y , A4 ' L ? BYRON --UNITS IL& 2- B 3.'4.17 10/5/98 Revision K y a ,4 .~. -
RCS' Loop Isolatioh Valves. B 3:4.17 BASES H/W)' f, L - BACKGROUND (continued) ,
- c. :0ther automatic interlocks. all.of which are part' of
- the Reactor. Protection System (RPS). prevent opening'
( : the hot leg-loop isolation valve unless:the cold leg l l
- l; -loop isolation valve is fully' closed.
APPLICABLEJ . SAFETY ANALYSES During startup oflan isolated loop in.accordance with LCO 3.4.18 the cold. leg . loop isolation valve interlocks-and X operating procedures prevent opening:of the valve until.the. h - isolated < loop and~unisolated portion of the RCS boren . y concentrations and temperatures are within limits. 'his se ensures that any undesirable reactivity effect from the T ; isolated loop does not occur. ! 4
~The safety analyses assume a minimum SDM as an initial condition for-Design Basis Accidents (DBAs) (Ref.-1). .
- Violation of-the LCO. combined with mixing of:the. isolated l: loop coolant' into' the unisolated portion-of the RCS.- could result in the-SDM being less than that assumed in the safety analyses. ,
i
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1The above analyses are for DBAs'that establish the acceptance limits for the-RCS loop isolation valves. Reference,to the analyses for.these DBAs is used to assess-changes-to the RCS' loop isolation valves as they relate.to the acceptance limits.
-The boron concentration of- an isolated loop may affect SDM :
, and therefore RCS loop isolation valves satisfy Criterion 2 of 10 CFR 50;36(c)(2)(ii).
f DQ M B 3.4.17 - 2 10/8/98 Revision K BYRON-l UNITS 1&2 M
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L RCS Loop Isolation Valves B 3.4.17 V 7% : BASES M ? ' This LC0 ensures that a loop isolation valve that becomes [.LCO: ,
-closed.in MCDES 1. through .4 is fully isolated and the plant
~placed.in MODE.5. Loop isolation valves are used for 3 performing maintenance when the-plant is in MODE 5 or 6. and.
M startup of an isolated-loop is covered by LCO.3.4.18. w
. El- ;This-LC0~also ensures that loop isolation valves remain open-in MODES.I. 2, 3.'and 4. Closure of the loop isolation valves 'during these. MODES results in' the' Jotential. for an:
inadvertent startup of an isolated loop w1ich could result in.the SDM being less than assumed in the safety. analyses.
)
7 DAPPLICABILITY In NDDES 1 through 4. this LC0 is' applicable since
% unisolating an. isolated loop has not been analyzed. The y
; potential affects (with a boron concentration or temperature Wl- . less than that of the unisolated portion of the RCS) may A include an inadvertent criticality.
.. 1 -
1 In MODES 5 and 6 the SDM of the operating loops is large J enough to permit operation with isolated loops. In.these I
., MODES, controlled startup'of isolated loops is ossible without significant risk of inadvertent critica ity.
l l l i BYRON - ONITS 1 & 2 .B 3.4.17 - 3 10/1/98 Revision K
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L RCS Loop Isolation Valves > B-3.4.17
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BASES. ACTIONS: The Actions ~have been provided with a Note to clarify tnat ;
- all RCS loop isolation' valves for this LCO are treated as f separate entities each with separate Completion Times. .
L- .(i.e., the Completion Time is on a component basis). L l D . l If. power -is inadvertently restored'to one or more loop isolation valve operators. .the potennial exist; for ! -accidental isolation of a loop with a-subsequent inadvertent L startup of the isolated loop.' The loop isolation valves !- have motor operators. Therefore, these valves will maintain L their -last ' position when power is remo>ed from the valve o)erator. With power ap) lied to the' valve operators, only t1e interlocks prevent t1e valve from being operated. l Although operating procedures and interlocks make the l occurrence of this event unlikely, the prudent action is.to ! remove power. from the loop isolation vah'e operators. The l, Completion Time of 30 minutes to remove power from the loop
- i. : isolation valve operators is sufficient considering the..
u complexity of the task. L B.l. B.2. and B.3 L(N . . Should a loo) isolation valve be closed in MODES 1 through 4. tie affected loop must be fully isolated
~
1 immediately.and the unit placed in MODE 5 to preclude I inadvertent startup of the loop and the potential inadvertent criticality. Required Actions B.2 and B.3 require placing the unit in MODE 3 within 6 hours and MODE 5- ) within 36 hours. The allowed-Completion Times-are : reasonable. based on operating experience, to reach the .
- recuired ~ unit conditions from full power conditions in an
- - orcerly manner and without challenging plant systems. ,
l i i i !O I BYRONL UNITS 1 & 2 B 3.4.17 - 4 8/21/98 Revision A l
RCS Loop Isolation Vaives B 3.4.17 l
, BASES SURVEILLAtlCE SR 3 4.17 1 i REQUIREMENTS The Surveillance is performed at ieast once per 31 aays to ensure that the RCS loop 1 solation velves are open. v.ith ,
! power removed from the loop 1 solation valve operaters Tne 1 primary 1;nction of this Surveillance is to ensure that power is removed from the valve operators. since SR 3.4.4.1 of LCC 3.4.4. "RCS Loops -MODES 1 ano 2." ensures that tne loop 1 solation valves are open by verifying every 12 hours that all loops are operating and circulating reactor coolant. The Frequency of 31 days ensures that the required flow can be nade available. is based on engineering judgment, ana has proven to be acceptable. Operating experience has shown that the failure rate is so low that the 31 day Frequency is justified. REFERENCES 1. UFSAR. Section 15.4.4. l l 1 I i l BYRON - UNITS 1 % 2 B 3.4.17 - 5 8/21/98 Revision A
.t RCS Loops-Isolated B 3.4.18
~, B 3.4 REACTOR COOLANT SYSTEM (RCS)
L "' ) B 3.4 18 RCS Loops-Isolated BASES BtCKGROUND The RCS may be operated with loops isolated in MODES 5 and 6 ' in order to perform maintenance. While operating with a loop isolated. .there is potential for inadvertently opening the isolation valves in the isolated loop. In this event, the coolant in the isolated loop would suddenly begin to mix with the coolant in the unisolated portion of the RCS, This situation has the potential of causing a positive reactivity addition with a corresponding reduction of SDM if l a. The temperature in the isolated loop is lower than the temperature in the unisolated portion of the RCS.(cold water incident): or
- b. The boron concentration in the isolated loop is lower than the boron concentration required in the RCS to meet SDM (boron dilution incident).
-{
, j. As discussed in the UFSAR (Ref. 1), the startup of an isolated loop is done in a controlled manner that virtually V) =l eliminates any sudden positive reactivity addition from cold l water or boron dilution because:
- a. This LCO and plant operating procedures require that the boron concentration in the isolated loop be maintained higher than the required SDM boron concentration of the unisolated portion of the RCS.
. thus eliminating the potential for introducing coolant from the isolated loop that could dilute the boron
'I concentration in the unisolated portion of the RCS to 4 less than.the required SDM boron concentration:
T b. The cold-leg loop isolation valve cannot be opened
- unless the temperatures of both the hot leg and cold E leg of the isolated 100) are within 20'F of the
- unisolated portion of tie RCS. Compliance with the temperature requirement is ensured by operating procedures and automatic interlocks; and
'i > BYRON - UNITS 1 & 2 B 3.4.18 - 1 10/5/98 Revision K
RCS Loops-Isolated B 3.4.18 )
') BASES
.j. BACKGROUND (continued)
- c. Other automatic interlocks prevent openina the hot leg loop isolation valve unless the cold leg loop s
isolation valve is fully closed. All of the o interlocks are part of the Reactor Protection System. I
% l s-
@ APPLICABLE During startup of an isolated loop the cold leg loop
. n; SAFETY ANALYSES isolation valve interlocks and operating procedures prevent
' opening the valve until the isolated loop and unisolated h portion of the RCS boron concentrations and temperatures are s
D
'within limits. This ensures that any undesirable reactivity effect from the isolated loop does not occur.
The safety analyses assume a minimum SDM as an initial condition for Design Basis Accidents. Violation of this LCO could result in the SDM being reduced in the operating loops to less than that assumed in the safety analyses. l The boron concentration of an isolated loop may affect SDM i and therefore RCS isolated loop startup satisfies ! Criterion 2 of 10 CFR 50.36(c)(2)(ii). J
'LCO Loop isolation valves are used for aerforming maintenance ,
when the unit is in MODE 5 or 6. T1is LCO ensures that the
~ loop isolation valves remain closed until the differentials O of temperature and boron concentration between the
' unisolated portion of the RCS and the isolated loops are
% 'j within acceptable limits.
W ai b APPLICABILITY In MODES 5 and 6. the SDM of the unisolated portion of the Q RCS is large enough to permit operation with isolated loops. In these MODES, controlled startup of isolated loops is possible without significant risk of inadvertent criticality. In MODES 1. 2. 3, and 4. operation with isolated loops is not permitted. See LC0 3.4.17. "RCS Loop Isolation Valves." (;
. BYRON - UNITS 1 & 2 B 3.4.18 - 2 10/8/98 Revision K
RCS Loops-Isolated B 3.4.18-n . ,7w. BASESL i d. g LACTIONS AIlandB.1-Required: Action A.1 and Required Action B.1 assume that the-gE
-prerequisites.of the 1.C0 are not met and a loop isolation valve has:been. inadvertently. opened. Therefore, the Actions gj. require =immediate closure of isolation valves to preclude a-
.u . boron ~ dilution event or a cold water event.
q SR 3.4.18.1 SURVEILLANCE . REQUIREMENTS This Surveillance is performed to ensure that the temperature differential between the isolated loop and the .1 unisolated portion of the RCS is s 20 F. Perforning the j
- Surve411ance 30 minutes prior to opening the cold leg j isolation valve in-the isolated loop provides reasc% ble )
assurance, based on engineering judgment, that the-
. temperature differential will stay within limits until the cold leg isolation valve is 03ened. This Frequency has been shown to be acceptable througl operating experience. l SR 3.4.18.2 I
- (, ;m,) ^ m.
LC -7 To ensure that the boron concentration of the isolated loop 4 is greater than er equal to the boron concentration required G' in the RCS-to meet SDM. a Surveillance is performed.4 hours
- 1rior to opening either the hot or cold leg isolation valve.
T4 : 3erforming the Surveillance 4 hours prior to opening either-the hot or cold: leg' isolation valve provides reasonable 5- ' assurance the resulting boron concentration difference will I be'within acceptable limits when the -loop is unisolated. 1: This Frequency is acceptable due to the amount of time 1 : required to sample and confirm concentration results.
- 1. UFSAR. Section 15.4.4.
' REFERENCES i
l f'~ v(f. - . . B 3.4.18 - 3 10/1/98 Revision K BYRON - UNITS 1 & 2' 1
,_._m m - 4 m ._m... ___ .+_ _ .. - --.-a , .a__. 2 2. -e. _ . . _, _-- _ _ . . . a - _<mm-3.4 BRWD ITS
- O f
d T ] . O l i l r s O
RCS Pressure. Temperature. and Flow DNB Limits-3.4.1 ,A 3.4- REACTOR COOLANT SYSTEM.(RCS). >Q ,
- 3.4.1' RCS' Pressure Temperature, and Flow Departure from Nucleate Boiling-
'(DNB) Limits-Nt LCO 3.4.1 RCS DNB parameters for' pressurizer pressure.-RCS average
-temperature, and-RCS total flow rate shall be within the A ' limits-specified below:
.c Y a. Pressurizer pressure a 2219 psig; 3 be .RCS average temperature (T,y) s'591.2 F: and
. l I : (- -
- c. RCS total flow rate a 371.400 gpm.
4 bi ' NOTE .
';<id - [ Pressurizer pressure limit does not apply during: ,
C= THERMAL POWER ramp >-5% RTP per minute: or a.
- b. THERMAL POWER step > 10% RTP.
i
. I
( )j sAPPLICABILITY: MODE-1.
- v. '
' ACTIONS
? CONDITION' REQUIRED ACTION COMPLETION TIME -
.A. -One or more RCS DNB. A.1 Restore RCS DNB 2 hours '
. parameters not within- parameter (s):to-limits within limit.
18 .' Required Action and B.1 Be in MODE 2.- 6 hours-associated Completion Time-not met.
?
b i. , 7% -.
} .;
F ;BRAIDWOOD. ' UNITS 1 & 2 3.4.1 - 1 8/21/98 Revision K
;RCS Pressure. Temperature, and Flow DNB Limits-3.4.1 SURVEILLANCE REQUIREMENTS A).\
SURVEILLANCE FREQUENCY" SR ~3.4.1.li Verify pressurizer pressure is'a 2219 psig. 12 hours-
..x SR. 3.4.1.2 : Verify RCS'. average temperature (Tm) is -12 hours s 591.2 F.
SR. 3.4.-l.3. Verify RCS total flow rate-is 12 hours-
- =:371.400 gpm.
? SR' 3:4.1'.4l ..
--NOTE--
-- -Not required to be performed until 7 days-T- Lafter = 90% RTP.
g u- . . [
.. Verify by precision heat balance that~RCS 18 months total. flow rate is a 371.400 gpm.
]x.. '
t T 5g:
. e ,
V: , y 3 s
'Iw.
.r
^
LBRAIDWOOD-UNITS 1&2 3.4. 2 - 8/21/98 Revision K l
- ,1
)
]
. . . __ .. . _ ._ . _ . . _ . . . . . . . . , . _ _ . . - .~ , . _ . . _ - _ . . . . . _ _ _ _ _ _ - . . _ _ . _ .
l' ' ' RCS Minimum Temperature for Criticality l H' . 3.4.2 9*1 '3.4i REACTOR. COOLANT SYSTEM (RCS)-
~
R h' " =
- h. - -1 g ?3.4.2 RCS ' Minimum' Temperature .for. Criticality 4 . 4 W
Y .'.LCOL3i4.2.~
. s; : 1 Each:RCS. loop average temperature '(T,y) shall be a 550 F.
- y :
1 1M00E 1. .
*I y? :l' APPLICABILITY: MODE.2 with k,,, a 1.0.
m . 4 ACTIONS , CONDITION-- REQUIRED ACTION- - COMPL'ETION TIME ,, 1 1 A. T,,,i.n one or'more RCS ..A;1- Be in' MODE ~2 with 30 minutes i 1 cops not within k,,,-< 1.0. ilimit.
' SURVEILLANCE REQUIREMENTS-SURVEILLANCE- FREQUENCY-s
; SR.:i 3,4.2.1 Verify RCS T, in each loop a 550 F, 12 hours 4
i (, l: . L j ,- m BRAIDWOOD'-LUNITS'l-~& 2 3.4.2 - 1 8/21/98 Revision K t , - , . . . . - . .
- l. ,
RCS P/T Limits 3.4.3 - L i... - , M- 3.4 REACTOR C00LANT' SYSTEM (RCS) O L3.4.3 RCS Pressure and Temperature (P/T) Limits L >
'LCO-'3.4.3 RCS pressure. RCS-temperature, and RCS heatup-and cooldown.
. rates-shall be maintained within the limits specified in tne PTLR.
!- APPLICABILITY: At'all times. l ACTIONS t l CONDITION' -REQUIRED ACTION COMPLETION TIME-A. ---- NOTE --- A.1 Restore parameter (s) 30 minutes ls ' Required Action A.2 to within limits. !l shall be completed whenever this AND Condition is entered. A.2 Determiae RCS is' 72 hours
- . acceptabis for l-(v Requirements of LC0 continued operation.
not met in MODE 1. 2.
- 3. or 4.
l
-B. ' Required Action and B.1 Be in MODE 3. 6 hours
- - associated Completion Time of Condition A -AND H~ not met.
l B 2. Be in MODE 5. 36 hours L i ! (continued) !1 p BRAIDWOOD - UNITS l'&'2- 3.4.3 - 1 8/21/98 Revision A l<
RCS P/T Limits 3.4.3 IN- ,Q)
' ACTIONS (continuedi l- ' CONDITION REQUIRED ACTION COMPLETION TIME C. NOTE-- C.1 Initiate dction to Immediately
-Required Action C.2 . restore parameter (s)-
shall be completed to within limits, wnenever this Condition is entered. AND
~ .
. C.2 Determine RCS is Prior to
- -Requirements of LC0 acceptable for entering MODE 4'
! not met any time other continued operation. than in MODE 1. 2. 3. or 4.
' SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 - - -
. NOTE-Only required to be performed during RCS l heatup and cooldown operations and RCS '
inservice leak and hydrostatic testing. Verify RCS pressure. RCS temperature- and . 30 minutes RCS heatup and cooldown rates are within < the limits specified in the PTLR, l L ,g - ,) '
- BRAIDWOOD - UNITS-1 & 2 3.4.3 - 2 8/21/98 Revision A
4i j
. RCS Loops - MODES- 1 and 2 3.4.4 i
([]. L-3.4 REACTOR COOLANT SYSTEM (ES) l
- i. V '
!' 31.4 'RCS; Loops-MODES 1 and 2 LC0 -3.4.'4- Four RCS loops shall be OPERABLE and in operation. l-APPLICABILITY.: MODES 1 and 2. 1
'l
. ACTIONS CONDITION REQUIRED ACTION' COMPLETION 1IME I
'A.
Requirements of A.1 Be in MODE 3. '6 hours i LCO not met. 1 l
'l SURVEILLANCE REQUIREMENTS I
'b/ SURVEILLANCE FREQUENCY SR1 3.4'.4.1 Verify each RCS loop is in operation. 12 hours l. i I~ f
, BRAIDWOOD - UNITS-1 & 2 3;4.4 - 1 8/21/98 Revision A l
l
. .- . . .. . . . - . ~ . . _
RCS Loops -MODE 3 - , 3.4,5 1 7N 3.4 REACTOR COOLANT S STEM (RCS) l)
;" 3.4.5 RCS- Loops -MODE, 3 LCO 3.4.5 Two RCS loops shall be OPERABLE. and either:
- a. Two OPERABLE RCS loops shall be in operation when the .
Rod Control System is capable of rod withdrawal; or ! l
- b. One OPERABLE RCS loop shall~be in operation when the Rod j Control System i.s not capable of rod withdrawal.
NOTE 1' All reactor coolant pumps may be removed from~ operation for s 1 hour per 8 hour period provided: l j-a No operations are permitted that would cause reduction of the RCS boron concentration: and W ll - b. Core outlet temperature is maintained a 10 F below M4 saturation temperature.
)
' l' APPLICABILITY: MODE 3.
ACTIONS
. CONDITION REQUIRED ACTION COMPLETION TIME
..A. One required RCS loo) A .- 1 Place the Rod Control I hour not:in operation witi System in a condition Rod Control System incapable of rod capable of rod withdrawal.
withdrawal. (continued) li[ r 6 1
.;/ y -(/;
BRAIDWOOD . UNITS 1 & 2 3.4.5 - 1 10/8/98 Revision K l.: n l
RCS Loops-MODE 3
. 3.4.5 a '
ACTIONS,'(continued) CONDITION REQUIRED ACTION COMPLETION TIME L B.' No required RCS loop B.1 ' Suspend all , Immediately l- in' operation with Rod operations involving-
. Control System not. a. reduction of RCS capable of_ rod boron concentration.
. withdrawal, i AND l
B.2 Initiate action to immediately restore one RCS loop-to operation. C. Two required RCS loops C.1 Initiate action to Immediately i
.not'in operation with place the Rod Control
. Rod Control System System in a condition capable of rod incapable of rod withdrawal, withdrawal.
'QR AND hi . Required Action and C.2 Suspend all Immediately
. associated Completion- operations involving Time of--Condition A a reduction of RCS not met, boron concentration.
AND C.3- Initiate action to Immediately restore RCS loop (s) l to operation. D. One required RCS_ loop D.1 Restore required RCS 72 hours inoperable, loop to OPERABLE status. l (continued) l l O o
-BRAIDWOOD - UNITS 1-& 2 3.4.5 - 2 8/21/98 Revision A l . ~
g - - - 6m ., y3+r-y. re
RCS Loops-MODE 3 , 3.4.5 1 t) - ACTIONS (continued)- V- CONDITION REQUIRED. ACTION COMPLETION TIME E. Required ' Action and E.1 Be in. MODE 4. 12 hours
. associated Completion
-. Time of Condition D.
.not. met.
F. Two required RCS loops F.1 Initiate action to Immediately. inoperable, place the Rod Control - System'in a condition l incapable of rod -l withdrawal, , AND F.2 Suspend all Immediately operations involving a reduction of RCS boron concentration. , AND
.t F.3 Initiate action to Immediately restore one RCS loop to OPERABLE status.
- SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY j
- 'SR 3.4.5.1 Verify each required RCS loop is in 12 hours i operation. I e
(continued) i: b , - t,j-L BRAIDWOOD - UNITS 1 & 2 3.4.5 - 3 8/21/98 Revision A t
RCS Loops -MODE 3 ' 3.4.5 A f,3
' ~
SURVEILLANCE REQUIREMENTS?'(continued)
' SURVEILLANCE--
FREQUENCY
- e.e ,
,.- _i, ,_ .
'I t- .SR.3.4.5.2 ' Verify steam generator. secondary side. 12 hours' q r >
'- a narrow range water level .is 2: 18% for each 'i g._ required RCS 1 cop.
-I
-) .
.SRh3.4.5.3l Verify correct breaker alignment a'nd 7 days
- indicated power are available to each required pump;that.is not in operation.
's s'
l u - l i i ' Nlrf ,
.,f
[ / .{? e < 7 a I i:; jp ,
~
g w , !l4 j.y :* L-- Q< l
~
EBRAIDWOOD.+: UNITS l'& 2 3.4.5 - 4 8/24/98 Revision K g, ' nic
. I, n= :.: . . . - . . - -. . . . -
L f / , RCS' Loops - MODE 4 3.4.6 Y i 341 REACTOR COOLANT SYSTEM.(RCS) 3.4,6: .RCS' Loops.-MODE 4
'LC0-3.4.6 .Two loops consisting of any combination of RCS loops.and-Residual Heat Removal (RHR) loops shall:be OPERABLE, and'one a
.. ;0PERABLE loop shallibe in operation.
p-
--NOTES -- - -----
l~ , .1. LAll Reactor Coolant Pumps (RCPs) and RHR pumps may be removed fromLoperation-for s 1: hour per 8 hour period-providedi f :pJ :a. L No operations"are permitted that would cause
-o: reduction of the RCS boron concentration: and.
Ho b. Core outlet temperature is maintained'a 10*F below KV saturation temperature. w; o 2. .No-RCP shall be started with any-RCS cold leg temperature s 350 F'unless the secondary. side water tem)erature of each Steam Generator (SG) is < 50 F above
.eac1 of the PCS' cold leg temperatures.
APPLICABILITYf MODE-4. ACTIONS CONDITION' REQUIRED ACTION COMPLETION TIME
#' A~. No required loop in A.1 Suspend all Immediately
^
operation, operations . involving a reduction in RCS boron concentration. blQ A.2 Initiate action to Immediately restore one loop to operation. L (continued) > j I " gy ): H-BRAIDWOOD.- UNITS-1 & 2 3.4.6 - 1 10/8/98 Revision K j l i
.. .. . . _ _ - ~ . ~ . .- - - _ . . . - . - _. _._-_._ __._.
i u _
- RCS Loops-MODE' 4 3.4,6 .
.a
~
f'T.; ':' ACTIONS >(continued) NS ' CONDITION. REQUIRED ACTION. COMPLETION TIME:
.B. 'One . required-loop. 8,1 ' Initiate, action:to' .Imediately :
(inoperable. restore a second loop ,i to OPERABLE-status. ' 1
-AND 4
':i B.2- -NOTE----
.Only required if RHR
. loop is 0PERABLE',
w .
'Be in MODE 5. 24 hours'
-C; ?Two required loops C.1 . Suspend:all Imediately ,
; inoperable.- operations involving
- a reduction of RCS-boron concentration.
- -. 4#
AND
.C.2- Initiate: action to Immediatiely restore one loop to OPERABLE' status.
4 [SURVEILLANCEREQUIREMENTS
-SURVEILLANCE' FREQUENCY L :.SRO3.416.1. Verify required RHR'or RCS. loop is in . 12 hours h;- operation, a-
,: A1 D '
W s JSR f3.4 3.2 -VerifyJSG ~ secondary side narrow range water 12 hours . l-L e, p slevel is 2 '18% for each required RCS loop. g. 1 g ' ,ce - (continued) 7 y .M d LBRAIDWOODi-UNITSili&~2 3.4.6 - 2 8/21/98 Revision K - L. b n , . v 9 -- 2 a- w -
i RCS Loops -MODE 4 ' 3.4.6 SURVEILLANCERs0UIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.4.6.-3 Verify correct. breaker alignment and 7. days indicated power are available to each h required pump that.is not in operation. 1
- 1 1
l l l l.
\
o 1 l l O
.r J l
l I-10 ' BRAIDWOOD'- UNITS-1 & 2 3.4.6 - 3 8/21/98 Revision A
.~ ... .. ---_.---~-. -.... . -- .- - ..
RCS Loops-MODE 5. Loops Filled 3.4.7 E ?~Y 3.4' REACTOR-COOLANT SYSTEM 4(RCS) U' 3.4.7 RCS Loops-MODE:5. Loops Filled
'LCO 3.4.7- One Residual Heat Removal (RHR) loop shall be OPERABLE and:
in operation, and either: i
.a; 'One additional RHR-loop shall be OPERABLE: or !
3d' M b. The secondary side' water level of at least two Steam i 33., Generators (SGs):shall be a 18%.
'N- NOTES -
i
%'~1 H-
- 1. The RHR pump may be removed from operation for s I hour per 8 hour period provided:
l 4 a. No operations are permitted that would cause m -reduction of the RCS boron concentration; and c: y4 b. -Core outlet temperature is maintained a 10'F below , saturation temperature, 49h -
' 2 .' . One required RHR loop may be inoperable for s 2 hours -
for surveillance testing provided that the other RHR 9, : loop is OPERABLE and in _ operation.
! 4 V 3. No reactor coolant pump shall be started with any RCS cold leg temperature s 350 F unless the secondary side water temperature of each SG is < 50 E<above each of the RCS cold. leg temperatures.
- 4. All RHR loops may be removed from operation during
_ planned heatup to MODE 4 when at least one RCS loop is in operation. APPLICABILITY: -MODE 5 with RCS' loops filled. 4 4 l %. I c - BRAIDWOOD - UNITS l'& 2 3.4.7 - 1 10/8/98 Revision K
p- ,
))'- . . -
"P RCS Loops-MODE 5. Loops Filled o 3.4.7 3
l
- : ACTIONS
, ' y- : CONDITION' REQUIRED ACTION ' COMPLETION TIME A~. No_ required RHR loop .A.I' Suspend all Imediately Lin_ operation. ' operations involving a reduction.in RCS-boron concentration. .
AND-A.2 ' Initiate action to Imediately restore one RHR loop. to operation. o tk v i
' 'i B.10ne required RHR loop - B.1 Initiate action to Imediately-4 inoperable.
. restore required RHR E- 1 cop to OPERABLE
'l status. .
~ 1 c,
'i?
rW JN. (,,) ip gd
-C. - One;or:both required SG secondary side C.1 Initiate action to restore required SG Imediately
< .r
.b" water level (s) not- secondary side water within limits, level (s) to within k limits.
(continued)- p. BRAIDWOOD - UNITS 1 & 2 3.4.7 - 2 10/5/98 Revision K
;qj: l_t. '^
~ ^
RCS Loops-MODE 5. Loops Filled 3.4.7
- l
- N, '
(ACTIONSI(continuedP l <k[ , CONDITION l ' REQUIRED ACTION . COMPLETION TIME-
- p. .
- e I
Suspend all. Immediately
'l D. LTwo' required RHR loops D.1 'l Linoperable, operations involving a reduction of RCS- l
@- boron concentration. 'j Required'RHR 1 cop- AND l !
'ino3erable and one or.-
m - boti required.SG. . D.2.1- Initiate action to Immediately , secondary side water restore one RHR loop l level (s) not within to OPERABLE status. 4 limits. _ l
- 'B' .)
J D.2.2 Initiate action to Immediately restore required SG i ," > secondary side water , level (s) to'within ! limits. + , p I
,7^( .- J
^
L ): p -SURVEILLANCE REQUIREMENTS- - SURVEILLANCE FREQUENCY i
. GQ 0
/c .SR .3.4.7.1 ' Verify' required RHR . loop is 'in operation. 12 hours t t.y .
- t] .
k- 1SR 3.4.7.2' . Verify SG secondary side' narrow range water 12 hours
%; level is a 18% in required SGs. ,
m
, SR '3.4.7.3 . Verify correct breaker alignment and - 7 days indicated power are available to each
- required RHR.' pump that is not in operation. .
.;[
; 3,
- v :.
~BRAIDWOOD - UNITS 1:& 2.- 3.4.7 - 3 10/6/98 Revision K
-- n -
RCS Loops-MODE 5. Loops' Not Filled 3.4.8 - I [Y
.f3.4 - 1 REACTORTC00LANT: SYSTEM -(RCS)
$3.4.8 ' RCS. Loops -MODE- 5. - L' oops'.Not Filled .
LCO.-3,4.8.' . Two. Residual Heat Removal (RHR): loops'shall be OPERABLE and L one. 0PERABLE RHR. loop. shall. be in operation.
.hf x . NOTES
;0~ 1. All RHR pumps may be removed from. operation for s 1 hour L ' provided:L
% i
-V .a. No operations are permitted that would cause a
% reduction of the RCS boron concentration:
N -b. The core outlet temperature is maintained 2 10 F i g below saturation temperature:.and
- c. No draining ' operations are permitted that would i' further reduce.the RCS water volume.
- 2. One RHR loop.may.be inoperable for s 2 hours for. H surveillance testing provided that the other RHR loop is j
.0PEFfBLE and in operation. !
t j-~/ i.
-APPLICABILITY: MODE 5 with RCS loops not filled. .]
ACTIONS
' CONDITION' REQUIRED ACTION COMPLETION TIME A. No required 'RHR loop A.1 Suspend all Immediately in operation. operations involving a reduction in RCS boron concentration.
AND A.2 Initiate action to Immediately restore.one RHR loop ,m to operation. (continued) L j'% N,/.! ' BRAIDWOOD '. UNITS;1.& 2' 3.4.8 - 1 10/8/98 Revision K , b i e r # y v, _
, .. . - ~ . .- - . . . - . . . . ~ . . . . . ~ . . . . . . - . - . . - . - . ... . .- -
t - j RCS Loops-MODE 5. Loops Not Filled
- 3.4.8 f,D).- _ ACTIONS (continued)-
'A" CONDITION REQUIRED ACTION COMPLETION TIME
~
B; 0ne required RHR loop B.1 Initiate action to Immediately inoperable, restore RHR loop to OPERABLE status. C. Two reouired RHR. loops C.1 Suspend all Immediately
. inoperable. operations involving reduction in RCS boron concentration.
AND i C.2 Initiate action to Immediately restore one RHR loop to OPERABLE status'. l l
' E ;
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.8.1 Verify required RHR loop is 'in operation. 12 hours SR' 3.4.8.2 Verify correct. breaker alignment and 7 days indicated power are available to each -1
-required RHR pump that is.not in operation.
- )
- BRAIDWOOD - UNITS'1 & 2-3.4.8 - 2 8/21/98RevisionA
-0> . - - - - -r - --
,--;e,. , , .2
+
[ ! Pressurizer. -. L- 3.4.9 [] . , 3.4LREACTOR'COOLANTSYSTEM(RCS)- v
-3.4.9sPressurizer LCO- 3.419 :The pressurizerLshall be OPERABLE with:
. -. cJ - a. Pressurizer water level s 92%; and- -
9 ' T b, Two groups of pressurizer heaters OPERABLE with the
- T capacity of each group a 150 kW and capable of being
'9 ' powered from redundant Engineered Safety Features (ESF) !
-. Oj . power _ supplied buses.
.g APPLICABILITY: MODES 1, 2. and 3.
.l ACTIONS' CONDITION REQUIRED' ACTION COMPLETION TIME A. Pressurizer water A.1 Be in MODE 3. 6 hours
. level not within ap): '. . limit. AND v-A.2 Fully insert.all 6 hours i rods.
AND A.3 Place Rod Control 6 hours System in a condition incapable of rod withdrawal. , AND A.4 Be in MODE 4. 12 hours
-B. One or more required B.1- Restore required 72' hours
. groups of pressurizer groups of pressurizer heaters. inoperable. heaters to OPERABLE status.
. (continued)
,.. ) BRAIDWOOD - UNITS 1 & 2 3.4.9 - 1 8/21/98 Revision K
- l
- I Pressurizer a 3.4.9 l l
ACTIONS (continued) ! D]'
- i.
' CONDITION REQUIRED ACTION COMPLETION TIME ;
1 l C. Required Action and C.1 Be in MODE 3. 6 hours !
~
associated Completion l Time'of Condition B AND ! L ' not met. . .
]
C.2 Be in MODE 4. 12 hours j l t i. [ SURVEILLANCE REQUIREMENTS SURVEILLANCE - FREQUENCY l SR 3.4.9.1 Verify pressurizer water level 'is s 92%. 12 hours . SR 3.4.9.2: Verify capacity of 'each re uired group of 18 months - pressurizer heaters is a 150.kW. SR '3.4.9.3 Verify required pressurizer heaters are 18 months capable of being. powered from an ESF power supply. I e (gj..
\
~BRAIDWOOD - UNITS 1 & 2 3.4.9 - 2 8/21/98 Revision A
r Pressurizer Safety. Valves !
'3.4.10 l l I
7 N- 3.4 . REACTOR COOLANT SYSTEM (RCS)- l L /- .l 3.4.10 Pressurizer Safety. Valves-
~
j l LCO-.-3.4.10 Three pressurizer safety valves shall be OPERABLE with lift' settings a-2460 psig and's 2510 psig. NOTE
.i P The lift settings are not required to be within the L?. LCO limits during MODE 3:for the purpose of setting the T pressurizer safety valves under ambient (hot) conditions.
Tl . This exception is allowed for 54 hours following entry into Ml MODE 3 provided a preliminary cold setting was made prior .to H heatup. o [. - ----
. APPLICABILITY: MODES 1. 2. an'd.3.
-ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME l f% )
'A. 6ne pressurizer safety. A.1 . Restore valve to 15 minutes valve inoperable. OPERABLE status.
B. ' Required Action and B.1 Be in MODE 3. 6 hours associated Completion-Time not met. BlQ 08- B.2 Be.in MODE 4. 12 hours
.Two or more pressurizer safety valves inoperable.
o
- l. / n .
l BRAIDWOOD - UNITS 1 & 2 3.4.10 - 1 8/21/98 Revision K 2
s L Pressurizer Safety Valves , 3.4.10 [N ' A,,/s
. SURVEILLANCE REQUIREMENTS.
~'
.. . SURVEILLANCE ~ FRE0UENCY, p y.
- 1. o. ..
ii < l.9,- . SR 314:10.1 zVerify each pressurizerf safety valve is In accordance with.-the
~
cc 2 OPERABLE in accordance with the Inservice
-A .. Testing Program. Following ' testing. lift Inservice Ep settings ~shal1 be within- 1%. Testing Program ,
\ 4-7, l. i y 4
,12',
k, P I r l u b 7 73:. Q )- UNITS 1 & 2 3.4.10 - 2 8/21/98 Revision K-4Y .BRAIDWOOD i'
w, , . , ,..
, - . - . . - - - . . , . - . . . - . . . . , ~ - , - . - - . - . . .
i I -
' Pressurizer PORVs; p 3.4.11' i ,.
s . .
. 3.41: REACTOR 7C00LANT SYSTEM (RCS)-
Z/N) R A"
' ^
~
[ L3.4111'. Pressurizer Power Operated l Relief Valves'-(PORVs) - i.- t,: ' LC013.4;11.. :Each;PORV and associated block. valve shall be:0PERABLE. l
.n APPLICABILITYi iMODESil. 2, and?3,
- y
, : ACTIONS 1 L --
.. ---NOTES- .
.1l Separate Condition entry is.; allowed for'each PORV and'each block' valve.-
22 ' LCO 3.0.4'is.not applicable. i CONDITION. REQUIRED ACTION. iCOMPLETION TIME y!? .A. L0ne or.mo're PORVs'.. 'A.1 ' Close and maintain l-1 hour g 2~ ! inoperable.and capable )ower.to associated- !o e W offbeing manually and alock valve. lf M 1w . automatically. Cycled, - - (continued)
-l r 3 s.
- j. ,
L \ h EM "
.tj
~
(- BRAIDWOOD - UNITS IT&'2 3.4.11 - 1 10/5/98 Revision K L
# e , _-. - -
. ~ . .
..-....7..
ug + y' ,
-u ) *
.L Pressurizer PORVs--
L 'i, , 13,4.11'
- b. 1 #'y, .
, //~T & FACTIONS- (continued)- by , Rf0 $u . CONDITION- REQUIRED ACTION COMPLETION TIME' l L., $ ::B; Lone PORViinoaerable' B.1: L Close associated 1 hour- ? t'
'and.not capa)1e of ; block valve.
[ being manually cycled. - o AND s 14 -
.QR:
A B.2 - NOTE-
.2 One:PORV ino)erable Not required-if N and not capa)1e of associated PORV M , being automatically-- remains capable of
-: 4 cycled. being. manually 1 cycled.
] ;
-e : Remove power from 1 hour associated block- J
.. valve.
MD-B.3 Restore PORV to' 72 hours
. -- 0PERABLE-status.
, yq. ; , LJ 18 C, 'One.blockLvalve C.1- Place associated PORV- 1 hour inoperable. in' manual controT. AND C.2 . Restore block valve 72' hours ' to OPERABLE. status. L:
..~ '
-Be in MODE 3;
? D. Required Action and D.1 6 hours O associated Completion N ,
Time of Condition A.' AND M B. ~or- C rat met. '
.4 -
O .~ 2 Be in MODE 4. 12 hours 4a '.
- q>
g (continued) L
- ry .
, n., :
L.. *
- BRAIDWOOD - UNITS:1 & 2 3.4.11 - 2 8/21/98 Revision K a
.i n.
\
Pressurizer PORVs ' 3,4411
- r. t lflf.,- _ - _, ..-
?ACTIONSL- '(cohtinued)/
f[~
- t '
,~ CONDITION) REQUIREDIACTION -COMPLETION TIME- .
c 4
~
Li
' CM E TwoLPORVstinoperable: E.1 Be..in MODE 3. .6 hours !
pg ' J c T . (and not; capable of' i~' K,E 'being manually cycled, AND
- .. t: #.. .
i w *Q.npab j
-QB ,
E.2 Be in MODE 14. 12 hours- j g JTwoIPORV's inoperable J'D Land notucapable of.
~o
, , %g, L Ay q ; cycled.
'being automatically.
s < , F t 4 !F. iTwo block valves. F.1 Restore one-block 2 hours- j (inoperable. valve'to OPERABLE status, p
. .t .
o L ?- T ..G. Required Action and' G.1 Be in MODE 3.- 6 hours. H
- P
's' *t ' ' associated Completion.
-Time of Condition F AND 1
l M not met. . .
'4 l-G.2 Be' in MODE 4. 12 hours
. SURVEILLANCE REQUIREMENTS _
SURVEILLANCE FREQUENCY
.. e s T -SR 3.4-11.1' .
NOTE i sVl ~ Not required ~to be met with bloci ralve _w clos'ed in accordance with the Rer ced p, WW;J Action of Condition B. LPerform a complete cycle of each block 92 days valve.
.pg (continued)
K/;, . .
< :BRAIDWOOD - UNITS'l-&'2 3.4.11 - 3 10/13/98 Revision K t
'; f .;
+ ~ , , . . , . . ,
r k - - Pressurizer PORVs L 1 3.4.11 y m I./ " SURVEILLANCE REQUIREMENTS'-(continued)- ! . SURVEILLANCE: FREQUENCY i
-SR 3.4111~.2 . .
. NOTE-
.Only' required to be performed in MODES 1.
-and-2.
' Perform'a complete cycle of each PORV. 18 months
~ SR 3.4:11.3' Perform a' complete cycle of each solenoid 18 months.
-air control valve and. check valve on the L , :- air accumulators in PORV control systems. ;
9- 1 4 i > 3 SR~'3.4.1114' Perform CHANNEL CALIBRATION-of PORV 18 months
& actuation instrumentation. .
l. j
- "h .-
L) l-g i l: I';:
- i d
L 4+
-BRAIDWOOD - UNITS l-& 2' 3.4.11 - 4 10/5/98 Revision K L
O ,. _ _ _ _ . . _. -- , --
+ . . . . . . . - - . .
, -LTOP_ System q i
3.4.12. ' u
-i f' '3.4. REACTOR' COOLANT SYSTEM-(RCS) l L')- h 3.~4.12; Low Temperature' Overpressure' Protection (LTOP) System
- 1
-g-
- v. .
61 LCO 13.4.12~- An LTOP: System shall be OPERABLE withi N ,
- a. A maximum'of one charging pump -(centrifugal) capable of
- injecting:into the RCS.
i b; No Safety Injection (SI) pumps Lcapable of[ injecting into ' the RCS-
- c. Each SI accumulator-isolated, whose pressure is greater than or. equal to the ma~ximum RCS pressure for the existing RCS cold leg temperature allowed by the P/T
' limit curves provided.in the'PTLR.,'and-d .' One of the following pressure' relief capabilities:
.1. Two Power Operated Relie' f Valves (PORVs) with lift settings:within the limits specified in the PTLR.
2 -. Two Residual' Heat Removal (RHR) suction relief
. valves with setpoints s 450 psig.
1 1 [')L u
- 3. -One PORV'with a lift setting within the limits specified,in the PTLR and one RHR suction relief valve with a setpoint s 450 psig'. or
~4, The'RCS depressurized and an RCS vent of T = 2.0 square inches.
W se . NOTE Operation in MODE 4 with all SI pumps and charging pumas u capable of injecting into the RCS is allowed when all~ RCS Ej - cold legs exceed 330 F.
-z .
G = MODES 4 and 5. M_[ APPLICABILITY: MODE 6 when the reactor vessel head is on.
- w if,
\ I
,3 n; v
'BRAIDWOOD - UNITS 1 & 2 3.4.12 - 1 10/5/98 Revision K
. . . . - . _ _ . ~ ~ . . . . - . - _ _ . . - , . _ . _ . _ _ _ . _ . _ . _ _ . . _ _ . . . _ . . , .
, w .,
LTOP System
. . 3.4.12
! go . -
V~S 11 ACTIONS:
% )? #g?.
NOTE : 4 LC0 3i014 is not applicable to the RCS pressure relief' capabilities. ,, zy. ,
, . 1 H CONDITION- , REQUIRED ACTION- COMPLETION' TIME.
- ,, . + -
A: :Two charging pumps- A.1' NOTE
- (centrifugal) capable Two charging pumps.
- of injecting-into the may be capable of RCS. injectingintothe E. RCS during pump swap'
- ;. @ operation for
~
- s 15 minutes.
C One charging pump..
-e -(positive-displacement) capable initiate action to Immediately of injecting into the. verify a maximum of i" RCS. one charging pump (centrifugal) is.
capable of injecting. ! . into the RCS.
- y%
B,J .
'B. .One or more SI pumps- B.1- : Initiate action to Immediately.
capable of injecting verify no SI pumps into the RCS.- are capable of injecting into the RCS.
;C; An accumulator not C.1 Isolate affected 1 hour-isolated when the- accumulator.
accumulator tressure
.is greater-tian or-equal to:the maximum .
- RCS-pressure for existing cold leg..
'3? , temperature allowed in the PTLR.
yL s . (continued) I' e h-VV hx , 4 BRAIDWOOD -' UNITS 1 & 2 3.4.12 - 2 8/21/98 Revision K
9, i,
),
- .. LTOPSystej a,~.u 7 s
- -ACTIONS (continued)
-CONDITION- REQUIRED ACTION . COMPLETION TIME D. Required Action and- D.1 Depressurize affected 12 hours l associated Completion. accumulator to less L Time of Condition C than the maximum RCS ,
! not met. pressure for existing ; cold leg temperature allowed in the PTLR. i t l l
'E. One required'RCS E.1 Restore required RCS 7 days relief valve. relief valve to inoperable in MODE 4. OPERABLE status.
I F. One required RCS- F.1 Restore required RCS 24 hours relief valve - relief valve to l inoperable in MODE 5- OPERABLE status. I or MODE 6 when the
., reactor vessel head is p)
L on. l G. Two required RCS G.1 Depressurize RCS and 8 hours relief valves establish RCS vent of inoperable. 2 2.0 square inches. ' Required Action and associated Completion Time of Condition ~D. E, or F not met. LTOP System inoperable for any reason other' ' than Condition A. B. , C. D. E. or'F. L l BRAIDWOOD - UNITS 1 & 2 3.4.12 - 3 8/21/98 Revision A
.,w, - ,- - -
',n _ LTOP' System 4
. 3.4.12 W:-- s. SURVEILLANCE' REQUIREMENTS i ).
- ESURVEILLANCE- FREQUENCY iSR 3i4.12.1- ; Verify no SI pump..ils capable. of injecting-12 hours
- v. into the RCS.
.; O ?
lv
$l ;:SR13.4.12.2- ~ Verify maximum a of.one charging pump 12 hours 1 . (centrifugal) 'is capable of inject ng Linto
-the RCS.
' SR- 3.4.12.3 -NOTE .
- ..Only required;to be met for accumulator whose pressure'is greater than or equal to the-maximum RCS pressure' for the existing RCS~ cold leg temperature allowed by the P/T limit curves provided.in the PTLR.
LVerify each accumulator is isolated. 12 hours'- , ;p%J yy ESRf3c4.12.4- ; Verify. required RCS vent a 2.0 square 12 hours for y* inches open. unlocked open vent valve (s) AND 31 days for locked open vent valve (s) L i le SR- 3.4 f12.5~ : Veri.fy RHR suction valves.are open for 'each 72 hours y' . required RHR-suction relief valve. 3:H p , 4l[ SR 03i4.12.61 Verify PORV. block valve. is open forleach 72 hours
' required PORV; e
L, (continued) M " LBRAIDWOOD/ ' UNITS;l & 3.4.12 10/5/98 Revision K L 2 9
_ . _ _ . .~ . _ . _ ~ _
,.shb
'hk e .
.LTOP SystemL 3;4.12-m
?"~is (SURVEILLANCE' REQUIREMENTS !(continued) H 3>' ; e. .<s SURVEILLANCE- FREQUENCY ,
-1 45Rd3.4.12-7.-
- --- .- NOTE ---
' ~ ~
'NotLrecuired'to be performed until 12 hours
.after: cecreasing RCS cold. leg -temperature -
j
'to s 350 Fc l
Perform a COT'on each required PORV!- 31 days rexcluding; actuation. -l l l SR 3.4.12.8 Per' form CHANNEL; CALIBRATION.for each 18 months' required PORV actuation channel.
.l 4
;s l
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I
'l
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1 41 l. tc i
). ,
V BRAIDWOOD - UNITS:1 & 2 3.4.12 - 5 10/5/98 Revision K L l'
RCS Operational LEAKAGE 3.4.13 A 34 REACTOR COOLANT SYSTEM (RCS)
'3.4.13 RCS.0perational LEAKAGE.
LCO 3;4.13- RCS operational LEAKAGE shall be limited to:
- i. a, :No pressure' boundary LEAKAGE:
- b. 1-gpm unidentified LEAKAGE:
- c. 10 gpm identified LEAKAGE:
- d. 600 gallons per day. total primary to secondary LEAKAGE through all Steam Generators (SGs); and
- e. 150 gallons per day primary._to secondary LEAKAGE through any one SG.
APPLICABILITY: MODES 1.-2. 3.-and 4. l ACTIONS-N xj CONDITION- REQUIRED ACTION COMPLETION TIME < 1 A. RCS' LEAKAGE not within A.1 Reduce LEAKAGE to 4 hours limits for reasons within limits. Other than pressure ; boundary LEAKAGE. l l l B. Required Action and B.1 Be in MODE 3. 6 hours
~ associated Completion
-Time of Condition A' AND
' not met.
B.2 Be in MODE 5. 36 hours DB Pressure boundary ,
~ LEAKAGE exists.
p , L i i
-l J
'BRAIDWOOD'- UNITS 1 & 2 3.4.13 - 1 8/21/98 Revision A ,
t
m- .. , < _lih' , lW! 6, J l
~
s
. ;RCS OperationaljLEAKAGE!
a 3l4.13z a I 1 [ Q~ 7 h h SURVEILLANCE REQUIREMENTS SURVEILLANCE-FREQUENCY. 1
'l u
" ' &.1
.o l
e 1 h D SR 03.14113.14 c . ~
.. . . . NOTE- . . -:--- .
,1y. 7 .LNotLrequired;tolbe' performed until: 12 hours- , y y, E1 # . lafter establishment:of steady ~ state. 11 -'
,p ' ' operation'.
l -
- i. Verify lRCSfoperational: LEAKAGE is within 72 hours-limits by,aerformance of RCS water? i .
1 inventory Jal'ance. - Lj R t u %. . ..
;SRf 3.4;13.2:' Verify steam generatorL ti2becintegrity is in In:accordance
'accordance with-the Steam. Generator Tube with-the Steam-
,j
-2 '. ,~~~ LSurve111ance Program. Generator Tube Surveillancel 1
~ Program:
i
. .i ;g l
w . (:{c} .
) 1
=
'r 94.![
^
.c qs , hp, - be h^)p it
'BRAIDWOOD -fUNITS 1 & 2 3.4.13 - 2 9/18/98 Revision K
,p
!c' +-
,.3 I
-: }-
- t. p
m . _.. , ._ - _ . . _ . - . - _ . - _ _ _ _ . _ _ . . . _ . . _
.r . .
l ' p; , l 4
,: RCS Ply Leakage s > 3.4.14 '
$ M ' REACTOR C00L' ANT SYSTEM (RCS) i [LE ..'
' ; df.. 3.4.14i.RCS PressureLIsolatioh Valve.(PIV) Leaka (g ;
L i :LC04 l 3.4?14 ~ _ Leakage from each RCS~ PIV shall be within limits. -
?
.u h : APPLICABILITY: .
. MODES-l',~2,.-3. and 4 , .
' ACTIONS' jay 19
-,4 L ;- . NOTES---s - - - Q T. [$1; :: Separate Condition entry is allowed'for each flo
,~
~
cy;- 'inoperaale by an' inoperable PIV. m 4 --- - - C0'NDT '0N REQUIRED ACTION- COMPLETION TIME
.!0ne. or mu, i. flow: paths' -
NOTE - - lp gt
+
~A.
l,with11eakage from,one Each va lve used to satisfy' AM ..or more'RCS PIVs'not- Required Action A.1 and' within limit. Required-Action A.2 must have been verified to meet-SR 3.4;14.1 and be.in the reactor coolant 3ressure boundary or.the ligh pressure-portion of the system. x A.1 Isolate.the high 4 hours-pressure portion of the affected system from the low pressure
' portion by use of one closed manual, de-ehergized power operated. ,
.. :de-activated
. automatic. or' check.
< v s valve.
[ a.e,> ; -AND L (continued)- '/] : KJ ,
- BRAIDWOOD; . UNITS =1'& 2 3.4.14 - 1 8/21/98 Revision K
'y
.< ;f4
~
,ci <
RCS PIV. Leakage 3.4.14
.. i - ,
LACTIONS
%-a : :y : CONDITION- REQUIRED ~ ACTION ' COMPLETION TIME-q, '
i' i L.. b A L(continued) A.2. Isolate the high pressure portion.of
. 72 hours w .h. - <
the.affected system. 1 L
$ ~
from the low pressure. portion:by;use of a t a second closed manual. R -- 1de-energized power Q.. operated.
^- de-activated automatic -or check I valve.
B .: -Residual Heat: Removal- B.1 ' Isolate the affected .4 hours
-(RHR) System suction flow path by'use of
' isolation valve-
-. one de-energized-interlock 1 function power' operated valve.
!' , inoperable. (( -"
'Ci Required Action and- C.1' Be in MODE 3, 6 hours associated Completion Time-not met, AND L' C.2 Be in MODE 5. 36 hours-
'f} .
M ft ' 10/1/98 Revision-K
-BRAIDWOOD - UNITS l & 2 3.4.14 - ?
- [f . , I r . + - . , , - . , . . , , - , - +n. . . = ,
w ' _
..L--
~
;RCS_PIV Leakage:
- 3.4.14-r- 4 ,
w _.,s ... . .. , W' ? SURVEILLANCE:REOUIREMENTS= '
- SURVEILLANCE' FREQUENCY
, d SR: 3d.14;1' .
.L -
NOTES
- 1, ;0nly re. quired _ to be performed in
. MODES :1 and 2;
, 22; :RCS'PIVs actuated during the- J u
performance of this. Surveillance are! -
.not'. required to be tested more than- 1 m once if a: repetitive testing l loop cannot be' avoided.
P' '
- 3. Not required to be performed for-RH8701A and B.and RH8702A-and'B on the-Frequency required following valve actuation or flow through the~ valve.
s r eg g; _ , i zy 8 f . Verify leakage from each RCS PIV is equivalent to;s:0.5 gpm per nominal inch of In accordance with'the p) ei valve size up to'a maximum of 5 gpm at an Inservice. _ ;-q Q Z': JV ' RCS pressure = 2215~psig and s 2255-psig.
~ ~
Testing Program. and f )? 18 months M. AND'
' Prior to -
enter.ing_ MODE 2, whenever the unit has been in MODE-5 for 2 7 days. if leakage testing has not been performed once within the r ;: previous 9" 9 months
- p '
V AND
.y 4 '
(continued) i
; m:-
}.. ,
4 , !xf. [l ' 'f in [ . BRAIDW0001-;UNITSi l.i '& 2 i3'4 14 - 3.
. -10/5/98 Revision K !
\; 1 , u,_
, fj
, ' ' y +
\ -y' +
- RCS:PlV Leakage- ,
'3.4.14' .l
~
[g(.. es
' SURVEILLANCE RE UIREMENTS (continued)
SURVEILLANCE . FREQUENCY- .
.i J
;y SR.R 3.4.14.1.'. .(continued) Within 24 hours following valve 1 actuation due :)
. E.
to automatic or , l manual. action-or flow ~through, the' valve' ;
, I l
.SR '3 4.14'2s . Verify RHR System suction ~ isolation valve
. 18 months-
, Linterlock 3revents the valves from being )
opened wit 1 a simulated or actual-RCS~ j pressure signal = 360 psig.
- 1 l!
s ( , l i
) ', .;f% ; ,
p.s
. w/ -
, . .BRAIDWOOD - UNITS 1E&.2 3.4.14 - 4 8/21/98 Revision K y' r
.s..,
_ k, i , _ . _ . . _ . . -
- - - . ; , .. m ~ - .. . - - . ~ ~ ~ -
. ,5 .a , - - . ~
}. , '. .y
; ,:;8 RCS' Leakage Detection Instrumentation u 3.4.15 !
o E3;4xREACTORCOOLANTSYSTEM(RCS) ifi q )*; .. . .. . .
;3i4il5LRCSLeakageDetectionInstrumentation:
.,LC0 3.4 151 . 'Thelfollowing RCS leakage detection instrumentation shall'be
, ~ OPERABLE: l 1
N 2;l a. :0ne containment sump monitor:<and w' ' w ' '
. . b' . One containment' atmosphere' radioactivity monitor
- (gaseous or' particulate).
~
1 W ! MAPPLICABILITY: ' MODES 1. 2.' 3. and 4. 1 ' lw ,
$+ -ACTIONS-l CONDITION REQUIRED' ACTION' COMPLETION TIME-5.A . Required' containment? . .. NOTE (sump monitor LC0 3.0.4 isnot. applicable.
Linoperable. --
.7;,9 ya .s v6' A.1- NOTE-W ,Not required to be -
;T$ performed until m 19- 12 hours after
, W . establishment of steady' state TU. '
l y Ml - operation. Perform SR 3.4.13.1. Once per: 24 hours AND M '
'A . 2 Restore required -30 days y ~
' containment sump 4 monitor to OPERABLE status.
e I N t: Q 1;l
.e,
'(continued).
s L ,~ , ^g
- y. [
- L. '
, JBRAIDWOODJ-UNITSL1L&2l 3.4.15 - 1 .-10/1/98 Revision K
@ s s
';.n l,
; V. , -
f
.RCS Leakage _ Detection' Instrumentation u 3.4.15
. .1 WN' / ACTIONS-- (continuedF
.:C0'N DITION REQUIRED ACTION -COMPLETION-TIME B. LRequired containmenti - - - - -
NOTE atmosphere LCO 3.0.4 is not applicable. radioactivity monitor - Linoperable. B.1.1 Analyze grab samples Once per-of the containment 24 hours 7 . atmosphere. X-
?o- gg .
V- B.1.2 . NOTE--- .
!Q Not required to be g . performed until- '
i 4 :: 12 hours'after Y establishment of i
%g - .
steady state
~l- operation.
Perform SR 3.4.13.1. Once per ,. x 24 hours ;
-( h.
U MD B.2 Restore required 30 days containment- .! atmosphere I radioactivity monitor to OPERABLE status, i C. Required Action and C.1 Be in MODE 3. 6 hours associated Completion Time-not met. AND C.2 Be in MODE 5. 36 hours .e
. D. jAll required monitors D.1 Enter LCO 3.0.3. Immediately
[l inoperable. L i-l' . 17y 4 D BRAIDWOOD -~ UNITS 1 &.2 3.4.15 - 2 10/1/98 Revision K g, l . . ,) [ --. 3 I / \ c3
'l
.~ . .
l RCS Leakage Detection Instrumentation 3.4.15
.' SURVEILLANCE REQUIREMENTS
-SURVEILLANCE- FREQUENCY SR 3.4 15.1~ Perform CHANNEL CHECK of the required 12 hours
. containment atmosphere radioactivity ;
L' monitor. 1
.t:
E SR 3c4.15.2 Perform COT of the required containment 92 days atmosphere radioactivity monitor. l l SR 3.4.15.3 Perform CHANNEL CALIBRATION of the required 18 months containment sump monitor. 1 I l SR 3.4.15.4 Perform CHANNEL CALIBRATION of the required 18 months 1 containment atmosphere radioactivity 1 monitor. l f% ~J l l l p
.I
(> BRAIDWOOD'- UNITS 1 & 2 3.4.15 - 3 8/21/98 Revision A g y - g - e. w g -y c ,,,.c-. ---
, .. ~. - . - ,. . . . . .. .. .-
; .P . It 6#
RCS Specific' Activity 3.4.16.
- [
L314-REA'CTORCOOLANTl SYSTEM (RCS) j L N'M 13.4116 .LRCS Specific' Activity. 1 L. :LC01 3<4.161 The; specific: activity of the reactor coolant-shall be within i ' the followingulimits-1 t (a. (Dose: Equivalent?I-131 specific activity 51.0 uCi/gm: 'l and-
- Jrg .b.. .Gro'ss' specific' activity s'100/2 uCi/gm.
Wi2 k' l W MODES 1'and'2.- .
&g[: APPLICABILITY: MODE 3 with RCS average temperature (Tm) = 500 F.
ACTIONS-
' CONDITION REQUIRED ACTION COMPLETION TIME AL DOSE-EQUIVALENT I-131 . NOTE .
' specific activity: LC0 3.0.4-is not applicable.
- > 1.0 'pC1/gm.' - - - -
'(w~)-.
'v i A .1 ' Verify DOSE Once per 4 hours EQUIVALENT I-131-
- specific activity within the acceptable region of Figure.3.4.16-1.
6.N_Q i A.2 Restore DOSE 48 hours EQUIVALENT I-131 . specific activity to - within limit. L i. (continued) L f~} y. BRA'IDWOOD '. UNITS 1 & 2 3.4.16 - 1 8/21/98 Revision K
- o
. : RCS Specific Activity -
3.4.16
< ACT: DNS :(continued)
- CONDITION REQUIRED ACTION ~ COMPLETION TIME
'B; Required Action and B .1- Be in MODE 3 with 6 hours
. , associated Completion T,4 < 500*F.
Time.of Condition A.
.not . met ,.
. 0E
. DOSE EQUIVALENT I-131 specific activity-in :
'the
, unacceptable region of
- Fi gure -. 3. 4.16-1.
C. . Gross specific C.1 ' Be in MODE '3 with 6 hours activity.not.within T,y < 500*F. limit. f~N);; bi 1 1 SURVEILLANCE REQUIREMENTS I SURVEILLA.NCE FREQUENCY
- SR 3.4.16.1
. Verify reactor coolant gross specific 7. days.
activity 5 100/E gCi/gm. (continued) l
. s) '
(BRAIDWOOD - UNITS l'&-2 3.4.16 - 2 8/21/98 Revision A
RCS Specific Activity 3.4.16 ' g i SURVEILLANCE'RE0VIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.4.16.2 NOTE b Only required to be performed in MODE 1.
]
.s 7 Verifv reactor coolant DOSE EQUIVALENT 14 days T I-131" specific. activity 5 1.0 pCi/gm.
f6 AND H Xl-
" Between 2 and 6 hours after a THERMAL POWER change of a-15% RTP within a 1 hour period
~
-SR '3.4.16.3 -
. . -NOTE -
Not required to be performed until 31 days
.after a minimum of 2 effective full ~O' - -
power days and 20. days of MODE 1 operation have elapsed since the reactor was last subcritical for a 48. hours. g Determine E from a reactor coolant sample 184 days taken in MODE 1 after a minimum of
-[ 2 effective full aower. days and 20 days of g MODE 1 operation lave elapsed since the e reactor was last subcritical for q! 2.48 hours.
1 C
- BRAIDWOOD - UNITS 1 & c 3.4.16 - 3 8/21/98 Revision K
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~RCS. Specific Activity i p,yA l3.4.16-
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- Figure 3.~4.16-1-(page 1 of 1). i
., : Reactor Coolant DOSE EQUIVALENT I-131 Specific ~ Activity
, , Limit Versus _ Percent of LRATED THERMAL POWER' n:a _,
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BRAIDWOOD.- UNITS 1:& 2 3.4.'16-4 9/18/98 Revision 8
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RCS Loop Isolation Valves i 3.4.17~ 3.4 REACTOR' COOLANT SYSTEM (RCS)
- 3.4.17 RCS Loopilsolation Valve - I l
LC0;;3.4.17
~
.Each RCS hot and cold leg loop isolation valve shall be open with power removed from each isolation valve operator.
~
L .
. i
~ APPLICABILITY-:
MODES 1. 2. 3.;and 4. -l I
-ACTIONS i E- ----- - -
. NOTE --
l
. Separate Condition entry is allowed for each-RCS loop isolation valve. l l
CONDITION REQUIRED ACTION COMPLETION TIME i A. Power available to one' A.1 Remove power from 30 minutes l
-or more loop isolation '. loop isolation valve valve operators. operators.
~
lg l l := l B. NOTE B.1 Maintain ' valve (s) Immediately ! All Required Actions- closed. l L .shall be completed- j E whenever this AND-- 1 Condition-is' entered. -
- B.2- Be in MODE 3. 6 hours -)
t. o Ono or more RCS loop AND isolation: valves'
.: closed B.3 Be in MODE 5. 36 hours i.
L.. , l L < l
~ i cBRAIDWOOD - UNITS 1 & 2 3.4.17 - 1 8/21/98 Revision A f-L l-.
.- -. - - . . . . - . . - .. . ...~ - -. - . .- . - . - - - - . . - -
l RCS Loop Isolation Valves 3.4.17 l I f t
'" ? SURVEILLANCE ~ REQUIREMENTS
. SURVEILLANCE FREQUENCY i' SR 3.4.17.1 Verify each RCS.. loop iselation valve is 31 days
'open and power is removed from each loop
-isolation valve operator.
f l l i O l i l 4 l lJ L. > - - '.vO 1 t B'RAIDWOOD - UNITS 1 &'2 3.4.17 - 2 8/21/98 Revision A
. .. . .- , , . . . . . .. . - , - . . . . ~ ~ . -. . .. .
RCS Loops-Isolated
- 3.4.18 i
~j
.3.4f REACTOR COOLANT SYSTEM'(RCS)-
'.7"J 3.4.18 .RCS Loops-Isolated-l r l LC0; 3.4.18_ ' Each RCS isolated . loop shall remain isolated with:
- a. The hot and; cold leg loop stop isolation valv'es closed
. if boron concentration of the isolated loop is--less than 7 the required SDM boron concentration of the unisolated
,, p ' portion of the RCS: and d b. The cold leg loop stop-isolation valve closed if the cold leg temperature of the isolated loop is > 20 F' 4 below the highest cold leg temperature of the unisolated 4 portion of the'RCS. APPLICABILITY: MODES 5 and 6.
, ACTIONS
. CONDITION- REQUIRED ACTION COMPLETION TIME fy M A. _ Isolated loop hot _or A.1 Close hot and cold Immediately cold-leg isolation leg isolation valves'.
valve'open with t nron concentration
' requirement not inet.
B. Isolated loop cold leg B.1 Close cold leg Immediately
. isolation valve open i. solation valve, with temperature-requirement not met.
i L ? f} SJ u -BRAIDWOOD - UNITS-I & 2- 3.4.18 - 1 10/5/98 P.evision K
, 4 .- . - - ,
. . . . . - . . - - . . . _ - . - . ~ . - - . _ . - . . - - -. ..
s
'RCS Loops-Isolated i 3.4.18 7u g TSURVEILLANCEREQUIREMENTS-
-:-I e -
- Y:. SURVEILLANCE FREQUENCY
.7 y
, iSR. 3.4 18.1- . Verify cold leg ' temperature of- isolated Within j? -loop is is 20 F below the highest cold leg
~
_30 minutes
- tenperature of the unisolated portion of prior to
.the.RCS, - opening.the I cold .: leg y ' isolation valve? .
i in the isolated
- i. loop M
x
%- ~
%"'SR'3.4.18.2 ' '
Verify boron concentration of isolated loop Within 4 hours-is greater than or. equal.to the required prior.to l 1 SDM boron concentration of the unisolated opening the hot
,- . portion of the RCS. or cold leg ]i
^
isolation valve !
. in the isolated I loop'
.i i
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> 1 E'?
y Q.)y[ . JBRAIDWOOD:- UNITS 1 & 2 i 4.18 - 2 10/5/98 Revision K L. , u
; 1' -
7.!-
r RCS Pressure. Temperature, and Flow DNB Limits i B 3.4.1 (N B 3.4 REACTOR-COOLANT SYSTEM (RCS) l B'3.4 1 RCS Pressure. Temperature. and Flow Departure ~fror. Nucleate Boiling (DNBT Limits ( [ . BASES b BACKGROUND. These Bases address requirements for maintaining RCS pressure.L temperature and flow rate'within limits assumed in the safety analyses. The safety analyses (Ref 1) of 4 normal operating conditions and anticipated operational l m, - occurrences assume initial conditions within the normal steady state envelope. .The limits ) laced on RCS pressure, temperature, and flow rate ensure tlat the departure from
'l. nucleate boiling.(DNB) will.be mer for each of the transients analyzed.
The RCS pressure limit is consistent with operation within the nominal operational envelope. Pressurizer pressure indications are averaged to come up with a value for comparison _to the limit. A lower pressure will cause the-reactor. core to approach DNB limits.
- ,_y limit is Et The RCS coolant consistent with fullaverage power temperature operation w (T,d)hin the nominal b).. operational . envelope. Indications of temperature are averaged to. determine a value for comparison -to the limit.
A higher average temperature will cause the core to approach DNBLlimits. The RCS flow rate normally remains constant during an-
- operational fuel cycle with all pumps running. The minimum-RCS flow limit corresponds'to that assumed for DNB analyses.
Flow rate indications are averaged to come up with a value
- for comparison to the limit. A lower RCS flow will cause '
the. core to approach DNB limits. Operation for significant periods of time outside these DNB limits increases.the likelihood of a fuel cladding failure in a DNB limited event. h n. i, ,)- BRAIDWOOD UNITS.1 & 2 B 3.4.1 - 1 10/13/98 Revision K t
+ , , , - , , , , ,
w,.. ;- a - n -.- - ~ . . i ! RCS.P'ressure. Temperature, and Flow DNB Limits B 3,4.1 f~ L}i; : BASES
-APPLICABLE. 'The requirements of this LCO represent the initial SAFETY. ANALYSES' conditions for DNB limited transients analyzed in the plant safety' analyses (Ref. 1). The safety analyses have shown
':go that transients-initiated from the limits of this LCO will l result.in meeting the DNBR criterion of a 1.4. This.is the acceptance limit for the RCS DNB parameters. Changes to the-
.N- unit that could impact these parameters must be assessed for
-their impact on.the DNB criteria. The transients analyzed (Q$6'~ l- for include loss of coolant flow events and dropped or stuck rod. events. A key assumption for the analysis of these events is that the core power distribution is within the limits of LCO 3.1.6. " Control-Bank Insertion Limits;" i LCO 3.2.3. " AXIAL FLUX DIFFERENCE (AFD):" and LCO 3.2.'4
# "OUADRANT POWER TILT RATIO (0PTR)."
o' . i
- j. Safety Analyses assumed a value:of 2207 psia (2192.3 psig).
P This value is bounded by the LCO value of 2219 psig assuming ,
- g l a measurement accuracy of less than 26.7 psi. I Safety Analyses assumed a value of 588.4 F for the vessel average temperature. In addition, the analyses assumed the calculated error (including the 4*F dead band for the rod control system) for the temperature is 8.74 F (2a random
- (~Y. error of 7.6 F plus the 1.14*F bias error). The value !
d assumed in the non-Revised Thermal Design. Procedure ! (non-RTDP) analyses is -8 F. +9.5 F. For the RTDP analyses. ; a value of 7.6 F with a bias of +1.5 F is assumed. 'l Safety Analyses-assumed a total RCS flow rate of-358.800 gpm. This value is bounded by the LCO value of. 371.400 gpm assuming a flow measurement uncertainty of 3.5%. 1 This 3.5% flow measurement uncertainty assumed in the analyses included errors from all sources including fouling .
'in the venturi . The use of 3.5% flow error is acceptable'if i actual uncertainty is unknown. At the time analyses were i performed. tne flow accuracy was unavailable. Subsequent calculations determined the error to be less than 3.5%. !
- Any fouling that might bias the flow rate measurement greater than 0.1% can be detected by monitoring and trending various plant performance parameters. If detected, either the effect of the fouling shall be quantified and
. ' compensated for in the'RCS flow rate measurement or the venturi shall be cleaned to eliminate the fouling.
l p, X.J BRAIDWOOD - UNITS 1 & 2 B 3.4.1 - 2 10/13/98 Revision K
RCS Pressure. Temperature, and Flow DNB Limits B 3.4.1 \. .. . " 'T ' ' BASES' . V APPLICABLE'. SAFETY ANALYSES (continued)
'The RCS DNB parameters' satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).
T LLCOL .This LCO specifies limits on the monitored process W variables-pressurizer pressure. RCS average T1 temperature-(Tm). and RCL total flow rate-to ensure the l core operates within the limits assumed in the safety V : analyses; . Operating within these limits will-result in I
-- meeting the DNB design criterion in the event ~of a DNB 9 limited' transient.
;~
4 A Note has been added to indicate the limit on pressurizer
% is not applicable during short term operational: transients such as a THERMAL POWER ramp increase > 5% RTP'per min te or
.O". a THERMAL POWER step increase > 10% RTP. These conditions represent short term perturbations where actions to control pressure variations night be counterproductive. Also. since-they typically represent transients initiated from power levels < 100% RTP, an increased Departure from Nucleate .
Boiling Ratio (DNBR) n:argin exists to offset the temporary I J .;V pressure. variations. U
;Another set of limits on DNB related parameters is provided I in SL 2.1.1 " Reactor Core SLs." LC0 3.4.1 re) resent-s the .
. initial conditions of the safety analysis whic1 are far more ~!
restrictive than.the Safety. Limit (SL).. Should a violation of this LC0 occur, the operator must check whether or not an SL may have been exceeded. APPLICABILITY In MODE 1, the limits on 3ressurizer pressure, RCS coolant average temperature, and RCS total flow rate must' be maintained during steady state' operation in order to ensure DNB design criteria will be met in the event of an unplanned
- loss of forced coolant flow or other DNB limited transient.
In all other MODES, the power level is low enough that DNB is not a concern. 1 3 <u.
>BRAIDWOOD - UNITS 1 & 2 B 3.4.1 - 3 ~ 8/22/98 Revision K
RCS Pressure. Temperature, and Flow DNB Limits B 3.4.1 BASES v , ACTIONS .A_1 RCS pressure and RCS average temperature are controllable pl and measurable parameters. With one or both of these j parameters not within LC0 limits, action must be taken to restore-parameter (s). RCS total flow rate is not a controllable parameter and is
.y not expected to vary during steady state operation. If the o indicated RCS total flow rate is below the LCO limit power 1 must be reduced, as required by . Required Action B.1. to q
restore DNB margin and eliminate the potential for violation of the accident analysis bounds. u 5l The 2 hour Completion Time for restoration of the parameters provides sufficient time to adjust unit parameters, to determine the cause for the off normal condition, and to restore the readings within limits, and is based on plant operating experience. B.1
- If Required Action A.1 is not met within the associated Completion Time, the unit must be brought to a MODE in which lV )
the LCO does not apply. To achieve this status the unit must be brought to at least MODE 2 within 6 hours. In MODE 2. the reduced power condition eliminates the potential for violation of the accident analysis bounds. The Completion Time of 6 hours is reasonable to reach the required unit conditions in an orderly manner. SURVEILLANCE SR 3.4.1.1
~REQUIREMENTS Since Required Action A.1 allows a Completion Time of 2 hours to restore parameters that are not within limits.
- the 12 hour Surveillance Frequency for pressurizer pressure is sufficient to ensure the pressure can be restored to a normal operation, steady state condition following load changes end other expected transient operations. The 12 hour interval has been shown by operating practice to be sufficient to regularly assess for potential degradation and to verify operation is within safety analysis assumptions.
( s'
'J BRAIDWOOD - UNITS 1 & 2 B 3.4.1 - 4 8/22/98 Revision K
RCS Pressure. Temperature, and Flow DNB Limits B 3.4.1 h& : BASES. U SURVEILLANCE REQUIREMENTS (continued) SR 3.4.1.2 f r . l Since Requi_ red Action A.1 allows a Completion Time of I. 2 hours-to restore parameters that are not within limits. I. <the-12 hour Surveillance Frequency for RCS average temperature (T is sufficient to ensure the temperature can be restoref)o t a normal operation, steady state condition following load changes and other expected , transient operations. The 12 hour interval has been shown ' by operating practice to be sufficient to regularly assess ! for potential degradation and to verify operation is within l safety analysis assumptions. l SR 3.4.1.3 E m The 12 hour Surveillance Frequency for RCS total flow rate lc is performed using the installed flow instrumentation. The required minimum RCS flow rate is met with a 95% indicated flow rate. The 12 hour interval has been shown by operating practice to be sufficient to regularly assess potential degradation and to verify operation within safety analysis assumptions. i l: 1
'4.. I i
l
- . l L
i i (-
.d,. ;
;BRAIDWOOD ~ UNITS 1 & 2 B 3.4.1 - 5 8/21/98 Revision A I'
y RCS. Pressure Temperature. and Flow DNB Limits 4 B 3.4.l. '#(7 U BASES
' SURVEILLANCE REQUIREMENTS (continued)-
a 1e SR- 3.4 L 4
"[p Measurement. of RCS total . flow rate by performance..of a
-!allows precision the calorimetric installed RCS heat. balance flow once everyto:be instrumentation 18 months calibrated and verifies the actual. RCS flow rate.is greater than or equal' to .the minimum required RCS flow rate.
7 The Frequency of-18: months reflects the im'p ortance of
]t -
verifying flow after a' refueling outage when the core has-h' -.been altered which may have caused'an alteration of flow 3' resistance. This SR is modified by a Note that allows entry into MODE 1. without having. performed the SR. and placement of the~ unit in'the best condition for performing the .SR. The: Note states that the SR is not required to be performed until-
.7-days after = 90% RTP. This exception is appropriate since the-heat balance requires the unit to be at a minimum of 90% RTP to obtain the stated RCS flow accuracies.- -The
- Surveillance shall'be performed within 7 days after~ reaching
,s . - 90% RTP.
ll ). a x./ REFERENCES 1 L1. UFSAR,; Chapter 15. g ( 0j%
- V
,BRAIDWOOD'--UNITS 11&.2- B 3.4.1 8/22/98 Revision K
RCS Minimum Temperature for Criticality B 3.4.2
.B 3.4 REACTOR COOLANT SYSTEM (RCS)
.B 3.4.2 RCS Minimum Temperature for Criticality BASES' BACKGROUND LThis LCO is based upon meeting several major considerations before the reactor can be made critical and while the reactor is critical.
'The first consideration is Moderator Temperature Coefficient l
(MTC). LCO 3.1.3. " Moderator Temperature Coefficient (MTC)." In the transient and accident analyses, the MTC is assumed to be in a range from slightly positive to neg6tive and the operating temperature is assumed to be within the nominal operating envelope while the reactor is critical. The LCO on minimum temperature for criticality helps ensure the unit is operated. consistent with these assumptions. The second consideration is the protective instrumentation. t. L Because certain protective instrumentation (e.g., excore neutron detectors) can be affected by moderator temperature, a temperature value within the nominal operating envelope is chosen to ensure proper indication and response while the " ,,-s i. reactor is critical. V. The third consideration is the pressurizer operating L characteristics. The transient and accident analyses assume that the pressurizer is within its normal startup and operating range (i.e., saturated conditions and steam bubble present). It is also assumed that the RCS temperature is within its normal expected range for startup and power operation. Since the density of the water, and hence the response of the pressurizer to transients, depends upon the initial temperature of the moderator, a minimum value for ,. moderator temperature within the nominal operating envelope j -is chosen. L The fourth consideration is that the reactor vessel is above its minimum nil ductility Reference temperature when the reactor is critic 61. This parameter is also assured through compliance with LCO 3.4.3. "RCS Pressure and Temperature (P/T) Limits." l L i 4 ..
- .BRAIDWOOD - UNITS 1 & 2 B 3.4.2 - 1 8/21/98 Revision A l.
l
l RCS Minimum Temperature for Criticality I B 3.4.2 l l '\
' BASES l
APPLICABLE 'Although the RCS minimum temperature for criticality is not
- SAFETY ANALYSES' 'itself an initial condition assumed in Design Basis i Accidents (DBAs) the closely aligned temperature for Hot i
! Zero Power (HZP) is a process variable' that is an initial i
- condition of DBAs. such as the Rod Cluster Control Assembly l (RCCA) withdrawal. RCCA ejection. and main steam line break accidents performed at zero power that either assumes the failure of, or presents a _ challenge to, the integrity of a fission product barrier.
All low power safety analyses assume initial RCS loop temperatures creater than or equal to the HZP temperature of i 557 F (Ref. 1),. This minimum temperature for criticality ' limitation provides a small band, 7*F. for critical operation below HZP.. This band allows critical operation below HZP during unit startup and does not adversely affect any safety analyses since the MTC is not significantly affected by the.small temperature difference between HZP and the minimum temperature for criticality. l 1 The RCS minimum temperature for criticality satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO _ Compliance with the LCO ensures that the reactor will not be. made or maintained critical (k 2 1.0) at a temperature i lessthanasmallbandbelowtI5eHZPtemperature,which'is I ca ssumed in the safety analysis. Failure to meet the ! requirements of this LCO may produce initial conditiont ; inconsistent with the initial conditions assumed in the l safety analysis. APPLICABILITY .In MODE 1 and MODE 2 with k ,, a 1.0. LCO 3.4.2 is applicable since the reactor can only be critical (kg , a 1.0) in these MODES. i r-i- nvO L BRAIDWOOD - UNITS 1 & 2- B 3.4.2 - 2 8/21/98 Revision A L
l RCS Minimum Temperature for Criticality l B 3.4.2 I l BASES
) APPLICABILITY (continued)
The special test exception of LCO 3.1.8. " MODE 2 PHYSICS l TESTS Exceptions." permits PHYSICS TESTS to be performed at I s 5% RTP with RCS loop average temperatures slightly lower : than normally allowed so that fundamental nuclear I characteristics of the core can be verified. In order for ! nuclear characteristics to be accurately measured. it may be l 7 necessary to operate outsidd the normal restrictions of this ; i LCO. For example. to measure the MTC at beginning of cycle. 9 it is necessary to allow RCS loop average temperatures to
- . which may cause RCS loop average fall below T*t 'o* fall below the temperature limit of this temperatures
' LCO.
. ACTIONS A,l l
If the parameters that are outside the limit cannot be restored, the unit must be brought to a MODE in which the . LC0 does not apply. To achieve this status, the unit must l be brought to iODE 2 with k ,, < 1.0 within 30 minutes. ,
-s Ra)idreactorshutdowncanbereadilyandpractically l
) aclieved within a 30 minute period.
u The Completion Time is reasonable, based on operating experience, to reach MODE 2 with k ,, < 1.0 in an orderly manner and without challenging pla,nt systems. SURVEILLANCE SR 3.4.2.1 REQUIREMENTS RCS loop average temperature is required to be verified a 550 F once every 12 hours. The SR to verify RCS loop average temperatures every 12 hours is frequent enough to prevent the inadvertent violation of the LC0 and takes into account indications and alarm.s that are continuously available to the operator in the control room. REFERENCES 1. UFSAR. Section 15.0.3. ,'V) BRAIDWOOD - UNITS 1 & 2 B 3.4.2 - 3 8/22/98 Revision K
RCS P/T Limits B 3.4.3 l B 3.4 REACTOR COOLANT SYSTEM (RCS) B 3.4.3 RCS Pressure and Temperature (P/T) Limits 1
. BASES BACKGROUND All components of the RCS are designed to withstand effects l of. cyclic' loads due to system pressure and-temperature
! changes'. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. This LC0 limits the pressure and temperature changes during RCS heatup and cooldown within the design ! assumptions and the stress limits for cyclic operation. The PTLR contains P/T limit curves for heatup. cooldown. Inservice Leak and Hydrostatic (ISLH) testing. and data for the maximum rate of change of reactor coolant temperature (Ref. 1). Each P/T limit curve defines an acceptable region for normal operation. The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and f compared to the applicable curve to determine that operation is within the allowable region. The LC0 establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the Reactor Coolant Pressure Boundary (RCPB). The vessel is the component most subject-to brittle failure, and the LC0 limits apply to the entire RCS-(except the pressurizer). The limits do not apply to the pressurizer, which has different design characteristics and operating functions. 10 CFR 50, Appendix G (Ref. 2), requires the establishment l of P/T limits for specific material fracture toughness requirements of the RCPB materials. Reference 2 requires an adequate margin to brittle failure during normal operation, anticipated operaticnal occurrences, and system hydrostatic tests. It mandates the use of the American Society of Mechanical Engineers (ASME) Code, Section Ill. Appendix G (Ref. 3). i L
.BRAIDWOOD - UNITS 1 & 2 B 3.4.3 - 1 8/21/98 Revision A
RCS P/T Limits B 3.4.3 ! "b r'] . BASES BACKGROUND (continued) l ! .The neutron embrittlement effect on the material toughness is reflected by increasing the Nil Ductility Reference
. Temperature (RTup) as exposure to neutron fluence increases.
The actual shift in the RTun of the vessel material will be , established periodically by removing and evaluating the l irradiated reactor vessel material specimens, in accordance with ASTM E 185 (Ref. 4) and Appendix H of 10 CFR 50 (Ref. 5). The operating P/T limit curves will be adjusted. ' as necessary, based on.the evaluation findings and the recommendations of Regulatory Guide 1.99 (Ref. 6). The P/T limit curves are composite curves established by . superimposing limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive. At any specific pressure, temperdture, and temperature rate of change, one location within the reactor vessel will dictate the most restrictive limit. Across the span of the P/T limit curves, different locations are more ! restrictive, and, thus, the curves are composites of the l most restrictive regions. ) l -
\
The heatup curve represents a different set of restrictions j than the cooldown curve because the directions of the thermal gradients through'the vessel wall are reversed. The thermal gradient reversal alters the location of the tensile stress between the outer and inner walls during heatup and cooldown, respectively. The criticality limit curve includes the Reference 2 requirement that it be a 40 F above the heatup curve or the cooldown curve, and not less than the minimum permissible temperature for ISLH testing. However, the criticality curve is not operationally limiting; a more restrictive limit exists in LC0 3.4.2, "RCS Minimum Temperature for Criticality." l i l' LO L BRAIDWOOD - UNITS 1 & 2 B 3.4.3 - 2 8/21/98 Revision A L
- _ . _ . . _ __.m._. . _ . _ . _ _ _ _ _ . _ . _ _ . _ _ . _ . _ _ _ _ _ _ . _ _
RCS P/T Limits B 3 4.3 BASES
' BACKGROUND-(continued) l The consequence of violating the LC0 limits is that the RCS has been operated under conditions that can result in brittle, failure of the RCPB. possibly leading to a nonisolable leak or-loss of coolant accident. In the event these limits are exceeded. an evaluation must be performed to determine the effect on the structural-integrity of the RCPB components. The ASME Code. Section XI. Appendix-E , (Ref. 7), provides a recommended methodology for evaluating
- l. an operating event that causes an excursion outside the limits.
I-
' APPLICABLE The P/T limits are not derived from Design Basis Accident (
SAFETY ANALYSES (DBA) analyses. They are prescribed during normal operation to avoid encountering pressure. temperature. and temperature rate of change conditions that might cause undetected flaws ! to propagate and cause nonductile failure of the RCPB. an L unanalyzed condition. Reference 1 establishes the L methodology for determining the P/T limits. Although the L P/T limits are not derived from any DBA. the P/T limits are 7 acceptance limits since they preclude operation in an unanalyzed condition, RCS P/T limits satisfy Criterion 2 of ; 10 CFR 50.36(c)(2)(ii). LCO The two elements of-this LCO are:
- a. The' limit curves for heatup, cooldown, and ISLH L testing: and 1
- b. Limits on the rate of change of temperature. l 1
The LC0 limits -epply to all components of the RCS. except l~ the pressurizer. These limits define allowable operating regions and permit a large number of. operating cycles while providing a wide margin to nonductile failure. u l !O L BRAIDWOOD - UNITS 1 & 2 B 3.4.3 - 3 8/21/98 Revision A F - - . _- _.
RCS P/T Limits B 3 4.3-i BASES ~ , ' LCO (continued) The limit's for the rate of change of temperature control the ! - thermal gradient through the vessel wall and are used as L inputs for calculating the heatup, cooldown, and ISLH testing P/T limit curves. Thus, the LCO for the rate of o change of temperature restricts stresses caused by thermal l gradients-and also ensures the validity of tne P/T limit curves.
' Violating the LCO limits places the reactor vessel outside of the bounds of the stress analyses and can increase
. stresses in other RCPB components. The consequences depend on several factors, as. follow:
i
- a. - The. severity of the departure from the allowable operating P/T regime or the severity of the ra'e t of change of temperature:
!' b. . The length of time the limits were violated (longer violations allow the temperature gradient in the thick l vessel walls to become more pronounced): and
- f. c. The existences, sizes, and orientations of flaws in
- the vessel material.
APPLICABILITY The RCS P/T limits-LC0 provides a definition of acceptable operation for prevention of nonductile failure in accordance with 10_CFR 50. Appendix G (Ref. 2). Although the P/T limits were developed to provide guidance for operation during heatup or cooldown (MODES 3. 4. and 5) or lSLH testing their Applicability is at all times in keeping with the concern for nonductile. failure. The Ap)licability includes MODE 6 and conditions with no fuel in t1e reactor vessel. This 3rovides continued prevention of nonductile failure even w1ile the reactor is "defueled" so that the RCS is acceptable for operation when fuel is returned to the reactor vessel. The limits do not apply to the pressurizer. l i f1 .. , QQ D L - BRAIDWOOD - UNITS 1 & 2 B 3.4.3 - 4 8/21/98 Revision A i g y'** 4 w' 4 "
I ,. RCS P/T Limits
)
B 3.4.3 BASES (~T (v APPLICABILITY (continued) During MODES 1 and 2. other Technical Specifications provide 1
. limits for operation that can be more restrictive than or )i can supplement these P/T limits. LCO 3.4.1. "RCS Pressure. !
Temperature, and Flow Departure from Nucleate Boiling (DNB) J l Limits:" LC0 3,4.2. "RCS Minimum Temperature for L Criticality;" and Safety Limit 2.1 " Safety Limits:" also j L provide operational restrictions for pressure, temperature and maximum pressure. Furthermore. MODES 1 and 2 are above the temperature range of concern for nonductile failure, and
- stress analyses have been performed for normal maneuvering l profiles; such as power ascension or descent.
l l ACTIONS A.1 and A.2 l Operation outside the P/T limits during MODE 1. 2. 3. or 4 ' must be corrected so that the RCPB is returned to a l condition that has been verified by stress analyses. The 1 30 minute Completion Time reflects the urgency of restoring i the parameters to within the analyzed range. Most l violations will not be severe, and the activity can be ' accomplished in this time in a controlled manner. l ' Besides restoring operation within limits, an engineering i evaluation'is required to determine if RCS operation can l continue. The evaluation must verify the RCPB integrity-remains acceptable and must be completed before continuing operation. Several methods may.be used. including comparison with pre-analyzed transients in the stress-analyses, new analyses or inspection of the components. For the vessel beltline only. ASME Code. Section XI. Appendix E (Ref. 7), may be used to support the evaluation. l-
'The 72 hour Completion Time is reasonable to accomplish the
-evaluation. The evaluation for a mild violation is possible within this time, but more severe violations may require l special, event specific stress analyses or inspections. A favorable evaluation must be completed before continuing to operate-.
BRAIDWOOD - UNITS 1 & 2 B 3.4.3 - 5 8/21/98 Revision A l l l
.. . - _ . - = _ - - - - - . - - - . - - _ - - . _ - - . . . - . . -
RCS P/T Limits B 3.4.3 L BASES ACTIONS (continued) i Condition A is modified by a Note requiring Required i Action A.2 to be completed whenever the Condition is l entered. The Note emphasizes the need to perform'the ' evalJation of the effects of the excursion outsioe the allowable _ limits. _ Restoration alone per Recuired Action A.1 is. insufficient because higher than analyzec stresses may l have occurred and may have affected the RCPB integrity, 1 B.1 and B.2
'If a Required Action and associated Com)letion Time of-Condition A are not met, the unit must Je placed in a' lower MODE because either the RCS remained in an unacceptable P/T region for an extended period of increased-stress or a sufficiently severe event caused entry into an unacceptable region. Either possibility indicates a need for more careful examination of the event, best accomplished
, . with the.RCS at reduced pressure and temperature. In reduced pressure and temperature conditions, the possibility of propagation with undetected flaws is decreased.- , If the required restoration activity of Required Action A.1 !- cannot be accomplished within 30 minutes. Required Action _B.1 and Required Action B.2 must be implemented to reduce pressure and temperature. If the required evaluation for continued operation cannot be accomplished within 72 hours.or the results are indeterminate or. unfavorable, action must proceed to reduce
. pressure and temperature as specified._in Required Action B.1 and Required Action B,2. A favorable engineering evaluation must be completed and documented before returning to
. operating pressure and temperature conditions Pressure and temperature are reduced by bringing the unit to MODE 3 within 6 hours and to MODE 5 within 36 hours.
The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging plant systems, e i I b ~ 8/21/98 Revision A
~BRAIDWOOD - UNITS 1-& 2 B 3.4.3 - 6
_ , , . , -_. . .__ .~ _
RCS P/T Limits B 3.4.3 i BASES ACTIONS:(continued) C.1 and C.2 Actions must be initiated immediately to correct operation
.outside of the P/T limits at times other than when in MODE 1., 2. 3. or.4. so that the RCPB is returned.to a condition that hat, been verified by stress analysis. Tne immediate Completion Time reflects the urgency of initiating action to restore the parameters to within the analyzed-range. Most.. violations will not be severe, and the activity can be accomplished in this time in a controlled. manner.
Besides restoring operation within limits, an evaluation is required to determine if RCS operation can continue. The evaluation must verify that the RCPB integrity remains. acceptable and must be completed prior to entry into MODE 4. Several methods may be.used, including comparison with pre-analyzed transients in the stress analyses, or inspection of the components. For the vessel' beltline only. ASME Code. Section XI.
=
Appendix E (Ref. 7), may be.used to support the evaluation. !(>v ) Condition C is modified by. a Note requiring Required Action C.2 to be completed whenever the Condition is entered. _The Note emphasizes the need to perform the L evaluation of the effects-of the excursion outside~the I allowable limits. Restoration alone per Required Action C.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity. l l l l L r. Lo 8/21/98 Revision A. BRAIDWOOD - UNITS'1 &-2 B 3.4.3 - 7 f:, _ , _ _ , . _ . _ , - .
__ , _ . . ~ _ . . _ _ _ . . _ _ . .. _ - - . - . . ._ RCS P/T Limits L B 3.4.3 ~i x . 1 !D BASES ! ~ Q.
- SURVEILLANCE SR 3.4.3.1 L- REQUIREMENTS ,
Verification that operation is within .the PTLR~ limits is required'every 30 minutes when RCS 3ressure and temperature conditions.are undergoing planned c1anges. This Frequency 1 is considered-reasonable.in view of the control room i indication available to monitor RCS status. Also, since temperature rate of. change limits are specified in hourly L increments. 30 minutes permits assessment and correction for ! minor deviations within a reasonable . time. 1 Surveillance.for heatup. cooldown, or ISLH testing may be l ' , f ' discontinued when.the definition given in the relevant plant procedure for ending the activity is satisfied. d
.L i This SR'is modified by a Note that only recuires th'is SR to l be performed during system heatup, cooldown, and ISLH <!
testing. This SR is.not required during critical operations because the combination of LC0 3.4.2 establishing a lower i bound and the Safety Limits establishing an upper bound will 1
< provide adequate controls to prevent a change in excess of H 100*F prior to entry into the performance condition of
'g heatup and cooldown operations. ' k.] l REFERENCES 1. WCAP-14040. " Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves." June 1994.
- 2. 10 CFR 50... Appendix G.
- 3. ASME. Boiler and Pressure Vessel Code. Section III.
. Appendix G.
- 4. ASTM E 185-82. July 1982.
'5. 10 CFR 50. Appendix H.
- 6. Regulatory Guide 1.99. Revision 2. May 1988.
- 7. ASME. Boiler and Pressure Vessel Code. Section XI.
Appendix E. n W LBRAIDWOOD. " UNITS 1 & 2 B 3.4.3 - 8 8/22/98 Revision K L, p'.'
RCS Loops-MODES 1- and 2 B 3,4.4 { B 3.4 REACTOR COOLANT SYSTEM (RCS). B 3.4.4. RCS Loops-MODES 1 and 2 i I BASES I ; BACKGROUND The primary function of the RCS is removal of the-heat . generated in the fuel due to the fission process. and , ' transfer of this heat, via-the Steam Generators (SGs) to l the secondary. plant. l l- The secondary-functions of.the RCS include:
- a. Moderating the neutron energy level to the thermal state, to increase the probability of fission; b .. Improving the neutron economy by acting as a reflector:
- c. Carrying the soluble neutron poison.-boric acid:
- d. Providing a second barrier against fission product release to the environment: and
'.O e .- Removing the heat generated in the fuel due to -fission ,. product decay following a unit shutdown. The reactor coolant is circulated through four loops connected in parallel to the reactor vessel, each containing L an SG ~a Reactor Coolant; Pump (RCP). and appropriate flow and temperature. instrumentation for both control and protection. The reactor vessel contains the clad fuel. The
~
SGs 3rovide the heat sink to the isolated secondary coolant. The RCPs circulate the coolant through.the reactor vessel L
.and SGs at a sufficient rate to ensure proper heat transfer and prevent fuel damage. This forced circulation of the reactor coolant ensures mixing of the coolant for proper boration and chemistry control, l
&q i _BRAIDWOOD - UNITS 1-& 2- B 3.4.4 - 1 8/21/98 Revision A t L
RCS Loops-MODES 1 and 2 B 3.4.4 h l . p' BASES L . APPLICABLE . Safety analyser contain various assumptions for the design SAFETY ANALYSES bases. accident initial. conditions including RCS pressure. RCS temperature, reactor power level. core parameters, and safety system setpoints. The important aspect for this LCO is the reactor coolant forced flow rate. which is
. represented by the number of RCS loops in service.
Both transient and steady state analyses have been performed to establish the effect'of flow on the Departure from Nucleate Boiling'(DNB). The transient and accident analyses for the plant have been performed assuming four RCS loops are in operation. The majority of the plant safety analyses are based on initial conditions at high core power or zero power. The accident analyses that are most important to RCP operation are the .four Sump coastdown single pump locked rotor, single pump (bro wn shaft or coastdown), and rod withdrawal events (Ref. 1). The Uncontrolled Rod Cluster Control Assembly Bank Withdrawal from a Subcritical or Low Power Startup Condition and the spectrum of Rod Cluster Control Assembly Ejection events were analyzed assuming only two of four RCPs in operation. This conservatively bounds operation in the lower modes. Analyzing these transients with only two RCPs in operation will result in a lower Departure from Nucleate Boiling Ratio (DNBR) thus producing more limiting results. Steady state DNB analysis has been per 'ormed for the four RCS loop operation. For four RCS loo) operation, the steady state DNB analysis, which generates t1e pressure and temperature Safety Limit (SL) (i.e.. the DNBR limit) assumes a maximum power level of 118% RTP. This is the design overpower condition for four RCS loop operation. The value for the accident analysis setpoint of the nuclear overpower (high flux) trip is 109% RTP and is based on an analysis assumption that. bounds possible instrumentation errors. The DNBR limit defines a locus of pressure and tem)erature points that result in a minimum DNBR greater t1an or equal
.to the critical heat flux correlation limit.
~
l LO BRAIDWOOD - UNITS 1 & 2 83.4.4-2 8/21/98 Revision A
I. - RCS Loops-MODES 1 and 2
~~
B~3.4.4 f~}; A ,,
. BASES'
.~~' APPLICABLE: SAFETY ANALYSES (continued)
-The unit is designed to operate with all RCS loops in- .
operation to maintain DNBR above the SL. during-all normal L . operations and' anticipated transients. By ensuring heat-1 ^ transfer in the. nucleate-boiling region, adequate heat
' transfer!is provided between the fuel cladding and the'
$' reactor coolant. l l 1 .RCS = Loops-MODES 1 and.2- satisfy Criterion 2 of 1
~ '
1 10 CFR 50.36(c)(2)(ii). j LCO. LThe purpose of this LCO is to require -an adequate: forced flow rate -for core heat " removal . Flow is represented by the . number of RCPs in operation for removal?of heat'by~the-SGs; j To meet safety analysis acceptance criteria for DNB. .four . t pumpsLare required at rated power. !
- (
'1 .
"An OPERABLELRCS loopLconsists of an OPERABLE RCP 'in o)eration.providing-forced flow for heat transport and an j
2 0)ERABLE-SGlin accordance with the Steam Generator Tube < Surveillance Program.
%gf _
t h '
.i f
k 3 u E , [N/$
.B 3.4.4 - 3 '8/22/98 Revision K
-[' ~BRAIDWOODI- UNITS 1 & 2.
I JL I 4 l sMy , s , -- < p.
RCS Loops-MODES 1 and 2 B'3.4.4 i . 8ASES.- APPLICABILITY In M0 DES 1 and 2. the reactor is critical and thus has the potential to produce maximum THERMAL POWER. Thus, to ensure that the assumptions of the accident analyses. remain valid, all RCS loops are required to be OPERABLE and in operation in these MODES to prevent-DNB and core damage. The decay. heat production rate is.much lower than the full
- power heat rate. As such. the forced circulation flow and
! heat' sink requirements are reduced for. lower. noncritical i MODES as indicated by the LCOs for MODES 3. 4. and 5. I i , , Operation in other MODES is covered by: j LCO 3.4.5 "RCS -Loops -MODE 3": l LC0 3.4.6, "RCS Loops -MODE 4": L LCO 3.4.7. "RCS Loops -MODE 5. Loops Filled": ,
- LCO 3.4.8. "RCS Loops-MODE 5. Loops Not Filled".
LCO 3.9.5. " Residual Heat Removal (RHR) and Coolant ' u Circulation-High Water Level" (MODE 6); and l LC0 3.9.6. " Residual Heat Removal (RHR) and Coolant < l Circulation-Low Water Level" (MODE 6). ACTIONS A.1 If the requirements of the LC0 are not met..the Required
- Action is to reduce power and bring the unit to MODE 3.
This lowers power level and thus reduces the core heat removal needs and minimizes the possibility of violating DNB limits.
-The Completion. Time of 6 hours is reasonable; based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging safety systems.
,J, BRAIDWOOD --UNITS-1 & 2 B 3.4.4 - 4 8/21/98 Revision A L y t
.(.
- ,- , .-n--- w -- - > , = = - - - . , , , , , -- =, y .,
! l I RCS Loops-MODES 1 and 2 l B 3.4.4 ! , 1 l - (N BASES
\s.-) !
SURVEILLANCE SR 3.4 4.1 REQUIREMENTS This SR requires verification every 12 hours that each RCS loop is in operation. Verification may include flow rate. temperature, or pump status monitoring, which helps ensure ! that forced flow is providing heat removal while maintaining l l the margin to DNB. The Frequency of 12 hours is sufficient l j considering other indications and alarms available to the l l operator in the control room to monitor RCS loop i performance.
)
1 l REFERENCES 1. UFSAR, Chapter 15, i l l f-~ l N.]x l l l l l l
<s v
BRAIDWOOD - UNITS 1 & 2 B 3. .4 <4 - 5 8/21/98 Revision A l l
RCS Loops-MODE 3 l B 3.4.5 l Wg 'B 3.4' REACTOR COOLANT SYSTEM (RCS)- l
' B 3.4.5 RCS Loops -MODE 3 BASES
- q l
In' MODE 3. the primary function of the reactor coolant is BACKGROUND removal of decay heat and transfer of this heat, via tne Steam Generator (SG) to the secondary plant fluid. A ; secondary function of the reactor coolant is to act as a I carrier for soluble neutron poison, boric acid. l The reactor coolant is circulated through four RCS loops. connected in parallel to the reactor vessel, each containing an SG a Reactor Coolant Pump (RCP), and appropriate fiow. pressure, level, and temperature. instrumentation for control, protection, and indication. The reactor vessel contains the clad fuel. The SGs provide the heat sink. The i RCPs circulate the water through the reactor vessel and SGs l at a sufficient rate to ensure proper heat transfer and I prevent-fuel damage. In MODE 3. RCPs are used to provide forced circulation for heat removal during heatup and cooldown. The MODE 3 decay l- , []. heat removal requirements are low enough that a single RCS ,b loop with one RCP running is sufficient to remove core decay l heat. However, two RCS loops are required to be OPERABLE to l ensure redundant capability for decay heat removal. APPLICABLE Uhenever the Rod Control System is capable of rod l SAFETY ANALYSES withdrawal (i.e. , the Reactor Trip Breakers (RTBs) are in !~ the closed position and the Control Rod Drive Mechanisms (CRDMs) are energized) an inadvertent rod withdrawal from i subcritical, resulting in a power creursion is possible j (Ref. 1). Such a transient could be caused by a malfunction l of the rod control system. In additioc. the possibility of f a power excursion due to the ejection of an inserted control rod is possi.51e with the breakers closed or open. Such a L transient could be caused by the mechanical failure of a L CRDM. L 1 L f c l
- O BRAIDWOOD - UNITS 1 & 2 B 3.4.5- 1 8/21/98 Revision A i
i
.--,m . , _ , _. . > - . - - + - .
. _ - . . . _ . . _ - _ _ . _ ~ ___ . _ . . . _ . . _ . . _ _ _ _ _ . . . _
. . . ._. __ _ _ ~ _ .
- RCS Loops-MODE 3 B 3.4.5-
- BASESl!
APPLICABLE. SAFETY ANALYSES'(continued) Therefore, in MODE 3 with the Rod Control System capable of rod withdrawal, accidental control rod withdrawal from subcritical is postulated and requires at least two RCS loops to be OPERABLE and in oceration to ensure that the
? accident analyses. limits are met. For those conditions wnen
. the Rod Control System is not' capable of rod withdrawal, two-RCS loops are required to be OPERABLE. but only one RCS loop is required to be in operation to be consistent with MODE 3 accident analyses.
7 Failure to provide decay heat removal may result.in challenges to a fission product barrier. The RCS loops are part of the primary success-path that functions or actuates , to prevent or mitigate a Design Basis Accident or transient that either assumes the failure of..or 3 resents a challenge to.- the integrity of a fission product Jarrier. RCS Loops-MODE 3 satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LC0 The purpose of this LCO is to require that at least two RCS' ' b' L/- loops be OPERABLE. In MODE 3 with the Rod Control System L capable of rod withdrawal, two RCS loops are required to be
.in operation due to the postulation of a aower excursion .
because of an inadvertent control rod wit 1drawal. The required number of RCS-loops in operation ensures that the y Safety Limit criteria will be met for all of the postulated L accidents. i l When the Rod Control System is not capable of rod
. withdrawal, only one RCS loop in operation is necessary to L
b ensure removal of decay heat from the core and bomogenous L'
- boron concentration throughout the RCS. An additional RCS loop is required to be OPERABLE to provide backup. forced flow capability
(: . , l C r i L. t . ~ i ] : BRAIDWOOD - UNITS 1 & 2 B 3.4.5 - 2 8/21/98 Revision A l 4.s , - e e 4 - .A. . . , , ., .--,,,,,.n-. ..,v.- m,,,w,..w-- -ww e . . , - . - ,
RCS Loops-MODE 3 B 3.4.5' 1 BASES ['] V-'
'LCO (continued)l r ;
The Note permits all RCPs to be removed from operation (i .e. , not' in operation)' for-s 1 hour per 8 hour period. The purpose-of the Note is to perform tests that are designed'to validate various accident analyses values. One ; of theseLtests is validation of the pump coastdown curve' ' used as input to a-number.of accident analyses including a ';
-loss of flow accident. This test is generally performed in '
MODE 3 during the initial startup testing program, and'as i such should only be performed once. .If, however, changes are made to the RCS that would cause:a change to the flow J characteristics of the RCS. the input values of the . . ccastdown curve must be revalidated by conducting the test 1 again. Another test performed during the startup testing l program is the validation of rod drop times during cold conditions, both with and without flow. The no flow test may be 3erformed in MODE 3, 4. or 5~ and requires that the pumps Je stopped for a short period of time. The Note permits the stopping of the pumps in order to perform this test and validate the assumed analysis . values. As with the validation of the pum coastdown curve, o pY this test should be 3erformed only once un ess the . flow-t characteristics of tie RCS are changed. The 1 hour time ; d period-specified is adequate to perform the desired tests. I and operating experience has shown that boron stratification is not a problem during this short period with no forced flow. ;
. Utilization of the Note is permitted 3rovided the following i fj conditions are met, along with any otler conditions imposed !
by proceduresi
- a. No operations are permitted that would dilute the RCS !
.. . boron concentration, thereby maintaining the margin to
.A; criticality. Boron reduction is prohibited because a L .,, : uniform concentration distribution throughout the RCS y cannot be ensured when in natural circulation: and D b. Core outlet. temperature is maintained at least 10 F below saturation temperature, so that no vapor bubble
~
%y H may form and possibly cause a natural circulation flow
[ obstr'uction. s yx ' .\,
;BRAIDWOOD . UNITS'1 & 2- B 3.4.S- 3 10/5/98 Revision K 6
, - - . .r,. . . - .. * - , ,
# I
. RCS Loops-MODE 3 -
B 3.4.5
.i
,. j f* : BASES ~. ;
d TLC 0:(continued)= i< 'An OPERABLE RCS loop consists of one OPERABLE RCP and one i C
.0PERABLE SG in accordance with the Steam Generator Tube
* = Surveillance Program. which has the minimum water level specified .in SR 3.4.5.2. : An RCP is OPERABLE if it is ,
1' ; capable of being powered and is' able to. provide forced flow ' if required. j .
..t
. iAPPLICABILITY- - In. MODE 3. this LC0 ensures forced circulation of the ;
-reactor coolant to1 remove decay heat from the core and'to !
provide proper. boron mixing. The most stringent. condition 1 of the LCO. that,is, two RCS loops OPERABLE and two RCS " loops .in operation applies 1to MODE 3 with the Rod. Control i
' System' capable of rod withdrawal.. The least stringent i condition, that is, two RCS loops OPERABLE and one RCS loop
'in operation.- applies.'to MODE 3 with the Rod Control System i not. capable of rod withdrawal.
Operation in other MODES is covered by: y.-# -LC0 3.4.4. "RCS Loops-MODES 1 and 2": l
' LC0 3.4.6. "RCS Loops -MODE 4":.
'k ):.
LCO 3.4.7. "RCS Loops-MODE 5. Loops Filled":. i I LCO 3.4.8. "RCS' Loops -MODE 5.' Loops Not Filled"; 1 LCO 3.9.5, " Residual: Heat Removal (RHR) and Coolant Circulation-High' Water Level" (MODE- 6): and LC0 3.9.6.'" Residual Heat Removal (RHR) and-Coolant
+ . Circulation-Low Water Level" (MODE 6).
i ' (:. g I [. oy BRAIDWOOD UNITS 1 & 2 83.4.5 8/22/98 Revision K l s e e < ,n - u -- n , --
RCS Loops-MODE 3 B 3.4.5
/'N BASES O
ACTIONS A.1 l If the required RCS loop is not in operation, and the Rod Control System is capable of rod withdrawal. the Required Action is to place the Rod Control System in a conmtion i incapable of rod withdrawal (e.g.. disable all CRDMs by l opening the RTBs or de-energizing the motor generator (MG) ! sets). When the Rod Control System is capable of rod withdrawal. it is postulated that a power excursion could occur in the event of an inadvertent control rod withdrawal. This mandates having the heat transfer capacity of two RCS loops in operation. If only one loop is in operation, the Rod Control System must be rendered incapable of rod withdrawal. The Completion Time of 1 hour to defeat the Rod Control System is adequate to perform these operations in an orderly manner without exposing the unit to risk for an undue time period. i B.1 and B.2 i If no RCS loop is in o)eration with the Rod Control System not capable of rod wit 1drawal, except as permitted by the Note in the LCO section, all operations involving a
/ reduction of RCS boron concentration must be suspended, and V; action to restore one RCS loop to operation must be immediately initiated. Boron dilution requires forced '
circulation for proper mixing. i The immediate Completion Time reflects the importance of maintaining operation for heat removal. The action to restore must be continued until one loop is restored to operation. l r
,\ ;v)
BRAIDWOOD - UNITS 1 & 2 B 3.4.5 - 5 8/21/98 Revision A
RCS Loops-MODE 3 B 3.4.5 fN BASES U' ACTIONS (continued) C .1. C . 2. and C . 3. . l If no'RCS loop is in operation with the' Rod Control System l capable of rod withdrawal. except as permitted by the Note .
. in the LCO section. or.if the Required Action and associated I Completion Time of Condition A are not met. action must be
.. initiated to place the Rod Control System in a condition
- incapable of rod withdrawal (e.g., disable all CRDMs by
- opening the RTBs or de-energizing the MG sets).
Additionally. all operations involving a reduction of RCS L boron concentration must be suspended, and action to restore , one of the RCS loops to operation must be immediately
~
1 initiated. Boron dilution requires forced circulation for ! proper. mixing and disabling the CRDMs removes the i possibility of an inadvertent rod withdrawal. The immediate Completion Time reflects the. importance of maintaining. operation for heat removal. The action to restore must-be continued until one loop is restored to operation.
.- M If one required RCS loop is inoperable, redundancy for heat removal is lost. The Required Action is restoration of the i required RCS loop to OPERACLE status within the Completion Time of 72 hours. This time allowance is a justified period l
to be without the redundant nonoperating loop because a single le in operation has a heat transfer capability L greater tnan that needed to remove the decev heat produced l - in the: reactor core and because of the low probability of a L failure in the remaining . loop occurring during this period. p l L.1 If the Required Action and associated Completion Time of ;
- Condition D are not met the unit must be brought to MODE 4.
In MODE 4. the unit may be placed on the Residual Heat Removal System. The additional Completion Time of 12 hours is compatible with required operations to achieve cooldown and depressurization from the existing unit conditions in an L orderly manner and without challenging plant systems. L pp y I ' BRAIDWOOD - UNITS 1 & 2 B 3.4.5- 6 8/21/98 Revision A e
% /
s .- - _
o r . -RCS Loops-MODE 3
~ B 3.4:.5 Cc fBASESc
\,
'~ )* ACTIONS (continued):
2 FL1. F.2.'and F 3 If two required RCS' loo)s'are inoperable, action must be initiated.to place the' Rod Control System in a condition incapable'of rod withdrawal.(e.g., disable all CRDMs by 1 opening the RTBs or de-energizing the MG sets). .All operations involving a reductinn of RCS boron concentration must be suspended,.and action to' restore one of the RCS loops to OPERABLE status must be initiated. Boron dilution-Trequires forcedtcirculation for proper mixing, and disabling the1CRDMs' removes the possibility of an' inadvertent rod
, withdrawal:. The~immediate Completion Time reflects the-
'importance of maintaining the capability for heat removal.
The action'to restore.must1 be continued until one loop is-restored to OPERABLE status. E 1SURVEliLANCE : SR 3.4.5'.1
. REQUIREMENTS .
This SR rennires verification every 12 hours that- the required opt. Jting loops are in operation. Verification may T, include flow rate,. temperature, and pump status monitoring. [_ J ' which helps ensure that forced flow is providing heat
. removal. The Frequency of'12 hours is sufficient considering other indications and alarms available to the operator in the control room to monitor RCS loop >
. performance. ,
s . SR 3.4.5.2 3 SR 3.4.5.2 requires verification of required SG OPERABILITY. I' o- SG OPERABILITY is verified by ensuring that the secondary s t- l side narrow range water level is = 18% for each required RCS
- loop. If the SG secondary side narrow range water level is 7 l. < 18%. the tubes may become uncovered and the associated k
loop.may not be capable of providing the heat sink for removal of the decay heat. The'12 hour Frequency is considered adequate in view of other indications available F -in-the control room to alert the operator to a loss.of SG level. .jm. k.f
'BRAIDWOODj-UNITS 1&2 B 3.4.5 - 7 8/22/98 Revision K t
--w , ,
- y. .. . -. -. - . . . - , - - - . - . , . .. . . . .- ..-
RCS . Loops -MODE 3 I B 3.4.5
, i ;"~% ': BASES' -
i)' ..
. SURVEILLANCE REQUIREMENTS-(continued)
F .; SR' 3.4.5.3
^ N;
" Verification that the required RCPs are OPERABLE ensures i that safety analyses-limits are met. -The requirement-also. :)
Ln' -a): ensures that an additional RCP can be:placed-in operation- ,
. if needed, to maintain decay heat removal and reactor l
1 coolant circulation. Verification is performed by verify'ing. j properLbreaker alignment and power availability to the ; Lrequired RCP. The Frequency of.7 days is considered j reasonable in-view of other administrative controls-available and has been shown to'be acceptable'by operating
., s ,;.
= experience.
~
REFERENCES- UFSAR..Section 15.4.1. l
- q ' '
I J%~~j1/ : , 1 l i' Li 4 ,s A 7y-
,BRAIDWOOD:- UNITS 11.& 2
. B 3.4.5 - 8 10/5/98 Revision K 4
RCS Loops -MODE 4 B 3 4.6 L(~N B 3.4 REACTOR COOLANT SYSTEM (RCS) q) B 3.4.6 RCS Loops -MODE 4 BASES BACKGROUND In MODE 4. the primary function of the reactor coolant is the removal of decay heat and the transfer of this heat to either the Steam Generator (SG) secondary side coolant or the component cooling water via the Residual Heat Removal (RHR) heat exchangers. The secondary function of the reactor coolant is to act as a carrier for soluble neutron poison, boric acid. In MODE 4. the reactor coolant is circulated through at ; least two of the four RCS loops connected in parallel to the reactor vessel, each loop containing an SG. a Reactor , Coolant Pump (RCP) and appropriate flow, pressure, level. l and temperature instrumentation for control. 3rotection and indication. The RCPs circulate the coolant t1 rough the reactor vessel and SGs at a sufficient rate to ensure proper hen transfer and to prevent boric acid stratification. In MODE 4. RHR loops can be used in lieu of RCS loops to (m kJ i provide forced circulation. The intent of this LCO is to provide forced flow from at least one RCP or one RHR loop for decay heat removal and transport. The flow provided by 1 one RCP loop or RHR loop is adequate for decay heat removal, j The other intent of this LC0 is to require that two paths be - available to provide redundancy for decay heat removal. 1 1 i l I APPLICABLE In MODE 4. circulation of the reactor coolant increases the SAFETY ANALYSES time available for mitigation of the accidental boron dilution event. The RCS and RHR loops provide this circulation i ! RCS Loops-MODE 4 satisfies Criterion 4 of l 10 CFR 50.36(c)(2)(ii). 1 2 j \ k._.,/ BRAIDWOOD - UNITS 1 & 2 B 3.4.6 - 1 8/21/98 Revision A i.
RCS Loops-MODE 4 m B 3.4.6 f'T - , BASES L)
-LCO- The purpose of this LC0'is to require that at least two
-loops be-0PERABLE in MODE 4 and that:one of these loops be tin. operation.. The LC0 allows the two. loops that are '
. required to be OPERABLE to consist of any combination of RCS loops and RHR loops. Any one loop in operation'provides enoughLflow to remove the decay heat from the core with forced circulation. -An additional-loon is. required to be
-~0PERABLE to provide redundancy for heat remova'.
. Note-1 permits all RCPs and RHR pumps to be removed from ;
operation for s 1 hour per 8 hour period. The purpose of. the Note .is to permit tests that are designed to~ validate various accident analyses-values. One of;the. tests - performed during the startup testing program is the . 1 validation of rod drop times during cold conditions. both !
- with and~without' flow. The no flow test may be perfonned in , 4 MODE 3. 4',1or 5 and requires that the pumas be stopped for a ,
, .short period of time. The Note permits t1e stopping of the <
pumps'in order to perform this test:and validate the assumed: analysis values If necessary,:this test may also'be?
- conducted after.the initial startup. testing program. 'The-1 hour time period is adequate to perform the test, and operating experience has shown that boron' stratification is 1,,
<s_,J not~ a problem during this short period with nofforced flow, Utilization of Note 1.is permitted provided the.following!
conditions are met along with any other conditions imposed by procedures: l 'a , - No operations are permitted that would dilute the.F.CS beran concentration, therefore maintaining the margin ,
- to criticality. Boron reduction is prohibited because y . a uniform concentration distribution throughout tne .
,3 RCS cannot be ensured when in natural circulation; and
- Y 4 b, Core outlet temperature is maintained at least 10 F l
M below saturation temperature, so that no vapor bubble l may: form and possibly cause a natural circulation flow L obstruction. o [y') BRAIDWOOD - UNITS l'& 2 L B 3.4.6 - 2 10/5/98 Revision K-S
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RCS Loops-MODE 4 B 3.4.6 LBASES; LLC 0 (continued) Note 2~ requires that the secondary side water temperature of
- each SG be < 50*F above each of the RCS cold leg temperatures before the start of an RCP witn.any RCS cold leg temperature:s 350 F. This restraint is to prevent a low temperature'ovarpressure event due to a thermal transient when an RCP 1s started.
- An OPERABLE RCS loop comprises an OPERABLE RCP and an l OPERABLE SG which has the minimum water. level specified in .
SR 3.4.6.2-Siihilarly'for the RHR System. an OPERABLE RHR loop'is l comprised of an OPERABLE RHR pump capable of providing L' forced flow to an OPERABLE RHR heat exchanger. RCPs and RHR j pumps are OPERABLE if they are capable of being powered _and are-able to provide forced flow if required.: 1 4 1
~ APPLICABILITY. In MODE 4, this LC0 ensures forced circulation.of the-reactor: coolant to remove decay heat from the core and to provide proper boron mixing. One loo) of either RCS or RHR provides sufficient circulation for t1ese purposes..
OL nv However, two loops consisting of any combination-of RCS and l
- RHR inops are required to be OPERABLE to provide adequate -l redun(ancy for decay heat removal. l Operation in other MODES is covered by:
LCO 3.4.4. "RCS Loops-MODES 1 and 2": i LCO 3.4.5, "RCS Loops-MODE 3": E- LC0 3.4.7 "RCS Loops-MODE 5, Loops Fiiled": LC0 3.4.8. "RCS Loops-MODE 5 Loops Not Filled";
< LC0 3.9.5, " Residual Heat Removal (RHR) and Coolant H: Circulation-High Water Level" (MODE 6); and l l
l LC0 3.9.6 " Residual Heat Removal (RHR) and Coolant
- . Circulation-Low Water Level" (MODE 6). -i l
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- BRAIDWOOD. UNITS 1 &'2 B 3.4.6 - 3 8/21/98 Rt 'ision A y
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l l RCS Loops-MODE 4 l B 3.4.6 l
. l 1
[] v BASES I ACTIONS A.1 anc A 2 If no loop is in operation, except during conditions permitted by the Note in the LCO section, all operations involving a reduction of RCS boron concentration must be suspended and action to restore one RCS or RHR loop to operation must be immediately initiated. Boron dilution , requires forced circulation to provide proper mixing and l preserve the margin to criticality. The immediate ! Completion Times reflect the importance of maintaining I operation for decay heat removal . B.1 and B.2 l If one required RCS '< ;HR loop is inoperable and only one required loop remains UPERABLE. the intended redundancy for i heat removal is lost. Action must be initiated to restore a second RCS or RHR loop to OPERABLE status'. The immediate , Completion Time reflects the importance of maintaining the ; availability of two paths for heat removal. l If the one required OPERABLE loop is an RHR loop and if the m required loop is not restored to OPERABLE status. the unit I i must be brought to MODE 5 within 24 hours. Bringing the V unit to MODE 5 is a conservative action with regard to decay heat removal. With only one RHR loop OPERABLE. the intended redundancy for decay heat removal is lost and, in the event of a loss of the remaining RHR loop it would be safer to initiate that loss from MODE 5 (s 200*F) rather than MODE 4 (200 to 350*F). The Completion Time of 24 hours is a reasonable time, based on operating experience. to reach 1 MODE 5 from MODE 4 in an orderly manner and without challenging plant systems. C.1 and C.2 If no loop is OPERABLE. all operations involving a reduction of RCS boron concentration must be suspended and action to restore one RCS or RHR loop to OPERABLE status must ce initiated. Boron dilution requires forced circulation to provide proper mixing and preserve the margin to criticality. The immediate Completion Times reflect the importance of maintaining the capability for decay heat removal. p
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BRAIDWOOD - UNITS 1 & 2 . B 3.4.6- 4 8/21/98 Revision A y x
RCS Loops-MODE 4 B 3 4.6
~s
. BASES s
SURVEILLANCE SR 3.4.6.1 REQUIREMENTS This.SR requires verification every 12 hours that the required operating RCS or RHR loop is in operation. Verification may include flow rate, temaerature or pump status monitoring, which helps ensure tlat forced flow is providing heat removal. The Frequency of 12 hours is sufficient considering other indications and alarms available to the operator in the control room to monitor RCS and RHR loop performance. 1 SR 3.4.6.2 i SR 3.4.6.2 requires verification of required SG OPERABILITY,
-' s' SG OPERABILITY is verified by ensuring that the sec.ondary .
i side narrow range water level is a 18% for each required RCS Jl loop. If the SG secondary side narrow range water level is
* < 18%. the tubes may become uncovered and the associated El loop may not be capable of providing the heat sink necessary for removal of decay heat. The 12 hour Frequency is considered adequate in view of other indications available in the control room to alert the operator to the lot.s of SG level,
,i., \ ! LJ SR 3.4.6.3 Verification that the required pum) is OPERABLE ensures that an additional RCS or RHR pump can 3e placed in operation. if i needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to the required pump. The Frecuency of 7 days is considered reasonable in view of other acministrative controls available and has been shown to be acceptable by operating experience. REFERENCES None.
~_-l BRAIDWOOD - UNITS 1 & 2 B 3.4.6- 5 8/22/98 Revision K
RCS Loops-MODE 5. Loops Filled B 3.4.7
'Q t B 3.4 REACTOR COOLANT SYSTEM (RCS) -LJ B 3.4.7 RCS Loops-MODE 5. Loops Filled BASES MCKGROUND In MODE 5 with the RCS loops filled, the primary function of the reactor coolant is the removal of decay heat and transfer this heat either to the Steam Generator (SG) secondary side coolant via. natural circulation (Ref. 1) or the component cooling water via the Residual Heat Removal (RHR) heat exchangers. While the principal means for decay heat removal is via the RHR System, the SGs via natural circulation are specified as a backup means for redundancy.
Even though the SGs cannot produce steam in this condition, they are capable of being a heat sink due to their large contained volume of secondary water. As long as the SG secondary side water is at a lower temperature than the reactor coolant. heat transfer will occur. The rate of heat transfer 1s directly proportional to the temperature difference. The secondary function of the reactor coolant is to act as a carrier for soluble neutron poison, boric acid. O In MODE 5 with the RCS loops filled. the reactor coolant is
.'d circulated oy means of two RHR loops connected to the RCS.
each loop containing an RHR heat exchanger, an RHR pump, and appropriate flow and temperature instrumentation for control, protection, and indication. One RHR pump circulates the water through the RCS at a sufficient rate to prevent boric acid stratification. The number of loops in operation can vary to suit the operational needs. The intent of this LCO is to 3rovide forced flow from at least one RHR loop for decay leat removal and transport. The flow provided by one RHR loop is i adequate for decay heat removal. The other intent of this ! LCO is to require that a second path be available to provide
- redundancy for heat removal.
The second path can be another OPERABLE RHR loop or two OPERABLE SGs to provide an alternate method for decay heat removal via natural-circulation. (3 s._, BRAIDWOOD - UNITS 1 & 2 B 3.4.7 - 1 8/21/98 Revisica A
RCS Loops-MODE 5. Loops Filled B 3.4.7 BASES.
. APPLICABLE- In MODE 5. RCS circulation increases the time available for SAFETY ANALYSIS mitigation of an accidental boron dilution event. The RHR
. loops provide this circulation and have been identified as- l importat contributors to risk reduction.
RCS Loops-MODE 5. Loops Filled, satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii). w ! o '- ;
, i
- 2 7 LC0 The surpose of this 1.C0 is to require that at least one of' r G [' the RHR. loops be OPERABLE and in operation with an
, ' additional RHR . loop OPERABLE or two SGs with secondary side water -level = 18%. One RHR loop provides sufficient forced l M, r,- circulation to Jerform the safety functions of the reactor i coolant under t1ese conditions. An additional RHR loop is
~@*i required.to be OPERABLE to meet single failure y considerations. 'However, if the standby RHR loop is not g OPERABLE. an acceptable alternate method is two SGs with ct 4 - their secondary side water levels a 18%. Should the Q~ [ operating RHR loop fail, the SGs via natural circulation could be used to remove the decay heat.
-Note 1-permits all RHR pumps to be removed from operation
[}--
%. l 5 1 hour per 8 hour period. The purpose of the Note.is to permit tests designed to validate various accident analyses .
values. - One of the tests perfonned during the startup I testing program is.the validation of rod drop times during I cold conditions, both with and without flow. The no flow -l test may be performed in MODE 3. 4. or 5 and requires that I the' pumps be stopped for a short period of time. The Note j permits stopping of the pumps in order to perform this test i and validate the assumed analysis values. If changes are j made to the RCS that would cause a change to the flow l
' characteristics of the RCS. the input values must be i I
revalidated by conducting the test again. The 1 hour time period is adequate to aerform the test, and operating. experience has shown t1at boron stratification is not likely L during this short period with no forced flow. ,
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h ! l l p Lj Q)- l BRAIDWOOD - UNITS l'8'2 B 3.4.7 - 2 10/13/98 Revision K
m , j RCS Loops-MODE 5. Loops Filled l B 3.4.7 _ p l BASES u . LCO:(continued). Utilization of Note 1 is permitted provided the following conditions ~are met, along with any other conditions imposed
~by procedures:
- a. No. operations are permitted that would dilute the RCS :'
boron concentration. therefore maintaining the'_ margin-to: criticality. -Boron reduction is prohibited because
$:r a-uniform concentration distribution throughout the
> 'p-RCS cannot be ensured when in natural circulation: and wi uT b. Core outlet temperature is maintained at least 10 F. ,
L$h below saturation temperature, so that no vapor bubble ' may . form and possibly cause a natural circulation flow obstruction.
' Note 2. allows one RHR loop to be inoperable for a period of 5 2 hours, provided that the other RHR loop is OPERABLE-and
'in operation. This permits periodic surveillance tests to ,
be performed on the-inoperable loop when such testing is ' safe and possible. Note 3 requires that the secondary side water temperature of.
,9 )
"i i- each SG'be < 50*F above each of the RCS cold leg. temperatures before the start of a Reactor Coolant' Pump-(RCP) with an RCS cold leg temperature s 350*F. This. restriction is to prevent a low temperature overpressure event due to a thermal transient when an RCP is started. ' ~ Note 4 provides for an orderly transition from MODE 5 to l MODE 4 during a planned heatup by permitting removal of RHR
- loops from operation when at least one RCS loop'is in o)eration iThis Note provides for the transition to MODE 4'
.wlere an RCS loop is permitted to be in operation and E replaces the'RCS circulation function provided by the RHR
. loops.
L An.0PERABLE.RHR. loop is comprised of an OPERABLE RHR pump capable of providing forced flow to an OPERABLE RHR heat
, exchanger; RHR pumps are OPERABLE if they are. capable of being powered and ar= able to provide flow if required. An L OPERABLE' SG via natu, ' circulation has greater than 'or
[ f equal to the minimum water level specified in SR 3.4.7.2 and is'otherwise capable of providing the necessary heat' sink via natural circulation. A)q _. p/ p 1BRAIDWOOD.- UNITS 1 & 2: B 3.4.7 - 3 10/5/98 Revision K
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RCS Loops-MODE 5. Loops Filled B 3.4.7 1 c ~l BASES J
' APPLICABILITY- In MODE 5 with RCS loops filled..this LC0 requires forced ,
circulation of'the reactor coolant to remove decay heat from , the core and.to provide proper boron mixing. One loop of !
-RHR provides sufficient circulation for these purposes. 1 However, one additional RHR loop is required to be OPERABLE.
or the secondary side water level of at least two SGs is i required to be a 18%. ; I Operation in other MODES is covered by: ! LCO 3.4.4. "RCS Loops-MODES 1 and 2": LCO 3.4.5. "RCS Loops -MODE 3": LCO 3.4.6, "RCS Loops-MODE 4": LCO 3.4.8. "RCS Loops-MODE 5. Loops Not Filled": i LC0 3.9.5 " Residual Heat Removal (RHR) and 'oolant i Circulation-High Water Level" / MODE 6); and l LC0 3.9.6. " Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level" (MODE 6). I ACTIONS A.1 and A.2
, ) l- If no required RHR loop is 'in operation, except during v' conditions permitted by Note 1. all operations involving a reduction of RCS boron concentration must be suspended and action to restore one RHR loop to operation must be immediately initiated. Boron dilut' ion requires forced circulation to provide proper mixing and preserve the margin to criticality. The immediate Completion Ti p s reflect the s importance of maintaining operation for decay heat removal.
.o
- c. l, B.1 and C.1 b
y If the required RHR loop is inoperable or the required SG(s) x have secondary side water levels < 18%. redundancy for heat removal is lost. Action must be initiated immediately to H restore either the required RHR loop to OPERABLE status or
.s to restore the required SG secondary side water level (s).
The Required Actions will restore an available alternate heat removal path. The immediate Completion Times reflect the importance of maintaining the availability of two paths for heat removal.
3 J BRAIDWOOD - UNITS 1 & 2 B 3.4.7 - 4 10/5/98 Revision K
4 RCS Loops-MODE 5. Loops Filled B 3.4.7 Q u BASES l ACTIONS (continued) D.1. D.2.1. and D.2.2 )
- If two required RHR loops are inoperable or the required RHR loo) and one or.both.SG secondary side water levels are not
- wit 1in limit (s). all operations' involving a . reduction of RCS .
~
2 boron concentration must be sus) ended and action to restore r!- one RHR loop to operation must 3e immediately initiated or 4 initiate. action to restore required SG secondary side water-
' u) level to within limits. Boron dilution requires forced 4 4 : circulation to provide proper mixing and preserve the margin j -
to criticality. The immediate Completion Times reflect the importance of maintaining operation for decay heat removal. tSURVEILLANCE: . SR -3.4'.7.1 l REQUIREMENTS This SR requires verification every'12 hours that the required operating:RHR loop is in operation. ' Verification- 1 may include flow rate. temperature, or pump status ; monitoring, which' helps ensure that' forced flow is providing i
. heat removal. The Frequency of 12 hours is sufficient !
/ e 4'b - considering other indications and' alarms available'to the = i hf o g-operator'in the control room to monitor RHR loop performance.
.0 V SR -3.4.7.-2 l
?
M Verifying- that at least two SGs are OPERABLE by ensuring -1
? their secondary side narrow range water levels are a 18% -!
4 ensures an alternate decay heat removal method via natural' circulation in the event that the second RHR loop is not l OPERABLE. If both RHR loops are OPERABLE. this surveillance L is not needed. The 12 hour Frequency is considered adequate in view of. other indications available in the control room l
$ to alert the operator to the loss of SG 1evel.
. r-
[ . 4
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'BRAIDWOOD .- UNITS 1-& 2 B 3.4.7 - 5 10/5/98 Revision K y ---h , 4
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9
. RCS Loops-MODE 5. Loops Filled -
B'3.4.7
, BASES KhL ~ ~
,i SURVEILLANCEfREQUIREMENTS(continued)
SR 3:4. 7 ' :3 ' Verification that a second RHR pump is OPERABLE when
; required _ ensures that.an additional pum) can be placed in -
.., . : operation. if needed. to maintain. decay. leat . removal and -
reactor coolant circulation. Verification is performed by verifying pr_oper breaker alignment and power available to the.RHR pump; If- secondary side water level'is a 18% in at-
-least two SGs. this surveillance'is not needed. The Frequency of>7 days,is considered reasonable in view of 1 -other administrative controls available and has'been shown
~to'be acceptable by, operating experience; REFERENCES -1. NRC Information Notice 95-35, " Degraded Ability:of,, '
Steam Generators to Remove Decay Heat by Natural
-i Circulation." August 28. 1995.
e f l i ! 1 V .l 1.; I f 'r 6 I[(. . .rf] ~. Q. BRAIDWOOD=-UNITSL1_'&'2 : B 3.4.7 - 6 '10/5/98 Revision K w .-
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l l RCS Loops -MODE 5. Loops Not Filled l B 3.4.8 I
/3 B 3.4 REACTOR COOLANT SYSTEM (RCS) l B 3.4.8 RCS Loops -MODE 5. Loops Not Filled BASES
' BACKGROUND In MODE 5 with the RCS loops not filled, the primary ,
function of the reactor coolant is the removal of decay heat ' generated in the fuel, and the transfer of this heat to the component cooling water via the Residual Heat Removal (RHR) I heat exchangers. The Steam Generators (SGs) are not i available as a heat sink when the loops are not filled. The I secondary function of the reactor coolant is to act as a 1 carrier for the soluble neutron poison, boric acid. i In MODE 5 with loops not filled, only RHR pumps can be used , for coolant circulation. The number of pumps in operation I can vary to suit the operational needs. The intent of this l LC0 is to provide forced flow from at least one RHR pump for decay heat removal and transport, and to require that two paths be available to provide redundancy for heat removal. ' ^
/ APPLICABLE 'n MODE 5. RCS circulation increases the time available for
, b' SAFETY ANALYSES mitigation of an accidental boron dilution event. The RHR loops provide this circulation and have been identified as l important contributors to risk reduction. The flow provided i by one RHR loop is adequate for heat removal and for boron mixing. RCS loops in MODE 5. Loops Not Filled. satisfies Criterion 4 l of 10 CFR 50.36.(c)(2)(ii). i l
,im 'N BRAIDWOOD - UNITS 1 & 2 B 3.4.8 - 1 8/21/98 Revision A
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RCS Loops-MODE 5. Loops Not Filled B 3.4.8-77 1 BASES: l jf l LC0~ Thepurhose"ofthisLC0isto.requirethatatleasttwoRHR loops b OPERABLE and one of these-loops be in operation. 0 An OPERABLE loop is one that has the capability of transferring heat from the reactor coolant at a controlled W rate. Heat cannot be removed via'the RHR System unless V ; forced flow is used. 'A minimum of one running RHR pump f meets the LC0 requirement-.for one loop in operation. An additional RHR loop ~is required to be OPERABLE to meet
'l
~H: I g single failure: considerations. l
- m. 1 Note 1 permits all RHR pumps to_be removed-from operation for.s 1. hour. Utilization of Note-1-is permitted provided i the following conditions are met. along with any other ;
conditions imposed by! procedures. ! l
'v a. No operations are permitted that would dilute the RCS l 0 boron concentration,'therefore maintaining the. margin
-g to criticality. ~ Boron reduction is prohibited'because' '
y a uniform concentration distribution throughout the
-g RCS cannot be ensured when in natural circulation:
h b. Core. outlet temperature is maintained at least 10 F l WY 3 ) below saturation temperature,'so that no vapor bubble may form and possibly cause a. natural circulation flow l UK obstruction: and , T l va :c. No draining operations are permitted that would j further reduce the RCS water volume. M
.t
~ Note 2 allows one.RHR loop to-be inoperable for a aeriod of
.s 2 hours, provided that the other loop is OPERABLE and in operation. This permits periodic surveillance tests to be
~
performed on the inoperable loop when these tests are safe q and possible.
- An40PERABLE RHR loop is comprised of an OPERABLE RHR pump capable of providing. forced flow to an OPERABLE RHR heat I. .x exchanger. ' RHR pumps- are OPERABLE if they are capable of
!' " being powered and are able to provide flow if required. i-
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-BRAIDWOODL-UNITS 1&2 B 3.4.8 - 2 10/8/98 Revision K
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r RCS Loops-MODE 5. Loops Not Filled-B 3.4.8 BASES' APPLICABILITY In MODE 5 with loops not filled. this LC0 requires core heat removal and coolant circulation by the RHR System.- Operation in other MODES is covered by: LCO 3.4.4; "RCS Loops -MODES 1 and 2": LCO 3.4.5. "RCS Loops -MODE 3": LCO 3.4.6. "RCS Loops -MODE 4": LCO 3.4.7. "RCS Loops -. MODE 5. Loops Filled"-: LCO 3.9.5. " Residual Heat Removal (RHR) and Coolant
. Circulation-High Water Level" :(MODE 6): and LCO 3.9.6. " Residual Heat Removal'(RHR) and Coolant Circulation-Low Water Level" (MODE 6).
ACTIONS A'.1 and A:2 If no RHR 1003 is in operation, except during conditions-3ermitted by Vote 1. all operations involving a reduction of RCS boron concentration must be suspended-and action to restore one RHR loop to operation must be immediately initiated. Boron dilution requires forced circulation to .O provide proper mixing and preserve the margin to criticality. The immediate Completion Times' reflect the . l importance of maintaining operation for decay heat removal. .j D.J. I If only one RHR loop.is OPERABLE. except during conditions permitted by Note 2. redundancy for decay heat removal is lost and action must be initiated immediately to restore a second loop to OPERABLE status. The immediate Completion ! Time reflects the importance of maintaining tha availability of.two paths for heat removal. O BRAIDWOOD'- UNITS 1 & 2 B 3.4.8 - 3 8/21/98 Revision A
RCS Loops-MODE 5, Loops Not Filled B 3.4.8 (~} BASES V ACTIONS (continued) j C.1 and C.2 I If no required RHR loops are OPERABLE. all operations ! involving a reduction of RCS boron concentration must be suspended and action must be initiated immediately to l restore an RHR loop to OPERABLE status. Boron dilution requires forced circulation to provide proper mixing and preserve the margin to criticality. The immediate Completion Times reflect the importance of maintaining the capability for heat removal. SURVEILLANCE SR 3.4.8.1 REQUIREMENTS This SR requires verification every 12 hours that the required operating RHR loop is in operation. Verification may include flow rate, temperature, or pump status monitoring, which helps ensure that forced flow is providing heat removal. The Frequency of 12 hours is sufficient considering other indicaticrs and alarms available to the operator in the control room to mcaitor RHR loop performance. (u } SR 3.4.8.2 Verification that a second RHR pump is OPERABLE ensures that an additional pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. l Verification is performed by verifying proper breaker j alignment and power available to the required pumps The i Frequency of 7 days is considered reasonable in view of W other administrative controls available and has been shown i to be acceptable by operating experience. j [ REFERENCES None. l l l l i 1
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!.h BRAIDWOOD - UNITS 1 & 2 B 3.4.8 -4 8/21/98 Revision A ,
_ , Pressurizer
>z .
B 3.4.9 f^^r B 3.4 REACTOR COOLANT SYSTEM (RCS). C./ B 3.4.9 Pressurizer BASES-LBACKGROUND' - The pressurizer provides a' point in the RCS where liquid and vapor are maintained in equilibrium under saturated ) conditions.for pressure control purposes to prevent bulk i boiling in the remainder of the RCS, Key functions include maintaining required primary system pressure during steady state operation. and limiting the pressure changes caused by reactor coolant thermal' expansion and contraction during 1 normal load transients.' The pressure control components addressed by this l LCO include the pressurizer water level, the required i heaters, and their controls and Engineered Safety Features l (ESF) power supplies. Pressurizer safety valves and I pressurizer' power operated _ relief valves are addressed by LCO 3.4.10. " Pressurizer Safety Valves." and LCO 3.4.11.
" Pressurizer Power Operated Relief Valves (PORVs)."
respectively.
,.g j ) The intent of the LC0 is to ensure that a steam bubble V exists in the pressurizer during MODES 1. 2. and 3 to minimize the consequences of potential overpressure transients. The presence of a steam bubble is consistent with analytical assumptions. Relatively small amounts of
.noncondensible gases can inhibit the condensation heat transfer between the pressurizer spray and the steam, and i diminish the spray effectiveness for pressure control.
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BRAIDWOOD -. UNITS 1 & 2 B 3.4.9 - 1 8/21/98 Revision A
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Pressurizer B 3.4.9 n 4 'y
- BASES:
} BACKGROUND (continued).
-Electrical immersion heaters' located in the lower section
'of:the pressurizer vessel. keep the water-in the pressurizer at saturation temperature and maintain a constant operating
, , pressure. .A minimum required available capacity of
. pressurizer heaters ensures that the RCS. pressure can be a maintained. :The. capability to maintain and control system
- pressure is important for maintaining subcooled conditions
-in _the RCS and ensuring the capability to remove core decay heat'by either. forced or natural circulation of reactor coolant. Unless adequate heater capacity is available, the
' hot; high pressure condition cannot be maintained .
. indefinitely and stil1 provide the required subcooling '
"q margin in the: primary system. LInability'to control the J . : system pressure and maintain subcooling under conditions of
~
natural circulation flow in the primary system could lead to A: .a loss of single phase natural circulation and decreased . LO: ,
. capability:to remove core decay heat.
L The pressurizer-heaters are powered from the non-Class 1E 7l . buses. ?The pressurizer heaters are non-safety related. V Plant design' includes'a total-heater capacity of 1800 kW q
.:C that'is' divided into.four groups, with separate controls for !
.t w'H, : ,
the proportional:and backup grouas. The non-Class 1E ESF buses servicing the pressurizer 1 eaters can be powered from
\ - y'$ '
the Unit ~ Auxiliary Transformer, the System Auxiliary . ' Transformer (SAT), or the emergency diesel generator by
. closing the ESF. to non-ESF:crosstle breaker.
i
, APPLICABLE. .
In MODES 1. 2.-and 3. the LCO requirement for a steam bubble SAFETY' ANALYSES. is reflected implicitly in the accident analyses. Safety analyses performed for lower MODES are-not limiting. All analyses performed from a critical reactor condition assume the existence of a steam bubble and saturated conditions in 1 'the pressurizer. In making this assumption, the analyses neglect the small fraction of noncondensible gases normally present. Safety analyses presented in the UFSAR (Ref. 1) do not take
. credit,for-pressurizer heater operation: however, an-
; implicit initial condition assumption of the safety analyses
'is that the.RCS.is-operating at normal pressure.
3g n a ny BRAIDWOOD' , UNITS'l & 2. B 3.4.9 - 2 9/18/98 Revision K 4 1
Pressurizer l B 3.4.9 I$ BASES APPLICABLE SAFETY ANALYSES (continued) The maximum pressurizer water level limit which ensures that a steam bubble exists in the pressurizer satisfies Criterion.2 of 10 CFR 50.36(c)(2)(ii). Although the heaters are not specifically used in accident analysis, they provide the capability to maintain subcooling in the long term during loss of offsite power, as indicated in NUREG-0737 (Ref. 2), and thus, satisfy Criterion 4 of 10 CFR 50.36(c)(2)(ii). LC0 The LCO requirement for the pressurizer to be' OPERABLE with a water volume s 1656 cubic feet, which is equivalent to 5 92%, ensures that a steam bubble exists. Limiting the LC0 maximum operating water level preserves the steam space for pressure control. The LCO has been established to ensure the capability to establish and maintain pressure control for steady state operation and to minimize the consequences of potential overpressure transients. Requiring the presence of a steam bubble is also consistent with analytical assumptions. f^)
's The LCO requires two groups of OPERABLE pressurizer heaters, each with a capacity a 150 kW, capable of being powered from 4y[ redundant ESF power supplied buses. Since the only safety I function for pressurizer heaters is in a loss of offsite a
power condition, normal power is not required for Q OPERABILITY. The minimum heater capacity required is s sufficient to maintain the RCS near normal operating pressure when accounting for heat losses through the
. O. pressurizer insulation. By maintaining the pressure near the operating conditions, a wide margin to subcooling can be obtained in the loops. The value of 150 kW is derived from generic evaluation of Westinghouse pressurizer heat loss calculations (Ref. 3).
q G BRAIDWOOD - UNITS 1 & 2 B 3.4.9 - 3 8/22/98 Revision K
16 h Pressurizer
~B 3.4.9 L
c l IM x/ - BASES . i
- APPLICAB' ILITY The need for pressure control is most pertinent when core heat can cause the greatest effect on RCS temperature, i resulting in the greatest effect on pressurizer level and RCS pressure control, Thus applicability has been ,
designated for MODES 1 and 2. The applicability 1s also l: provided for MODE 3. The purpose is to prevent solid water RCS operation during heatup and cooldown to avoid rapid pressure rises caused by normal operational perturbations, i e such as reactor' coolant pump startup. I l
.In MODES 1. 2. and 3. there is need to maintain the availability of pressurizer heaters, capable of being powered from an ESF power supply In the event of a loss of 1 offsite power, the initial conditions of-these MODES give ;
the greatest demand for maintaining'the RCS in a hot-
- pressurized condition with loop subcooling for an extended period. For' MODE 4. 5. or 6. it is not necessary to control ]
pressure (by heaters) to ensure loop subcooling for heat transfer when the Residual Heat Removal (RHR) System is in !- service, and therefore, the LCO is not applicable. I
..v O ACTIONS A.1. A.2. A 3. 6nd A 4 I 1
Pressurizer water level control malfunctions or other plant ) evolutions may result in a pressurizer water level above the :l' nominal upper limit, even with the unit at steady state conditions. In MODE 1~at > 10% RTP (P-7), the unit will trip since the upper limit of this LCO is the same as the Pressurizer Water Level-High Trip. If the. pressurizer water level is not within the limit. I action must be taken to bring the plant to a MODE in which ! the LCO does not apply. To achieve this status. within 6 hours the unit must be brought.to MODE 3. with all rods fully inserted and incapable of withdrawal. Additi6nally. the unit must be brought to MODE 4 within 12 hours. This takes the unit out of the applicable MODES. The allowed Completion Times are reasonable. based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging plant systems.
,~
O BRAIDWOOD - UNITS 1 & 2 B 3.4.9-4 8/21/98 Revision A
.. . . ~
i Pressurizer
- B 3.4.9
. (~~3 - BASES qj ACTIONS (continued)
EL1 If_the required groups of pressurizer heaters are inoperable. restoration is required within 72 hours. The Completion Time of 72 hours is reasonable considering the g anticipation that a demand caused by loss of offsite power 4 would be unlikely in this period. Pressure control may be t' maintained during this time using the remaining pressurizer "y heater capability. C.1 and C.2 If Required Action B.1 and its associated Completion Time are not met, the unit must be brought to a MODE in which the LC0 does not apply. To achieve this status. the unit must be brought to MODE 3 within 6 hours and to MODE 4 within 12 hours. The allowed Completion Times are reasonable. based on operating experience, to reach the required unit - conditions from full power conditions in an orderly manner and without challenging plant systems. p V SURVEILLANCE SR 3.4.9.1' m REQUIREMENTS '
? This SR requires that during steady state operation Ie pressurizer level is maintained below the nominal upper limit to provide a minimum space for a steam bubble. The h
Surveillance is performed by observing the indicated level. , The Frequency of 12 hours corresponds to verifying the ' 3arameter each shift. The 12 hour interval has been shown
)y operating practice to be sufficient to regularly assess level for any deviation and verify that operation is consistent with the safety analyses assumption of ensuring that a steam bubble exists in the pressurizer. Alarms are also available for early detection of abnormal level indications.
i 3Q ,) BRAIDWOOD - UNITS 1 & 2 B 3.4.9 - 5 8/22/98 Revision K
1 Pressurizer o B 3.4.9 l t .
- BASES q.)
SURVEILLANCE REQUIREMENTS-(continued) SR 3 4.9.2 The SR is satisfied when-the power supplies are demonstrated
- o. to be capable of producing the minimum power and the 7 associated pressurizer heaters are verified to be a 150 kW.
This is performed by energizing the heaters and measuring . li M circuit current. The Frequency of 18 months is considered 1
.p adequate to detect heater degradation and has been shown by g operating experience to be acceptable.
SR 3.4.9.3 This Surveillance demonstrates that the heaters can be manually transferred from the normal non-ESF power supply to the ESF power supply and energized. The Frequency of l 18 months is based on a typical fuel cycle and is consistent with similar verifications of ESF power supplies. REFERENCES 1. UFSAR, Chapter 15.
') 2, NUREG-0737, " Clarification of TMI Action Plan
\> Requirements." November 1980.
- 3. Westinghouse Owners Group Study. " Emergency Power o Supply Requirements for the Pressurizer Heaters."
I transmitted via B. L. King to C. Reed TMI-0G-83. September 26. 1979. b
'H l n
- i
- '%.)
BRAIDWOOD - UNITS 1 & 2 B 3.4.9- 6 9/18/98 Revision K
Pressurizer Safety Valves 1 B 3.4.10 1 p Lj B 3.4 REACTOR COOLANT SYSTEM (RCS) , j l B 3.4.10 Pressurizer Safety Valves l l BASES l l BACKGROUND The pressurizer safety valves provide, in conjunction with the Reactor Protection System, overpressure protection for l the RCS. The pressurizer safety valves are totally enclosed pop type, spring loaded, self actuated valves with backpressure compensation. The safety valves are designed l to prevent the system pressure from exceeding the system i Safety Limit (SL). 2735 psig. which is 110% of the design pressure. Because the safety valves are totally enclosed and self actuating, they are considered independent components. The l relief capacity for each valve. 420.000 lb/hr. is based or, i postulated overpressure transient conditions resulting from a complete loss of steam flow to the turbine. This event results in the maximum surge rate into the pressurizer. i which specifies the minimum relief capacity for the safety valves. The discharge flow from the pressurizer safety valves is directed to the pressurizer relief tank. This D [V discharge flow is indicated by an increase in temperature downstream of the pressurizer safety valves or increase in the pressurizer relief tank temperature or level. Overpressure protection is required in MODES 1. 2. 3. 4 ; and 5: however. in MODES 4 and 5. and in MODE 6 with the reactor vessel head on, over)ressure protection is provided by operating procedures and ]y meeting the requirements of i LCO 3.4.12. " Low Temperature Overpressure Protection (LTOP) l System." The upper and lower pressure limits are based on the 1% l tolerance requirement (Ref.1) for lifting pressures above 1 1000 psig. The lift setting is for the ambient conditions l associated with MODES 1, 2. and 3. This requires either that the valves be set hot or that a correlation between hot and cold settings be established. l (D V BRAIDWOOD - UNITS 1 & 2 B 3.4.10 - 1 8/21/98 Revision A
Pressurizer Safety Valves l B 3.4.10 i l /] V BASES i 1 BACKGROUND (continued) l The pressurizer safety valves are part of the primary j success path and mitigate the effects of postulated ! accidents. OPERABILITY of the safety valves ensures that 1 the RCS pressure will be limited to 110% of design pressure. The consequences of exceeding the American Society of Mechanical Engineers (ASME) pressure limit (Ref. 1) could include damage to RCS components, increased leakage, or a requirement to perform additional stress analyses prior to resumotion of reactor operation. APPLICABLE All accident and safety analyses in the UFSAR (Ref. 2) that I SAFETY _ ANALYSES require safety valve actuation assume operation of three l pressurizer safety valves to limit increases in RCS 1 pressure. The overpressure protection analysis (Ref. 3) is . also based on operation of three safety valves. Accidents ! that could result-in overpressurization if not properly l terminated include: '
- a. Uncontrolled rod withdrawal from full power (s b. Loss of reactor coolant flow;
- c. Loss of external electrical load;
- d. Loss of normal feedwater;
- e. Loss of all AC power to station auxiliaries:
- f. Locked rotor: and
- g. Feedwater line break.
Detailed analyses of the above transients are contained in Reference 2. Safety valve actuation is required in events c. d. and e (above) to limit the pressure increase. Compliance with this LCO is consistent with the design bases and accident analyses assumptions. Pressurizer safety valves satisfy Criterion 3 of 10 CFR 50.36(c;(2)(ii). O O BRAIDWOOD - UNITS 1 & 2 8 3.4.10 - 2 8/21/98 Revision A
Pressurizer Safety Valves B 3.4.10 /N BASES d LCO The three pressurizer safety valves are set to open at the RCS design pressure (2500 psia), and within the ASME specified tolerance, to avoid exceeding the maximum design pressure SL. to maintain accident analyses assumptions. and to comply with ASME requirements. Tne upper and lower pressure tolerance limits are based on the 1% tolerance reauirements (Ref.1) for lifting pressures above 1000 psig. The limit protected by this Specification is the Reactor Coolant Pressure Boindary (RCPB) SL of 110% of design pressure. Inoperab,lity of one or more valves could result in exceeding the SI. if a transient were to occur. The l consequences of exceeding the ASME pressure limit could l include damage to one or more RCS components. increased l leakage, or additional stress analysis being required prior j to resumption of reactor operation. The Note allow:: entry into MODE 3 with the lift settings I outside the LCO limits. This permits testing and ' examination of the safety valves at high pressure and temperature near their normal operating range. but only ) after the valves have had a preliminary cold setting. The cold setting gives assurance that the valves are OPERABLE near their design condition. Only one valve at a time will l O V be removed from service for testing. The 54 hour exception is based on 18 hour outage time for each of the three l valves. The_18 hour period is derived from operating experience that hot testing can be performed in this time frame. , APPLICABILITY In MODES 1. 2. and 3. OPERABILITY of three valves is required because the combined capacity is required to keep reactor coolant pressure below 110% of its design value during certain accidents. MODE 3 is conservatively included, although the listed accidents may not require the safety valves for protection. The LC0 is not applicable in MODES 4 and 5 and in MODE 6 with the reactor vessel head on because Low Temperature Overpressure Protection (LTOP) is provided. Overpressure protection is not required in MODE 6 with reactor vessel head detensioned. /~S U BRAIDWOOD - UNITS 1 & 2 B 3.4.10 - 3 8/21/98 Revision A
Pressurizer Safet) Valves B 3.4.10 I BASES - (N) ACTIONS .A_1 With one pressurizer safety valve inoperable, restoration must take place within 15 minutes. The Completion Time of 15 minutes reflects the importance of maintaining the RCS Overpressure Protection System. An inoperable safety valve coincident witn an RCS overpressure event could challenge the integrity of the pressure boundary. B.1 and 8.2 If Required Action A.1 and its associated Completion Time are not met or if two or more pressurizer safety valves are inoperable, the unit must be brought to a MODE in which the requirement does not apply. To achieve this status.. the unit must be brought to at least MODE 3 within 6 hours and to MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the recuired unit conditions from full power conditions in an orcerly manner and without challenging plant systems. In MODE 4, overpressure protection is provided by the LTOP System. The change from MODE 1. 2. or 3 to MODE 4 reduces
-, the RCS energy (core power and pressure), lowers the /) \~2 potential for large pressurizer insurges, and thereby removes the need for overpressure protection by three !
pressurizer safety valves. l l SURVEILLANCE SR 3.4.10.1 ; REQUIREMENTS ! SRs are specified in the Inservice Testing Program. Pressurizer safety valves are to be tested in accordance with the requirements of Section XI of the ASME Code (Ref. 4), which provides the activities and Frequencies necessary to satisfy the SRs. No additional requirements are specified. The pressurizer safety valve setpoint is 1% of a nominal 2485 psig. BRAIDWOOD - UNITS 1 & 2 B 3.4.10 - 4 8/21/98 Revision A
L Pressurizer Safety. Valves . B 3.4.10 l 1: 1 l-l
/'~') BASESL i l \s,/
{ 1 L REFERENCES, 1. ASME. Boiler and Pressure Vessel Code. Section Ill. l l
- 2. UFSAR. Chapter 15. l l
L 3. WCAP-7769. Rev. 1. June 1972. I L ' 1
- 4. ASME. Boiler and Pressure Vessel Code. Section XI. i l
1 i i I L ; 1 1 1 I l
/~~ 1
'L ,,, c - \ ! l
- -l L
l 1 i i l l 1 l l l l l l [
. h' % i l . -( /
! BRAIDWOOD - UNITS 1 & 2 8 3.4.10 - 5 8/21/98 Revision A g
1 Pressurizer PORVs l 8 3./ 11 l B 3.4 REACTOR COOLANT SYSTEM (RCS) [~} v
'B 3.4.11 Pressurizer Power Operated Relief Valves (PORVs) !
BASES ! BACKGROUND The pressurizer is equipped with two types of devices for l
. pressure relief: pressurizer safety valves and PORVs. The '
PORVs are air operated valves that are controlled to open at a specific set pressure when the pressurizer pressure increases and close when the pressurizer pressure decreases. l The PORVs may also be manually operated from the control room. ' Block valves, which are normally open. are located between the pressurizer and the PORVs. The block valves are used to isolate the PORVs in case of excessive leakage or a stuck open PORV. Block valve closure is accomplished manually using controls in the control room. A stuck open FORV is, j in effect. a small break Loss Of Coolant Accident (LOCA). As such, block valve closure terminates the RCS depressurization and coolant inventory loss. The PORVs and their associatec block valves may be used by (f,) plant operators to.depressurize the RCS to recover from i l certain transients if normal pressurizer spray-is not available. Additionally, the series arrangement of the PORVs and their block valves permit performance of surveillances on the valves during power operation. The PORVs may also be used for feed and bleed core cooling in the case of multiple equipment failure events that are not within the design basis, such as a total loss of feedwater. m The PORVs. their block valves, and their ;ontrols are 1 powered from the vital buses that normally receive power
; from offsite power sources. but are also capable of being
? powered fr.om emergency power sources in the event of a loss 4 of offsite power. Two PORVs and their associated block l valves are powered from two separate safety trains (Ref. 1).
/
BRAIDWOOD - UNITS 1 & 2 B 3.4.11 - 1 8/22/98 Revision K
Pressurizer PORVs B 3.4.11 A-t j BASES BACKGROUND (continued) The unit has two PORVs. each having a relief capacity of l 210.000 lb/hr at 2350 psia. The functional design of the l PORVs is based on maintaining pressure below the Pressurizer l Pressure-High reactor trip setpoint following a step reduction of 50% of full load with steam dump. In addition, the PORVs minimize challenges to the pressurizer safety valves and also may be used for Low Temperature Overpressure Protection (LTOP). See LC0 3.4.12. " Low Temperature Overpressure Protection (LTOP) System." APPLICABLE. . Plant operators employ the PORVs to depressurize the RCS in l SAFETY ANALYSES response to certain unit transients if normal pressurizer I spray is not available. For the Steam Generator Tube l Rupture (SGTR) event, the safety analysis assumes that I manual operator actions are required to mitigate _the event. If a loss of offsite power is assumed to accompany the ; event, normal pressurizer spray is unavailable to reduce RCS l pressure. The PORVs are assumed to be used for RCS
'depressurization, which is one of the steps performed to ,m equalize the primary and secondary )ressures in order to
'( ) terminate the primary to secondary areak flow and the V radioactive releases from the affected steam generator. PORVs are also credited for automati'c pressure control during recovery from an inadvertent safety injection (SI). While the automatic pressure control is assumed to function for this event, the PORV(s) can initially be in " manual" if available for the operator to place in " auto" upon ! recognition of the inadvertent SI (Ref. 5). Automatic
? operation of the PORVs to control reactor coolant system
, pressure reduces challenges to the pressurizer safety valves :
2 during an inadvertent SI at power. ' i . v'. {} , l BRAIDWOOD - UNITS 1 & 2 B 3.4.11 - 2 8/22/98 Revision K
Pressurizer PORVs B 3.4.11 1 N BASES i, T l
'^' '
. APPLICABLE SAFETY ANALYSES (continued)
The PORVs are also modeled in safety analyses for events that' result in increasing RCS pressure for which Departure from Nucleate Boiling Ratio (DNBR) criteria are critical (Ref. 2). By assuming PORV actuation, the primary pressure remains below the high pressurizer pressure trip setpoint: thus, the DNBR calculation is more conservative. As such, this actuation is not required to mitigate these events, and this PORV automatic operation is, therefore, not required to r support an assumed safety function. I Pressurizer PORVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). i
! LCO The LCO requires both manual and automatic operation of the ;
j PORVs and manual operation of their associated block valves
. jp to be OPERABLE to mitigate the effects associated with an SGTR and inadvertent'SI. However, placing the PORV(s) in
" manual" is acceptable, and the PORV may continue to be OPERABLE if the automatic circuitry remains OPERABLE and the
_ ,,i operator remains capable of returning the PORV(s) to " auto." By maintaining two PORVs and their associated block valves i A OPERABLE, the single failure crit:rion is satisfied. An ! w- OPERABLE block valve may be either open, or closed and 1 { energized with the capability to be o]ened, since the required safety function is accomplis 1ed by manual operation. Although typically open to allow PORV operation, the~ block valves may be OPERABLE when closed to isolate the flow path of an inoperable PORV that is capable of being 9 manually.and automatically cycled (e.g. as in the case of excessive PORV leakage). Similarly, isolation of an OPERABLE PORV does not render that PORV or block valve ino)erable provided the relief function remains available M wit 1 manual and automatic action. An OPERABI . PORV is required to be capable of manually and automatic :ly opening and closing, and not experiencing excessive seat leakage. Excessive seat leakage, although not associated with a specific acceptance criteria, exists when conditions dictate closure of the block valve to limit leakage. t,,,! v BRAIDWOOD - UNITS 1 & 2 B 3.4.11 - 3 8/22/98 Revision K
l
- -Pressurizer PORVs
* ,B 3.4.11 j7 . BASES-LC LC0 (continuedF I Satisfying the LC0 helps minimize' challenges to fission.
product barriers. L ! i. APPLICABILITY -In. MODES'1. 2, and 3. the PORV and its block valve' are - required to be OPERABLE to limit the potential for a'small break'LOCA through the flow path. The.most likely'cause for a PORV small break LOCA is a result.of a- pressure increase transient.that causes the PORV to automatically open.
. Imbalances in the energy output of the core and heat removal-by the secondary system can cause-the RCS pressure to increase to'the PORV. opening setpoint. The most rapid.
increases will occur at the higher operating power and-Q pressure conditions of MODES 1.and 2..-The PORVs are also
.: required to be OPERABLE in MODES '1. 2. and 3 for manual 9 actuation to. mitigate a. steam generator tube rupture event-J -and for automatic actuation to mitigate an inadvertent SI jv event.
. Pressure increases are less prominent in MODE 3 because the core input; energy is reduced, but the RCS pressure is high.
l(g). Therefore, the LC0 is applicable in MODES 1. 2. and 3. The
~ ' " LC0 is not applicable in MODE 4. 5. Lor 6. when both. pressure and core energy are decreased and the pressure surges become
.much less significant. ' LC0' 3.4.12 addresses the PORV requirements in MODES 4 and 5. and in MODE 6 with the -
' reactor vessel head in place.
f ,. jy ,V' l BRAIDWOOD : UNITS 1 & 2 B 3.4.11 - 4 8/22/98 Revision K
Pressurizer PORVs , L B 3.4.11 l f"% BASES ds tv
- r. ACTIONS An ACTION Note 1 has been added to clarify that all i
, g pressurizer PORVs and block valves are treated as separate )
.. entities, each with separate Completion Times (i.e.. the '
Completion Time is on a component basis). The exception for l g- LC0 3.0.4, Note 2. permits entry into MODES 1, 2. and 3 to i perform cycling of the PORVs or block valves to verify their l y% e OPERABLE status. Testing is not performed in lower MODES. W A.1 2 PORVs may be inoperable but capable of being manually and i
~? automatically cycled (e.g., excessive seat leakage). In 4'
this condition, either the PORVs must be restored or the l flow path isolated within 1 hour. The associated block valve is required to be closed but power must be maintained to the associated block valve since removal of power would render the block valve inoperable. This permits operation of the unit until the next refueling outage (MODE 6) so that maintenance can be performed on the PORVs to eliminate the problem condition. B.1. B.2 and B.3 I" v-
) If one PORV is inoperable and not capable of being manually cycled, it must be either restored, or isolated by closing the associated block valve and remveing the power to the associated block valve. 'The Completion Times of 1 hour are reasonable, based on challenges to the PORVs during this time period, and provide the operator adequate time to correct the situation. If the inoperable valve cannot be restored to OPERABLE status, it must be isolated within the specified time. Because there is at least one PORV that remains OPERABLE. 72 hours is provided to restore the inoperable PORV to OPERABLE status. If the PORV cannot be restored within this time, the unit must be brought to a MODE in which the LCO does not apply, as required by Condition D.
Required Action B.2 is modified by a Note stating that removing power from the block valve is not required unless d the associated PORV is inoperable due to being incapable of f being manually cycled. In the event the PORV is inoperable
's solely due to inoperable automatic cycling capability, it is desired for power to be retained to the block valve to allow
{ more readily accessible manual relief capabilities.
%,d BRAIDWOOD - UNITS 1 & 2 B 3.4.11 - 5 8/22/98 Revision K
l. Pressurizer PORVs l B 3.4.11 rT v BASES l . ACTIONS (continued) ! 'C.1. and C.2 If one block' valve is inoperable. then it is necessary to either restore the block valve to OPERABLE status within the Completion Time of 1 hour or place the associated PORV in , manual control. The. prime importance for the capability to close the block valve is to isolate a stuck open PORV. Therefore, if the block valve cannot be restored to OPERABLE status within 1 hour, the Required Action is to place the PORV in manual control (i .e. , closed) to preclude its automatic opening for an. overpressure event and to avoid the potential for a stuck open PORV at a time that the block
. valve is ino)erable. The Completion Time of 1 hour is reasonable. Jased on the small potential for challenges to the system during this time period and provides the operator time to correct the situation.
Because at least one PORV remains OPERABLE. the operator is permitted a Com)letion Time of 72 hours to restore the inoperable bloc ( valve to OPERABLE status. The time allowed to restore the block valve is based upon the Completion Time for restoring an inoperable PORV in Condition B. since the j ,) PORVs may not be capable of mitigating an event if the
- d
' inoperable block valve is not full open. If the block valve is restored within the Completion Time of 72 hours, the power will be restored, and the PORV restored to OPERABLE status. If it cannot be restored within'this additional time, the unit must be brought to a MODE in which the LCO does not apply, as required by Condition D.
D.1 and D.2
% If the Required Action of Condition A. B. or C is not met.
then tne unit must be brought to a MODE in which the LC0 7 does not apply. To achieve this status, the unit must be
- w. A brought to at least MODE 3 within 6 hours and to MODE 4 o~ g within 12 hours. The allowed Completion Times are
{u reasonable. based on operating experience, to reach the
,,. required plant conditions from full power conditions in an
. orderly manner and without challenging plant systems. In 4
( ,9 l MODE 4. 5. and 6 with the reactor vessel head on, automatic See LCO 3.4.12. g PORV OPERABILITY may be required.
/
V BRAIDWOOD - UNITS 1 & 2 B 3.4.11 - 6 9/18/98 Revision K
.- P'ressurizer PORVs B 3.4.11 BASES:
. ACTIONS (continued)-
g E.1 and E.2 7- If two PORVs are inoperable and not capable 'of being j
. ::- manually and automatically cycled. Condition B and its '
.. J associated. Required Actions would already be entered. The 54 Required Actions would either restore at least one valve l
-- within the Completion Time of 1 hour or isolate the flow '
3ath by closing and removing the power to the associated alock valves. The Completion Time of I hour 'is reasonable. based on the small potential for challenges to the system-during this time and provides the operator time to correct E the' situation. If no PORVs are restored within.the 8 Completion Time, then the unit must be brought to a. MODE'in M-Ng R which the LCO does not apply. To achieve this status, the unit must be brought to at least' MODE 3 within 6 hours and Ref to MODE 4 within 12 hours. The allowed Completion Times are , reasonable, based on operating experience..to reach the J recuired unit conditions from full power conditions-.in an
- orcerly manner and without challenging plant systems. In ;
i MODE 4. 5. and 6 with the reactor' vessel head on, automatic l PORV OPERAblLITY may be required. See LC0 3.4.12.
.p-m) . .O M 1.$ If two block valves are inoperable, it is necessary to !
W restore at least one block valve within 2 hours. The 4.' ^ Completion Time is reasonable, based on the small potential for challenges to the system during this time and provide the operator time to correct the situation.
- s. G.1 and G.2-
- o i 'If the Required Actions of Condition F are not met, the unit y3 must. be brought to a MODE in which the LCO does not apply.
To achieve this status, the unit must be brought to at least Q t-A MODE 3 within 6 hours and to-MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating
.p' . experience, to reach the required unit conditions from full
'TE power conditions in an orderly manner and without XZ challenging plant systems. In MODE 4, 5. and 6 with the
-reactor vessel head on, automatic PORV OPERABILITY may be f5 Y TK: . required. See LCO 3.4.12.
L l!
.BRAIDWOOD - UNITS 1.& 2 B ' 3.4.11 - 7 10/13/98 Revision K
1 , Pressurizer PORVs B 3.4.11 'W BASES Q TSURVEILLANCE- SR 3 4.11.1
<p-REQUIREMENTS
- Block valve cycling verifies that the valve (s) can be opened Q and closed if needed. The basis for the Frequency of M
u. 92 days is the ASME-Code. Section XI (Ref. 3). JU The' Note modifies this SR by stating that it is not required to be met with the block valve closed in accordance with the
-l Required Actions of.this LCO. If the block valve is closed i to isolate an inoperable PORV that is incapable of being dj manually and automatically' cycled, the maximum Completion i *t Time to restore the PORV and open the block valve is 72 hours, which is well within the allowable limits (25%) to extend the block valve Frequency of 92. days. Furthermore, i these test requirements would be completed by the reopening of a recently closed block valve upon restoration of the ,
PORV to OPERABLE status (i .e. . completion of. the Required i
, Actions fulfills the SR).
SR 3.4.11.2 i SR 3.4.11.2 requires a complete cycle of each PORV. 9~ Operating a PORV through-one complete cycle ensures that the ,U f}- }y PORV can be manually and automatically actuated for mitigation of an SGTR and inadvertent SI. The Frecuency of
& 18 months is based on a typical refueling cycle anc, industry accepted practice.
The Note modifies the SR to-allow entry into_and operation in MODE 3 prior to performing the SR. This allows the test to be performed.in MODE 3 under operating temperature and pressure conditions prior to entering MODE 1 or 2. In accordance with Reference 4. this test should be performed in MODE 3 or 4 to adequately simulate operating temperature and pressure effects on PORV operation. SR 3.4.11.3 Operating the solenoid air control valves and check valves on,the air accumulators ensures the PORV control system , actuates properly when called upon. The Frequency of l- -18 months is based on a typical refueling cycle and the Frequency of the other Surveillances used to demonstrate PORV OPERABILITY. 4 tl'(
'BRAIDWOOD ; UNITS 1 & 2 B 3.4.11 - 8 9/18/98 Revision K 1 .
tr- e ~ =w -
=rma
Pressurizer PORVs B 3.4.11 BASES c f") j SURVEILLANCE REQUIREMENTS (continued) l SR 3.4.11.4 l
.SR 3.4.11.4 is the performance of a CHANNEL CALIBRATION.
A CHANNEL CALIBRATION is performed every 18 months. Or approximately at every refueling. CHANNEL CALIBRATION is a i complete check of the instrument loop, including the sensor. The test verifies that the channel responds to measured 3.t. parameter within the necessary range and accuracy. !
@ CHANNEL CALIBRATIONS must be performed consistent with the 4 assumptions of the plant specific setpoint methodology. The p difference between the current "as found" values and the y' previous test "as left" values must be consistent with the I drift allowance used in the setpoint methodology.
The Frequency of 18 months' is based on the assumption of an 18 month calibration interval in the determination of the magnitude of equipment drift in the setpoint methodology. I
~
1
/) REFERENCES' 1. Regulatory Guide 1.32. February 1977.
y 2. UFSAR. Section 15.2. I sg 3. ASME. Boiler and Pressure Vessel Code. Section XI. Ml
- 4. Generic Letter 90-06. " Resolution of Generic Issue 70.
" Power Operated Relief Valve and Block Valve Reliability.' and Generic Issue 94. " Additional Low l Temperature Overpressure Protection for Light Water Reactors." pursuant to 10 CFR 50.54(f) June 25. 1990.
T 5. UFSAR. Section 15.5.1. u hl-N i l n ( %.) l BRAIDWOOD - UNITS 1 & 2 B 3.4.11 - 9 9/18/98 Revision K
LTOP System B 3.4.12 l u V-Q- B 3.4 REACTOR COOLANT SYSTEM (RCS) l B 3.4.12 Low Temperature Overpressure Protection (LTOP) System BASES BACKGROUND The LTOP System controls RCS pressure at low temperatures so the integrity of the Reactor Ccolant Pressure Boundary , (RCPB) is not compromised by violating the pressure and temperature (P/T) limits of 10 CFR 50. Appendix G (Ref.1). The reactor vessel is the limiting RCPB component for demonstrating such protection. The PTLR provides the maximum allowable actuation logic setpoints for the pressurizer Power Operated Relief Valves (PORVs) and the maximum RCS pressure for the existing RCS cold leg temperature during cooldown. shutdown, and heatup to meet the Reference 1 requirements during the MODES in which LTOP is necessary. The reactor vessel material is less ductile at low temperatures than at normal operating temperature. As the vessel neutron exposure accumulates, the material toughness decreases and becomes less resistant to pressure stress at g low temperatures (Ref. 2). RCS pressure, therefore, is maintained low at low temperatures and is increased only D; within the limits specified in the PTLR. The potential for vessel overpressurization is most acute when the RCS is water solid, occurring only while shutdown; a pressure fluctuation can occur more quickly than an operator can react to relieve the condition. Exceeding the RCS P/T limits by a significant amount could cause brittle cracking of the reactor vessel. LC0 3.4.3. "RCS Pressure and Temperature (P/T) Limits." requires administrative control of RCS pressure and temperature during heatup and cooldown to prevent exceeding the PTLR limits. i r (O l LJ B 3.4.12 - 1 8/21/98 Revision A BRAIDWOOD - UNITS 1 & 2
i i'
- 1. LTOP System B 3.4.12 BASES-BACKGROUND (continued):
This LC0 provides RCS overpressure protection by having a minimum coolant-input capability and having adequate pressure relief capacity. Limiting coolant input capability requires al1 Safety Injection (SI) pumps and all but one charging pump (a centrifugal charging pum]) incapable of injection into the RCS and isolation of tie SI accumulators. The pressure relief capacity requires either two redundant RCS rel.ief valves' or a depressurized RCS and an RCS vent of sufficient size. One RCS relief valve or the open RCS vent
-is the overpressure protection device that acts to terminate an increasing pressure event.
With minimum coolant input capability, the' ability to l provide core coolant addition is restricted. The LCO does not require the makeup control system deactivated or the SI actuation circuits blocked. Due to the lower pressures in ; the LTOP MODES and the expected core decay heat levels, the makeup system can provide adequate flow via the makeup control valve. If conditions require the~use of more than one centrifugal charging pump for makeup in the event of loss of inventory. then pumps can be made available through manual actions. V~ The LTOP System for pressure relief consists of two PORVs with reduced lift settings, or two Residual Heat Removal (RHR) suction relief valves, or one PORV and one RHR suction relief valve, or a depressurized RCS and an RCS vent of sufficient size. Two RCS relief valves are required for redundancy.- One RCS relief valve has adequate relieving capability'to prevent overpressurization for the required coolant input capability. l I
/]
BRAIDWOOD'- UNITS.1 & 2' B 3.4.12 - 2 8/21/98 Revision A
LTOP System B 3.4.12 l . (~N . BASES O. BACKGROUND (continued) PORV Recuirements As designed for the LTOP System. each PORV is signaled to i open if the RCS pressure approaches a limit determined by the LTOP actuation logic. The LTOP actuation logic monitors both RCS temperature and RCS pressure and determines when a condition not acceptable in the PTLR limits is approached. The wide range RCS temperature indications are auctioneered to select the lowest temperature signal. The lowest temperature signal is processed through a function generator that calculates a pressure limit for that temaerature. The calculated pressure limit is then compared wit 1 the indicated RCS pressure from a wide range pressure channel. If the indicated pressure meets or exceeds the calculated value, a PORV is signaled to operi. The PTLR presents the PORV setpoints for LTOP. The setpoints are normally staggered so only one valve opens during a low tem 3erature overpressure transient. Having the setpoints of bot, valves within the limits in the PTLR ensures that the Reference 1 limits will not be exceeded in O-V any analyzed event. When a PORV is opened in an increasing pressure transient. the release of coolant will cause the pressure increase to slow and reverse. As the PORV releases coolant, the RCS pressure decreases until a reset pressure is reached and the valve is signaled to close. The pressure continues to decrease below the reset pressure as the valve closes. l l l \ l(O i LJ BRAIDWOOO - UNITS 1 & 2 B 3.4.12 - 3 8/21/98 Revision A l
LTOP System B 3.4.12 ' (D. BASES C/' BACKGROUND (continued) RHR Suction Relief Valve Reauirements During LTOP MODES. the RHR System is operated for decay heat removal and low pressure letdown control. Therefore. the l' RHR suction isolation valves are open in the piping from the
.RCS hot legs to the inlets of the RHR pumps. While these valves are open'. the RHR suction relief valves are exposed to the RCS and are able to relieve pressure transients in the RCS.
The RHR suction isolation valves must be open to make the RHR suction relief valves OPERABLE for RCS overpressure mitigation. The RHR suction relief valves are spring loaded. bellows type water relief valves with pressure tolerances and accumulation limits established by " l Section III of the American Society of Mechanical Engineers (ASME) Code (Ref. 3) for Class 2 relief valves. RCS Vent Reauirements Once the RCS is ~dearessurized, a vent exposed to the containment atmosp1ere will maintain the RCS at containment
.U(]. ambient pressure in an RCS overpressure transient. if the relieving requirements of the transient do not exceed the capabilities of the vent. Thus, the vent path must be capable of relieving the flow resulting from the limiting LTOP mass or heat input transient, and maintaining _ pressure below the P/T limits. The required vent capacity may be provided by one or more vent paths.
For an RCS vent to meet the flow capacity requirement, it requires removing a pressurizer safety valve removing a PORV's internals, and disabling its Ncck valve in the open position, or similarly establishing any comparable vent. The vent path (s) must be above the level of reactor coolant. so as not to drain the RCS when open. l g V BRAIDWOOD - UNITS 1 & 2 B 3.4.12 - 4 8/22/98 Revision K
LTOP System . B 3.4.12 I BASES o/N APPLICABLE Safety analyses (Ref. 4) demonstrate that'the reactor vessel SAFETY ANALYSES is adequately protected against exceeding the Reference 1 P/T limits. In MODES 1. 2. and 3, the pressurizer safety valves will prevent RCS pressure from exceeding tne Reference 1 limits. In MODE 4 and belet overpressure prevention falls to two OPERABLE RCS relief valves or to a depressurized RCS and a sufficient sized RCS vent. Eacn of these means has a limited overpressurc relief capability. The actual temperature at which the pressure in the P/T limit curve falls below the pressurizer safety valve setpoint increases as the reactor vessel material toughness decreases due to neutron embrittlement. Each time the PTLR ! curves are revised, the LTOP System must be re-evaluated to l ensure its functional requirements can still be met using I the RCS relief valve method or the depressurized and vented RCS condition. The PTLR contains the acceptance limits that define the LTOP requirements. Any change to the RCS must be evaluated I against the Reference 4 analyses to determine the impact of the change on the LTOP acceptance limits. D Transients that are capable of overpressurizing the RCS are -[d categorized as either mass or heat input transients. examples of which follow: Mass Inout Tvoe Transients , 1
- a. Inadvertent safety injection; or
- b. Charging / letdown flow mismatch. ,
l Heat Inout Tvoe Transients
- a. Inadvertent actuation of pressurizer heaters;
- b. Loss of RHR cooling; or
- c. Reactor Coolant Pump (RCP) startup with temperature asymmetry within the RCS or between the RCS and steam generators.
O d BRAIDWOOD - UNITS 1 & 2 B 3.4.12 - 5 8/21/98 Revision A
LTOP System B 3.4.12 BASES (~N \ ) APPLICABLE SAFETY ANALYSES (continued) The following are required during the LTOP MODES to ensure that mass and heat input transients do not occur, which either of the LTOP overpressure protection means cannot handle:
- a. Rendering all SI pumps and all charging pumps but one centrifugal charging pump incapable of injection;
- b. Deactivating the accumulator discharge isolation valves in their closed positiens; and
- c. Disallowing start of an RCP if secondary temperature is more than 50 F above primary temperature in any one loop. LC0 3.4.6. "RCS Loops -MODE 4. " and LCO 3.4.7.
"RCS Loops-MODE 5. Loops Filled." provide this protection.
The Reference 4 analyses demonstrate that either one RCS relief valve or the depressurized RCS and RCS vent can maintain RCS pressure below limits when only one centrifugal charging pump is actuated. Thus the LCO allows only one
,_s centrifugal charging pump OPERABLE during the LTOP MODES.
f ') Since none of the overpressure protection methods can handle V the pressure transient need from accumulator injection, when RCS temperature is low, the LCO also requires the accumulators isolation when accumulator pressure is greater than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed in the PTLR. The isolated accumulators must have their discharge valves closed and the valve power supply breakers in their open positions. p G BRAIDWOOD - UNITS 1 & 2 B 3.4.12 - 6 8/21/98 Revision A
LTOP System B 3.4.12 T'N BASES Q,i . APPLICABLE SAFETY ANALYSES (continued) PORV Performance The fracture mechanics analyses show that the vessel is protected when the PORVs are set to cpen at or below the limit shown in the PTLR. The setpoints are derived by analyses that model the performance of the LTOP System, assuming the limiting mass addition transient of one centrifugal charging pump injecting into a water solid RCS or the limiting heat input transient of the startup of an idle RCP with the secondary water in the steam generator s 50 F above the RCS cold leg temperatures. These analyses consider pressure overshoot and undershoot beyond the PORV opening and closing, resulting from signal processing and valve stroke times. The PORV setpoints at or below the derived limit ensures the Reference 1 P/T limits will be met. The PORV'setpoints in the PTLR will be updated, as necessary, when the P/T limits are revised.. The P/T limits are aeriodically modified as the reactor vessel material touginess decreases due to neutron embrittlement caused by neutron irradiation. Revised limits are determined using !,,J 'v neutron fluence projections and the results of examinations of the reactor vessel material irradiation surveillance specimens. The Bases for LC0 3.4.3 discuss these examinations. The PORVs are considered active components. Thus, the failure of one PORV is assumed to represent the worst case, single active failure. O V BRAIDWOOD - UNITS 1 & 2 B 3.4.12 - 7 8/21/98 Revision A
1 i LTOP System l B 3.4.12 l h O BASES APPLICABLE SAFETY ANALYSES (continued) RHR Suction Relief Valve Performance The RHR suction relief valves do not have variable pressure and temperature lift setpoints like the PORVs. Analyses must show that one RHR suction relief valve with a setpoint s 450 psig will pass flow greater than that required for the limiting LTOP transient while maintaining RCS pressure less than the P/T limit curve. Assuming all relief flow requirements during the limiting LTOP event, an RHR suction relief valve will maintain RCS pressure to within the valve rated lift setpoint. plus an accumulation s 10% of the rated lift setpoint. As the RCS P/T limits are decreased to reflect the loss of l toughness in the reactor vessel materials due to neutron I embrittlement, the RHR suction relief valves must be analyzed to still accommodate the design basis transients .
-for LTOP. l The RHR suction relief valves are considered active components. Thus, the failure of one valve is assumed to p represent the worst case single active failure.
RCS Vent Performance With the RCS depressurized analyses show a vent size of 2.0 square inches is capable of mitigating the allowed LTOP overpressure transient. The capacity of a vent this size is greater than the flow of the limiting transients for the LTOP configuration, maintaining RCS pressure less than the maximum pressure on the P/T limit curve. The RCS vent size will be re-evaluated for compliance each time the P/T limit curves are revised based on the results of the vessel material surveillance. The RCS vent is passive and is not subject to active failure. The LTOP System satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). l t Ch v-BRAIDWOOD - UNITS 1 & 2 B 3.4.12 - 8 8/21/98 Revision A l l l
LTOP System B 3.4.12 TV . ( ) &o BASES-l' Q LCO This LC0 requires that the LTOP System is OPERABLE. The s .LTOP System is OPERABLE when the minime coolant input and
. qF pressure relief capabilities are;0PERAk t. Violation of hi this .LC0 could. lead to the loss of low temperature overpressure mitigation capability and violation of the Reference 1 limits as a result of an operational transient.
To limit the coolant input capability, the LC0 requires no SI pumps and a maximum of one charging pump (centrifugal) be capable of. injecting into the RCS and all accumulator discharge. isolation valves be closed and-de-energized (when accumulator pressure-is greater than or equal to the maximum 1',' ' , RCS pressure for the existing RCS cold leg temperature N. allowed in the PTLR). x Nl The LCO is modified by a note that permits the o)eration in MODE 4 with all SI pumps and charging pumps capa)le of RCS
*i injection whenever all RCS cold legs exceed 330 F. This is
~
-q I necessary to allow transition between MODES 3 and 4.
yi -
-. l The elements of the LC0 that provide low temperature overpressure mitigation through pressure relief are:
; a. Two OPERABLE PORVs:
- b. Two~ OPERABLE RHR suction relief valves:
- c. One OPERABLE PORV and one OPERABLE RHR suction relief valve; or
- d. A depressurized RCS and an OPERABLE RCS vent.
A PORV is OPERABLE for LTOP when its block valve is open, its lift setpoint is set to the limit required by the PTLR and testing proves its ability to open at this setpoint, and motive power is available to the two valves and their control circuits. . An RHR suction relief valve is OPERABLE for LTOP when its RHR suction isolation valves are open its setpoint is s 450 psig, and testing has proven its ability to open at this setpoint. y-k, BRAIDWOOD - UNITS 1 & 2 B 3.4.12 - 9 10/1/98 Revision K
LTOP System B 3.4.12
,/^) BASES G
LCO (continued) An RCS vent is OPERABLE when open with an area of a 2.0 square inches. Each of these methods of overpressure prevention is capable of mitigating the limiting LTOP transient. 1 APPLICABILITY This LCO is applicable in MODES 4 and 5. and in MODE 6 when the reactor vessel head is on. The pressurizer safety valves provide overpressure protection that meets the Reference 1 P/T limits above 350 F. When the reactor vessel head is off, overpressurization cannot occur. LC0 3.4.3 provides the operational P/T limits for all MODES. LCO 3.4.10. " Pressurizer Safety Valves." requires the ; OPERABILITY of the pressurizer safety valves that provide overpressure protection during MODES 1. 2. and 3. Low temperature _ overpressure prevention is most critical during shutdown when the RCS is water solid, and a mass or heat input transient can cause a very rapid increase in RCS
.; g)
V pressure resulting in little or no time available to allow operator action to mitigate the event. L (a BRAIDWOOD - UNITS 1 & 2 B 3.4.12 - 10 8/21/98 Revision A
. LTOP System 4
B 3.4.12 N
' BASES L)
ACTIONS The' Actions are modified by a Note. that in'd icates that the - y provisions of LCO 3,0.4 are not: applicable to the.RCS n pressure relief capabilities (PORVs and RHR suctica relief ~
.2 . valves.or! vent of a 2.0 square inches with the RCS-
) depressurized). As a result. MODE changes are allowed wnen m one or more of these capabilities are inoperable. This
~
allowance is provided because the Required Actions have been
*: C determined.to provide ~an acceptable level of safety.
A.1 and B.1 With'two centrifugal charging pumps capable of injecting Linto the RCS, or one positive displacement charging pump capable of injecting into the RCS,' or any SI' pump capable of
' injecting into the RCS RCS overpressurization is possible.
The requirement to immediately initiate action (except
.. during charging pump swap operation) to restore restricted
._ coolant inaut capability to the RCS reflects the urgency of 0- removing tie RCS from this condition.
E t Required Action A.1 is modified by a Note that permits two
# charging pumps capable of RCS injection for s 15 minutes to.
allow for pump swaps. 3,-s[. L g C.1 and D.1 An unisolated accumulator requires isolation.within 1 hour. This is-only required when the accumulator pressure is at or more than the maximum RCS pressure for the existing temperature allowed by the P/T limit curves. If the Required Action and associated Completion Time of Condition C are not met Required Action D.1 must be.
-performed in the next 12 hours.- Depressurizing the accu:Lulators below the LTOP limit from the PTLR prevents an accumulator pressure from exceeding the LTOP limits if the accumulators are fully injected.
The Completion Times are based on operating experience that
- these activities can be accomplished in these time periods and on engineering evaluations indicating that an event requiring LTOP is not likely in the allowed times.
- l. .
-N v.-
.BRAIDWOOD - UNITS 1 & 2 B 3.4.12 - 11 9/18/98 Revision K m
l LTOP System B 3.4.12 7 BASES l (O 1 l ACTIONS (continued) i l L1 l 1 L In MODE 4, with one required RCS relief valve inoperable. l the RCS relief valve must be restored to OPERABLE status ! within a Completion Time of 7 days. Two RCS relief valves I in any combination of the PORVS and the RHR suction relief l valves are required to provide low temperature overpressure j mitigation while withstanding a single failure of an active i component. I The Completion Time considers that only one of the RCS , relief valves is required to mitigate an overpressure j transient and that the likelihood of an active failure of the remaining valve path during this time period is very low. F.1 The consequences of operational events that will I overpressurize the RCS are more severe at lower temperature ! (Ref. 5). Thus. with one of the two RCS relief valves I c .3 inoperable in MODE 5 or in MODE 6 with the head on. the ! i i Com)letion Time to restore two valves to OPERABLE status is !
'd 24 1ours.
The Completion Time represents a reasonable time to investigate and repair several types of relief valve failures without exposure to a lengthy period with only one OPERABLE RCS relief valve to protect against overpressure events.
/
7~ ( BRAIDWOOD - UNITS 1 & 2 B 3.4.12 - 12 8/21/98 Revision A
LTOP System .i B 3.4.12 l BASES ACTIONS (continued) G.1 1 The RCS must be depressurized and a vent must be established within 8 hours when: ,
- a. Both required RCS relief valves are inoperable; or
- b. The Required Action and associated Completion Time of .
Condition D. E. or F 1s not met; or J
- c. The LTOP System is-inoperable for any reason other -
than Condition A. B. C. D. E. or F. l The vent must be sized a 2.0 square inches to ensure that l the. flow capacity is greater than that required for the i worst case mass input transient reasonable during the l , applicable MODES. This action is needed to protect the RCPB from a low temperature overpressure event and a possible brittle failure of the reactor vessel. The Completion Time considers the time required to place the
.g- unit in this Condition and the relatively low yobability of i an overpressure event during this time period 'due to I increased operator awareness of administrative control requirements.
l 4-i
- i. !
BRAIDWOOD - UNITS-1 & 2 B 3.4.12 - 13 8/21/98 Revision A
- ~ . - .
LTOP System B 3.4.12 l . BASES f'} ss l SURVEILLANCE SR 3.4.12.1. SR 3 4.12.2. and SR 3.4.12 3 REQUIREMENTS l- To minimize the potential for a low temperature overpressure event by limiting the mass input capability, all SI pumps l and all charging pumps but one centrifugal charging pump are verified incapable of injecting into the RCS. anc tne accumulator discharge isolation valves are verified closea and de-energized. l The SI pumps and charging pumps are rendered incapable of l injecting into the RCS through removing the power from the pumps by racking the breakers out under administrative control. An alternate method of LTOP control may be
- employed using at least two independent means to prevent a mass addition event such that a single failure or single action will not result in an injection into the RCS. This may be accomplished through the pump control switch being placed in pull to lock and at least one valve in the discharge flow path being closed. This latter method is appropriate when the SI pump needs to be available for mitigation of the effects of a loss of decay heat removal event (Ref. 6). Another alternate method of LTOP control may be utilized when a pump must be energized for testing or
?s
! for filling accumulators to assure positive control of the
- V) capability for injectim by the pump. This may be accomplished by closing the isolation valve and removing power from the valve operator, or by secur m a manual l isolation valve in the closed position. These methods are I acceptable provided that an OPERABLE flow path exists from the RWST to the RCS.
The Frequency of 12 hours is sufficient. considering other indications and alarms available to the operator in the control room, to verify the required status of the equipment. i SR 3.4.12.3 is modified by a Note stating that accumulator isolation is only required to be met for an accumulator if its pressure is greater than or ecual to the maximum RCS pressure.for the existing RCS colc leg temperature allowed by the P/T limit curves provided in the PTLR. l l~b lV BRAIDWOOD - UNITS 1 & 2 B 3.4.12 - 14 8/21/98 Revision A
LTOP System B 3.4.12 l BASES A(-j-SURVEILLANCE REQUIREMENTS (continued) , SR 3.4.12,4 The RCS. vent of i 2.0 square. inches is proven OPERABLE.by verifying its:open condition either: ,
- a. Once every,12 hours for a valve that cannot be locked.
- b. 0nce every-31 days for a valve that'is locked, sealed; E or secured in position. A removed pressurizer safety' jj valve fits this category. ,
-4 The passive vent arrangement must only be open to be' OPERABLE. This' Surveillance is required to-be performed if
{ the vent is being used to satisfy--the pressure relief requirements of LC0 3.4.12.d.4. SR ~3.4.12.5
- Each required RHR suction relief valve shall be demonstrated is;
- ' OPERABLE by verifying its RHR suction isolation valves are
-f open. This Surveillance is only required to be performed if 4 . tie RHR suction relief valve is being used to satisfy.this
- - (,*) . ' g - 'LCO.
v-The'RHR suction isolation valves. RH8701A and RH87018 for ; relief valve RH8708A. and RH8702A and RH87028 for relief i valve RH8708B. are verified to be opened every-72 hours. ] The Frequency is considered adequate in view of other I administrative controls such as valve status indications available to the operator in the control room that verify the RHR suction valves remain open, j c, , L 2, The ASME Code. Section XI (Ref. 7), test per Inservice 3 Testing Program verifies OPERABILITY by proving proper 1 A relief valve mechanical motion and by measuring and, if l { required, adjusting the lift setpoint. i I I' t s % a\) i IBRAIDWOOD,- UNITS 1 S 2 B 3.4.12 - 15 10/5/98 Revision K i y
1 L LTOP-System B 3.4.12-73 ' BASES
- &)
SURVEILLANCE REQUIREMENTS.(continued). ! SR 3:4.12.6
' The PORV block valve must be verified open every 72 hours to provide tne. flow path for each required PORV to perform its i function when ' actuated. . The valve must be remotely _ verified
.open in the main control room.
The' block valve is a remotely controlled, motor. operated' ! valve.' .The power to the valve operator is not required ! removed, and the manual operator'is:not required locked in the inactive position. Thus, the block valve can be closed-in the event the PORV develops excessive leakage or does not close (sticks open) after relieving an overpressure ! situation.
. j 3 The-72 hour Frequency is considered adequate in view of .
4 .other: administrative controls available to the operator in 1 1 1 -the control room, such as valve position indication, that
- W verify that the PORV block valve remains open.
N
. @! -SR 3.'4.12.7 JO \~~f
- Performance'of a COT is required within 12-hours after
' decreasing RCS temperature to s 350 F'and every 31 days on ;
each required PORV to verify and. as necessary, adjust its ' lift- setpoint. The COT wi_11 verify the setpoint is within the allowed maximum limits in the PTLR. PORV' actuation
'C
{ could depressurize the RCS and is not' required.
'N
~
The 12 hour. Frequency considers the unlikelihood of a low
- *perature overpressure event during this time. <
, -w A Note indicates that this SR is not required to be performed until 12 hours after decreasing RCS cold leg temperature to s 350*F. _The COT cannot be performed until in the LTOP MODES when the PORV lift setpoint can be reduced to the LTOP setting. The test must be performed within
-12 hours after entering the LTOP MODES.
,1f~). BRAIDWOOD - UNITS 1 & 2 B 3.4.12 - 16 10/5/98 Revision K r', e -< - , - . -
ww LTOP. System .
-B 3.4'.12 j
/'%:L t
.'BASES:-
'qsw f.!SURVEILLANCEREQUIREMENTS.-(continued)?
, SR 3!4 12:8-
~
- ;PefformanceofaCHANNELCALIBRATION'on'eachrequired'PORV.
~ actuation channel is required every 18 months.'?o ' adjust the.
;whole channel.so1that-it responds and the val:ve opens within i OtheLrequired range.and: accuracy.to known input.
1
REFERENCES:
- 1. 110C'R50.fAppendix':G.
FL
~
L 2. : Generic Letter-88-11.
-4
- 3. ASME,-Boiler'and Pressure Vessel Code; Section III;
~4 :UFSAR,. Chapter 15.
-j
+w $ 5. Generic Letter 90-06,
} on ,
6 '.' Safety Evaluation Report, dated August.31, 1990. , 3w ; :t l
- 7. ASME. Boiler and Pressure Vessel Code ' Section XI.
?%. .
i N' - l H 1 1 f
-(. -
,,. : ; -i . :q ; ' ?BRAIDWOOD'- UNITS 1-&-2' B 3.4.12 - 17 10/5/98 Revision K [. p ,
I RCS.0perational LEAKAGE. B 3.4.13 l L LO a 8 3.4 REACTOR COOLANT SYSTEM (RCS) B 3.4.13: RCS Operational LEAKAGE-BASES BACKGROUND Components-that contain or transport.the coolant to or from-the reactor core make up the RCS. Component joints are made l by welding ' bolting, rolling, or 3ressure loading. Valves isolate connecting systems from t1e RCS.
.During plant life, the joint and valve interfaces can produce varying amounts of reactor coolant LEAKAGE. through
.either normal operational wear or mechanical deterioration.
The purpose of the RCS. Operational LEAKAGE LCO is to limit system operation in the presence of LEAKAGE from these l sources to amounts that do not compromise safety. This LC0 specifies the types and amounts of LEAKAGE. 10 CFR 50, Appendix A. GDC 30 (Ref.1), requires means for ; detecting and, to the extent practical, identifying the
. source of reactor coolant LEAKAGE. Regulatory Guide 1.45 (Ref. 2) describes acceptable methods for selecting leakage 7
detection systems. The leakage detection instrumentation is I discussed in Section 3.4.15. t The safety significance of RCS LEAKAGE varies widely depending on its source, rate. and duration. Therefore. detecting'and monitoring reactor coolant LEAKAGE into the containment area is necessary. Quickly separating the identified LEAKAGE from the unidentified LEAKAGE is , necessary.to provide quantitative information to the . ; operators =, allowing them to take corrective action should a l
. leak occur that is detrimental to the safety of the facility- l and the public. i l
- A-limited amount of leakage inside containment is expected '
! from systems that cannot be made 100% leaktight. Leakage from these systems should be detected, located (identified). , and . isolated in such a manner, if possible. .to not interfere l with detection of unidentified RCS leakage. l l 4 5 L .BRAIDWOOD - UNITS 1 & 2 B 3.4.13 - 1 8/21/98 Revision A
V RCS Operational LEAKAGE L B 3.4.13 iin ,1 <: BASES' I
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BACKGROUND (co'ntinued) l This LC0 deals with orotection of the Reactor Coolant i
~ Pressure Boundary (RCPB) from degradation and the core from inadequate cooling, in addi_ tion to preventing the accident -
-analyses radiation release assumptions from being exceeded.
.The consequences of violating this LCO ' include.the ' ,
- possibility of a. Loss Of Coolant Accident- (LOCA). However; the abi.lity .to monitor leakage provides advance warning to l 3ermit unit shutdown before a LOCA occurs. This advantage :
las been shown by " leak before break" studies. l
' APPLICABLE! Except for primary to secondary LEAKAGE. the safety analyses SAFETY ANALYSIS. do not address operational. LEAKAGE -However, other operational LEAKAGE is related to the safety analyses for .
. LOCA: the amount of leakage can affect the_ probability of L such an event. The safety analysis for an event resulti.ng in. steam discharge to the atmosphere assumes 1 gpm primary to secondary. LEAKAGE as the initial condition. Primary to secondary LEAKAGE is a factor in the dose releases outside containment resulting from a Steam Line ,fT Break (SLB) accident because such leakage contaminates the- ? N/' ~ f secondary fluid. Other accidents or transients involve L secondary steam release to the atmosphere. such as a Steam Generator Tube Rupture (SGTR). The SGTR is more limiting than the SLB for site radiation releases. The UFSAR (Ref. 3) analysis for SGTR assumes the contaminated secondary fluid is released for a limited time r- via-the steam generator PORV.. After a tube rupture occurs, reactor coolant'immediately begins flowing from the primary system into the. secondary side of the ruptured steam
. generator causing the RCS pressure to decrease until a
,. reactor trip occurs on-low pressurizer pressure. The analysis assumes a loss of Offsite Power occurs coincident with the reactor trip causing the Reactor Coolant Pumps to L trip and the main condenser to become unavailable when the i:; circulating water pumps are lost, p .n. b.
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RCS 0perational LEAKAGE B 3.4.13
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BASES V ' '
' APPLICABLE SAFETY ANALYSES (continued)
I After the reactor trips. the core power quickly decreases to I decay heat levels. The steam dump system cannot be used to-
- dissipate the core decay heat .due to the unavailable condenser. Therefore, the secondary pressure increases in the Steam Generators-(SGs) until the steam. generator PORVs
. open at which time the ruptured steam generator PORV is
- assumed to fail in the open position. The ruptured SG failed PORV .is-isolated when the-block. valve is manually closed twenty minutes 'after the PORV first opened. The 1 gpm primary to secondary LEAKAGE is relatively inconsequential to the results of this analysis. .
-The dose consequences resulting from the SLB accident are l well within'the limits defined in 10 CFR 100.
1 To support the use of sleeving techniques for steam generator tube repair, the Unit 1 primary to secondary . leakage limits are conservatively reduced from 500 gpd for I any single steam generator and 1 gpm. total to 150 gpd for any single steam generator and 600 gpd total (Ref. 4)'.
-y ' The RCS operational LEAKAGE satisfies Criterion 2'of .i i 10 CFR 50.36(c)(2)(ii).
p LC0: RCS operational LEAKAGE shall be limited to;
- a. Pressure Boundary LEAKAGE l
.No pressure boundary LEAKAGE is allowed, being i indicative of material deterioration. LEAKAGE of this I type is unacceptable as the leak itself'could cause l further deterioration, resulting in higher LEAKAGE. !
Violation of this LCO could result in-continued degradation of the RCPB. LEAKAGE past seals, valve seats, and gaskets is not pressure boundary LEAKAGE. l n ! L H
]
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RCS Operational LEAKAGE B 3.4.13 BASES ! LCO.(co'ntinued) L
- b. Unidentified LEAKAGE One gallon per minute (gpm) of unidentified LEAKAGE is allowed as a reasonable-minimum detectable amount tha:
l the containment air monitoring and containment sump j' discharge flow monitoring equipment can detect within ! a reasonable time period, Violation of this LCO could result in continued degradation of the RCPB. if the-LEAKAGE is from the pressure boundary,
- c. Identified LEAKAGE ,
i Up to 10 gpm of identified LEAKAGE is considered
- allowable because LEAKAGE is from known sources that do not. interfere with detection of. unidentified LEAKAGE and is well within the capability of the RCS l Makeup System. Identified LEAKAGE includes LEAKAGE to h the containment,from specifically known and located sources but~does not include pressure boundary LEAKAGE.or controlled Reactor Coolant Pump (RCP) seal leakoff (a normal function not considered LEAKAGE). ,
Violation of this LCO could result in continued degradation of a component or system. e d. Primarv to Secondarv LEAKAGE throuah 611 Steam l ! Generators (SGs) 1 l Total primary to secondary LEAKAGE amounting to l 600 gallons per day through all SGs not isolated from the RCS produces acceptable offsite doses in the SLB l
- accident analysis. Violation of this LC0 could exceed ,
the offsite dose limits for this accident. Primary to ) secondary LEAKAGE must be included in the total ' g allowable limit for identified LEAKAGE. l 3 o I L 1 l
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RCS Operational LEAKAGE B 3.4.13
'3 BASES V
LCO (continued)
- e. Primary to Secondarv LEAKAGE throuch Any One SG The 150 gallons per day limit on one SG is based on the assumption that a single crack leaking tnis amount would not propagate to a SGTR under the stress conditions of a LOCA or a main steam line rupture. If leaked through many cracks. then the cracks are very small. and the above assumption is conservative.
LCO 3.4.14. "RCS Pressure Isolation Valve (PIV) Leakaae." measures leakage through each individual Pressure Isolation Valve (PIV) and can impact this LCO. Of the two PIVs in series in each isolated line, leakage measured through one PIV does not result in RCS LEAKAGE when the other is leak tight. If both valves leak and result in a loss of mass from the RCS the loss must be included as identified LEAKAGE. APPLICABILITY In MODES 1. 2. 3. and 4. the potential for RCPB LEAKAGE is r3 greater due to RCS pressure. l In MODES 5 and 6. LEAKAGE limits are not required because the reactor coolant pressure is far lower. resulting in lower stresses and reduced potentials for LEAKAGE. ACTIONS A.1 , 1 Unidentified LEAKAGE. identified LEAKAGE. or primary to l secondary LEAKAGE in excess of the LCO limits must be l reduced to within limits within 4 hours. This Completion Time allows time to verify leakage rates and either identify unidentified LEAKAGE or reduce LEAKAGE to within limits before the reactor must be shut down. This lequired Action is necessary to prevent further deterioratior, of the RCPB. ~ (. Q) BRAIDWOOD - UNITS 1 & 2 B 3.4.13 - 5 8/21/98 Revision A
RCS Operational LEAKAGE B 3.4.13
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BASES ACTIONS (continued) L l and 6.2 If any pressure boundary LEAKAGE exists. or if unidentified LEAKAGE. identified LEAKAGE. or primary to secondary LEAKAGE cannot be reduced to within limits within 4 hours. the reactor must be brought to lower pressure conditions to reduce the severity of the LEAKAGE and its potential 7 consequences. It should be noted that LEAKAGE past seals r and gaskets is not pressure boundary LEAKAGE. The unit must
> be brought to MODE 3 within 6 hours and MODE 5 within 36 hours. This action reduces the LEAKAGE and also reduces 5 the factors that tend to degrade the pressure boundary.
The allowed Completion Times are reasonable based on operating experience. to reach the required unit conditions from full power conditions in an orderly manner and without
' challenging plant systems. In MODE 5. the ]ressure stresses acting on the RCPB are much lower. and furtier deterioration is much less likely.
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RCS Operational LEAKAGE B 3.4.13
~i BASES J i SURVEILLANCE SR 3.4.13.1 REQUIREMENTS Verifying RCS LEAKAGE to be within the LCO limits ensures the integrity of.the RCPB is maintained. Pressure boundary LEAKAGE would at first appear as unidentified LEAKAGE and can only be positively identified by inspection. It should be noted that LEAKAGE past seals. valve seats, and gaskets is not pressure boundary LEAKAGE. Unidentified LEAKAGE and identified LEAKAGE are determined by performance of an RCS water inventory balance. Primary to secondary leakage is a component of the gross leakage as determined by the performance of an RCS inventory balance. Primary to
-secondary leakage is quantified by analysis of the radionuclides present in secondary feedwater. steam, or condensate, or the noncondensible gaseous effluent.
The RCS water inventory balance must be performed with the reactor at steady state operating conditions and near o)erating pressure. Therefore. a Note is added allowing tlat this SR is not required to be performed until 12 hours after establishing steady state operation. The 12 hour allowance provides sufficient time to collect and process c ., all necessary data after stable plant conditions are
/
) established.
O Steady state operation is required to perform a proper inventory balance since calculations during maneuvering are not useful. For RCS operational LEAKAGE determination by water inventory balance, steady state is defined as stable RCS pressure 3 2150 psig), tem 3erature, power level. 3ressurizer and makeup tank levels, ma(eup and letdown, and RCP seal injection and return flows.
.T.
eg An early warning of pressure boundary LEAKAGE or m unidentified LEAKAGE is provided by the systems that monitor
% the containment atmosohere radioactivity and the containment
$ suma level. It should be noted that LEAKAGE past seals and gascets is not pressure boundary LEAKAGE. These leakage d detection systems are specified in LCO 3.4.15. "RCS Leakage W Detection Instrumentation."
The 72 hour Frequency during steady state operation is a reasonable interval to trend LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents.
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i RCS Operational-' LEAKAGE B 3.4.13
'S ? BASES-
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SURVEILLANCE REQUIREMENTS (continued) SR '3.4.13.2 This SR provides the means'necessary to determine SG OPERABILITY'. The requirement to demonstrate SG tube l integr y in accordance with the Steam Generator Tube l Survei ance Program emphasizes the importance of SG tube integrity, even though this Surveillance cannot be. performed
-at normal-operating conditions.
1
~ REFERENCES 1. .10 CFR 50.. Appendix A. GDC'30.
- 2. Regulatory Guide 1.45. May-1973.
- 3. UFSAR, Chapter 15.
l- 4. Safety Evaluation Report, dated May 7, 1994. i f.x . 1 l- - L i
, f.
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RCS PIV Leakage B 3.4.14 /' B 3.4 REACTOR COOLANT SYSTEM (RCS) b B 3.4.14 RCS Pressure Isolation Valve (PIV) Leakage BASES BACKGROUND 10 CFR 50.2. 10 CFR 50.55a(c). and GDC 55 of 10 CFR 50. Appendix A (Refs.1. 2. and 3). define RCS PIVs as any two normally closed valves in series within the Reactor Coolant Pressure Boundary (RCPB) which separate the high pressure RCS from an attached low pressure system. During their lives, these valves can produce varying amounts of reactor coolant leakage through either normal operational wear or mechanical deterioration. The RCS PIV Leakage LCO allows RCS high pressure operation when leakage through these valves exists in amounts that do not compromise safety. The PIV leakage limit applies to each individual valve. Leakage through both series PIVs in a line must be included as part of the identified LEAKAGE. governed by LC0 3.4.13.
"RCS Operational LEAKAGE." This is true during operation only when the loss of RCS mass through two series valves is determined by a water inventory balance (SR 3.4.13.1). A f known component of the identified LEAKAGE before operation
( begins is the least of the two individual leak rates determined for leaking series PIVs during the required surveillance testing: leakage measured through one PIV in a , line is not RCS operational LEAKAGE if the other is ! leaktight, i l Although this specification provides a limit on allowable l PIV leakage rate, its main purpose is to prevent overpressure failure of the low pressure portions of connecting systems. The leakage limit is an indication that the PIVs between the RCS and the connecting systems are degraded or degrading. PIV leakage could lead to overpressurization of the low pressure piping or components. Failure consequences could be a Loss Of Coolant Accident (LOCA) outside of containment. an unanalyzed accident. that could degrade the ability for low pressure injection. d,a BRAIDWOOD - UNITS 1 & 2 B 3.4.14 - 1 8/21/98 Revision A
l RCS PIV Leakage I B 3.4.14 [ +
- BASES BACKGROUND (continued)
The basis for this LCO is the 1975'NRC " Reactor' Safety
)
Study" (Ref. 4) that identified potential intersystem LOCAs as a significant contributor to the risk of core melt. A ' subsequent study'(Ref. 5)' evaluated various PIV l configurations'to _oetermine the probability of intersystem l LOCAs. PIVs.are provided to isolate the RCS from tne following connected systems:
- a. ResidualHEatRemoval(R.HR) System:
- b. Safety Injection (SI) System; and
- c. Chemical and Volume Control System.
Violation of this LCO could result in continued degradation of a PIV, which could lead to overpressurization of a low pressure system-and the loss of the integrity of a fission product barrier. l L ' APPLICABLE' Reference 4 identified potential intersystem LOCAs as a
<[u . SAFETY ANALYSES significant contributor to the risk of core melt. The dominant accident sequence in the intersystem LOCA category is the failure of the low pressure portion of the RHR System outside of containment. The accident is the result of a i postulated-failure of the PIVs. which are part of the RCPB.
and the subsequent pressurization of the RHR-System downstream of the PIVs from the RCS. Because the low 4 pressure portion of the RHR System is designed for 600 psig. I overpressurization failure of the RHR low pressure line ~ ; could result in a LOCA outside containment and subsequent risk of core melt. Reference 5 evaluated various PIV configurations, leakage testing of the valves, and operational changes to determine the effect on the probability of intersystem LOCAs. This
~
study concluded that periodic leakage testing of the PIVs can substantially reduce the probability of an intersystem LOCA. RCS PIV leakage satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). a iBRAIDWOOD - UNITS 1 & 2 B 3.4.14 - 2 8/21/98 Revision A
RCS PIV Leakage B 3.4.14
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O ' BASES
. V LCO RCS PIV OPERABILITY protects the low pressure systems attached to the'RCS from potential failure due to overpressurization. This protection (i.e.. RCS PIV OPERABILITY) is provided by both the leak tight PIVs and the
~
RHR System suction isolation valve interlocks.
< RCS PIV leakage .is identified LEAKAGE into closed systems
~ connected to the RCS. Isolation valve leakage is usually on the order of drops per minute, Leakage that increases significantly suggests that something is operationally wrong and corrective action must be taken. The LCO PIV leakage limit is 0.5 gpm per, nominal inch of valve size with a maximum limit of 5 gpm. The previous criterion of 1 gpm for all valve sizes imposed an unjustified penalty on the larger valves without providing. information on potential valve degradation and resulted in higher personnel radiation exposures. A study concluded a leakage rate limit based on valve size was superior to a single allowable value (Ref. 6). Reference 7 permits-leakage testing at a lower pressure l differential than between the specified maximum RCS pressure , and the normal pressure of the connected system during RCS j operation (the maximum 3ressure differential) in those types 1 of valves in which the ligher service pressure will tend to j diminish the overall leakage channel opening. In such j . cases, the observed rate may be adjusted to the maximum ! pressure differential by assuming leakage is directly . proportional to the pressure differential to the one half power-. L 4 (N
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'RCS PIV Leakage. ,
, B 3.4.14 l fy"y BASES l T jf
~ I
- LC0 (continued): -
The!followingvalvesareRCSPIVs: Valve Number Function- i SI8900A . .B. C D Charging /SI check valve ; SI8815 Charging /SI backup check ; valve-S18948A. B.-C. D Accumulator check valve. SI8956A.-B. C. D Accumulator backup check-. valve SI8818A. B. C. D RHR cold leg check' valve l SI8819A.'B. C. O SI cold leg check valve- j SI8949A. B.'C. D. ~SI hot leg check. valve SI8905A, B. C. D SI hot leg backup checkivalve SI8841A. B RHR hot leg check valve RH8701A B RHR suction Motor;0perated Valve (MOV) RH8702A, B RHR suction MOV l APPLICABILITY In MODES:1. 2. 3. and 4..this LCO ap) lies because the PIV-L'(l-
-leakage potential is greatest when tie RCS is-pressurized. I u -1 InLMODES 5 and 6. leakage limits are not provided because the lower reactor coolant pressure results in a reduced potential for leakage and for a LOCA outside the containment.
i
-l ACTIONS The Actions are modified by two. Notes. Note 1 provides j clarification that separate entry into a Condition is i allowed for each flow path. .This is allowed based upon the l T.. functional inde'pendence of the' flow path. . Note 2 requires
? 1:'
an evaluation.of affected' systems if a PIV is inoperable. The leakage may have affected system operability, or isolation of a leaking flow path with an alternate valve may i -have degraded the ability of the-interconnected system to perform its safety function. 1 1 i, g.
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RCS PIV Leakage B 3.4.14
- ". BASES i )
ACTIONS (continued) A.1 and A 2 The flow path must be isolated by two valves. Required Actions A.1' and A.2 are modified by a Note that the valve used for isolation must meet the same leakage requirements as the PIVs and must be within the RCPB or the high pressure portion of the system. i Required Action A.1 requires that the isolation with one ! valve must be performed within 4 hours. Four hours provides ! time to reduce leakage in excess of the allowable limit and i to isolate the affected system if leakage cannot be reduced. The 4 hour Completion Time allows the actions and restricts , the operation with leaking isolation valves. i Required Action A.2 specifies that the double isolation , barrier of two valves be restored by closing some other i
,t valve qualified for isolation or restoring one leaking PIV. I w The 72 hour Completion Time after exceeding the limit
{ considers the time required to complete this Action and the low probability of a second valve failing during this
, period.
i ,) - ! j- y ! The inoperability of the RHR System suction isolation valve interlock could allow inadvertent opening of the valves at RCS 3ressures in excess of the RHR Systems design 3ressure. If tie RHR System suction isolation valve interlocc is inoperable, operation may continue as long as the affected RHR suction penetration is closed by at least one de-energized. power operated valve within 4 hours. This Action accomplishes the purpose of the interlock function. O BRAIDWOOD - UNITS 1 & 2 B 3.4.14 - 5 8/22/98 Revision K
7, RCS PIV Leakage B 3.4.14
)
fy BASE 3 ! I ACTIONS'(continued) C.1 and C.2 If the' Required Actions and associated Completion Times of Conditions A and B are~not met, the unit must be brought to a MODE in which the requirement _does not apply. To achieve
.this' status. the unit must be brought to MODE 3 within 6 hours and MODE 5 within 36 hours. This Action may reduce the leakage and also reduces the potential for-a LOCA outside the containment. The allowed Completion Times are reasonable based on operating experience, to reach the:
recuired unit conditions from full-power conditions in an orcerly manner and without challenging plant systems. SURVEIL' LANCE- SR 3.4.14.1 REQUIREMENTS
' Performance of leakage testing on each RCS PIV 'or isolation valve used to satisfy Required Action _ A.1 and Required Action A.2 is required ~to verify.that leakage is below the specified limit and to identify each. leaning valve. The leakage limit of 0.5 gpm per inch of nominal valve! diameter - (_i up to 5 gpm maximum applies to each valve. Leakage testing
, v' requires a stable pressure condition.
-For'two PIVs in series, the leakage requirement applies to
.each valve individually and not to the combined leakage-across both valves. If the PIVs are not individually leakage tested. one valve may have failed completely and not be detected if the other valve in series meets the leakage rec ui rement. In this situation, the protection provided by-recundant valves would be lost.
co o Testing is to be performed every 18 months. a typical J refueling cycle, if the plant does not go into MODE 5 for at y least 7 days. The 18 month Frequency is consistent with ei 10 CFR 50.55a(g) (Ref. 8) as contained in the Inservice . Testing Program, is within the frequency allowed by the
,*Y - American Society of Mechanical Engineers (ASME) Code.
l: Section XI (Ref. 7). l
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m . RCS PIV Leakage l 1 B 3.4.14 l
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r ; IBASEs J I. l
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SURVEILLANCE REQUIREMENTS-(continued) l 1
- l. l Testing must also be performed prior to entering MODE 2 .!
whenever the unit has been in MODE 5 for a 7 days if leakage l testing has not been performed once within'the previous !
-9~ months. 1 L The . leakage testing is typically performed at the RCS .
' pressure associated with MODES 1 and 2. This permits f
leakage testing at high differential pressures with !
-o stable conditions. However, test pressures less than l i 2235 psig but greater.than 350 psig are allowed. When i
.; -measured at these reduced pressures.. observed leakage must H be adjusted for the actual test pressure up to 2235 psig !
y assuming the leakage to be directly proportional.to pressure i a - differential to the one half power. , This SR is modified by three Notes. Note.1 allows entry- ) into MODESL3 and 4 to establish the necessary differential- ' pressures and' stable conditions to allow for performance of this Surveillance. Note 1 is applicable to all Frequencies of this Surveillance.
'In addition, tercing must be performed once after.the. valve u(9 L/
has been opened by flow or exercised to ensure tight reseating. -PIVs disturbed.in the performance of this Surveillance should also be tested-unless it has'been
- established (per Note 2) .that an . infinite testing loop.
cannot' practically be avoided. Testing must be performed within 24 hours after the valve has been reseated if in MODE 1 or 2. or prior to entry. into~ MODE 2 if not in MODE 1 or 2.at the end of the 24 hour period. Within 24 hours is a-reasonable and practical _ time limit for performing this test after. opening or reseating a valve. Note 3 exempts the RHR suction isolation valves (RH8701A
- and B and RH8702A and B) from the specified Frequency of this. testing since these MOVs'are not subject to the same
-failure characteristics as a check valve that has actuated due to flow, s
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fs RCS PlV. Leakage B 3 4.14 e-;, (yG
. BASES' SURVEILLANCE REQUIREMENTS-(continued) .
< SR 3.4.14.2-1The interlock setpoint- that prevents the RHR System suction isolation valves from being opened is set so the actual RCS pressure must be < 360 psig'to open the valves. This-E setpoint' ensures the RHR design pressure will not be
- exceeded and the RHR relief valves will not . lift.- The
'18 month Frequency is based on the need to perform the Surveillance under conditions that' apply during a unit-
. outage. The 18 month. Frequency is also acceptable based on consideration of the design reliability (and confirming operating experience) of the equipment.
REFERENCES' 1. 10 CFR 50.2.
- 2. - 10 CFR 50.55a(c).
- 3. - 10 CFR 50. Appendix A, Section V. GDC 55.- '
4 WASH-1400 (NUREG-75/014). Appendix V. October.1975.
.(,-~l.-
^~ / 5. NUREG-0677. May-1980.
- 6. - EG&G Report. EGG-NTAP-6175.
- p 7. ASME. Boiler and Pressure Vessel Code. Section XI. ,
F-
-l.- .8. 10 CFR 50.55a(g).
'i '
1 Ji M D
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RCS Leakage Detection Instrumentation ; B 3.4.15 B'3.4 REACTOR COOLANT SYSTEM (RCS) ; B 3.4.15 RCS Leakage Detection Instrumentation l BASES l BA'KGROUND GDC 30 of Appendix A to 10 CFR 50 (Ref. 1) requires means for detecting and. to the extent practical, identifying the i location of the source of RCS LEAKAGE. Regulatory ; Guide 1.45 (Ref. 2) describes acceptable methods for j selecting leakage detection systems. l Leakage-detection systems must have the capability to detect l l- significant Reactor Coolant Pressure Boundary (RCPB). I degradation as soon after occurrence as practical to L minimize the potential for propagation to a gross failure. ) l- Thus, an~early indication or warning signal is necessary to l L permit proper evaluation of all unidentified LEAKAGE. j l Industry practice has shown that water flow changes of 0.5 ) to 1.0 gpm can be readily detected in contained volumes by l i monitoring changes in water level, in flow rate, or in the ! operating frequency of a pump. The containment sump. used L - to collect unidentified LEAKAGE. is instrumented to alarm L( 'N for leakages of 1.0 gpm. This sensitivity is acceptable for detecting increases in unidentified LEAKAGE. The reactor coolant contains radioactivity that, when released to the containment, can be detected by radiation monitoring instrumentation. Instrument sensitivities of 10] pCi/cc radioactivity for particulate monitoring and of 10' pCi/cc radioactivity for gaseous monitoring are practical for these leakage detection systems. Radioactivity detection systems are included for monitoring both particulate and gaseous activities because of their l sensitivities and rapid responses to RCS LEAKAGE. An increase in humidity of the containment atmosphere would indicate release of water vapor to the containment. Dew 1 Joint temperature measurements can thus be used to monitor lumidity levels of the containment atmosphere as an indicator of potential RCS LEAKAGE. , 4 l
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l RCS Leakage Detection Instrumentation B 3.4.15 BASES l BACKGROUND (continued) Since. the humidity level is influenced by several factors, a
. quantitative evaluation of an indicated leakage rate by this means may be questionable and should be compared to observed increases in 11auid flow into or from the containment sump Humidity. level monitoring is considered most useful as an
. indirect alarm or indication to alert the operator to a potential problem. -Humidity monitors are not required by this LCO.
Air temperature and pressure monitoring methods may also be used to infer unidentified LEAKAGE to the containment. Containment temperature and pressure. fluctuate slightly. during unit operation, but a rise above the normally indicated range of values may indicate RCS leakage into the containment. The relevance of temperature and pressure measurements are affected by. containment free volume and, for temperature, detector location. Alarm . signals from these instruments can be valuable in recognizing rapid and
-sizable leakage to the containment. Temperature and pressure monitors are not required by this LCO.
-o O APPLICABLE. The need to evaluate the severity of an alarm or an SAFETY ANALYSES indication is important to the operators, and the ability'to compare and verify with indications from other systems is necessary. The system response times and sensitivities are described in the UFSAR (Ref, 3).
The_ safety significance of RCS LEAKAGE varies widely I depending on its source. rate, and duration. Therefore. detecting and moni_toring RCS LEAKAGE into the containment area is necessary. Quickly separating the identified LEAKAGE from the unidentified LEAKAGE provides quantitative information to the operators allowing them to take corrective action should a leak occur detrimental to the safety of the plant and the public. RCS leakage detection instrumentation satisfies Criterion 1 of 10 CFR 50 36(c)(2)(ii). 1 O BRAIDWOOD - UNITS 1 & 2 B 3.4.15 - 2 8/21/98 Revision A
~. _ _ . _. _ _ . . _m _ _ . . _ _. . ._ _ .
RCS' Leakage Detection Instrumentation B 3.4.15
/~' BASES
- L)y LCO- One method of protecting against large RCS leakage derives
'from the ability of instruments to rapidly detect extremely small leaks. This LC0 requires instruments of diverse monitoring principles to be OPERABLE to orovide a high degree-of confidence that extremely smal'l' leaks are detected in time tc allow actions to place the unit in a safe condition. when RCS LEAKAGE indicates possible RCPB degradation.
The LC0 is satisfied when monitors of diverse measurement means'are available. Thus, the containment sump monitor, in combination with a gaseous or particulate radioactivity monitor, provides an acceptable minimum. The containment floor drain sump flow monitor (RF008) and the reactor cavity sump flow monitor. (RF010) are utilized to fulfill the containment sump monitor requirement. For the containment atmosphere radioactivity monitor, the PR011A (particulate) or PR011B (gaseous) monitor satisfies the LC0 requirement. IAPPLICABILITY. Because of: elevated RCS temperature and pressure in MODES.1. L>t 2, 3. and 4. RCS leakage detection instrumentation is required to.be OPERABLE. In MODE 5 or 6. the temperature is to be s 200 F and pressure is maintained low or at atmospheric pressure. Since the temperatures and pressures are far lower than those for MODES 1. 2. 3. and 4. the likelihood of leakage and crack propagation are much smaller. Therefore, the requirements of this LC0 are not applicable in MODES 5 and 6. ? (py E(/
, BRAIDWOOD - UNITS 1 & 2 B 3.4.15 - 3 8/22/98 Revision K g
1
'f
.RCS Leakage Dete' c tion Instrumentation i B 3.4.15 .
h BASES l ACTIONS? 'A'.~1 and A.2 m With the required containment . sump monitor. inoperable. 'no
~o ther. form'of sampling-can provide.the equivalent information::however, the containment atmosphere-radioactivity monitor will provide indications of changes in
. leakage...Together with the atmosphere. monitor. the periodic a surveillance -for RCS water inventory balance,-- SR 3.4.13.1.
-~
must' be performed at.anL increased frequency of 24 hours to l'. l 9? 3rovide information that-is' adequate to detect leakage . A iote is added allowing that SR 3.4.13.1-is not required to-be performed until 12 hours after establishing steady state M- operation'(stable RCS pressure : temperature, power ~ level.. Q pressurizer and makeup-tank levels, makeup and letdown and
.:q -_RCP seal; injection and return flows). The 12 hour allowance provides sufficient time to collect and process all necessary data-after. stable plant conditions are
' established.
Restorationfof the. required sump monitor to OPERABLE-status
' within a Completion Time _of 30-days is required to regain the function after the monitor's failure. This time is L acceptable, considering-.the Frequency and adequacy of the ?,\ RCS water inventory balance' required by Required Action A.1, Q
Required Action A.1 and Required Action A.2 are modified by a Note that indicates that'the provisions of LCO 3.0.4 are not applicable. As a result, a MODE change is allowed when the containment sump monitor is inoperable. This allowance is,provided because other instrumentation is available to monitor RCS. leakage.
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BRAIDWOOD-UNITSI&2 B 3.4.15 - 4 10/5/98 Revision K h
RCS Leakage Detection Instrumentation B 3.4.15 O BASES
'~'
ACTIONS'(continued) B.1.1. B.1.2. and 8.2 With both gaseous and particulate containment atmosphere radioactivity monitoring instrumentation channels inoperable, alternative action is required. Either grab
, samples of the containment atmosphere must be taken and 9 analyzed for gaseous and particulate radioactivity or water inventory balances, in accordance with SR 3.4.13.1. must be 4 performed to provide alternate periodic information.
I With a sample obtained and analyzed or water inventory-2 balance performed every 24 hours, the reactor may be operated for up to 30 days to allow restoration of the / required containment atmosphere radioactivity monitors. The 24 hour interval provides periodic information that is adequate to_ detect leakage. A Note is added allowing that SR 3.4.13.1 is not required to be performed until 12 hours after establishing steady state operation (stable RCS Ed pressure, temperature, power level, pressurizer and makeup yg tank levels, makeup and letdown, and RCP seal injection and ml 1, return flows). The 12 hour allowance provides sufficient (,T :. e time to collect and process all necessary data after r
->;t stable plant conditions are established. The 30 day Com]letion Time recognizes at least one other form of leacage detection is available.
Recuired Action B.1 and Required Action B.2 are modified by a hote that indicates that the provisions of LC0 3.0.4 are not applicable. As a result, a MODE change is allowed when the gaseous and particulate containment atmosphere radioactivity channel is inoperable. This allowance is provided because other instrumentation is available to monitor RCS leakage. w) BRAIDWOOD - UNITS 1 & 2 B 3.4.15 - 5 8/22/98 Revision K
f l RCS Leakage Detection Instrumentation B 3.4.15 I , ('T BASES Q ,1 l ACTIONS (continued) ! C.1 and C.2 . If a Required Action and associated Completion Time of Condition A or B is not met, the unit must be brought to a i MODE-in which the requirement does not apply. To achieve this status, the unit must be brought to at least MODE 3 I within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full i power conditions in an orderly manner and without I challenging plant systems. j 1 RJ. With all required monitors inoperable, no means of I monitoring leakage are available, and immediate actions, in accordance with LC0 3.0.3. are required. I SURVEILLANCE SR 3.4.15.1 REQUIREMENTS (s)' SR 3.4.15.1 requires the performance of a CHANNEL CHECK of the required containment atmosphere radioactivity monitor. The check gives reasonable confidence that the channel is operating properly. The Frequency of 12 hours is based on instrument reliability and is reasonable for detecting off normal conditions. SR 3.4.15.2 SR 3.4.15.2 requires the performance of a COT on the required containment atmosphere radioactivity monitor. The test ensures that the monitor can perform its function in the desired manner. The test consists of exercising the digital computer hardware using data base manipulation and injecting simulated process data to verify OPERABILITY of alarm and trip functions. The test verifies the alarm setpoint and relative accuracy of the instrument string. The Frequency of 92 days considers instrument reliability. and operating experience has shown that it is proper for detecting degradation. (^N
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BRAIDWOOD - UNITS 1 & 2 B 3.4.15 - 6 8/21/98 Revision A
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RCS Leakage Detection Instrumentation B 3.4.15
.[~ BASES SURVEILLANCE REQUIREMENTS (continued)
L SR 3.4.15.3 and SR 3.4.15 4 These SRs require the performance of a CHANNEL CALIBRATION for each of the required.RCS leakage detection instrumentation channels. The calibration verifies the , accuracy of the.1nstrument string. including the instruments.
- i. located inside. containment. -The Frequency of -18 months is a typicalLrefueling cycle and considers channel reliability.
- Again, operating experience has proven that this Frequency is acceptable.
REFERENCES 1. 10.CFR 50, Appendix A.'Section IV. GDC 30.
- 2. . Regulatory. Guide 1.45,
- 3. UFSAR. Section 5.2.5; i
'l O 1 l
[; O ;
'BRAIDWOOD - UNITS 1 & 2 B 3.4.15 - 7 8/21/98 Revision A l'
RCS Specific Activity B 3.4.16 -[~1 B 3.'4 REACTOR COOLANT SYSTEM (RCS) B 3.4.'16 RCS' Specific Activity BASES BACKGROUND The maximum dose to the whole body and the thyroid that an individual at the site boundary can receive for 2 hours. during an accident is specified in 10 CFR 100 (Ref.1). The , limits on specific activity ensure that the doses are held ; to a small fraction of the 10 CFR 100 limits during analyzed , transients and accidents. l The RCS specific activity LCO limits the allowable concentration level of radionuclides in the reactor ~ coolant. The LC0 limits are established to minimize the offsite radioactivity dose consequences in the event of a Steam Generator Tube Rupture (SGTR) accident. The LCO contains specific activity limits for both DOSE EQUIVALENT I-131 and gross specific activity. The allowable levels are intended to limit the 2 hour dose at the site boundary to a small' fraction of the 10 CFR 100 dose guideline limits. The limits in the LCO are standardized. [,T based on parametric evaluations of offsite radioactivity , Nd-dose consequences for typical site locations. ! The parametric evaluations showed the potential offsite dose levels for a SGTR accident were an appropriately small ! fraction of the 10 CFR 100 dose guideline limits. Each evaluation assumes a broad range of site applicable > atmospheric dispersion factors in a parametric evaluation. I l' f \ APPLICABLE The LCO limits on the specific activity of the reactor ;
' SAFETY ANALYSES coolant ensures that the resulting 2 hour doses at the site !
a boundary will not exceed a small fraction of the 10 CFR 100 ! s dose guideline limits following a SGTR accident. The SGTR l
- safety analysis (Ref. 2) assumes the specific activity of :
the reactor coolant at the LCO limit and an existing reactor 0 coolant Steam Generator (SG) tube leakage rate of 1 gpm. The safety analysis assumes the specific activity of the ,
,'i secondary coolant at its limit of 0.1 pCi/gm DOSE EQUIVALENT i s- I-131 from LC0 3.7.3. " Secondary Specific Activity." )
l
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w/ l BRAIDWOOD - UNITS 1 & 2 B 3.4.16 - 1 10/1/98 Revision K I
RCS Specific Activity B 3.4.16 D tV BASES' l APPLICABLE SAFETY ANALYSES (continued) ! The analysis for the SGTR accident establishes the acceptance limits for RCS specific activity. Reference to this analysis is used to assess changes to the unit that could affect RCS sptc1fic activity, as they relate to the acceptance limits. The analysis is for two cases of reactor coolant specific activity. One case assumes specific activity at 1.0 uCi/gm DOSE EQUIVALENT I-131 with a concurrent large lodine spike that increases the I-131 iodine release rate from the fuel to the coolant to a value 500 times greater than the release rate corresponding to the initial _ primary system iodine concentration. The second case assumes the initial reactor coolant iodine activity at.60.0 pCi/gm DOSE EQUIVALENT I-131 due to a pre-accident iodine spike caused by an RCS transient. In both cases, the noble gas activity in the reactor coolant assumes 1% failed fuel, which closely equals the LCO limit of 100/E uCi/gm for gross specific activity. An SGTR event causes a reduction in reactor coolant' inventory. The reduction initiates a reactor trip from a
. low pressurizer pressure signal or an RCS Overtemperature a AT signal.
n- l If a coincident loss of offsite power occurs, the steam dump valves close to protect the condenser. The rise in pressure in the ruptured SG discharges radioactively contaminated steam to the atmosphere through the SG power operated relief valves and the main steam safety valves. The unaffected SGs remove core decay heat by venting steam to the atmosphere
.until the cooldown ends.
The safety analysis shows the radiological consequences of an SGTR accident are within a small fraction of the Reference 1 dose guideline limits. Operation with iodine specific activity levels greater than the LCO limit is permissible. if the activity levels do not exceed the limits shown in Figure 3.4.16-1, in the applicable s)ecification, for more than 48 hours. The safety analysis las concurrent and pre-accident iodine spiking levels up to 60.0 pCi/gm DOSE EQUIVALENT I-131. BRAIDWOOD - UNITS 1 & 2 B 3.4.16 - 2 8/21/98 Revision A
m 1 , RCS-Specific Activity '
, B-3.4.16 D LBAS'ESJ 4'h . ' ~
' :APPLICABLELSAFETY: ANALYSES (continued)
- TheLremainder of the above limit permissible iodine levels shown cin Figure 3.4.16-1 are acceptable because of the low- -
0l - probability.of'a SGTR' accident occurring-during the- ,
- established'48 hour time limit. The occurrence of an SGTR-
~
s
; - accident- at these permissible ' levels- could . increase the. site - -
t ' boundary dose levels. .but!still' be.within 10 CFR :100 dose- -
' guideline limits.
T' - The: limits;on RCS-specific activity are also used for: 1' 4 . establishing' standardization in radiation shielding and.- y plant.' personnel radiation protection practices. 1 kJ RCS specific activity satisfies Criterion 2 of
, _ 10.CFR 50.36(c)(2)(ii). 1 LLC 0 ~The? specific-iodine activity:is limited to 1.0 pC1/gm DOSE
' EQUIVALENTzl-131. The gross specific activity in the
. reactor coolant _is limited.to the number of pCi/gm. equal to 100 divided by E (average-disintegration energy of the sum g -
of the average beta and gamma energies of the coolant p M : nuclides). The limit on DOSE EQUIVALENT I-131 ensures the - .x 2 hour thyroid dose:to an individual at the site boundary
.during' the Design Basis Accident (DBA) will be a small- ,
fraction of the allowed thyroid dose. The limit.on gross ! specific activity ensures the 2 hour whole body dose to-an individual at the site boundary during the DBA will be a 1 L small- fraction-of the allowed whole body dose. : L . The SGTR accident analysis (Ref. 2):shows that the 2 hour E . site boundary dose levels.are within acceptable limits. , Violation of the LC0'may result in< reactor coolant L radioactivity levels that could, in the event'of an SGTR. L . lead to' site' boundary doses that exceed the 10 CFR 100 dose L guideline limits. l' i g f; . y. l BRAIDWOOD - UNITS 1 & 2 B 3.4.16 - 3 8/22/98 Revision K 4 % y r . -. m , w w _ . ~ - -g..
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RCS Specific Activity B 3.4.16 y
-O' BASES
- d. .
APPLICABILITY - In MODES 1 and 2. and in MODE 3 with RCS average temperature a 500 F. operation within the LCO limits for DOSE EQUIVALENT I-13! rd gross specific activity are necessary to contain i the pote,tial consequences of an SGTR to within the l acceptable site _ boundary dose values. l l-
-. For operation in MODE 3 with RCS-average temperature.
< 500 F. and in MODES 4 and 5. the release of radioactivity in the event.of a SGTR.is unlikely since the saturation
. pressure of.the reactor coolant is below the lift pressure settings of the main steam safety valves. ;
4
' ACTIONS A.1-and A;2 With the DOSE EQUIVALENT I-131 specific activity greater than the LC0 limit, samples at intervals of 4 hours must be
. taken to demonstrate that the limits of Figure 3.4.16-1 are 1 l not exceeded. The Completion Time of 4 hours provides sufficient time to obtain and analyze a sample. Sampling is '
done to continue to provide a. trend.
.O
~v The DOSE EQUIVALENT I-131 s)ecific activity must be restored to within limits within 48 lours. The Completion Time of-48 hours'is required, if the limit violation resulted from '
. normal iodine spiking.
A Note to the Required Actions excludes the MODE change , restriction of LC0 3.0.4. This exception allows entry into ' the a3plicable MODE (SL while relying on the ACTIONS even i thoug1 the ACTIONS may eventually require unit shutdown. I I This exception is acceptable due to the significant conservatism incorporated into the specific activity limit. ! the low probability of an event' which is limiting due to exceeding this limit, and the ability to restore transient specific activity excursions while the unit remains at, or l proceeds to power operation. l l l h 1 V 1 l- l
- O 8/21/98 Revision A
. .BRAIDWOOD.- UNITS 1 & 2 B 3.4.16 - 4
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_ _ - _ _ _ .. _ _ _ _ _ . ._ _ ~ _ _ _ ._-. _ .. _ ... _ _ . .__ _ __ RCS Specific Activity B 3.4.16 BASES ACTIONS (continued) ell > If the Required Action and associated Completion Time of ! Condition A is not met or if the DOSE EQUIVALENT I-131-specific activity is in the unacceptable region of Figure 3.4.16-1, the reactor must be brought to MODE 3 with
.RCS average temperature < 500 F within 6 nours. .The Completion Time of 6 hours is reasonable, based on operating experience to reach MODE 3 below 500*F from full power conditions in an orderly manner and without challenging plant systems.
L1 ! With the gross specific activity in excess of the allowed , limit the unit must be placed in MODE 3 with RCS average ; temperature < 500'F. This action lowers the saturation pressure of the reactor coolant _below the setpoints of the ; main steam safety valves and prevents venting the SG to the environment in an SGTR event. The Completion Time of L, 6 hours is reasonable, based on operating experience, to i, - reach MODE 3 below 500 F from full power conditions in an l
) . orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.4.16.1
' REQUIREMENTS SR 3.4.16.1 requires performing a gamma isotopic analysis as a measure of the gross specific activity of the reactor l; coolant at least once every 7 days. A gross radioactivity L analysis consists of the quantitative measurement of the total specific activity of the reactor coolant except for radionuclides with half lives < 10 minutes and all ,
- radiciodines. The total specific activity is the sum of the ,
- j. degassed beta-gamma activity and the total of all identified L gaseous activities in the sample within 2 hours after the L sample was taken. Determination of the contributors to the i gross specific activity are based upon those energy peaks j identifiable with a 95% confidence level. The latest-available data may be used for pure beta emitting l-L radionuclides. This Surveillance provides an indication of
!4 any increase in gross specific activity. r BRAIDWOOD - UNITS 1 & 2 8 3.4.16 - 5 8/21/98 Revision A l
RCS ' Specific' Activity L B 3.4.16
,jy l BAS'E'S '
SURVEILLANCE ~ REQUIREMENTS (continued) : Trending the results,of th'is Surveillance ~ allows proper remedial action to be taken before reaching the LCO limit : under normal operating conditior.s. -The Surveillance is '! applicable in. MODES 1 and 2, and in MODE 3'with RCS average
~
. temperature a 500*F. The 7 day Frequency considers the l
, unlikelihood of.a gross fuei failure during the time.
SR 3.4.16.2 ' i This. Surveillance is performed in MODE 1 only to ensure iodine remains within_ limit ~ during normal operation and'
, following fast power changes when fuel-failure is more apt to occur. The 14; day Frequency.is adequate-to trend changes in the iodine activity. level. considering gross activity is i ug pl
= monitored every 7 days. The Frequency. between 2.and ,
+ 6 hours after a ower change 2 15% RTP within a 1 hour.
l . period, is estab ished because' the iodine levels peak 1during A - this time following fuel failure: samples at other times gl .would provide inaccurate results. SR 3.4.16.3 , D A radiochemical analysis for E determination is' required d every 184Ldays (6 months) witti the unit operating.in MODE 1 equilibrium conditions. The E determination directly relates to the LCO and is required to verify unit operation
.within the specified gross activity LC0 limit.: The analysis for-[ is a measurement of.the average energies per
- disintegration for isotopes with half lives longer than ;
10 minutes;_ excluding iodines. The Frequency of=184 days recognizes =E does not change rapidly. This SR has been modified by a Note that indicates sampling is required to be performed wirnin 31' days after a minimum of,2 effective full power days sni 20 days of, MODE 1 , operation have elapsed since the reactor was last subcritical .for at least 48 hours; This ensures that the radioactive materials are at equilibrium so the analysis for
~
[ is representative and not skewed by:a crud burst or other similar abnormal event. 9 L ,
/- ,
[k/ ! BRAIDWOOD . UNITS 1 & 2 B 3.4.16 - 6 8/22/98 Revision K
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RCS Specific Activity B 3.4.16
/N BASES N)
REFERENCES- 1. 10 CFR 100.11, 1973.
- 2. UFSAR. Section 15.6.3.
- 3. Safety Evaluation Report dated May 7. 1994.
- 4. Safety Evaluation Report. dated August 18. 1994
- 5. Safety Evaluation Report, dated November 9. 1995.
V i 1 l l h
') 'w BRAIDWOOD - UNITS 1 & 2 B 3.4.16 - 7 8/21/98 Revision A
a 1 RCS Loop Isolation. Valves-
~
B 3.4.17
/~i B 3.4 REACTOR COOLANT SYSTEM (RCS).
B 3.4.17 RCS Loop Isolation Valves' BASES
.'BkCKGROUND'- The RCS may be' operated with loops isolated in order'to perform maintenance. While operating with a loop 1solated' . ,
there is potential. for inadvertently opening the . isolation 1 valves in the isolated loop. In this event, the coolant in l
.the isolated loop would suddenly begin to mix with the j
.l.' coolant in the unisolated portion of the RCS. This- .
. situation has the potential of causing a positive reactivity i addition with a corresponding reduction of SDM if: .
1
- a. The temperature. in the isolated loop is lower than the i T temperature in the unisolated portion of the RCS (cold ]
water incident); or ! l
- b. The boron concentration in the isolated loop is lower I than the boron concentration required in the RCS to meet'SDM~(boron dilution. incident).
As discussed in the UFSAR (Ref. 1) the startup of-an
- f. _} ' 'j- isolated loop is- performed-in a controlled manner that Y virtually eliminates any sudden positive reactivity addition-from cold water or boron dilution because: i
- a. LC0 3i4.18. "RCS Isolated Loop Startup." and plant I operating procedures require that the boron
'3 concentration in the isolated loop be maintained l higher than the required SDM boron concentration ~of E the unisolated portion of the RCS. thus eliminating T the potential for introducing coolant from the A isolated loop that could dilute the boron q concentration in the unisolated portion of the RCS to N= less than the required SDM boron concentration:
- b. The cold leg loop isolation valve cannot be opened unless the temperatures of both the hot and cold legs of the isolated loop are within 20*F of the temperatures of the hot and cold legs of the unisolated portion of the RCS (compliance is ensured by operating procedures and automatic interlocks); and f
I-p O ! .BRAIDWOOD . UNITS 1 & 2 . B 3.4.17 - 1 10/5/98 Revision K n -
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. .- f RCS.-LoopLIsolation Valses-B 3.4;17 M BASES-BACKGROUND (continued)-
- c. '0ther' automatic interlocks' all' of which are'part of~
the Reactor. Protection: System (RPS). prevent opening-
, -the. hot leg loop' isolation'valvelunless the cold. leg o- gloop isolation valve is fully closed. '
f a Hl
'R fAPPLICABLE . .
Daring startup;of.an isolated loop in accordance.with '
^; . : SAFETY ANALYSES 'LC0 3.4.18 'the cold leg' loop isolation. valve interlocks and 1; operating procedures prevent; opening of the valveLuntil the
'4 -
isolated loop and:unisolated portion of the RCS boron
';- k ; concentrations and. temperatures are within limits. This ;
. ensures that any undesirable reactivity effect' from the .
A. H- isolated: loop does'not occur. The safety analyses assume.'a minimum. SDM as an initial
, condition for Design Basis Accidents-(DBAs) (Ref. 1).
. . . Violation of the LCO.Lcombined with mixing of.the isolatad .
T loop coolant into?the unisolated portion of'the RCS; could i result in the SDM being less than that assumed;i.n.the. safety- ,
- analyses.
[j. The above analysesiare for DBAs that establish the' 'l V ' acceptance limits 'for:the RCS loop isolation valves.
- H Reference to-the analyses for these DBAs is.used to assess
. changes to the RCS loop isolation valves as they relate'to .
- the acceptance limits. ,
F The boron concentration of ~an isolated loop may-affect SDM 3- Land therefore RCS loop isolation valves satisfy Criterion 2 '
.of 10 CFR 50.36(c)(2)(ii).
4 .k u ?O 4BRAIDWOOD.- UNITS 1 &-2 8 3.4.17 - 2 10/8/98 Revision K
-RCS Loop Isolation Valves l B 3.4.17 g A/ : BASES
.i 10- ,
%/ / . l
% 1 w ,LCD- 'This LC0 ensures that a loop isolation valve that becomes .
g: closed in MODES 1 through 4 is. fully isolated and the plant 1
- p. ' placed 'in MODE 5. - Loop isolation valves are used for S performing maintenance when the plant is in MODE 5 or 6, and I q ;startupofan.isolatedloopiscoveredbyLC03.4.18. ]
hT.' l- This LCO also ensures that loop isolation valves remain open :I in MODES.1, 2. 3 and 4. Closure of the loop isolation l
. valves during these MODES results in the )otential for an inadvertent startup of an isolated loop w1ich could result
.in the SDM being.less than assumed in the safety analyses. ;
l I 2--APPLICABILITY'. 'In MODES'1 through 4. this LCO is applicable since Li unisolating an isolated -loop has not been analyzed. The
- N C potential affects (with 6 oron concentration or temperature '
$l 7,4 less than that of the unisolated portion of the RCS) may.
include an inadvertent criticality.
'In MODES 5 and 6. the.SDM of the operating loops is large enough to permit operation with isolated loops. In these MODES.. controlled startup of isolated loops is possible
' h. without significant risk of inadvertent criticality. Q i l l 1 l
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l i
; sq.
Q), BRAIDWOOD - UNITS 1 & 2 B 3.4.17 - 3 10/1/98 Revision K 7 i y
RCS Loop. Isolation Valves.
.B 3.4.17 BASES-ACTIONS 'The Actions.have' been_ provided with a Note to clarify that all .RCS loop isolation valves for this LCO are treated as separate entities, each with separate Completion Times. !
(i.e. the Completion Time is on a component basis). 6_1
.If power is inadvertently restored to one or more loop isolation valve operators, the potential exists for accidental isolation of a loop with a subsequent inadvertent ,
startup of the isolated loop. The loop isolation valves ' have motor operators. Therefore, these valves will maintain their last position when power is removed from the' valve 0)erator. With power ap) lied to the valve operators. only tie-interlocks prevent.tle valve from being operated. : Although operating' procedures and interlocks make'the l 1 occurrence of this event unlikely the prudent action is to l remove power from the loop isolation valve operators. ..The
' Completion Time of 30 minutes to remove power from the loop ,
isolation valve operators is sufficient considering the i complexity of the task.- . B;1. B.2. and B.3
)
O Should a loop isolation valve be closed in MODES 1 I
'through 4. the affected loop must be fully isolated i immediately and the unit placed in MODE 5 to preclude j inadvertent startup of the loop and the potential I inadvertent criticality. Recuired Actions B.2 and B.3 l require placing the unit in F0DE 3 within 6 hours and M0"E 5 within 36 hours. The allowed Completion Times are reasonable. based on operating experience, to reach the recuired unit conditions from full power conditions in an orcerly manner and wii.oout challenging plant systems.
4 l i O : BRAIDWOOD - UNITS 1 & 2 B 3.4.17 - 4 8/21/98 Revision A L e, ,- ,
RCS Loop Isolation Valves B 3.4.17 (~i L! BASES SURVEILLANCE -SR 3.4.17.1 REQUIREMENTS The Surveillance is performed at least once per 31 days to ensure that the RCS loop isolation valves are open. with power removed from the 1000 isolation valve operators. The
~
primary function of this Surveillance is to ensure that power is removed from the valve operaters, since SR 3.4.4.1 of LC0 3.4.4. "RCS Loops-MODES 1 and 2." ensures that the loop isolation valves are open by verifying every 12 hours that all loops are operating and circulating reactor coolant. The Frequency of 31 days ensuras that the required flow can be made available, is based on engineering judgment, and has proven to be acceptable. Operating experience has shown that the failure rate is so low that the 31 day Frequency is justified. , REFERENCES' 1. UFSAR. Section 15.4.4. D (V , l l l
)
l l 1 L ' , (* ' j A,.s) i BRAIDWOOD - UNITS 1 & 2 B - 3. 4.17 - 5 8/21/98 Revision A i J-l
.RCS . Lonps - Isolated . l B 3.'4.18
,^ ,. . .
7y -B-3.4-LREACTOR COOLANT, SYSTEM:(RCS) 4fB ..B 3!4.18. -- RCS: Loops - Isolated
- BASES-LBACKGROUND? The RCS may.be operated with loops isolated in MODES 5 and 6 in order to perform maintenance. While operating with-a -
. loop isolated, there is: potential for inadvertently opening the isolation valves in-the isolated loop. In this event.
.the coolant in-the isolated loop would suddenly begin to mix with the coolant in the unisolated portion of the RCS. This situation has the potential'of causing a' positive reactivity
. addition with a corresponding reduction of SDM if T a. The temperature in the isolated loop is lower than the '
l temperature in the unisolated portion of the RCS (cold water incident): or
- b. 'The boron. concentration in the isolated loop'is -lower than'the boron concentration required in the RCS to meet SDM-(boron dilution incident).
As' discussed in the UFSAR (Ref. 1), the startu) of an f' isolated loop is done in a controlled manner taat virtually V)- l '- eliminates any. sudden positive reactivity addition from cold l water or' boron dilution because: a.- -This'LC0 and plant operating procedures require that i the boron concentration in the isolated loop be
, J maintained higher than-the required SDM boron
.T concentration of the unisolated portion of the RCS.
3 thus eliminating the potential for introducing coolant
.) from the isolated loop that could dilute the boron y concentration in the unisolated portion of the RCS to q less than the required SDM boron concentration:
- b. The cold leg loop isolation valve-cannot be opened unless the temperatures of both the hot leg and cold leg of the isolated loo) are within 20*F of the unisolated portion of tie RCS. Com)11ance with the temperature requirement is ensured )y operating procedures and automatic interlocks: and
[ Nq h LBRAIDWOOD- UNITS 1 &'2 B 3.4.18 - 1 10/5/98 Revision K o i y
RCS Loops-Isolated B 3.4.18 m BASES . l l
~
' BACKGROUND (continued) s c. Other automatic interlocks prevent opening the hot leg G loop isolation valve unless the cold leg loop i isolation valve is fully closed. All of the k interlocks are part of the Reactor Protection System.
N I\{ APPLICABLE During startup of an isolated loop, the cold leg loop
% SAFETY ANALYSES isolation valve interlocks and operating procedures prevent
@ opening th' valve until the isolated loop and unisolated s portion of the RCS boron concentrations and temperatures are within limits. This ensures that any undesirable reactivity effect from the isolated loop does not occur.
The safety analyses assume a minimum SDM as an initial condition for Design Basis Accidents. Violation of this LC0 could result in the SDM being reduced in the operating loops to less than that assumed in_the safety analyses. The boron concentration of an isolated loop may affect SDM and therefore RCS isolated loop startup satisfies , Criterion 2 of 10 CFR 50.36(c)(2)(ii). ! 1
- J LC0 Loop isolation valves are used for )erforming' maintenance when the unit is in MODE 5 or 6. T11s LC0 ensures that the 1 loop isolation valves remain closed until the differentials of temperature and boron concentration between the unisolated portion of the RCS and the isolated loops are 2l within acceptable limits.
t H} D APPLICABILITY In MODES 5 and 6. the SDM of the unisolated portion of the
% RCS is-large enough to permit operation with isolated loops.
In these MODES. controlled startup of isolated loops is possible without significant- risk of inadvertent criticality. In MODES 1. 2. 3. and 4. operation with isolated loops is not permitted. See LCO 3.4.17. "RCS Loop Isolation Valves." m
%.]
BRAIDWOOD - UNITS 1 & 2 B 3.4.18 - 2 10/8/98 Revision K i
3
~, . RCS Loops-Isolated i B-3.4.18- j 9N ! BASES
' j
-(b
,.< ACTIONS iAi1 and B.1' Required Action A.1 and Required Action B.1 assume'that the -
H
-prerequisites of the LCO are not met and'a loop isolation
. valve-has been. inadvertently opened. Therefore, the Actions j .f .. '
N 4- require immediate closure-of isolation valves to preclude a - N# , m-boron. dilution event or a cold water event. a
?
-SURVEILLANCE' SR 3.4.18.li 1 REQUIREMENTS This: Surveillance is performed to ensure that the ,
; temperature' differential between the isolated loop and the unisolated portion of the RCS is s 20 F. Performing the Surveillance 30 minutes prior to . opening the cold leg fi ' isolation valve in the isolated loop provides: reasonable
.g
. assurance, based on engineering ' judgment, that the.
xtemperature differential'will stay within limits until the e cold- leg isolation- valve -is 'o)ened. This Frequency has been 4 , shown'to be: acceptable througl operating experience. j V SRt 3:4.18.2 ! g L[ !
,9'
- To ensure that the boron concentration of the isolated loop . .
p -isLgreater than or equal to the boron concentration required i F in the'RCS'to meet!SDM, a Surveillance is performed:4 hours 3rior.to opening either the hot or cold leg isolation valve. t Lllp )erforming the Surveillance 4 hours prior to opening.either : the hot or cold leg isolation-valve provides reasonable i
' t- ' assurance the resulting baron concentration difference will
~
<- be within acceptable limits when the loop is unisolated This Frequency is acceptable due to the amount of time required to sample and confirm concentration results.
REFERENCES 1. UFSAR, Section.15.4.4. 1 T T: .
~BRAIDWOOD - UNITS i l' & 2 B 3.4'.18 - 3 10/1/98 Revision K ll
4 e a 4 d, A & nmm. a - s.A4 ~ * ,,% +^9 9- A z-,44 u - d - -rL--4 d ,--a-eiAA--mma-34 T] BYRON CTS MARKUPS O l f -
,. .s, ,. n. . . . . . .an.-.--~.--.~ ~ . - . - . - . . -
. - . . . . . . . . -l c o . '5 4 ; 2. '- . . - . . . . . . . - . . . ~ .i
.SecA on 3' l t _ . j 2tAewd 2.' eno i s.J-r ' svo M ' (Y
;- ([R[.^.CTIVITY CON 7%CL 3737C^3)
!$82
' Q%,:.. 2(.5MINIMLIM TEMPERATURE FOR CRITICALITY
. l
' LIMITING CONDITION FOR OPERATION
- t. '
^
- L CO ? .4.2.
N The Rehetor: Coolant' System lowest operating loop temperature'(T"V9)
..shall be greater than or~ equal to 550*F. 1 I
APPLICABILITY': -MODES l'and'.2# ACTION. 1 C.J A 'With' a Reactor Coolant System operating loop temperature (1,yg) less than . 4 550*F,:r: n;r; C.,.g _t; s tn'- :: ' :
- t --i tt' - n 2: m = a.:be-inEE}
F -fr =--- .within the r.axt 15 minutes. A, MittJ wWh Ken c 1. of l g SURVEILLANCE REQUIREMENTS'
- - -)
< J
.SR 34.2.. l _
The Reactor Coolant System temperature (T"V9);shall be determined to . be greater than~or. equal'to 550*F:
-- e . "ithin 15 ; inst;; prier t: ;;ti;c' .;; r;;;t;r ;riti;;1ity, ;nd ,
o O~ At least once oeff minu es who "the. reactor is critical ~and~ the) o l b. ]'l ' Reasftor ant Co stem avg. is ss thedr 557*F 'th the ava-Tref , i-a_ D[viatio Ala not reset./ 44 4 i {/4 N 1 L @ b r.s l m 1 Q A L i: 1 f.- f I LCO .- i , ' Arru(A0aiW
- -#With K g greater than or equal to 1.
(*S;; ip;;i;l I;;t Ex;eptier.; Sp;;ifi.. 'm. 3.10.3.) ]
- O
' BYRON l- UNITS 1 & 2- 3/4 1-6 U
LC O 3.2.2. Lco 3.4. I POWER DISTRIBUTION LIMITS A3 mUJ1 6bi3.227
% h o c & Seg.,12 d LIMITING CONDITION FOR OPERATION ,
ACTION (Continued)
- b. Within24hoursofinitiallybeidgoutsidetheabovelimits, veri rough (incore flux mapping and)RCS total flow rate compariso at the c inationof}Ffy and,RCS total flow rate are resto ed o within the ove limits, or reduce THERMAL POWER to ss than 5% of RATED THERMAL WER within the next 2 hours; an
- c. Identify and correc the cause of the out- limit condition prior to increasing THERMAL ER above the uced THERMAL POWER limit required by* ACTION a.2. an r b. e; subsequent POWER OPERATION may proceed provided that the ination of ,Fh and(indicated RCS total flow rate are demo rated, th hhcore flux mapping and l
[ RCS total flow rate e arison, to be w1 'n the region of acceptabie operation defined Specification 3.2.3 pr to exceeding the fol-lowing THERMAL ER levels:
- 1. A no
- al 50% of RATED THERMAL POWER,
- 2. nominal 75% of RATED THERMAL POWER, and Within 24 hours of attaining greater than or equal to 95% f
/
RATED THERMAL POWER. \ SURVEILLANCE REQUIREMENTS { 8i'
- we n.. -, u.wm , n , . ,,u c.eJ 5 e3
... e combination of indicated RCS total flow rate (and sbW ldetermined to n the region of acceptable operat f Specification 3.2.3:
- a. Prior to operatio o RMAL POWER after each fuel loadin .
;l i '
Atleast once oer 31 Effective Full Power Days. el 5(3.4.l.3(4.2.2.9 The indic.a.dp_d_RCS_to_tal. flow tate _shall be verified to t,e within the; fregien ;f e;ceptable ep;r:thn Of Sp::if t::ti;n 2.2.3'at least once per 12 hours H' " W " A' ' T a e l 5t34.1.4:2.1O The RCS total flow rate shall be determined by precision j
==: heat balance ===H=p Nr y measurement ?-'e- '- - N etier of Nih L -- -
% ^ ya nnier enanistina twa nenujgjang cf Sp jfje.T ;3 A n } r" 301 Spplic*ble, f2Mie Themeasurementinstrumentationshallbecalibratedwith17sevencayspiorto
' ""O the pe ormance of e calorimetric flow measurement. ' rior to the p cision A hea alance meas ement, at leas wo of the four f ater flow m er /
v uris shall visually inspe ed and, if foulin s found, all enturis/ f all be clea d. f 3/4 2-9 AMENDMENT NO. 49 BYRON - UNITS 1 & 2 6 t'-
~ '
. 9,,, , ,
{ )k
- TABLE :3. 2-1 h
.E.
h 1 DNS-PARAMETERS-
^
- e. .-PARAMETER -LIMITS
.x _
L3 LCO 3.4.I.b Indicated Reactor Coolant System T,,, .< 591.2*F.L ~ w -
-. i.
l.CO 3.4.1. a. Indicated Pressurizer Pressure * > 2219'psig*
. n q
4 k-3
't s ;
n '!
'tt
~
O' I i y k I k
.I .
- Limit not applicable'during eitheria Ti1ERMAL POWER ramp In M gmi'!
LCO MOTE +- excess of 5% of RATED TilERMAL POWER per minute or.a TilERMAL POWER step in excess -of 10% of- RATED TilERMAL. POWER. Q 0 '! t ta ;
?I' Ni!
p :u - b
- r. .!
i _ _ _ _ - - _ _ _ - - - _ . _ __- ___ __ -__.=--_____---- _ -__- __--__-________-_ _ -
--.__-.-_.-___ - _--_-_ - .___---_-_____ - - - -_-____-_____ L ___-- - ___ - __.L_______:__
' 3.q **me= Ten ' *wnwT SYsPrw [ Rt Q ---
t c,o 3 q 'y i r_ _ ___c ::1 M s.=a ec5 t .,. - ecen) . g, LIMITINo CONDITTON FOR OPERATION
~ Lt.c 74.T 2 . 0 . *,. . :
. At least two .cf the reactor coolant loops hi. a
( .,e.,4 shall be. e --
. OPERABLE with two reactor _ coolant loops in operation when the N---tz: L 15%
.C,_1- :. 1-- --- ; . _: cir :f)and one reactor coolant loop in operation when the
(:f- ^;.a;.....;;z-
- Cb
- 't. --( b =,7 +4 ISAC4A
. . ... c. .. W 4 <e A w ... A e ]
C lant A and its associated steam r and
- b. Reactor coo nd its asso steam generator and I
reactor coolant pump,: ;
- c. Reactor Coolant Loo its associa aam generator and
- reactor c .puep, and t . c. Coolant Loop D_ and its ' associated steam oenerator nd
, eactor coolant nuno./ APPLICAEILITY: ' MODE 3. C E2'ISE8 - 4 c tJo (., Ld ? '4-R l '4 ggh ~s'. With less than the above required reactor coolant loops OPERABLE, rw I
. restore -dhe required loops to OPERAELE status within 72 hours or be 1A-14
, cota C < fBCSMUTDOWN7Ithin th'e%zt 12 hours 1 p l Coot. A Y* With only on* ***ctor coolant loop in operation and the@pector-+ed
~~~ ' ' " ' --
22:it1- ; within 1 hour 5:n t2;f L3 ( o e c &A C.i . J ..c.n 6 % % .=.a w ..a. m ,.n,A ,, w ,X)
/, tWith no reactor coolant loop in operation,4 suspend all operations .
g MS1 LAB > involving a reduction-in boren eencontration of the Reactor coolant 1 04- ( fd (4 systaajand immediately initiate corrective action to return the] i- *r 4 r, v , a t * ,veguired reactor coolant loop to operation. '
- eg*j [ "# '% i1,ma v. u.
- SURVEILLANCE REOU1REiuWii I s p .f.T -
3 4.a.1.2.1 At-least the above required reactor coolant pumps, if not in : operation, shall be determined OPERABLE once per 7 days by verifying correct ' {
~ breaker alignments and indicated power availability. I l s. 's 4 6.1 -
, 4.0.1. .- 'The required steam generators shall be determined OPERABLE by I j' verifying,secpnA gy # da Jow ranaeyster level to be greater than or equal' g a' to 18% i'1' !?- ~ + ' --ia- *a
~~'- =qat least once per 12 hours. >
wm 4.0.1. k-~ ~~
~ _ i s a ...
The required coolant loops shall be verified in operation (-f riree-]
- n eti- 7 1- 77
- :: :::;at least once per 12 hours. .
W) . @w
*All Reactor coolant pumps may be kleenereikeed for up to I hour provided
_ ...a &,4n....g (1) no operations are permitted that would cause dilution of the Reactor coolant system boron concentration, and-(2) core outlet temperature is maintained _at least 20*F below saturation temperature. 1104
'(** rf 5--eiki T::t :G i--- *a-ai'ia=**aa BYRON.- UNITS 1 & 2 3/4 4-2 AMENDMENT NO. 96
' 6:a V
- .. - . . . . . . _ . ~ _ . . . . . . . . . . . . - . . - _ . . - . . - . . . . . - . . .. . . .
., p, . REAETCMt CDOLhMT SYSTEM su m ii.LANet Riculaim nis
$p. 3 Al . L.3 ' . .
4.4.1.0.1 The required reactor coolant pump (s) and/or RJut. pumps, . if not in
- l 1 - I ' .'_ -operation, shall he 4etermined. OPERABLE once per 7 days by verifying correct )
, { breaker alignments and indicated power availability.
. $L M 1,4 . .
4.4.1.2.0 The. required steam. generator (s) shall be determined' OPERABLE-by
~
verifying _,secondag,giAe,rygr_romw34.ter 1evel to be greater than or equal. y to 18t' G it "rr "-it 1 p.... -- .. 1- $,lat least once per 12 hours. ? )
'g 1.q .(,,l "j " "W%%# .
d i e 4 . 4.1. 3.1 it 102 t "' reactor coolant or RHR loop shall be verified in I
. operation at least once per 12 hours.'
, I 1 I
. M ,
i 9 V f i 6
- u. .
anon -: UNITS :1 & 2 3/4 4-4 AMENDMENT No. % Rev K
m._. _ _ . . _ - . _ . _ . . . _ _ . _ . _ _ _ . _ . - _ _._. _ _ .. _ _ . _ _., _._ LCO 3.4. 7 "Q LOOPS FZZ. LED j L. 3,4,7 Rcs Loa Ps - Mo0E G, i
, LTMYTINC CONDTTTON FOR OPERATION .~'
LCD 3$."1 0.4.;.4.; At least one residual heat removs1 (Rict) loop shall be OPERABLE and in, operation *, and eithers .9 c.- oo 5,4.7. 4 -e. One additional RER loop shall be OPERA 5 led, or hhh { sh oi E 3.V 7. 6 -kn - The escondaryl side narrow range water level of at least two steam 3R.3A~i.7 . generators shall be gr - 1st ((43rfor y=1t 1 ora,er so) QH4 (Cyg# f* cd 5 AppticABILITY:
~
5 with reactor' coolant loops filled ##. - t% l
&GZZ.Qlis _
Svec 4 r.s-Mf 4rith gneed <Q-4he Riot loopg inoperable and with less than the require p. N p' g g g . steam generator level,.immediately initiate corrective action to return the inoperable RER loop _to OPERABLE statu.s or restore the required steam sene tor as soon as possible. esence 4_, opEE A BLE daTu s b7. tritdr RER loop in operation, suspend all operations involving a 5 A reduction in boron concentration of the Reactor coolant system and immediately initiate corrective action to return the required RRR loop to operat.Jon. (Ni 4 (gry.E"[' 3' # 0 b stavritLAmer nuuin .nd
-~
- .5A 3. Y. 7.7 4.4.;.4.1.1 L 7
' The secondary side water level of at least two steam generators y3 ,i, when required shall be determined to be within limits at least once per 12 (E $ .
q, (khc reqinreal O3 7 4.,.i.4.;.0kSt
' .icirrel ti- rer-*er :::1:_' y20) 2 r RNR at least loop once shall per be12 determined hours. to be in operation 5k 34,7,3 4:Asm M-58 % eg,n b m e 2 hou r perio
*The RHR pus @ nay beff::::: i :flfor up to 1 hour provideds (1) no operations eco :are permitted.that would cause dilution of the Reactor Coolant System boron I note I concentration, and (2) core outlet temperature is maintained at least 10*F Vr f
belon saturation temperature. focitew A 5i LCO' #0ne \RHR loop may be inoperable for up to 2 hours for surveillance testing 'Uj 140TF 2J provided the other Ram loop is cPERAaLE and in operation.
##A reactor coolant pump shall not be started with one or more of the Reactor g LM 'co61 ant system cold leg temperatures less than or equal to 350'F unless the
/jDIE3 secondary water temperature of each steam generator-is less thsn 50*F above each of the Reactor coolant system cold leg temperatures.
.LCO NOTE 4 ]INSEA.T" 3.tl- 5 c.) y u
BYRON - UNITS 1 & 2 3/4 4-5 AMENDHENT NO. 96 ile o K
. . .- ._ _, . _ . _ _ . _ .. ~ . _ _ _ . _ _ . _ _ _ . . _ . . . _ . . _ _ _ . . _
4 (. g h CTS INSERT (S)
- SECTION 3.4 !
f~Y). w sj -,. t LC0 3.4.7-
-[ INSERT 3.4' 5A : ( A.9)' - j it l[; Deleted::in:.Revis4on'K1
, . INSERT 3.4' 5B . (M 1)' ~
I 1 i . SURVEILLANCE FREQUENCY H
-SR 3,4.7.3 -Verify correct breaker alignment and 7 days indicated sower are available to each-
..c . required.RiR pump.
l y~y. w; p
~
.'I'NSERT- 3.4 5C1 -(L33 )
. NOTE- -
4' . All-RHR loopsLmay be removed from op'eration during planned heatup to MODE 4 when at least one.RCS loop is in operation.
+
7
- /
l
- c
.I.
' j' _ .Q 9/22/98 Revision K s f'. c t lt
1
. CTS INSERT (S) I g] -SECTION 3.4 l
. .w/
LC0 3.4.7l q
; INSERT < 3.4 5D '(Ah)'
. 4 .. CONDITION- REQUIRED ACTION COMPLETION TIME-
~
i = D.
- Two~ requiredLRHR ' loops D.1 Suspend.all Immediately .l
. inoperable. . operations involving-
. . . a reduction of RCS
-@ boron concentration. ,
1
; Required RHR:1 cop . MLD
'ino)eratne and:one'or . .
1
. bot 1 required SG. D . 2.1 ' Initiate.. action.to. Immediately
. secondary: side water.
restore one RHR loop 11evel(s) not within: to OPERABLE status. ilimits. g ; D.2.2 . Initiate action to Immediately TIN restore required SG
-NA -
secondary side water
. level (s) to within
. limits.
f k i-u , l l
- .'l .
Q 10/6/98 Revision K l;
LLO 2.4.B. L M g REACTOR COOLANT SYSTEM.(Rcd
;p p [O LO = UT ;0 = L;;7 ;;07 7:LLE 34.6 2 RC5 Laps - Moht 5 iimp 64 Filleci
> LIMITING-CONDITION FOR OPERATION
' LLO .2.4. s E'E'E E'E Two residual heat removal (RHR) loops shall be OPERABLE" and at
~least one:RHR' loop lshall be in-operation.**'
' APPLICABILITY:' MODE'5 with. reactor coolant loop; not filled.
ACTION: - 244 B' i ,X. :Withlless't,han the above required RHR loops OPERABLE, immediately-initiate corrective ~ action.to return the,requi*ed RHR loops to i co,a c ~ RA C.1. OPERABLE' status as'.soon as possible. Gnd A RA A:lf. 'f With no-RHR loop in operation, suspend all operations involvingg y steduction'in boron concentration of the Reactor Coolant Systemiand C immediately initiate corrective action to _ return the required Ci?
' C.,J A RA A.1 loop to operation, t.e d : C,
- i2rt er 34- Ah "
iSURVEILLANCE REQUIREMENTS gMg; Me' rebire/ .
- '- r iAt %;;t :q;jRHR loop shall be determined to be in operation d:r:::: ::: r::::: ::::::= at least once per 12 hours.
' .5R .3 4 S.1 - I TWE RT .3.4-6 E
~
ico MTE1l*0ne -RHR ' loop may' be inoperable for up to 2 hours for surveillance testing provided the other RHR loop _is OPERABLE and in ooerat' . semond from occuYoi Asn
-lcoNort1**The RHR pump may.be d::re ;he}_ tor up to 1 houred: prov (1) no operations are permittedithat would cause dilution of the Reactor Coolant System boron b concentration, and'(2) core' outlet temperature is maintained at least 10*F 9 below saturation temperature.: , ,
y [-
. [(n 4 clrLala - opme.as areperinWthsf wisld fuh, l t r<du & 3cs wak volume.. / }j*
v
.BYRONf-; UNITS 1' & 2 3/4 4-6 I'
i Rco t'.
LCo 3.q a g A. Jg REAC OR COOLANT SYSTEM (RLD _ 3,4:d9 m: . ;;. -- mP, SHUTDOWN LLL.t L..e swQ LIMITING CONDITION.FOR OPERATION tcc, 3 a .t f . iL 1 .If an RCS loop islisolated, maintain the hot leg and cold leo stoo g cme iussiac sorm C valves closed - 6 [heboronconcentrationoftheisolatedloopis yyff
,ur D. a
- a. . G G bor*on concentration of thy },,w .a c.;
- =2.
than er :q='. C k
- b. The temperature of the cold leg of the isolated Toop is within 20*F of.the highest cold leg temperature of the operating loops.
APPLICABILITY: MODES 5 and 6. ces ACTION: JWit M require of the above ification rto(satisfied..#o noj gg .open ei the hot ] r cold,,_ leo stop.. ves.f r l L su,r
@ A- 7A
. SURVEILLANCE REQUIREMENTS t
'T ~ *
-$3 u .\S.1 ,
t 4:iFT:31:F:B The. isolated loop cold leg temperature shall be determined to be within 20*F of the highest cold leg temperature of the operating loops within Se 30 a..,t minutes prior to opening the cold leg stop valve. L L'm The boron concentration of an isolated loop shall be determined to be greater than or equal to the boron concentration of the4lime ;- s ecr within) hours prior to opening either the hot leg or cold' leg stop valves of. an /1solated loop. 3 4 Lg uvuuo s o,w IE Qq, [ew pc mn ar me x v<mosa w Q L. % O
-BYRON - UNITS 1 & 2 3/4.4-8 RW E
f J 3 f
' CTS INSERT (S) $o .f , 'SECTION 3.4-1V.)
[
~
-LCO 3.4.'18 3
- r
. INSERT. 3.4 8A L(L,) ~
g .. Jl, CONDITION l REQUIRED. ACTION' . COMPLETION TIME
^
mR -
$ LA Isolated' loop hot;or-
.-cold lea isolation
.A.1~ Close hot and' cold leg' isolation valves.
.lmmediately
'j j <
- valve'open withLboron-N '
g concentration
- requirement'not_ met.
g.
~ . .. .
3 B, Isolated loop. cold leg' B.1 ' Close cold leg Immediately.
;. isolation valve.open- isolation valve.-
with-temperature ,' requirement not met. 73 AJ p ; I b T 9/15/98 Revision K [. [: p 1
Lc0 3 A. IO 1 .- 3zlfREACTOR COOLANT SYSTEM (&csl l 1 1
< 3 4.10 Pressur.w 5deh %lvd !
- k)s. .
LIMITING CONDITION FOR OPERATION LCO .3 4.10 EC2;@ All pressurizer Code safety valves shall be OPERABLE with a lift setting 2of 2485 psig IL@e_ APPLICABILITY: MODES 1', 2, and 3. ACTION: l M A -With one pressurizer Code safety valve inoperable, either*rcstore the l j .. " . . . . inoperable valve to OPERABLE status within 15 minutes ortbe in at least HOT) I CwlB / STANDBY within 6 hours and in at least HOT SHUTDOWN within the following / l 16 hours./ , Cud B ( INS 6 R.Y .3.4 - 10 8 ' )- SURVEILLANCE REQUIREMENTS- [ ' .5 E. 3A.10.1 - . N No additional requirements other than those required by I j l Specification 4.0.5. H g gf ggg ) . [be, wI4hi n 1.l'lo . J el
. o lCO NOTE : 'IIN5ERT ,
g$ [J3.4-10A~ NQ *i ec to ambien condifions fth[valy(I) Q l*The/ lift fetti g pres re s,hdll respo I aynomigal opgratin tempefatur nd p ssurp[ _ / - / / /
~
J, 1 BYRON -UNITS 1 & 2- 3/4 4-10 9 k
i CTS INSERT (S) k: 6A SECTION 3.4 LC0 3.4.10 1 l IINSERT'3.41A8 (L 3 ).- zT ,-- ---- NOTE .
., iThe 11ft settings are not . required to be.within the LCO l;mits during y MODE 3 for the purpose of setting the pressurizer safety valves under o- ambient-(hot) conditions. This exception is allowed for 54 hours .
e -following entry into MODE-3 provided-a preliminary cold setting was f made prior to heatup. , l
- INSERT '3.4 108 - (M3 )
' CONDITION- REQUIRED ACTION ~ COMPLETION TIME- l
.. l l
-[: } .g : B'.- ... B ,1 - Be in MODE 3. 6 hours l DB - eND LTwo'or more. B.2 Be in MODE 4. 12 hours ,
' pressurizer. Safety- l valves inoperable.
p 1
. I kI) 8/22/98 Revision K
+ i i
f
- + - + - ~
L
' 0 0 'i . '* .Ct ,
i
?gREACTORCOOLANTSYSTEM(RCS) 3.y$C/?.4.:D PRESSURIZER o
LIMITING CONDITION FOR OPERATION $, Lcog T The pressurizer shall be OPERABLE with at least two groups of t..>. ; l
. pressurizer heaters each having a capacity of at least 150 kWa and a water 6 level .of less than or. equal to 92%. ..
od ca96\e. ok big powered 4een <<a. A.-) a APPLICABILITY: MODES 1, 2, and 3.
,,,3,d,q Mrs puerupphed WM
. ACTION:
cod g ' g. -With less than two groups of pressurizer heaters OPERABLE, restore at l least two groups of pressurizer heaters to OPERABLE status within '
' 72 hoursfor be in at least HOT STANDBY within the next 6 hours and inj e,'.
.y -HOT SHUTDOWN within the following 6 hours.J.
y der leve\ och& Loj)---h ; c, , < g With the 3ressurizer 5therwise ;ae;,2r;tlfb'e in at least HOT STANDB p m th: 1:::ter tM : tr= M r; x xiwithin 6 hours and in HOT
)
SHUTDOWN within the following 6 hours. du\'k hsert d toAs, od 9 6e bl. C[4.\ \ SURVEILLANCE-REOUIREMENTS [' N'* I" * *MS'*P^Mc ** d M*4 l g ; ges . .c .9. : 4 '
' *1 The pressurizer water level shall be determined to be within its limit at least once per 12 hours.
L 2.:.::.c.t
'e . e . 3. 2) The capacity of each of the above required groups of pressurizer
. heaters 'shall betverifted ts enercizine the heate ud n : = rtr.; circutt , .
'~411r74i4t at least once each{ refueling interval. Q , j'3 g=
3r : . . 4. 3 ' 3 0.3.31 The. tre:: 6 ,n.n.,ei ::: ' /A Q -f N ttif pressurizer heaters tythe ESF power supply ,
.ishall be demonstrated OPERABLE at least once per 18 months [by ener; ::n; :n:',
r-b
- j V i
p BYRON -. UNITS 1 & 2 3/4 4-11 w AMENDMENT NO. 82 l FTes K
=. .- _ _ . . -.
LCO 3 4. I l 3A REACTOR COOLANT SYSTEM (Res)
. 3 A. fi Reww %* Ox*dod Mel hm libRN D/5.4.1 ."tutt Y%vty O, '
LIMITING CONDITION FOR OPERATION
, r
[ ko 3A. IIA Both power-operated relief valves (PORVs) and their associated block valves shall be OPERABLE. I' 4 APPLICABILITY: MODES 1, 2, and 3.
- --) nnter 3 a . ;z A ).Q N ***cdM g no+e I ~ ACTION:
led t h a A,. M c u a le J be m w all Co"a A x" With one or more PORV(s)_inoperablet ecen: ef :::::: he :::t u:k:cea within 1 hour Wh:- ::?cm tM pep!m te OPEP"LE stat = :-- close
/" the associated b'ock va've(s) with power maintained to the block 4
i Cad lve(s); otherwise be in at least HOT STAN05Y within 6 hours and l I'; Owsm r.*-o e,}in T SHUL'DOWN within the followina 6 hours. g % .n o rou % ,u,,w% a c,,a,, m ss sw'3 < ma ar. h on RV inoperable'" - "' W 9FtMn3r~m ~:::n
' ' u ther re: tere-th: "0RY to-OPERABLE--stat
" ?
thinIhourdTockvalveandremovepowerfromtheblo associate ~
"I r valve; ri restore the PORV to OPERABLE status withiner ?:r:= n Y cond D 72 hours or be in NOT STANDBY within the next 6 hours and < n HOT thin tne followingf ha ws w w <. s w s. .,
manwouy arrau,umahcativ e Uckd
- a' Cond E A , w Mi@ RVs inoperable M- - "e cr=r rnerewmacte --n
, within 1 hour Mthr re;t.a ;t 4&G Ar.e "0"V t: 0"E."^ar I
7rrlatus or close its associated block valve and remove 3 the block valve and be in HOT STANDBY within the next 6 hours and-in HOT SHUTDOWN within the following 6 hours. p CAC/F t. fWith one or more block valves inoperable, within 1 hourL ste m tM C k
/ manual control;. block-valve (s) to-OPERABLt-status- #) place its associated Vp i Restore at least one block valve to OPERABLE status 3m within the next hour if both block valves ire inoperable; restore "2 '
any remaining inoperable block valve to OPERABLE status within 72 Y' Comi c,- hours; otherwise, be in at least H0i STANDBY within the next 6 nours y and HOT SHUTDOWN within the following 6 hours.
%%g g. The provisions of Specification 3.0.4 are not applicable. ;
SURVEILLANCE REQUIREMENTS a v [ OAg , l' . 4. 8.1 ~ 2 3'X'er, te th re ' ' N '~ Spin.ifi s tk,i,4.0.0,)each PORV shall be demonstratiiid'0PERABLElt Teast once per 18 adhthi by:
- N N
.V m .o.q a. Performance of a CHANNEL CALIBRATION of the actuation instrumentation, and 2
,v
.5R 3.4.tr 3 Jr. Operating solenoid air control and check valves on associated air accumulators in the PORY control system through one complete cycle of full travel, and M 3 4.11.2. A coneratina the v lve through one complete cycle of full travel ahreifvaj g((2, ws>-TC ^ i O 'r, Ins. , t 3 4-:: e !
d.4.4.b Each block-valve shall be demonstrated OPERABLE at least once per - 92 davs_by_0Rttra.tiDg3ht_Nalve throuch one complete _ cycle of_ fulLtravel/unless i the block valve is closed f th ::=7 7 wMin order to meet the requirements / O$ (ofACTION_b.or_c._.ofSpecification.3.4.4.i
$R 34 ti i NcTii '
l BYRON - UNITS 1 & 2 3/4.:4-12 AMENDMENT NO. H . 44 hJd
V t CTS INSERT (S) T..N SECTION 3.4' L .}d L t ~ LC0 3.4.11-D . .- . INSERT 3.4 12C._-(A3 )
<bi ?With one PORVl inoperable because'of automatic actuation circuitry failure.
within 1 hourLeither restore the PORV to.0PERABLE. status-or close r.e-associated blockxvalve and maintain-power to the block. valve: restore the. i PORV'to OPERABLEsstat'us.within the following 72-hours or be in HOT STANDBY f within 6 hours and in HOT. SHUTDOWN within the-following 6 hours. 3 ~ Lc. With both:PORVs.' inoperable because of automatic. actuation circuitry-failure, within:1 hour either restore at least one PORV to OPERABLE status
~
,or close'its associated block valve and maintain power to the block valve i
H and be in' HOT STANDBY within the next 6 hours and in HOT SHUTDOWN-within
-the-following.6 hours.
-1 .:
LINSERT 3.412D -(A3 ) ,-t M I 1 - e, c so'rLautbmatic actuation circuitry failure
%/ 4 I ' &l Y
_d-
- f- v i
4 i-f Y 7;m . 2- , 9/1/98 Revision K t-p . y :~ a m: l 3 . l
SURVEILLANLt MtVVIMLMtM R LWm-' f 1) All tubes that previously had ' detectable tube e:all penetrations
- greater than 20 percent thtt have not been plugged or sleeved in the affected area, and all tubes that previously had detectable sleeve wall penetrations that have not been plugged, 6,. .
. 3
%. x .-
(} V 2) Tubes in those areas where experience has indicated potential problems,
- 3) A tube inspection (pursuant to Specification 4.4.5.4a.8) shall be performed on each selected tube. If any selected tube coes not permit the passage of the eddy current probe for a tube inspection, '
j
= ~ this shall be recorded and an adjacent tube shall be selected and A<WWA subjected to a tube inspection, WWCo For Westinghoese Model D4 steam generators, indications left in W 4e 4) service as a result of application of the tube support plate g to voltage-based repair criteria shall be inspected by bobbin coil prone during all future refueling outages, and
- 5) For Westinghouse Model D4 steam generators, tubes which remain in l service due to the application of the F criteria will be ispected, in the tubesheet region, during all future outages. .
- c. The tubes selected as the second and third samples (if required by Table .
4.4-2) during each inservice inspection may be subjected to a partial i tube inspection provided: 1) The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were - previously found, and ;
- 2) The inspections include those portions of the tubes where imperfec-tions were previously found.
, d. For Unit 1, through Cycle 8, implementation of the steam generator tube / tube support plate repair criteria requires a 100 percent bobbin coil inspection for hot-leg and cold-leg tube support plate
! intersections down to the lowest cold-leg tube support plate with known outside diameter stress corrosion cracking (ODSCC) indications. The determination of the lowest cold-leg tube support plate intersections having 005CC indications shall be based on the performance of at least a 20 percent random sampling of tubes inspected over their full length,
- e. For Westinghouse Model D4 and D5 steam generators, a random sample of at least 20% of the total number of laser welded sleeves and at least 20%
of the total number of TIG welded sleeves installed shall be inspected for axial and circumferential indications at the end of each cycle. In the event that an imperfection exceeding the repair limit is detected, an additional 20% of the unsampled sleeves shall be inspected, and if an imperfection exceeding the repair limit is detected in the second sample, all remaining sleeves shall be inspected. These inservice inspections will include the entire sleeve, the tube at the heat treated area, and the tube to sleeve joints. The inservice inspection for the sleeves is required on all types of sleeves installed in the Byron and Braidwood Steam Generators to demonstrate acceptable structural integrity.
-l (p) Y' 3/4 4-14 AMENDMENT NO.101 g BY'RON - UNITS 1 & 2
~ ~ _ .j PM \'-
REArTOR COOLANT SYSTEM A & mti subaT.o.) / g g g,g-
% Ooc O Snk 5eJ
--= -
1- SURVEILLANCE REOUTREMENTS (Continued) -, i 1 The results of each ' sample inspection shall be classified into one of the (-~ following three categories: Category Inseettien Results I i C-1 Less than 5% of the total tubes inspected are degradec 1 l tubes and none of the inspected tubes are defective. ! 1 i 1- C-2 One or more tubes, but not more than 1% of the total j tubes inspected are defective,.or between 5% and 10% of the total tubes inspected are degraded tubes. { .q
- 'C-3 More than 10% of the total tubes inspected are O
degraded tubes or more than 1% of the inspected l' tubes are defective.
- Note: In all inspections, previously degraded tubes or sleeves must exhibit significant'(greater than 10% of wall thickness)
,: further wall penetrations to be included in the above i-I ' percentage calculations.
i) 4.4.5.3 Insnection Frecuencies - The above required inservice inspections of !. steam generator tubes shall be performed at the following frequencies: c
- a. The first inservice ins)ection shall be performed after 6 Effective Full i-
' Power Months but within 24 calendar months of initial criticality or initial operation following a steam generator raplacement. Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If two consecutive inspections, not including the preservice inspection, i result in all inspection results falling into the C-1 category or:if two 1
consecutive inspections demonstrate that previously observed degradation L.
--has not continued and no additional degradation has occurred, the i inspection interval may be extended to a maximum of once per 40 months; i
- b. If the results of the inservice inspection of a steam generator
!L n conducted in accordance with Table 4.4-2 at 40-month intervals fall in , Category C-3, the ' inspection frequency shall be increased to at least once per 20 months. 'The increase in inspection frequency shall apply U l until the subsequent inspections satisfy the criteria of Specification l-
~4.4.5.3a.; the interval may then be extended to a maximum of once per 40 months; and
~
I L c. Additional, unscheduled inservice inspections shall be performed on each l
'} steam generator in accordance with the first sample inspection specified g
in Table 4.4-2 during the shutdown subsequent to any of the following }' conditions:- 1)- Reactor-to-secondary tube leaks (not including leaks originating L 3 from tube-to-tube sheet welds) in excess of the limits of
! g Specification 3.4.6.2c., or j _--
i-g i BYRON UNITS 1 & 2 3/4 4-15
%* J J
k f.* d b
-. . . - - . . - - ..- . - - . - . ~
. adEVEY1.fJJter REOUTREMENTS fcentinued4
[- 2) A coissie oscurr:nes greater then the Operating Daois Earthquaks, i Sc.c4m su l or f -
%. L A.~ e o i A condition IV loss-of-coolant accident requiring actuatien cf the O 3)
Engineered safety Features, or
- 4) 'A condition IV main steam line or feedwater line break.
s 4 4.4.5.4 heeentanea criteria
- a. As used in this specifications k- 1) Innerfection means an exception to the dimensions, finish or contour of a tube or. sleeve from that required by fabrication l drawings or specifications. Eddy-current testing indications l below 20% of the nominal tube or sleeve wall thickness, if l detectable, may be considered as imperfections;
- 2) Deeradation means a service-induced cracking, wastage, wear or !
l general corrosion occurring on either inside or outside of a tube
,_ - ; .rj sleeve;
- 3) Decraded Tube means a tube or sleeve containing unrepaired 2 imperfections greater than or equal to 20% of the nominal tube
. l Gr N'M#DOCb or sleeve wall thickness caused by degradation; 4 I feb C.o 4) e neeradation means the percentage of the tube or sleeve wall i thickness affected or removed by degradation; I!
1
.I tl
- 5) Etgigga means an imperfectio;. of such severity that it exceeds ii i
3 the plugging or repair limit. A tube or sleeve containing an unrepaired defect is defective;
- 6) Pluomine er Renair Lielt means the imperfection depth at or beyond j
which the tube shall be removed from service by plugging or j j repaired by sleeving in the affected area. The plugging or repair ;
' limit imperfection depth for tubing is equal to 40% of the nominal !
I' . wall thickness. For Westinghouse Model D4 and D5 steam generators, ) the plugging or repair limit' imperfection depth for laser welded j it - sleeves is equal to 40% of the nominal sleeve wall thickness, and ! for TIo welded sleeves is equal to 32% of the nominal sleeve wall j [ thickness. For Westinghouse Model D4 steam generators, this l l' definition does not apply to defects in the tubesheet that meet the { ' criteria for an F* tube; l
, 1 For Unit 1, through cycle 8, this definition /does not apply to tube j j
support plate intersections for which the voltage-based plugging
- criteria are being applied. Refer to 4.4.5.4.a.11 for the repair limit applicable to these intersections;
- 7) unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affoct its structural integrity in the event of an operating Basis Earthquake, a loss-of-coolant accident, or a . steam line or feedwater line break as spiecified in 4.4.5.3c., above;
,,. m -
Q BYROM - UNITS 1 & 2 3/4 4-16 AMENDMENT No.101 g 0 1 n .v t
~~ 5UMV L I L Lmm t- m vw msaw m . _ _ _ _ ___ ~~
8) Tube Insnection means an irispection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend
'to the top support of the cold leg. For a tube that has been repaired by sleeving, the tube inspection shall include the sleeved portion of the tube, and G.m p 4. G .h .e s Preservice insoection means an inspection of the full length of g) each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing. This inspection shall be perfonned prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
10) Tube Renair refers to a process that reestablishes. tube serviceability. . Acceptable tube . repairs for Westinghouse Model D4 or D5 steam generators will be performed by the following j processes: / a) Laser welded sleeving as described in a Westinghouse Technical Report currently approved by the NRC, subject to the limitations and restrictions as noted by the NRC staff, or
/ di TIG welded sleeving as described in ABB Combustion Engineering AN'd ,* Inc. Technical Reports: Licensing Report CEN-621-P, gg to ' Revision 00, "Comonwealth Edison Byron and Braidwood Unit I and 2 Steam Generators Tube Repair Using Leak Tight Sleeves, Ac DOC
- FINAL REPORT," April 1995, and Licensing Report CEN-627-P, Revision 00-P, " Verification of the Installation Process and 4 g' p jso Operating Performance of the ABB CEN0 Steam Generator Tube Sleeve for Use at Commonwealth Edison Byron and Braidwood Units 1 and 2,* January 1996, subject to the limitations and p restrictions as noted by the NRC Staff.
Tube repair includes the removal of plugs that were previously installed as a corrective or preventative meesure. . A tube inspection per 4.4.5.4.a.8 is required prior to returning previously plugged tubes to service.
- 11) For Unit I through Cycle 8, the Tube Sunnort Plate Plucoina limit is used for the disposition of an alloy 600 steam generator tube for continued service that is. experiencing predominantly axially oriented outer diameter stress corrosion cracking confined within the-thickness of the tube support plates. At tube support plate intersections, the plugging (repair) limit is based on maintaining steam generator tube serviceability as described below:.
a) Steam generator tubes, with degradation attributed to outside diameter stress corrosion cracking within the bounds of the cold-leg tube support plate with bobbin voltages less than or equal to the lower voltage repair limit (Note 1) .will be-allowed to remain in service. Steam generator tubes, with degradation attributed to outside diameter stress corrosion cracking within the, bounds of the hot-leg tube support plate with bobbin voltages less than or equal to 3.0 volts will be allowed to remain in service. O . AMENDMENT NO.101 a BYRON - UNITS 1 & 2 3/4 4-17 5 Rev t
'~ -
suwyt i u mn.t ne uu m t nt n i s n.m u "www' i Vaa ,
$rk 3,%
1.0+NDE+Gr( *) g (g,, g,7 9
- wn
) { Cl-4 *)
k1d*'d b y"M, y"M_gyM . CL L % 6,o fn ~OCL Where: 4W
- 6.o V,t . = upper voltage repair limit V ,t . = lower voltage repair limit Vt t = mid-cycle upper _ voltage repair limit based on time into cycle -
Vt = nid-cycle lower voltage repair limit based on V ot and time into cycle At = length of time since last scheduled insoection during and V tat wm which V.,,d. implemente CL = cycle length '(the time between two scheduled steam generator inspections)
= . structural limit voltage V,t Gr = average growth rate per cycle length .
/ NDE = 95-percent cumulative probability i f_
//
allowance for nondestructive examination uncertainty (i.e., a value of 20 percent has been approved by NRC)
' Implementation of these mid-cycle repair limits should follow the same approach as in TS 4.4.5.4.a.11.a. ;
4.4.5.4.a.11.b, 4.4.5.4.a.11.c and 4.4.5.4.a.11.d. l 4 l Note 1: The lower voltage repair limit is 1.0 volt for indications ; ! of outside. diameter stress corrosion cracking occurring at cold-leg tube support plate intersections. 7 ~ Note'2: The upper voltage repair limit for inoications of outside dianter stress corrosion cracking occurring at cold-leg tube l j suppart plate intersections is calculated according to the
~
methodology in Generic Letter 95-05 as supplemented. f ~ 12) F* Distance is the distance into the tubesheet of a
- Westinghouse Model 04 steam generator from the secondary face
-of the tubesheet or the top of the last hardroll, whichever is further into the tubesheet, that has been determined to be 1.7
- inches. j- 13) F* Tube is a Westinghouse Model D4 steam generator tube with ~ degradation below the F* distance and has no indications of degradation (i.e., no indication of cracking) within the F*
,o distance. Defects contained in an F* tube are not dependant -
2 on flaw geometry. [Q
' BYRON - UNITS 1 & 2 3/4 4-17b AMENDMENT NO. 1 f d
- 9N \l
vivwu 1 SUMTLILLmMLE MLuv a MLr1Ln i a u hve. 6
) 5)' If the calculated conditicnal burst probability based on the i projected end-of-cycle (or if not practical, using the actual f measur::d cnd-of-cycle) voltage distribution exceeds 1 x 10~3, >
notify the NRC and provide an' assessment of the safety h significance of the occurrence. (
- 6) Following-a steam generator internals inspection, if indications detrimental .to the integrity of the load path necessary to support the 3.0 volt IPC are found, notify the NRC and provide an assessment of the safety significance of the occurrence.
- e. .The results of inspections of. Westinghouse Model D4 steam generators F* Tubes shall be reported to the Comission prior to the resumption ~ of plant operation. The report shall include:
- 1) Identification of F* Tubes, and
- 2) Location'and size.of the degradation. i j
f' Au a me.o frL.w M 6 coc o w n O e 4 4 e f - mg
-V. I AMENDMENT NO.101 ,3
. BYRON - UNITS 1.& 2 3/4 4-17d e
hv K
.. . . . .. . _ . - . . . - ~ - . - . . . . - . . . . - . -
I l
. , _ . CTS INSERT (S) y' '/t . SECTION 3.4 LC0 3.4.15 INSERT 3.'4 20C 2(L5 L 22 .. and L24 )
. CONDITION' REQUIRED ACTION COMPLETION TIME-A. .. A.1 NOTE- -
7[ ~- Not required-to be performed until l 12 hours after g establishment of steady state w.- i l operation. l l Perform SR 3.4.13.1. Once per 24 hours AND
' A. 2 -
.7 , ... ..
Q , B. ... B.1.1 ... ... AND N B.1.2 NOTE l
? Not required to be M performed until
+ 12 hours after N establishment of steady state
- l. operation.
l ('- l .. l l 'l*- NJ 10/1/98 Revision K p 1i'
l Lm 2 4 11 i LC O .34.1-4
- ,, . .. .34 REACTOR COOLANT SYSTEM ( Ats')
~ d- *~'e e '
, .3 4.13 IONAL LEAKAGE-
.- i l LIMITING CONDITION FOR' OPERATION Reactor Coolant- System leakage shall be limited to: j i i Lco J 4 la a hr No PRESSURE BOUNDARY LEAKAGE,
)
ico S A.ll.b F. I gpm' UNIDENTIFIED LEAKAGE,' Leo 3A,13.,1 h. 600 gallons per day total reactor-to-secondary l leakage through all l .i steam generatorsfr.et b; ht;d fr;; the R :ctcr Coc hnt Sy:te ad-
' lCo 3 AM t- '
g50 gallons per day through any one steam generator _,) b,74g,Qg] LCO .3 413 c.h: ' 10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System, D Y ' "W
- r. e \
< 40gpmCONTROLLEDLEAKAGEataReactorCoolantSystempressureo 2235 i 20 psig, and- ---
.5R 3 4.14.'l A. leakageataReactorCoolantSystempressureof2235f20/psjg
. rom any React Coolant System Pressure Isolation Valve (sper c ified',
I' gvTam an.
- f 0.5gpm per nom *ual #uk of volve .ve. L9 hnmn.mv.MSyv)
APPLICABILITY: . MODES 1, 2, 3, and 4. 7_ ACTION:
- Lc.o,3 4 IS Co,a ti . m With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.
Coa A .b'. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage from Reactor' Coolant System Pressure Isolation Valves, reduce _t_he leaka e rate to within limits within 4 hourslor be in at least HOT STANDB
-~- -
_ eg g (Oiin th'e riest 6 hours and in COLD SHUTDOWN within the followina 0 hours.I: L, ?g [ .-} 1NSEPI .3.A- 2JA I'CO 2 A.14 5 e.- With any. Reactor Coolan System Pressure Isolation Valve leakage Mg
' C* d A- greater than thes above limit,E' === =+ ====g= =tr er - m"
; i:it; _; m , , , - ~
g
,ur4 or be in at .least~ HOT STANDBY within the nex Cmt C, 14--/b tours anc in C. .D 5_HLTDOWN within the followino 30 hours. (
' Acm Nous HIuSER.T 3.4- 21B [--Q
.-C w l 13 e d 2 h5CRT .3 4 llc $ g, 1
if 'NsDressGreAb$;ib[a h k g but greater than- a owed. !f p Observed leakage shall ba adj est_ s pressure up to 2235 psig l dc assuming the le knaa t: L uirectly proporational to pressure-dHfer-aHal to_/
#-' :L-iia power. /
BYRON . UNITS 1l& 2 3/4 4-21 AMENDMENT NO. 6'7 I b Y,.
,- - g.- --w,er .-4,...w ,m
CTS INSERT (S) i f~T .SECTION 3.4 J LC0 3.4.14 INSERT 3.4 21A . (L 7) .
-CONDITION . REQUIRED ACTION COMPLETION TIME l A. .One or more flow paths NOTE with leakage'from one' Each valve used to satisfy-or more RCS PIVs'.not ' Required Action A.'1 and
~w ithin limit. Required Action A.2 must have been verified to meet .
SR 3.4.14.1'and be in the reactor coolant 3ressure boundary or the.ligh pressure portion of the system. ' A.1 ' Isolate the high 4 hours - pressure portion of
<s, the affected system 1 -
from the low pressure L'd portion by use of one ; closed manual, ; de-energized power operated, de-activated automatic, or check valve. - AND i A.2 Isolate the high 72 hours pressure portion of. the affected system from the low pressure
@ portion by use of a
- - second closed manual, i9 de-energized power M operated, w de-activated
. it automatic, or check V valve.
l
, ,\.
10/1/98 Revision K ; r
i
,~ CTS INSERT (S)
/ F SECTION 3.4
.-p LCO 3.4.14
~
u INSERT 3.4 218 (Ag)- - r' A DNF-
. NOTES-- -- - ---
n 1. -Separate Condition' entry is allowed for 'each flow path. 2 .- Enter a)plicable Conditions and Required ^ Actions for systems made. inopera ale ~ by an -inoperr. ole. PIV. . INSERT 3.4 21C i (Mi): 1 CONDITION REQUIRED ACTION COMPLETION TIME ,:;j/ ; .. . ct B. Residual Heat Removal. B.1- ' Isolate the affected 4 hours
- N -
,. (RHR)oSystem suction- - flow path.by=use of:
. isolation valve one de-energized, Q power operated valve.
tr Dinterlock function-inoperable.
; f
F H Af 8/22/98 Revision K' C r a , e- - -g, % ---
, ..m-w *r>-
LCO 3. '!. S t,cc 2.- ' L'
.- 1 REACTOR COOLANT SYSTEM L c *- 2 2 ,
i
~
s L SURVEILLANCE REOUIREMENTS
\
9 3 A. r3.' N.:.;.E.II Reactor Coolant System leakages shairbrdemonstrated to be within each of the above limits by:
=:. ..;..it;rir.; t5: ceate;..-.,t et ey;,. = ya..... ...; p;.ru: 1:t:L'a
- redi:::tivity reaiter et 10 :t 03:0 ; r !? 5:0r:; J d m
I
't. Me..it: ring the r;;;t:r ::vity :::p discherg;, :nd th; centeir.;;Et ^
-ficer decir. : 7 diseherge end i.....te., et 1:::t e- e pe- !? heyes:
l Ik&essehc. Measurementc;theCONTROLLEDLEAKAGEtothereactorcoolantpump~j l l
/ g sg,;wpI seals when the Reactor Coolant System pressure is 2235 i 20 least once per 31 days with the modulating valve fully open.psig The at '
1 l l
' 3" M *,
l f provisions of Specification 4.0.4 are not applicable for entry into ,'
's 6 " b 3-{ j (MODE 3 or 4; J _
Lp {3MSEL" 3 & DF} ,
- f. Performance of a Reactor Coo ant System water inventory balance' at least once per 72 hours; and (L%
- e. "::itT % ; th T P.::.;ter ;;e.u F'ange Lea u fi 3ysi.m al lee 31 ea;; p;r
-2t h:ure.)
L(o 3.'J. ll! g"
'?.4.0.2.0 Each Reactor Coolant System Pressure Isolation Valve :p::if t:d in ., h,
, T:ble 0.?-1; shall be demonstrated OPERABLE]by feritying leakage to be within" i fitslimit:j WM a.
4 ft 1:::t ;;;; ;;d(18 months](hlauedana WNh se ISTNoyrow,G
'b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 7 days or more and if leakage testing has not been performed in the previous 9 months, lk l
t n-icr te retura%; the vme t: : rvice f:!!: win;; :: int:n:n;er)-.k L ae 9 er r::1:c;;;;t w;rk :n th: v:17:. ::# -
- d. Within 24 hours following valve actuation due to automatic or manual action or flow through the valve)except for valves RH 8701 A and B j R eMe 3 e--{and RH 8702 A and B.]
The pr;visicas of Specific; tic '.0.' are :t ;;110:51 fer ?"tP" i"te MOE 'O or-4-I ' \ g 9p,1 (Bsw 2A-& A,s sr.13&e 9 f1nstLT 3A-nQ L is-l BYRON - UNITS 1 & 2 3/4 4-22 AMENDMENT NO. 82 Rev d l
L 6 o I i
- I-
- ;-9 '
~ CTS-INSERT (S)
SECTION 3,4 3.) . LCO 3'4.13 . i y " INSERT 314 22A: (L .3 ) l l
~
l >
. SURVEILLANCE'- FREQUENCY f
4:SRo3.4.13.1 .. . NOTE--- , A .. Not. required to 'oe. performed until .12 hours H after establishment of steady state-Qll j operation.
- u Verify RCS ~ operational-LEAKAGE is within 72 hours
~ 1': '
limits by aerformance of RCS water inventory ]alance. n .; V k l 1 l. i ..
< 4
-ep
- k/
8/22/98 Revision K
- i n ,s--r e ,. .- e n ,m.. , . -w-e , ,- ,e:-, c,, s a , .., -
LCo 14.tto
~
3,4 AEAeron COoz.Arr sysrEx (RoS) 3,q,l(p;;;..= SPECIFIC ACTIVITY LIMITING CONDf?f0N FOR OPERATION 14.16 M The specific-activity of the reactor coolant shall ne limited tot
^gg g,q,g,ge, Less than or. equal to 1 microcurie per gram DOSE EQUIVALENT I-131**,
and M 3 4/,l(a.l-h.c lLess than or equal to 200/5 microCuries per gram of gross radioactivity. APPLICKEILITY: MODES 1, 2,[2, t. ...; : ) h6 MDD E ?. wl1% KCb M tr appl. Tt rY.F;O* ((C 2 5 OD'F ) ACTION: , 1 MODES 1,.2 and 3** ,. g,,g g g ,,g gg,3,g g ggg g CodD A *,. With the specific activity of the reactor cc31 ant greater than 1 _ gA Q,L ' 1 microcurie per gram DOSE EQUIVALENT I-131** for_more than 48 hours
- ' - during one continuous tiJoe interval /or exceeding the limit line l shown on F1~g~ure 33-17be "fii at Toast NOT STAND 3Y with .T less l,' OtJ0 6 than 500*F within 6 hours; and y l 1
ter With the specific activity of the reactor coolant greater than 100/E
- CooJO C., microcuries per gram, be in at least NOT STAND 8Y with Tg less than 500 F within 6 hours.
'O t
l l l l l l l
- i LLO 3.L{,lb hpflICObcl&
~
l *With T g greater than or equal to 500*F. 2.
.f
\ 1
**Foy Unit 1 thrfugh cycle 8, p actor coolant 40SE EQUIVALEN7 1-131 w b( limited tqf0.35 microcurfes par gram.f BYRON - UNITS 1 & 2 Rev E 3/4 4-27 AMENDMENT NO. 101
LCD 3.'/. Ho i-REACTOR COOLANT SYSTEM , LIMITING COND! TION FOR OPERATION 4 ACTION fContinuedi kDf)C 3 with AC.6 GHra4L \uuperabH 5.600*f Ls 3 MODES 1, 2,I2- ' - ' i) - O m A ,1,3 3, ,,,,1,1, ,,,1,1,7 ,, 3, ,,,,,,, ,,,1,,, ,,,,,,, ,3,, ; 4A S.l - ,1 microcurie per gram DOSE EQUIVALENT I-131* i; ;_;;;;; C.e., , 7-a-- - -*-- --- 100/0l !~
----2 perform ~the sampling and analysis requirements of
} Item 4.a) of Table 4.4-4 until the specific activity of the reactor coolant is restored to within its limits. .
.SURVETLLANCE REOUTREMENTS
@ The specific activity of the reactor coolant shall be determined to be within the limits by performance of the. sampling and analysis program of Table 4.4-4.
SR 54.IG.I J m SR 3M. IG..Z. i
} R 3 A.)&.3 l
l-
-f- aro Unit I thro gh cycle 8, r *
.. d 1 ited to C. microcuries rctorcoolantDO[EQUIVALENTI-13[w111*be gram. f i BYRON - UNITS 1 & 2
?.E 9 E 3/4 4-28 AMENDMENT NO. 101 l
1 < \ t. LCO 3W.Ho ,
. -- \
i i t. a I t ~
- E i a n D 16 O s 3 '
t l
- s 250 .
! E~ 1 - - i eml 8 'k im s 1 E
$200 ",
2 u i B: s ' UNACCEPTABLE j
@ ', OPERATION i A k gt) u
! p i
- z 150 (-
i 5 , f o u t- p i ! g k l < L [ a g 100 g i s A a 4 E t ACCEPTABLE i 7 OPERATION t b ' i^ .z ! m 50 l 4
- D 1 0
- W 4
m i sn V ' O O O ._ j 20 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMAL POWER ) FIGURE P.' (3. 4 . lt. - t ) L r__...___......__..., /l,
- DOSE EQUfVALENT l.131 REACTOR COOLANT SPECFIC ACTMTY U5IT VERSUS PERCENT l dt, OF b RATED THERMAL POWEC'J;7;; TriC F.2ACiG C-:-- -Ri erasii n. Asiivn , o ps ~ w
~
} ; emJr Gm. ;. a ; 3 l
- j. ..
f j'
.9 d j BYRON . UNITS 1 & 2 3/4429 AMENDMENT No.101
i LLO BM,W I i i d i
' I bL l_ / h
' ,s n , 1 E ' '
l en > , 3 s
/
S250 1 ,
~
i r , , E ;
/
- s ,
4
- s . .
o 200 li ,
4 g UNACCEPJ BLE I <
g _ OPERATION m u - I &
- tn ,
I H {, z 150 4 i i l 4 - O i j O < ! O 6 y
; y n -
- ; E 1 ,
t 1 i 4 'L i . E 100 s e E , i E i t y L n
.:. A i p 'k '
l g/ -
" /
l ! l2j' 50 , k 1 t;c n '
- h. i# h [ ACCEPTABLE k -
l {mwO - OPERATION a , O i
/
j 20 30 50 60 70 80 90 100 [ 40 j i. e/ PERCENTFIGURE OF RATED 3.4-2 THERMAL POWER
/
UNIT 1 T!!ROUOM CYCLE 8f j ' DOSE EQUIVALENT l 131 REACTOR COOLANT SPEQlFIC ACTIVITY LIMIT VERSUS
; ' \ PEpCENT OF RATED THERMAL POWER _WITH THEftEACTOR COOLANT SPECIFIC
'O 'C ACTIVITY >0.35 pC1/ GRAM DOSE EOUIVALENT l 131 ,/
\a l
@ c K.
] . BYRON - UNITS 1 & 2 3/4 4-29a AMENDMENT NO.101 a i I.
4 +g *'
&. Q '
h TABLE 4.4-4 REACTORCOOLANTSPECIFICACTIVITYSAMPLE
-AND ANALYSIS PROGRAM . TYPE OF MEASUREMENT SAMPLE AND ANALYSIS' MODES IN WHICH SAMPLE AND ANALYSIS -FRE0VENCY AND ANALYSIS REQUIRED
- 1. Gross Radioactivity Determination **-
SK J.'I.f4.le-(At least once per(N 1, 2, G h - 7 SR 3.*l. L.2 - 3
- 2. Isotopic Analysis for DOSE EQUIVA- Once per 14~ days. l~
LENT I-131 Concentration S R F,4,/4.5
- 3. Radiochemical for E Determination *** Once per 6 months *- 1
- 4. Isotopic Analysis for Iodine .470nceper.4 hours, if, 2f, 3K. ";. 5;' Lso Including I-131, I-133, and I-135 whenever.the specific.
activity exceed C Cond A R4 4*I [sact/ gram DOSE -+ 6R 3,4, % 2 a (EQUIVALENT I-131****j l (erlee/E;O/gr: o a: ef crete redte::tivit;.; 7 and Lyy jqr
' b70ne sample between 2 1, 2, ? , gt and 6 hours following eg, a THERMAL POWER change L ;
SR 5.4,lG, A -: ; exceeding 15 percent
! of the RATED THERMAL ,
~
POWER within a 1-hour uperiod. w: n N! k 8 6
*i :
6 RyanN IINTTS 1 A7 3 /4 4-3n AMFNnMFNT NO 77
L.co 3 Al.! G
~
Oc Rnikms /./ TABLE 4.4-4 (Continued) TABLE NOTATIONS f
\
MD A # Until the specific activity of the Reactor Coolant System is restored RA A l within its limits. _ _
- Sample to be takenn(3tdass fter a minimum Y of 2 EFPD and 20 days of POWER FR 34.lu.3 OPERATION have elapsed since reactor was last subcritical for NOT 48 hours or longer.
gross radioactivity analysis shall-consist of the quantita tve ' - me nt of the total specific activity of the reac coolant' ' except fo nuclides with half-lives less minutes and all radiciodines. I specific activ all be the sum of the ,, degassed beta-gamma activ otal of all identified gaseous activities in the sample n after the sample is taken and extrapolated bac en the sample was Determination of the contributo the gross specific activity shal ed upon those en saks identifiable with a 95% confidence level. ist available data may be used for pure beta-emitting radionuclides.
*** A radiochemical analysis for E shall consist of the quantitative \
measurement of the specific activity for each radionuclide, except i for radionuclides with half-lives less than 10 minutes and all radio-iodines, which is identified in the reactor coolant. The specific activities for thsse individual radionuclides shall be used in the determination of E far the reactor coolant sample. Determination of the contributors to E shall be based upon these energy peaks .( . identifiable with a 95% confidence level. -- - d*'** For' U~ ~ ~' lht' rohgfi~ i8, reacto(coolant DOSJ40VIVALE47'I-13p l3 w - be limited 0.35 microcurfes per gram.,7 : 5
\
S e.e. 0 0 ( -f w S e c h*in 1. 0 , n .. 9ea V-- BYRON - UNITS 1 & 2 3/4 4-31 AMENDHENT NO.101
(f/0.4.3 PRESSURE / TEMPERATURE MITS Lw 3,4,3 REACTOR COOLANT SYSTEM iM b 'd# " b LIMITING CONDITION FOR OPERATION Lco 3 4,3 - 3.4.9.1 The Reactor Coolant., System. emperature andygstuteand heatu and cooldown rates shall be%aintained' In"ac'cordificElif th th'e ' limits spec le in, S W .. ,, w n _ w w ~ - ~ ~ ' APPLICABILITY: At all times.
"' 7
- M' ACT M he c.ex, A With any of the limits hthe PTLR^/exceeded, reslate the tem rature and/or o o A A.) pressure to within the_lla1Lwit)fn 30 minutes;Jperfo 'n a)6ginee, ring uAn 5
( evaluMon to deft rnine the effects of i.hpeut-of-li conditioff on the ine_that the Reactor RAA4 (stmetural Coolant System intecrity ofacceptable remains theinActor Coolint operation for continued Svstem(/determ*or lbe in at leas (3TA~NOBririth1Trthe next 6 hours and reduce the RCS T and pressure to less f occiS 7 than 200*F and 500 psig nrespectively,withinthefoUowing30 hours.f coco c. (TNurf 3.4- M A
. SURVEILLANCE RE0VIREMENTS gu%
4 . 0 . 0 .1.-i The Reactor Coolant System temperature and pressure shall be o u n determined _to .be_withjn_the._Ilmits..at_ leas.t_once..p.er_3Luinutesfduring system) e,h heatup, coofd'own, and inseryjce leak and hydrostatic testing operations. F I4.4.9.1.2 The reactor vessel material irradiation surveillance specimens (' shall be removed and examined, to determine changes in material properties,
/ d by 10 CFR Part 50, Appendix H, in accordance with the schedule in l
/t e PTL i 1
y-- l z AMeset/ W 6 Q co c. G. 9 h
- T 0
?
1 l O - u ( , AMENDMENT NO BYRON - UNITS 1 & 2 3/4 4-32 m ./ [ Rev K.
sPRESSURTZER , LTMT,TNG CONDTTTON FoD OPERATTON _/i (,/ - 344.9.2 Th pressurizer temperature shall be limited tot
- a. A max um heatup of 100*F in any 1-hour period,
- b. A maximum ooldown of 200*F in any 1-hour period, an4/
- c. A maximum spr water temperature differential of 0*F.
APPLICABILITY: At all time ACTION: With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within the h ,id within 30 minutes; perform an engineering evaluation to determine theM facts of the out-of-limit condition on the structural integrity of the p,r/ssur er; determine that the pressurizer remains acceptabte for continued o ration or in at least HOT STANDBY within the next 6-hours 'and reduce the ressurizer pre ure to less than 500 psig within the following 30 hour
/
/
/
/
S_URVEILLANCE REOUTREMENTS s ID () 4.4.9.2 fdhe pressurizer temperatures shall be determinsd to be within e limits /at least once per 30 minutes during system heatup or cooldown. The spray' v.et.ar temperature dif ferential shall be determined to be withis. the limit at least crace per 12 hours during auxiliary spray operation. _ [ \ f o f m
. v _ ,_._, x BYR6N - UNITS 1 & 2 3/4 4-33 AMENDMENT No. 98 P_c o R.
& OvERPREssoRE PROTzcTION SYSTEMS LCO F. 4 .12.
[3,4.11 Lw TEMf'3MATVRrJ Seecif4ca 6on 5~.fodo tTMITING CONDITION FOR OPERATION LLO 3All.c oimA 64 3.% t'2..'5 1 O 3,4 d.d(fi.Tif?.F) At-least two overpressure protection f/NsEs.T J.4-34A) hdevices shall and be OPERAB each device shall be either: a. A residual heat removal (RRR) suction relief valve with a lift setting of less than or equal to 450 peig, or
- b. A power operated relief valve (PORV) with a lift setpoint that varies with RCS temperature which does not exceed the limit established in the PTLR.
APPLICABILITY: MODES 4, 5, and 6 with the reactor vessel head on. ACTION: g +--(I kJ5E F T S A 3'l B) Cok)0 E g With one of the two required overpressure protection devices inoperable in MODE 4, restore two overpressure protection devices to (,oMD 4 OPERABLE status within 7 dayofor depressurize and vent the RCS ] (through at least a 2 square inch vent within the next 8 hours. )
- p. )s'. With one of the two required overpressure protection devices inoperable in MODES 5 or 6, restore two overpressure protection CONO G devices to OPERABLE status within 24 hours (or vent the RCS through]
.----.{at least a 2 square inch vent within the next a hours. j)
(ogDGf. With both of the required overpressure protection devices inoperable, y O depressurize and vent the RCS through at least a 2 square inch vent within 8 hours. E 6O LC0 3 Al'Z ja'. 3 1:t :t: n:: = .::: ;; 2 r r 1, 5, er 23 verify the vent pathway 3 gg 3, y .r2..Q at least once per 31 days when the pathway is provided by a valve (s) '*d that is locked, sealed, or otherwise secured in the open position; otherwise, verify the vent pathway every 12 hours. { n the event either the PORVs, RHR suction relief valves, p M vent ed to mitigate an RCS pressure transie t, .<r-Special Report shall be d and submitted t insion pursuant to specification 6.9.2 within . The report shall describe the circumstances initi e transion , atffect of the PORVs, RHR suction valves, or RCS vents on the tra51'enty .and any etive action necessary to prevent recurrence. gg3 g f'. The provisions of specification 3.0.4 are not applicable, g O BYRON - UNITS 1 & 2 3/4 4-34 AMENDMENT NO. 98 Reo K
l
' CTS INSERT (S) lj' 'y
~~
SECTION 3.4 w..s 4 ' LCO 3.4.12
; Gl _ INSERT 3.'4 34A' (M,)'
~ ~
l LC0 3.4.12 : An LTOP System shall be OPERABLE with:
-:<g
- a. ...
- b. ..
- c. 'Each SI accumulator. isolated. whose pressure is greater than or equal to.the maximum RCS' pressure for the
-existing RCS cold leg temperature allowed by the P/T limit curves provided in the PTLR. and
.s
.d.
i SURVEILLANCE FREQUENCY
.; e _ SR 3.4.12.3 .
. . -NOTE-1/ 1 Only' required for accumulator whose
_/ pressure is greater than or- equal-to the maximum RCS pressure for the existing RCS
. cold leg-temperature ~ allowed by the P/T limit curves'provided in the PTLR.
. Verify each' accumulator is isolated. 12 hours f
5 e
& t A. 10/5/98 Revision K
i j
- )
~ CTS INSERT (S) fmi SECTION 3.4 A_[ -
i LCO 3.4.12 ) i l INSERT 3.4 348. (M,) l l CONDITION' REQUIRED ACTION COMPLETION TIME l l A .- Two charging . A.1- - -NOTE : (centrifugal) pumps- .Two charging pumps
-c ' capable of injecting may be capable of i
.. .into-the RCS.- injecting into the s RCS during pump, swap
- i. QR operation for L s 15 minutes.
2 One charging (positive - displacement) pump capable of injecting Initiate action to Immediately . into'the RCS. verify a maximum of one charging pump l (centrifugal) is ' capable of injecting
/q c .into the RCS. }q) i B. One or more SI pumps B.' Initiate action to Immediately capable of injecting verify no.SI pumps into the RCS. are capable of 1 injecting into the RCS.
1 C. An accumulator not C.1 Isolate affected 1 hour isolated when the accumulator. accumulator 3ressure is greater t1an or equal to the maximum RCS pressure for' existing cold leg temperature allowed in the PTLR. f% ,O, 8/22/98 Revision K
~
. . ~.. . , . . - . - . .. . . ~ . . .- . - . . . - . . . - . - - . - . ~ . . - -
l 1 CTS' INSERT (S)- (X SECTION 3.4 R
,l J
< n. .
LC0 3.4.12 l JREERT 3.4 34B (M7 ) (continued) l
-i
., LCONDITION' ~ REQUIRED ACTION COMPLETION TIME e 2 lD. Required Action and D,l' Depressurize affected 12 hours
. associated Completion' accumulator to less 1 Time of Condition C than the maximum RCS not met. pressure.for existing l .. cold-leg tem)erature allowed in tie PTLR.
i . 8
- f,
.. 8/22/98 Revision K
. . . . .- ~ . . - _ - ~ - . . ~ - . - - . . . . - . . . . . . . . . . - . . . ~ . . - . - . . _ ~ . . - - - . - . .
i' SURVEII.I.ANCE REGUimm.sEnaa Lco 3.4,I'2. l h Bach PORV shall be demonstrated OPERAELE[when the PORVs are being)
. gggeg > Jused for cola overpressure protection)Dy:
- _(j Q gnJ 5E A.T 34 - 35 A } so gg 3.4.t'2..'1. a. 'f Performance of an ANALOG CHANNEL OPERATIONAL TE on the PORV actuation channel, but excluding valve operation, at least once per 31 days when the PORV is required OPERABLE; and
' SR 3.u . il.S b.
Performance of a CHANNEL CALIBRATION on the Poiv actuation channel at least once per 18 months; and
.g g 3,q,q,ge. Verifying the PORV isolation valve is open at least once per 72 hours.
' 5 R. 3N.12 5 ,
- y E t 9;?_$ Each RNR suction relief valve shall be demonstrated OPERABLE when yd tne RHR suction relief valves are being used for cold overpressure protection 9 as follows:
q m [. For RHR suction relief valve (RH8708Bjverify at least once per 72 hours that valves (RH8702A and RH87 M rs open. x LN5 j ! f.. For RNR suction relief valve (ItH8708A) verify at least once per 72 hours that valvesfRH8701A and RN8701 Blare open. 1 N YA$ sh him A A = $na A Aw hg.rus b[b5m Ambws4 koheb a BYRON - UNITS 1 & 2 3/4 4-35 AMENDMENT NO. 98 2.c.o K.
l.. CTS INSERT (S) SECTION 3.4 { 1 LC0 3.4.12-L INSERT 3.~4-35A .(Lg):
.'l K.
-SURVEILLANCE: FREQUENCY-p SR;3.4.12.7: NOTE.
Not.recuired'to be performed.untill12 hours !? .after cecreasing RCS cold leg temperature L. to s 350 F, L f.. i p :- 1; L l u. l. l l' l l-L i - o) 8/27/98 Revision K e L r f >:
. . . . - - . ~ . _ , _ _ . . . _ - - , . - - - - .__
. - . . . . - . - -.. . . - .- ~ _-.- .. - . -. - - ,
.i-ik.rumAt twTecmTTY o /4.4.10 k l LIM!TINCsCONDITION FOP OPERATION
~
'Q NN !
-Q 3.4.10 The st'ructural' integrity of AsME Code Class 1, 2, and 3 compon ts
- shall be maintained in accordance with specification 4.4 10. x , APPLICABILITY: All MODES. i ACTION:
\
f N
\.
a. - ~ With the structuralNintegrity of any ASME Code less 1 component (s) not conforming to the'above requirements, r ore the structural integrity of the affected component (s) to the affected component (s) priortoincrea/ithinitslimitorisolate sing the Reactor Coolant system temperature above 200*J.
- b. \/[
With the structural integrity pf.any ASME Code Class 2 component (s) i not conforming to the above requirements, restore the structural integrity of the affectedjcomponent (s to within its limit or isolate \ 1 t
- the afiseted componentfe) prior to inc(reasing the Reactor Coolant
-system temperature above 200*F.
/ \\
; c.
With the structural integrity of any ASME Co[Se, Class 3 component (s) not conforising to the above requirements, restore the structural
- integrity of'the affected component (s) to within its limit or isolate the affected component (s) from service.
i ,/ 'NN d. Thep'rovisionsofspecification3.0.4arenotapplicablek f j ( /
- supvEILtkNCE REOUIREMENTS
\
) .10 Inadditiontotherequirementsofspecification4.0.5,/eachreactor coolant pump flywheel shall be inspected as follows: ~-
- a. Volumetric examination of the areas of higher stress concentration at i
the bore and keyways will be performed each 40 month period during 1 j 4 refueling or maintenance shutdowns coinciding with the service j inspection schedule as required by section XI of the ASME Code. j
- b. Visual examination of all exposed surfaces will be performed and a i
surface examination of the bore and keyway surfaces will be performed whenever the flywieels are removed for maintenance purposes, but not more frequently than once each 10 year interval. }. fr m, , i-i fQm) w & W T.O Q %C Q< 94.h w C.O i BYRON - UNITS 1 & 2 3/4.4-36 AMENDMENT No. 98 Qe o K.
. .. _ . - . - - ~ , . - -- - ..--~~ - -- - --
- - - x --- -- - -- ~~
! 3/1 4.11 mnw - =T SYS*EM VENTE T fn k .s 3 M -
4 LIMITING TTION FOR OPERFTION I
.s A 1
/ .
- (,/ . .3.4.11 At les one reactor vessel head vent path consisting of ym/ e valves in '
1 series powered emergency busses shall be OPERABLE and clo,se8. i / i APPLICABILITY: MODE 1, 2, 3 and 4. i. OS i-
- Nith the above reactor essel head vent p'ath inoperable, STARTUP and/or j POWER OPERATION may cent ue provided the inoperable vent path is main- '
tained closed with power r from /he valve actuator of all the valves
- j. in the-inoperable vent paths store'the inoperable vent path to OPERABLE l
t status within 30 days, or, be i OT STANDBY within 6 hours and in COLD l l i SNUTDOWN within the fc11owing 30 urs. ; + SURVETLLANCE REOUIREMENTS /
,. e j
1 l
/
4.4.11 Each reactor pEssel head vent path shall be amonstrated OPERABLE at ( least once per 18 mo)the by:
,/ .
Verif[ingallmanualisolationvalvesineachvon
- a. th are locked in tho' open position,
- b. Cycling each valve in the vent path through at least one c late cycle of full travel from the control room during OOLD 5 a
O. REFUELING, and or { c. Verifying flow through'the reactor vessel head vent paths during
- - venting operations at COLD SNUTDOWN or REFUELING.
i __ f i I 4 f 4
+
i sYRON - UNITS 1 & 2. ' 3/4 4-37 AMENDMENT NO. 98 N e
.~. _ . . - .n ...-...,... . ,.. . . _.- ._ _ _ . - . . . _ . . _ - . . _ _ _ . . . _ _ _ _ _ _ _ . - _ _ _ _ _ . _ . _ . _ . ~ - _ _ . . _ . . . _ .
- f. , s I l
\! c s 1
' -4
/ e
- -- i Lz. , .
CTS INSERT (S): y, l -v. SECTION 3.4 o L. ? LC0 3.4.12 2 - (_ - . , , , . 01NSERT'3.4 39A.
, 2 Deleted ' in L Revision 'K' .
a . . V . r'
^:
+
!;n
.1
' 1
-1
; INSERT 3.4 39B- 4 1
Deleted in Revi.sion:K-4 INSERT 3'.4 39C
"' ~ -
Deleted :in. Revision K , :f t .- k {t f r i 1
-i
<- l I; 1 1-(
) . + , ;
h A J p l l' ?{/ ffv -: 8/22/98 Revision K l .- s ( 4 4 v4* c. w, r* -,, .,. ,.- .. =em ..-.-e- , ,, . .w-...
~
i- ;, CTS INSERT (S) SECTION 3.4- i l( '.
.l
.N .
)
LC0 3.4.12 -l l INSERT 3.4 41A' Deleted in' Revision K i l l 1 i '( 5 1 t. l l L en s 8/22/98 Revision K
~
- +
l.c o : .3.3 . I Z. ' A '>. 2 COI333 , ,M,W / ! EMERGENCYE CORE C00 LING' SYSTEMS-s < 43/4.5.3= ECCS SUBSYSTEMS -'T <.350*F' ; 'p
' ava
%h . .. .
fhMeewJe.Sen.35.
; LIMITING CONDITION'FOR' OPERATION. \ _% .D Or L , 9 4 ;. ,3 e d
- m. u - -
1 ' ,. ? 4 325.3 9 Asia minimum,E one ECCS subsystem comprised of. the following.shall be 1' 1 l OPERABLE *' '1 e >
.a. -One'0PERABLE;. centrifugal charging pump,"-
g, , '. 2 1
'b.: .One OPERABLE RHR heat' exchanger,-
l 1
; , 'c. -0'ne' OPERABLE RHR pump,'and -
- d. - An OPERABLE. flow ' path capable- of taking suction from the refueling i w,* ! water storage; tank upon being manually realigned and transferring
^
". Esuction to:the containment-sump'during the recirculation phase of f ~ operation.=
! APPLICABILITY:.TMODE14.
t ACTION: j
, 'a . Withino.ECCS subsystem OPERABLE because'of, the inoperability'of ,
,, . either the centrifuga1'chargirig pump or the flow path from the : E(^\ refueling water storage tank, restore at least one ECCS subsystem to i V ' OPERABLE-status within 1. hour ~or be in COLD SHUTDOWN within the ne'xt
- 20. h'o'urs.
l ib. LWith no ECCS subsystem.0PERABLE because of the i: W M11ty of ; ,a < j' either'the RHR heat exchanger or RHR pump, restore at least one ECCS subsystem to OPERABLE' status.or maintain the Reactor Coolant-System , T,yg less..than 350*F by use of' alternate' heat-removal methods. y c. In.the event:the ECCS is actuated and injects water into the Reactor j ' Coolant System,.a Special Report shall'be prepared and submitted to i
, the Commission within 90 days', pursuant to Specification 6.9.2, -}
describingthecircumsta'ncesoftheactuationandthetotalaccumulatedj Li actuation cycles to date. The current value of the us' age factor for ; each affected Safety Injection nozzle shall be provided in this ' l[!- , Special Report whenever its value exceeds 0.70. f V Lco _3 A.12..a. [ M maximum of one. charging pump-shall be OPERABLE]and that pump shall b N - ' f . TcentrTfdga'l~cTarging a pump,~w"lienever thi~timpieTature of one or more of the i g , YoTE .M RCS cold legs is less;than or equal: to 330*F. ,2 d c . . - paut u in
' C0907S inA f.1 M n
.sy
, BYRON - UNITS 1 & 2-
=
n 3/4 5-7 l
, k\) h
. J _- m __ _ __. , - _. _. -
, - . , ~ . . . . - . - , - . ~ . , . - - . . - . _ . . . . . ~ . . - - . - - . . - . - - . . - - . -
)
' ' CTS INSERT (S)
Y. - SECTION 3.4 ; ty)Y .J l LCO 3.4.12 INSERT 3.5 7A (L4) > . i @ #l,; u. d* . CONDITION .' . REQUIRED ACTION COMPLETION TIME: A. ... A.1- . NOTE l
-Two charging pumps )
may be capable of l injecting into the' l RCS during-pump swap )
' operation for - ,
-s 15 minutes.
I
.g.
AI. - 1 l t "1 Y: a L-4 / y% dd 9/18/98 Revision K V f 9-
34 l
,e -
BRWD CTS MARKUPS s 4 fl-y I 4 0 ~O 1 ) t t 4 4 s' V a
. . . . .. . . - _ - - . .. . - . . - . _ ~ . .
% 6^ on 3. \
^
- 3,st ROired caoHnr CMrd(es) b
;<,g g- Pt S MINIMUM TEMPERATURE FOR CRITICAL!'Y LIMITING' CONDITION FOR OPERATION Leo 3.4 2 : ~
&-t-i-+)
shall.be~ Thegreater Reactor thanCoolant System or equal to 55lowest operating loop temperature (Tavg)
*F.
' APPLICABILITY: MODES 1 and 2 d I ACTION:
/ Com 4 -' With a Re_act_or. Coolant System operating loop temperature (Tavg) less than 550*F, e -; : ._ ;.p y;M ig. i.31 .ig.;. vi.nfn n - n= : en be in %
avy .. iET^.""" ithin the next 15 minutes. . Mope 2. 4.tk 4(4 < t.o) . SURVEILLANCE REQUIREMENTS 6 A 'J.4. 2 1 - (+-t-t-43 The Reactor Coolant -System temperature .(Tavg)'shall be determined to-
. be greater than or equal to 550*F:
p. ie 1 Q SitETE T5 cindE:: pri r t: :thi:ving ::3 : 1r,itt::] ty gp V b. At least_ o.pg.p / 0 minut .when e tea tor is ritic .and'thei fR ctor Cdolant ystem T is le s than 557*F w th the T -T
~
9 . . . . . .
... .a,,vg ref) _
eviatio/n Ala
~
not res t. / I Q i- .M d
' 14 :5. .
L {ll h o u r.5 & y h :f _ 4 E tro '-
- l. Mk u v.u. n.4
#With K,ff greater than or equal to 1.
5 -
~ C h 0 55.=['.I5=i J M5 s.'u.ue Sv"ii kai'w" 3 10 3v
(. a BRAIDWOOD - UNITS 1 & 2 3/4 1-6 [ L l - ReJ N L-
__ _. . .. . . ~ . - . - . . . . - . - - - - - - - -- 33 uc o 3.41 l , POWER DISTRIBUTION LIMITS - f %ed w % s. ,,; - LIMITING CONDITION FOR OPERATION / W' Lh?~' I t ACTION'(Continued) ! , i l Within 24 hours of initially /being outside the above limits, verifyh l through (incore flux mapping and)RCS total flow rate comparison that th ombination of Fh and[RCS total flow rate are restored to within the..above limits, or reduce THERMAL POWER to less than 5% of l RATEDTHERMAt40WERwithinthenext2 hours;and;.. ~ l , s. ' \
- c. Identify and corfect the cause of the out-of-limit condition prior i to increasing THERMAb40WER above the reduced THERMAL POWER limit I required by ACTION a.2. 'and/or b. above'; subsecuent POWER OPERATION s
may proceed provided that th'e't:ostiination of F A and indicated RCS x AH - - total flow rate are demonstr'ated, through incore flux mappingjand i RCS total flow rate comp'a rison, to be w~ithin the region of acceptable operation defined by' Specification 3.2.3 p'rior to exceeding the foi-lowing THERMAL POWER levels: N s
/
K'N x
" 1. A- nosiinal 50% of RATED THERMAL POWER,
,o 2/A nominal 75% of RATED THERMAL POWER, and N o N ,
/'/3. Within 24 hours of attaining greater than or equal to 95% o l
J. 4 (C RATED THERMAL POWER. m ' SURVEILLANCE REQUIREMENTS j G.I.i n(The pr;vhi:n: c' 5;;;ifiestien 4.0.0 :n ::t :.;;1i::51 ., . , (4--h31 combinationofindicatedRCStotalflowrate(andFh , determined to'be vithi.n the region of acceptable operati \ Specification 3.2.3: ..~~.. Li - (
- a. Prior to operation.above 5% of RATED THERMAL POWER after eac.h fuel A
. loading a nd ' W c 4r,44.t.5 C be- ~~%t'least once per 31 Effective Full Power Days. ^ ~ ~ _.~ ~ N 4 (4.2.1 33 The indicated RCS total flow rate shall be verified (t: 5: rithin theJ $ Thgi:- iflt :p'i'TE~id : ten ef~5Ffdift::ti:n 3.2.3}at least once per 12 hours j ch:n th: =:t r=:ntly obtained-value--of--f , obte4ced-per Specificetien 4.2.3.2.f j
- 1e =:m t: ::i zt.,
~
e [4.2.3.4- Tra ",CS tete fE~T&Y.ddedterE ihe1IEUN;ted to E C. LEE.L L% k I- (CALI""JTIDh' et isnt er;; ;;r 10 naths./ I 5r, M.I.2i '(4.2.15) The RCS total flow rate shall be determined by precision heat balance _ b measurementt,rier to ; Lletien e7 M 5I C TET W:!t:7 :::5 ft:1 10 di -D"" " ~~ N y - i;NO ' M ^' b N I #b ** ta = = c n n at-instrumentation shall be calibrated within-seven days prior tos l' %d -AL'f [S4 25f tM-calorinett-ic-. the-performance Ite.5 N.5flow-measurement.---Prior--to-the-prec4s-ien l11.iNI N 3 5 INei'i CIShib y 0 h::t b: lance-seasurementr-at-least-two-of--the-founfeedwater-f-low-seter i' ventur4s-sha144e visua14y-inspected -andrif-fouling-is-found, 211 :::t -45 ) 4 Qh:11 5: :le:ned./ 3/4 2-9 AMENDMEffT NO. 38 BRAIDWOOD - UNITS 1 & 2 y -
_ v4 - - ,
.~
, * * . =; N
_k i if )f:
~
~ '
'r
~ TABLE;3.2-1( ~-
- E..o-DNB-PARAMETERS:
-g:
PARAMET R- LIMIT 50
.- E Leo 5.4.f.g . Indicated Reactor Coolant ~ System T.* ,9 5 591.-2*F QJ -
7-g
- L c o . 8.4 I*d- Indicated Pressurizer Pr. essure- .
> 2219 psig*'
'M ro .
4 i g-
?
's N
) ts i-h t
r c. (*Limitnotapplicableduringeithera:THERMALPOWERrampin)- i I- M# ##I" " excess of 5% of RATED THERMAL POWl.R per' minute or'a THERMAL. - (POWERstepinexcessof10%of_RAlEDTHERMALPOWER. . j -w . g-t o
> c
- . tes ! g o=
- - . - - - . . - - _ - - - - - . _ _ _ - - _ _ _ _ _ . _ -- - - - - _ - _ _ _ _ _ _ . - - - . - - _ _ - _ _ _ _ - - - _ . _ _ ~ _ - _ _ _ _ . _ _ _ - - _ _ ~ . _ - - _ _ - _ _ -
- - ~ . - _- .. - - - . ~ ~ - - - . . - - . . . . . . . - ~ . . . . - - - . . - - - -
LCC 3 A.T 3 4 RrWR COOLANT SYSTEM e g 6 (39.5 R u, Leo p - N o0 E 3) A l' / LIMITING CONDITION FOP OPERATION I 3.4.1.2 At least two of the reactor coolant loops (P shall be 2 OPERABLE with two reactor coolant loops in operation when the C r_;; Tri2' 4
^ = ^'--^ and one reactor coolant loop in operation when the
; - - - -~ ~ ,n=
. . . , - = r - - r = : -- - - - n, *m(v.,A cwQ.e.e ye ,s,mar c.%\e .L)
T Reactor Coolant Loop A and its associated steam
%,i g k rea coolant pump, toranq h g,,46,a lb. Reactor Loop B and its a ed steam generator and reactor coolant p ,
- c. Reactor Coolant L and it asociated steam generator and reactor e pump, and
- d. r coolant Loop D and its associated ste rator andi 1
remeter coolant numo. # PLICABILITY: MODE 3. c } twm RX M_-N g p. g With less than the above required reactor coolant loops OPERABLE, rest..re the required loops to OPERABLE status within 72 hours or be
~
Ce& L - f in IiBT EMUTnoWN within the next 12 hours.) (ane A Mf With only one reactor coolant loop in operation and the @ easter-Osep)hs.4 2
.y..- .. .__...- m_. ....,, within 1 hour djiEN=Giiib : '
. O cc.4 c p t.t pl. With no reactor coolant loop i(n operation,4 suspena aAAe..cNe.operations
....-A.w .u e xw.u.J k w -
1 cuwt t s. e t involving a reduction in boron concentration of the Reactor Coolant (c.egt,c A A t a systamJand immediately initiate corrective action to return the) c,esM'^[]p(requiredreactorcoolant loop to operation. j eence d (cMO F - k L,w4 3.4- M L.a3 SURVEILLANCE REOUIREMENTS Q.*S% '^^^~ At least the abeve required reactor coolant pumps, if not in c,peration, shall be determined OPERABLE once per 7 days by verifying correct breaker alignsents and indicated power availability.
%% 51 The required steam generators shall be determined OPERABLE by
- 4. .i.4.e verifying secondaryysifgj ayrow rance water level to be greater than or equal to 184;T2P f:: ;..n * ;;i;; :: ;1e.e S at least once per 12 hours. N
~~
k 3 4.5".# As ' . d
$.0.1.2.0 The required coola loops shall be verified in operation '- ' ri :29 Gatin: - ::ter ::n--S at least once per 12 hours.
Cet*ed C=~ @<m4 2 hr 7 he.me pencA
*All Reactor coolant pumps may beI:Weeeeeeksedd for up to 1 hour &provided:
(1) no operations are permitted ~that would cause dilution of the Reactor Coolant system boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.
% -- s--r4-1 +--,. wire neinn.
$~ rifi--+ i-- ? m G BRAIDWOOD - UNITS 1 & 2 3/4 4-2 AMENDMENT NO. 87 b%
A. R ACTOR COOLANT SYSTEM kCo 3 y. L ,
/ l b ' SURVEILLANCE REOUIREMENTS
-fg344.s .
4.4.1.3.1 The required reactor coolant pump (s) and/or RHR pumps, if not in operation,'shall be determined OPERABLE once per 7 days by verifying correct breaker alignments andlindicated power availability. 4c ? wL 4.4.1.3.2 The required steam generator (s) shall be determined OPERABLE by verify c4AdaryA4eA14rrErange water level to be greater than or equal- a to 18% A F fer "-ft Ori:r te cyc+ O at least once per 12 hours. ~ 4gp El ^? -h As 2 4.4.1.3.3 at ::::: :::)reacto colant or RHR loop shall be verified in
-.operati 5 0 C G' m .9 :21st:r --^M*O at least once per 12 hours, 5
Am e 4 t i 4
.g
+
BRAIDWOOD - UNITS I & 2 3/4 4 4 . AMENDMENT NO. 87 Eev K
_ _ , . - . _ ~ , . _ _ _. .__.___._m _ _ . _ - _ . . _ . _ . . . _ _ . _ . _ _ _ . . _ _ . _ _ . _ tco 3i'. '/,7 - F T j,l}'REAcTORCD01ANTSYSTF5 j i .",._ _ N gDOPS FILLED
- L. J,V,7 RCS ls0Pt. Mo0E O g , LYMITING ODWDITION FOR OPTRATTOW Lc0 3.47 .
;.0.".0.i- At-least one residual heat removal (RNR) loop shall be OPERABLE and I in operation *, and either: e$
ooi r-1'3, 4 .7. d + One additional RNR loop shall be OPERABLE #, or j 3.M.7.b >, r) The secondary side narrow range water level of at least two steam e f m.o,6 seaarastra **11 w p 2a u = = === : ;=t== # - , d;p L. A 2. 1 APEICABlIJTY: MODE 5 with reactor coolant loops filleddf. I. 4 L EZ2Rt! t
- f. G2eavic.eD )
j a1 - Witt,4ne 6f-1rhe RER loops inoperable and with less than the required
. %g g g,,, return steam generator level, isesediately initiate corrective action to ~
O the inoperable RRR loop to OPERABLE status or restore the
. required steam rator MC j ,, hvi tte 6 $4.r- as+o- OPERast.E- as nossible. 5 +etv5 d . }
X. . seestrAo taan loop n operation, suspend all operatione involving a { gA reduction in heren concentration of the menetor coolant, system and immediately initiate corrective action to return the required RNR g
.. . loop to operation. . . ., g*rfX t. } T .. A 4-C Ms m.
- 1. \ l sunwrIff2ect hiuui= = n6 gt
'5 R 3.4."/.*2 ' The secondary side water level of at least two steam generators 0" 0.0. .t.i_1
-when required shall be determined to be within limits at least ones per 12 LA 3
'" k hours..
(m requiud Aw s g 3 N ,7 !-4.,..^ .;.;.; M;.". i_- ^. -.. '. RBR loop shall be determined to be in operation
) :L __ r = rrrr r :::: - lat least once per 12 hours. 4
[ removed from apamb [ par 2 ho# periDd M,3 3 R h.Q."13 ( ttJ5 EAT' 3.Y ~68
*The RsR pump y be '2:- :_M- d for up to 1 hour /provideds (1) no operations are permitted that would cause dilution of the Reactor coolant system boron t,f,3 @it ( coecentration, (2) core outlet temperature is maintained at lea o*r g below saturation temperature. Mof-h loop s. o pa rgh_j 4
(,topjole2.#one.asR-lo*P may be inoperable for up to a nours for surveillance to ng -f
"~%ided the other RNR loop ie OPERABLE and in operation. 1
( O ? W C M #A remeter eeolant pump shall not be.etarted with one er more of the Reactor (A25-) . Coolant system cold leg temperatures less than or equal to 350*F unless the secondary water temperature of each steam generator is less than 50*F above co Ac3 each of the Reactor cooient system cola leg temperatures.
@ Ajok.Q.dlMSE E 8 4-E [,y O; ..
AMENDMEW No. 87 L/ ' 3RAIDvo0D UNITS 1 s 2 3/4 4-5 hJ V _. _- . .4 . _ ._. . _ _
.. . . . . , _ . . . . . _ . . , . . . . . - ...~. - . - .. - - . . - ~ . - . - . - - . . .
t I ' '
,c, '_
l
~
l j !J
=-
+
CTS INSERT (S) l id"Y > SECTION 3.4 L V.-,, - m 1
. .i
~
p. LC0'3.4.7 R , c ;l: INSERT 3:4 5A' '.'(Ap ) - p , b 'll! 'Deletedlin Revision K
, .s _. . ,
-j L /
p u p': 1 !' I L iINSERT-3:4 5B ' .(M1 ); E ~ ~
. SURVEILLANCE FREQUENCY j u
., , a.
SR; 3.4.7.3. : Verify-correct' b'reaker alignment and 7' days- l L , - indicated power are available to'each ,
, :. required:RHR: pump;
.s
< p,c.: -3 m W -s
. . . , s.i ilNSERT 3.'4 5C -(Ly).
o / i 4 - NOTE- -- M > 4, - All RHR' loops may be. removed- from operation during planned heatup to
'l10DEJ4 when'at':least one RCS loop is in operation.
ym i l: il 1-1lV; 9/22/98 Revision K ,, . 'k ,. l ,
/ ,-n w e- w -- w r . m, --~ e
._., . .. - . ~ , _ . _ _ . .. _ .. _ . - . - ~ . . . . - - .~~. . . . . . . .
.y .
'f-l
. CTS INSERT (S)
?5p SECTION 3.4
%[. .-
f
- LC0 3.4.7
~
lI'NSERT 3.4-5D ' . ( Ay) 2 m ,. t - CONDITION- REQUIRED ~ ACTION COMPLETION TIME I Di : Two~ required RHR loops: D.1 ; Suspend all Immediately- { inoperable, operations involving
- l. a reduction of.RCS
;@- boron concentration.
Required RHR. loop AND ' 1
.ino3erable and one or' . 1
.. bot 1 required SG' D.2.1 Initiate action to- Immediately
-secondary side. water- restore one RHR loop level (s) not'within to OPERABLE status, !
. limits.
v g Initiate action to
~
D.2.2 Immediately D.. .'- restore required SG V . secondary side water. i level (s), to within ; limits. ,
.' I' 'r 4
l1
,i - A'Y ,,
Il + f,Nl'
- 'jg. - 10/6/98 Revision K r
l i V y -- w y n -
, . . - . ~ . . . . , - . .~_ ~~ - - - ~ - - - ~ ~ - -- - - - m -~ LC.b 3. 'li? i 'd g l M.i ' REACTOR COOLANT SYSTEM '(t46') . l {- ("0LO 5"L'T00t" - LOOPS MOT ." LL 0d3 'id R45. Lo*p+ - Mope. f, Leon N,t FuiM) , ~ LIMITING CONDITION FOR OPERATION' l I ! .Lco 5.4.7 (M-tt:tF Two' residual heat removal (RHR) loops shall be OPERABLE
- and at least one RHR loop shall beiin operation.** ' '
~
1 , APPLICABILITY: MODE 5 with reactor coolant loops not fil' led. I L . ACTION: co4 g ,g y. With less than the above required RHR loops OPERABLE,'immediately E c co 8 r/ c M e . A. . initiate corrective action to return the required RHR loops to OPERABLE status.as soon as possible.
- ' X. . With no RHR. loop in operation,-suspend all operations. involving ja G.oHD 4 Rt. A.I .
reduction in boron Concentration ~of the Reactor Coolant Systemfand
^
immediately. initiate corrective action to return the required RHR
.c '"O A M *
- A loop to operation.
~ cou p c ITas m 34-LA
' SURVEILLANCE REQUIREMENTS- 1
.1 Me resule,/[
;.Sif.v.t.1-(4.4.1.4.0 ^.t !:::t :::'RHR loop 'shall be determined to be. in operation ( '
H- fgcel: tin:: r:::t:r ::: :nt) at least once ~per 12 hours'. LA3
- S g. 3.4,f. O . (zusar 14 4 8) i 1
4
. Lao Norc- A *0ne RHR loop _ may. be inoperable for up to 2 hours for surveillance testing provided the other RHR loop,is OPERABLE and i peration. ;
-Crem on s % o p.m ~m Au t- i TLconars i **The RHR' pump may be th ;;Pd for up to 1 ho provided: (1) no operations o are permitted that would cause dilution of.the' Reactor Coolant System boron p H concentration, and (2) core outlet' temperature is maintained at least 10*F T below saturation temperature, 2 y
I g4
&(3)taNo drisdiq op !tabOMt ort, /**r-4td )Mdi fw%w red 4C.L._Wl(k@bd V0\V%
g
' BRAIDW OOD' -UNITS 1'& 2 3/4 4-6 Q.v t<-
s Lcc 3M S 4. I. '3.*4' REACTOR C0OLANT SYSTEM (acs)
- 3.us r= :=T:ce = sQus n.w u.pwQ j SHUTDOWN' l
- j. J.IMITINGCONDITIONFOROPERATION
. ao 1.u t t. N If an RCS loop is isolated, maintain the hot leg and cold leo stoo 4 valvesclosedwettt;QY) g) ,o
, pm s.,o a.' ron concentration of the isolated loop.is s.. 6. than er iq.el 'j r
g a # * *- boron concentration of thef:::r:U..e ivopy, and '
- b. The temperature of the cold leg of the isolated loop is within 20*F of the highest cold leg temperature of the operating loops.
APPLICABILITY: MODES 5 and 6. f th: it:;; ;er!'ic:ti:n ::t retir*4=d da aaQ cee g Q (ACTION:
~;;; ;fth:rftiith th: the N:t !:;r;;;f-r nnt: ::1d ?;; :te; ^;;?;;;.f l
, . . . . . W 'Ld g -SURVEILLANCE REQUIREMENTS' (3N-FA ,
l A )[ SF 1D S.t - ZZE3EEB .The isolated loop cold leg temperature shall be determined to be within 20*F of the highest cold leg temperature of the operating loops within 30 minutes' prior to opening the cold leg stop valve. l
% ! w it.: '
(CE'ET:ET The boron concentration of an isolated loop shall be determined to y be greater than or equal to the boron concentration of.the QTcritiae 'aaah within hours prior to openin either the hot leg or cold leg stop valves ' B 4 of an isolated loop. N j 4g ga} s w. wumw h
-~
sow 4 kW "cm " ~ ]e a l p .. . y BRAIDWOOD - UNITS 1 & 2 3/4 4-8 1
%aL
t' f
-- - .. CTS INSERT (S)
,1 % SECTION 3.4 SI
.f LCO 3.4.18
.y .. . . .
. .o- INSERT 3.~4 8A :(L,) '
?
A Mlj ; CONDITION: = REQUIRED ACTION- COMPLETION TIME a 3-A. : Isolated 1 cop hot._orL A.1 Close hot and cold Immediately : cold leg isolation L leg isolation valves. i valve open with boron- 1 concentration- 1 requirement not met. I
, 'N,**
T.. ..; . M - B. Isolated loop cold. leg ~. B.1 Close cold leg Immediately-
>? Y
- . isolation valvelopen isolation valve.
.with temperature
[; requirement not met. j ' 1 1
-; p. .
a i i l
,'g. j 1
L.t T i f:X j. ' 9/15/98 Revision K c
,,. . .. - - . . . . . . . . . ~ . - - . . . . . . . . . . . . . . -
_( l40 2.4.IO. l.j.q s.y REACTOR ' COOLANT SYSTEM (Rc.5) _ NM ' 3.4.to Pressurfeer . Safe.yy falvat N
.. LIMITING CONDITION FOR OPERATION
.i LCo J.4.fo C.O.2.D All, pressurizer Code safety valves shall be OPERABLE with a lift setting - l
'of 2485 psig. 1% .@~ -
APPLICABILITY: MODES'1, 2, and 3.
.' ACTION:
-(.oNP4. With one pressurizer ~ Code safety valve inoperable, either restore the inoperable'valvetoOPERABLEstatuswithin15minutesorfbeinatleastHOT]
con 9B[STANDBYwithin6hoursandinatleastHOTSHUTDOWNwithinthefollowing ~/
\ 6 hours,f .
CMY%< (rNs LLT ' ^3,4- tos) ~ v ' SURVEILLANCE' REQUIREMENTS
- i 6It 3.4.lo.1 M No' additional requirement's other than those required by 'T
~ Specification 4.0.5. (g g. , gg. ggg -
( Withir, t / % , .
} .
4 l d u,
'LCONOTL= (rm sun.T M ICA) i m
~ ~
Thfliftsftting essure hallcp[respon~ toambie[)2 condition,5/of thege) . t pt nomin,al:oper ing te erature/and pr sure.f h
)
4 u
,BRAIDWOOD - UNITS l'& 2 -
3/4 4-10 9.e V E
n L L. m m ,
- CTS INSERT (S)^
SECTION 3.4;
* (('
l.. LC0~3.4.10.
;l_ INSERT 3.4 iOA (L3 )
T , . .
. . NOTE ~
The lift settings: are .not. required -to.be-within the LC0 limits during h ' 4 : MODE 3'for the: purpose. of- setting the pressurize'r safety valves under i- . ambient:(hot) conditions. This: exception is allowed for 54 hours
-a following entry into MODE;3 provided a preliminary cold. setting was
-! 'made. prior to heatup.
c
< INSERT 3.410B ' (M3 )
w' 1 ' CONDITION REQUIRED ACTION COMPLETION TIME e.- . 9
- y
- B'. .. B.1 'Be~in. MODE 3. 6' hours 1 ma Two'or more- B.2 Be in MODE 4. 12 hours-
, pressurizer . safety-valves inoperable.
1
'h _ -
l j%' d 8/22/98 Revision K
% t- e y- y 7
_q LCO 3,4,9 Id REACTOR COOLANT ' SYSTEM [RCf)- 3,qi" '
- PRESSURIZER
- l. ~
<_y . LIMITING' CONDITION FOR OPERATION
-LCO 3A % j
~-a-4-M The pressurizer shall be OPERABLE with at least two groups of l jressurizer heaters each having a capacity of at-.least:150 kW and a.. water W. i level -of. less than or equal to 92%. f .w.j n i e a.d cap & c1 keg werea-(rendo4 h ,.
L APPLICABILITY:: MODES 1, 2, and 3. y
' engineered sede43 bahtes pomer59pliebsc A
ACTION: I LCte 8 g. With less than two groups of pressurizer heaters OPERABLE, restore at . least two arouos of pressurizer heaters to OPERABLE status within 72 hoursfor be in at least HOT STANDBY within the next 6 hours and in) Com C 4-{ HOT SHUTDOWN within the following g.6 hours.J .aerlev.t not waW WaQ Q, A y. ' With' the_ pressurizer att:=t:: ::::rr e; be in at least HOT STANDBY
- l. ;;n =: =n:r r:;; ;rntr: ::::lwi;hin 6 hours and in HOT SHUTDOWN within the following 6 tours.
l
, k\f kserk cn rds,, ad Was LA Cmdrol l As+em k e. cwAWini Weaph\e. of roA wiMrmd a
L
' SURVEILLANCE REQUIREMENTS se 3.a.9.L . .
;. .:.U The pressurizer water level shall be determined to be within its limit at.least-'once per'12 hours.
s c 3.4.9.2 - M.'.0.2] The capacity of each of the above recuired groups nf pressurizer ,
; heaters shall be ferified by er,e,;iz his wi. ....i... . . .d .. .a., ; r ;i :: r::n 'I
': rrst)at least once each[ refueling Interval.b <g g ,v,4g3 ga L4' SR 3.4.93 t e pM t +y o f }.-(an3 '
M.'.2.3) The m ... ... ...,t r pressurizer heaters to.tthe ESF power supply ,
' , shall be ' demonstrated OPERABLE at least once per 18 months;ey :::r;t:::: = : ',
lf.;;te. )
)
be pwe n d] u i
. t l
I' BRAIDWOOD - UNITS 1 & 2 3/4 4-11 AMENDMENT NO. 74 L
- j. ,
b/ o
( -. 3,4 REACTOR COOLANT SYSTEM (ac d g M,ll(3/4.4.4 EELIEI VALVES k L h u .e u b er Opu+ A AmeG h Ow.v.)) LIMITING CONDITION FOR OPERATION '
- s tul'ut GFF0 Both power-operated relief valves (PORVs) and their associated block 1 l
valves shall be OPERABLE. , _V l l
.g urc i APPLICABILITY: MODES 1, 2, and 3.
C " " 3*'" h n3 capale d beim muondeM abdlC g ACTION: , L Co d A y With one or more PORV(s) inoperablen::cruce c m t W m m aka w )
, within 1 hour EEliincen+te- te F-e(s) tr0FF"FMt3*a or close
" the associated block valve (s) with power maintained to the block e(s); otherwise be in at least HOT STANDBY within 6 hours and D __ : "in DOWN within the followi nn A hance <w c vanaavunub Mf($ws d cra WPb W5EE4 9-n.ce ORV inoperable kne t crecer :t u r_t M t.:::: = : ::::
} 2.far
< ' thin 1 hourC:it ;;r r::ter: t ;: PORY t: OPERe,n :t tu: ~
p
. / ;, ,1:_:s;;;),heOc16, 9 t e associated block valve and remove power/fro INf i m^,aAr ore the PORV to OPERABLE status within^3: fcP M an
, MN2. 0 72 h or be in HOT STANDBY within the next 6 hours and in HUT s Con SH ' c.a mwg%n g 3
within the following 6 haneccrin mtw wauw~, ant eveno % , PORVs inoperable (12 to cru::: cr.:r tt,qr : ::::iv :: ' l W o EE 1 m - g# O'
- E ,p
- ::::),
- 4etesJor within 1 hour close (;itbr itsre:ter:
associated block valve and r 4 ,r the block valve and be in HOT STANDBY within the next 6 hours and
' (n --O in HOT SHUTDOWN within the following 6 hours. U) 4 ,A t.oNO C/F d' pWith one or more block valves inoperable, within 1 hour (Estore the># E !
- remew = Oi5=2 stme e place its associated PORY in E4 i i- I' Mmanual contro1 Restore at least one block valve to OPERABLE status 3A '
I vo.iiin the ist hour 11 Loth block veiven are inuperable; restore any remaining inoperable block valve to OPERABLE status within 72 3
~
ilours; otherwise, be in et ieast HOT STANDBY wii.hin the next 6 hours U"' d and HOT SHUTDOWN within the following 6 hours. , f
% , , p g, ,ie, The provisions of Specification 3.0.4 are not applicable.
4 i~ SURVEILLANCE REQUIREMENTS (4.4.4.D (In ;dditi:n t; the requir:;;;t: ef S g:ific:ti:n 4.0.5, each PORV I i shall be demonstrated OPERABLE at Teid once~per~1B months by: a H
- a. Performance of a CHANNEL CALIBRATION of the actuation instrumentation, and k ;-
j l Se ? 4,u.3 - .E Operating solenoid air control and check valves on associated air accumulators in the PORV control system through one complete cycle of full travel, and W M.U.;L .r' ,0)erating the valve through one complete cycle of full travel Giuriq n-
$E24 u 2 ate <OD E s~ 3 or U -- .
m;.i.s.u.t - L5ert 34.c 3[ (EEFID Each block valve s 11 be demonstrated OPERABLE at least once per p 92 days by operating the val _ve through one complete cycle of full travelfdnless }
'Q M'f N l(the block valve is closed (ytthgowerMem6yed)in _ order to meet the requirements J Nn- lof ACTION b. or c. of Specification 3.4.4.] (.m AMENDMENT NO. E, 33 BRAIDWOOD - UNITS 1 & 2 3/4 4-12 f ; ': V.
CTS INSERT (S) Lf'i~ SECTION 3.4
%.J <
LC0 3.4.11 INSERT 3.4 12C .(A29)
- b. Withlone'PORV ino]erable because of automatic actuation circuitry failure, within 1 hour eitler restore-the PORV to OPERABLE status or close the
. ~ associated block valve and. maintain power to the block valve:. restore'the l3 :PORV to OPERABLE status within the following 72 hours or be in HOT STANDBY
- 4
~
within 6 hours and in HOT SHUTDOWN within the following 6 hours. e .I c. With both PORVs inoperable because of-automatic actuation circuitry M . failure, within 1 hour either. restore at least-one PORV to OPERABLE status or.close its associated block valve and maintain power.to the block valve L4 and be in. HOT STANDBY within the next 6 hours and-in HOT. SHUTDOWN within d the fallos.ing 6 hours. Y ~ I INSERT 3.412D - (A29) ph or-automatic actuation circuitry failure us . i) 9/1/98 Revision K
SURVEILLANCE RE00fREMENTS (continued)
- 1) All: tubes that previously had detectable tube ' wall penetrations j greater than 20% that have not been plugged or sleeved in the affected area, and all tubes that previously had detectable sleeve wall penetrations that have not been plugged, hp k 2) Tubes in those areas where experience has indicated potential I' problems,
- 3) A tube inspection (pursuant-to Specification 4.4.5.4a.8) shall be performed on each selected tube. If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recoroed and an adjacent tube shall be selected and I subjected to a tube inspection, .
- 4) For Westinghouse Model'D4 steam generators, indications left in ,
j service as a result of application of the tube support plate voltage- l based repair criteria shall be inspected by bobbin coil probe during l 4l all future refueling outages, and l l 5) For Westinghouse Model D4 steam generators, tubes which remain in . ' k f service due to the application of the F' criteria will be inspected, l AMeJnJD $ '
,in tho tubesheet region, during all future outages, l- 1 \
e i t;u D o t I The' tubes selected as the second and third samples (if required by Table ; j ' g' g c. 4.4-2) during each inservice inspection may be subjected to a partial tube ' 1 E r inspection provided-j 1) The tubes selected for these samples include the tubes from those 1 i areas of the tube sheet array where tubes with imperfections were previously found, and The' inspections include those portions of the tubes where ! !' 2) l imperfections were previously found. + 1 I l d. For Unit 1 Cycle 7, implementation of the steam generator tube / tube j f support plate repair criteria requires a 100-percent bobbin coil inspection for hot-leg and cold-leg tube support plate intersections down ! to the lowest cold-leg tube support plate with known outside diameter
- 3. stress corrosion cracking (00 SCC) indications. The determination of the lowest cold-leg tube support plate intersections having 00 SCC indications j- shall be based on the performance of at least a 20 percent random sampling of tubes inspected over their full length, t .
r t e. For Westinghouse Model D4 and D5 steam generators, a random sample.of at i : ! least 20% of the total number of laser welded sleeves and at least 20% of ' ! I the total number of TIG welded sleeves installed shall be inspected for axial and circumferential indications at the end of each cycle. In the ' event that an imperfection exceeding the repair limit is detected, an additional 20% of the unsampled sleeves shall be inspected, and if an !~ 4 imperfection exceeding the repair limit is detected in the second sample, all remaining sleeves shall be inspected. These intervice inspections l ! will include the entire sleeve, the tube at the heat treated area, and the . tube to sleeve joints. The inservice inspection for the slee"es is
.i required on all types of sleeves installed in the Byron and Braidwood Steam Generators to demonstrate acceptable structural integrity.
1 7_.y 3/4 4-14 DMENT NO. 92 BRAIDWOOD - UNITS 1 & 2 2e4 K
SURVEILLANCE REQUIREMENTS (Centinued) C,b)h T5 t T4.4.5.3 InsDection Frecuencies - The above required inservice inspections of steam generator tubes shall be performed at the folloeing frequencies: ,V(O -
- a. The first inservice inspection shall be performed after 6-Effective
' Full Power Months but within 24 calendar months of initial l
! criticality or initial operation following a steam generator {_- replacement. . Subsequent inservice inspections shall be performed at
- I intervals of not.less than 12 nor more than 24 calendar months after the previous inspection.- If two consecutive. inspections, not k 1-
{ including' the preservice inspection,- result in all inspection results ; r falling into the C-1 category or if two consecutive inspections i f~ :i demonstrate.that previously observed degradation has'not. continued . i- and no additional degradation has occurred, the inspection interval *j
' may be extended to a maximum of once per 40 months, i -
i i 1 - b. If.the results of the inservice inspection of 'a steam generator ! 1 L conducted in accordance with Table 4.4-2 at 40-month intervals fall l i- } in Category C-3, the inspection frequency shall be increased to at i i least once per.20 months. The increase in inspection frequency-shall apply until the subsequent inspections satisfy the criteria of. ' Specif 2 cation 4.4.5.3a.; the interval may then be extended to a i j; maximum of once per 40 months; and j e
< i
+
! c. Additional, unscheduled inservice inspections shall be performed on }
j aach steam generator in accordance with the first sample inspection i . f specified in. Table 4.4-2 during the shutdown subsequent to.any of
!- the following conditions:
L
- 1) Reactor-to-secondary tube leaks. (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of j_ Specification 3.4.6.2c., or
- 2) A seismic occurrence greater than the Operating Basis Earthquake, or
[ 3). A Condition IV loss-of-coolant accident requiring actuation of the-Engineered Safety Features, or i 1_ 4)_ A Condition IV main steam line or feedwater line break. s
- N
; M S wal G b 00
% Doc Cec G k s.b
^ O _ 2 BRAIDWOOD - UNITS 1 & 2 3/4 4-15 AMENDMENT NO.92 n, e
- r. 7..w m .a
** 3 M IP c 70 SURVEfLLANet REOUTREMENTS fcentinued) 4.4.5.4 Aeeentance criteria
- a. As used in'this specification
< 1) imoerfection means an exception to the dimensions, finish or contour of a tube or sicave from that required by fabrication drawings or specifications. Eddy-current testing indications below 20% of the nominal tube or sleeve wall thickness, if detectable, may be considered as imperfectiones 2)' Deeradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube or sleeve;
- 3) Dearaded Tube means a tube or sleeve containing unrepaired imperfections greater than or equal to 20% of the nominal tube or sleeve wall thickness caused by degradation;
- 4) t nearadation means the percentage of the tube or sleeve wall thickness affacted or removed by degradation; J
e - Defect means an imperfection of such severity that it exceeds gg,g 5) the plugging or repair limit. A tube or sleeve containing an LQ. 0 ; unrepaired defect is defective; ke ' DOC 3 Pluanina er Renair Limit means the imperfection depth at or 6) b beyond which the tube shall be removed from service by plugging N - or repaired by sleeving in the affected area. The plugging or , repair limit imperfection depth for tubing is equal to 40% of I
.I the nominal wall thickness. For Westinghouse Model D4 and D5 steam generators, the plugging or repair limit imperfection depth for laser welded sleeves is equal to 40% of the nominal sleeve wall thickness, and for TIG welded sleeves is equal to 32% of the nominal sleeve wall thickness. For Westinghouse Model D4 steam generators, this definition does not apply to defects in the tubesheet that meet the criteria for an F* tube.
- _For Unit 1 Cycle 7, this definition does not apply to the tube support plate intersections for which the voltage-based repair l criteria are being applied. Refer to 4.4.5.4.a.13 for the !
repair limit applicable to these intersections; l
- 7) ' unserviceable describes the condition _of a tube.if it leaks or contains a defect large enough to affect its structural integ-rity in the event of an operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break'as specified in 4.4.5.3c., above; 8). Tube Insnection means an' inspection of the steam generator tube from the point of entry (hot leg side) completely.around the U-bend to the top support of the cold leg. For a tube that has been repaired by sleeving, the tube inspection shall include the sleeved portion of the tube, and 4
i BRAIDWOOD - UNITS 1 & 2 3/4 4-16 AMENDMENT92 NO. g %
%# g Ee4 %
p um s.a w m.e .e . . ,., 6ee%.~ ,"A9< 66. SS SURVEILLANCE RE00fREMENTS (Continued) 9): Etg.pervice insoection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques: prior to service to establish a baseline condition h g of the tubing. This inspection shall be performed prior to initial
' -POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
- 10) Tube Renair refers to a process that reestablishes tube serviceability. Acceptable tube repairs for Westir.ghouse Model D4 or D5 steam generators will be performed by the following processes:.-
a) Laser welded sleeving as described in a Westinghouse Technical Report currently approved by the NRC, subject to the limitations and restrictions as noted by the NRC staff, or
'& b) TIG' welded sleeving as_ described in ABB Combustion Engineering Inc. Technical Reports: Licensing Report CEN-621-P,
.A) M <A L Revision 00, "Comonwealth Edison Byron and Braidwood Unit I
/ . M ^^ C 0 T and 2 Steam Generators Tube Repair Using Leak. Tight Sleeves, FINAL REPORT," April 1995, and Licensing Report CEN-627-P,
/ %,. Doc 4 / -
2 Revision 00-P, " Verification of the Installation Process and p 4e ! Operating Performance of the ABB CENO Steam Generator Tube Sleeve for Use at Commonwealth Edison Byron and Braidwood s Units 1 and 2," January 1996, subject to the limitations and restrictions as noted by the NRC-Staff. Tube repair includes the removal of plugs that were previously l' installed as a corrective or preventative meature. A tube
~
inspection per 4.4.5.4.a.8 is required prior to returning previously plugged tubes to service,
- 11) Locked-Tube Model Intersection means all steam generator hot-leg tube-to-tube support plate intersections which have been analyzed 4
to experience a tube support plate displacement less than 0.1 inches during accident conditions, excluding the following: 3 l a) All tube-to-tube support plate intersections where IPC can not be applied per Generic-Letter 95-05; b) All Flow Distribution Baffle intersections; f. c) All steam generator tube intersections adjacent to an l intersection that contains a corrosion induced dent greater than 0.065 inches; and All tube-to-tube support plate intersections that will be [ d) displaced more than 0.1 inches during accident conditions due-
- to failure of the steam generator internal structures.
p _ LO d i BRAIDWOOD'- UNITS 1 & 2 . 3/4 4-17 AMENDMENT NO. 92 9 CD hK
REACTOR COOLANT SYSTEM' NewM l SURVEILLANCE RE0UTREMENTS (Continuedi
- 12) Free-Soan Model Intersection means all steam generator tube-to-tuce O support plate. intersections to which the Locked-Tube Model does not
. apply and which meet the criteria of Generic Letter 95-05, excluding the following:
.i .
I a) All tube-to-tube' support plate intersections where IPC can not be applied per Generic Letter g5-05; and f b) All Flow Distribution Baffle intersections.-
.13) For Unit I through Cycle 7, the Tube Sucoort Plate Pluoaino Limit H is used for the disposition of an alloy 600 steam generator tuce for continued service that is experiencing predominantly axially.
- ' oriented outer diameter stress corrosion cracking confined witnin i W the thickness of the. tube support plates. At tube support plate l
- Abud w intersections, the plugging (repair) limit is based on maintaining
- gg steam generator tube serviceability as described below.
l l % Ooc - a) Steam generator tubes, with degradation attributed to outside l g44 diameter stress. corrosion cracking within the bounds of the Free-Span Model Intersections with bobbin voltages .less than [
' t, g'o or equal to the lower . voltage repair limit (Note' 1] will be
- allowed to remain in service. Steam generator tubes, with j degradation attributed to outside diameter stress corrosion 1 cracking within the bounds of the Locked-Tube Model. 1 Intersections with bobbin voltages less than or equal to 3.0 q volts will be allowed to remain in service.
'D - b) Stea:n generator tubes with degradation attributed to outside l diameter stress corrosion cracking within the bounds of the Free-Span Model Intersections with a bobbin voltage greater l than the lower. voltage repair limit (Note 1], will be repaired or plugged, except as noted in 4.4.5.4.a.13.d below. l c) . Steam generator tubes with degradation attributed to outside l diameter stress corrosion cracking within the bounds of the
. Locked-Tube Model Intersections with a bobbin voltage greater 1 than 3.0 volts will be repaired or plugged.
d). Steam generator tubes, with indications of potential --- l degradation attributed to outside diameter stress corrosion cracking within the bounds of the Free-Span Model Intersections with a bobbin voltage greater' than the lower voltage repair limit (Note 1] but less than or equal to the . upper voltage repair limit-(Note 2], may remain in service if a rotating pancake coil inspection does not detect degradation. Steam generator tubes, with indication of outside diameter stress corrosion cracking degradation within the bounds of the Free-Span Model Intersections with a bobbin I voltage greater than the upper voltage repair limit [ Note 2]
\ will be plugged or repaired.
oo b- q BRAIDWOOD - UNITS 1 & 2 3/4 4-17a ENDMENT NO. 82j M J d
. we
R REACTOR COOLANT SYSTEM l hu.k. e 5.5 j ' SURVEittANCE RE0VIREMENTS (Continu?O !:j ~ [e. Certain intersections as' identified in WCAP-14046, See: ion- al M L 4.7,-will be excluded from application of the voltage-eased g pi repair criteria as it'is determined that these intersections j may collapse.or deform following a' postulated LOCA - SSE l- event. Flow Distribution Baffle intersections are .also l excluded frcs application of the voltage-based repair [ criteria. ~
- f. If an unscheduled mid-cycle inspection is performed, the l following mid-cycle repair limits apply'instead of the limits l
1 identified in 4.4.5.4.a.13.a. 4.4.5.4.a.13.b and 4.4.5.4.a.13.d for outside diameter stress corrosion cracking l indications occurring in the steam generator Free-Span Model I ! Intersections. For outside diameter stress corrosion cracking- '
-% indications occurring in the steam generator Locked-Tube Mooelf
- /4y,ucia L *) Intersections, the limits in 4.4.5.4.a.13.a and 4.4.5.4.a.13.c l apply. The mid-cycle repair limits are determined from the M" ( 0 - $ .
following equations: l
- ( p oe. d'r I.- J y )
i b 50 ""
/ 1.0*NDE+Gr(C1-AU) CL p-m ;
Va= Va -(Vm -Vm) ( CL-A
- CL -)
L-;
}
! Where:
I k Vuu = upper voltage repair limit
'I tg = lower voltage repair limit g
(g Vot= mid-cycle. upper voltage repair limit based on :
.i time into cycle '
j Vgg= mid-cycle lower voltage repair based on Vot and time into cycle At .- length of time since last scheduled inspection during which V and V were implemented.- CL = cyclelength-($etimeietweentwoscheduled steam generator' inspections) _._.
! Vg = structural limit voltage.
Gr .= average growth rate per cycle length
,l NDE = 95-percent cumulative probability allowance for
/ nondestructive examination uncertainty (i.e., a
/ value .of 20 percent has been approved by NRC)
/~ Implementation'of these mid-cycle repair limits should follow
[ the same approach as in TS 4.4.5.4.a.13.a. 4.4.5.4.a.13.b, 4.~4.5.4.a.13.c and 4.4.5.4.a.13.d. L Note 1: The lower voltage repair limit is 1.0' volt for indications of L x outside diameter stress corrosion cracking occurring in the N Free-Span Model intersections.
\ w 7
_ . - . _ _) y BRAfDWOOD-UNITS?l.&2 3/4 4-17b AMENDMENTN0.82p 3 l-n y g CCV K
. _x . - - . _ _
; 3 4.w=> .a- . ne... r c, . e n , . 7
' PC h ST SURVEf t.f_ANCE R20DIREMENTS fcentinued)
Nsto 2: 'Th3 upper voltcg3 repair limit fer indications of outsida diameter stress corrosion cracking occurring in the Free-Span Model intersections is calculated according to the methodology
.in Generic 1etter 95-05 as. supplemented.
- 14) F Distanga is'the distance into the tubesheet of a Westinghouse Model D4 steam generator from the secondary face AAAx*a f. . of th. tub.sh.et or th. top of th. tast hardroll, whien.ver is L, Sc6- further-into the tubesheet, that has been determined to be 1.7 6.0 inches.
bD[ 15) F Tube is a Westinghouge Model D4 steam generator tube with l
. b b" degradation below the F distance and has no indications of 50 degradation (i.e., no indication of cracking) within the F*
distance. Defects contained in an F* tube art not dependant on flaw geometry.
- b. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair in the affected area all tubes exceeding the plugging or repair limit) required by
. Table 4.4-2.
4.4.5.5 Renort.a
- a. Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged or repaired in each steam generator shall be reported to the Comunission in a special Report pursuant to specification 6.9.2;
- b. The complete results of the steam generator tube inservice
]= inspection shall be submitted to the comunission in a Special Report i pursuant to specification 6.9.2 within 12 months following the
- completion of the inspection. This special Kaport shall includes
- 1) Number and extent of tubes inspected, t
- 2) 1,ocation and percent of wall-thickness penetration for each 1, .
indication of an imperfection, and i^ '
- 3) Identification of tubes plugged or repaired.
! c. Results of steam generator tube inspections which fall' into category c-3 shall be reported in a Special Report to the Commission pursuant i to Specification 6.9.2 within 30 days and prior to resumption of i plant operation. This report shall provide a description of l investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence. l-f, d. For implementation of the voltage based repair criteria to tube support plate intersections for Unit I through cycle 7, notify the ! staff prior to returning the steam generators to service should any a of the following conditions arises [O BRAIDWOOD - UNITS 1 & 2 3/4 4-17c ANENDMENT NO. 92 a wa
RE AL10k LUULM i m itn aceng 2., SURVEILLANCE REOUTREMENTS (Continued) bb e <~ *i 1 I
- 1) If esticated leakage based on the projected end-of-cycle (or if not practical, using the actual measured end-of-cycle) voltage
(~l distribution exceeds the leak limit (detennined from the (/ licensing basis dose calculation for the postulated main steamline break) for the next operating cycle.
- 2) If circumferential crack-like indications are detected at the tube support plate intersections.
1
- 3) If indicaticns are identified that extend beyond the confines l of the tube support plate. l
- 4) If indications are identified at the tube support plate elevations that are attributable to primary water stress corrosion cracking.
- 5) If cracking is observed in the tube support plates.
- 6) If any tube which previously passed a 0.610 inch diameter bobbin coil eddy current probe currently fails to pass a 0.610
- inch diameter bobbin coil eddy current probe.
- 7) If the calculated conditional burst probability based on-the projected end-of-cycle (or if not practical, using the actual measured end-of-cycle) voltage distribution exceeds 1 x 10'2, notify the NRC and provide an assessment of the safety significance of the occurrence.
(~T
- 8) Following a steam generator internals inspection, if C/ indications detrimental to the integrity of the load path necessary to support the 3.0 volt IPC are found, notify the NRC and provide an assessment of the safety significance of the occurrence.
- e. The resu'ts of inspections of Westinghouse Model D4 steam generators F Tubes shall be reported to the Comission prior to the resumption of plant operation. The report shall include:
- 1) Identification of F* Tubes, and
- 2) Location and size of the degradation.
(mM An A '~ h e kee DOC. Ce W~ ! s 'w) y AMENDMENT NO. 92 1 BRAIDWOOD - UNITS I & 2 3/4 4-17d
%vd Y ww
I' ] j
- j. >
l i .:- , x
-^
CTS INSERT (S) SECTION 3.4
-- /],(f ;
L .LC0 3.4.15~
- i. .
- -INSERT 3.4 20C . . 2(L
- .22L -' and ,L24).
l
/ CONDITION' ' REQUIRED ACTION COMPLETION TIME' -
1
.-A. . . . A.1 NOTE :
Not required to be
-.,g .
performed until -l 12 hoursLafter
-establishment of i steady.-state ;
'll -
_ operation. ! 1 Perform SR 3.4.13.1. Once per !
.. 24 hours 1 MQ !
c~t A.2 ... ... d i - y,
- 1. B .! ... B.1.1 ... ... _
f,[ AND
.t- ;
B .1. 2. .
-NOTE . i
, .y( . Not required to be
"'W performed until-
- 4. 4 -12 hours after establishment of' 7 steady state
?! operation.
l,^, 1 j. y.. 3
- i f
. 10/1/98 Revision K r
p +4 + .. . W a v v w , ,- -
co rusy
$,4 REACTOR COOLANT SlgIgg (Ac Q W 3M 3.4.13 TIONAL LEAKAGE -
LIMITING CONDITION FOR OPERATION
! co 3 s t5
.pF.WB Reactor Coolant System leakage shall be limited to:
Lco s.03.e. 4 No PRESSURE B0UNDARY LEAKAGE, Leo Im54 F. I gym UNIDENTIFIED LEAKAGE, i G l W #4 '8 d e'. 600 steamgallons permotrisolatedW5it day _ total reactor-to-secondary t'hE.Riactor.iC6olattt leakage _SJJfimMthro
~
generators LM :' M *< (TS0Tallins perlay ~throdgh ~ariy~6ne steam generator,) L " M a rt. .:. d". 10 gym IDENTIFIED LEAKAGE from the Reactor Coolant System, fe. 40 gym CONTROLLED LEA _KA_GE .at_aleac_ tor . Coolant Syst.es_ pressure of)~~ 2235 i 20 psig, and]
~.s.~'see ew 'n . :irch,3.r.;
voc %- sum 3 5 6A 3.414.1 f. a-eeadleakage at a Reactor Coolant System pressure of 2235 i 20 osi.g
,from any Reactor, Coolant System Pressure Isolation Valve (specifiedL "
inqabre va-h9 (j.r- n . per n.s.at
',w_'L ,-( utve ,ee q * % 4 r:r+ )
APPLICABILITY: N0 DES 1,2k3,and4. f- ACTION:
't Lcs .5.a :s mp3 a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.
cupt Jr. With any Reactor Coolant System leakage greater than'any one of the above limits, excluding PRESSURE B0UNDARY LEAKAGE and leakage from reduce the leakage Reactor Coolant System Pressure Isolation Valves,least HOT ~STAWY \
.a.te to_within_. limits .w_ithin 4 hoursfoFWWit CMP S within the next 6 hours and__in C0_LD_S_H1TDOWN_within_the following J r 30 hours.f~. o
. p ,.p. ,q x wa r 3.,-2, 4 h !y E With any Reactor Coolant Syst Pressure Isolation Valve leakage > -f
#"D ^ greater than the above limi -:-: h: W-: rn: .t: :=.un -
5$ gjt;; =='r txw or W~it least HOT STANDBY within the_n_ n g e capc; e6 hours and in. COLD SHUTDOWN within,the following 30 hours.[~ h=rson"Nt-+ W u m.r n -x s J
.38 0 E , 4 m y .,, e y n D
E)
~~
T [ Test- pressures-less than 22557p'sig~buigreatebhan 350 psig are-allow observed -leakage-shaWhe-adjusted for-the-actual-test-pressure- up to-2235
' - psig-assuming-the-leakage-to-be-directly-proporational-to-pressure-differentiti
~"~~~~~ ~~~-~~ - )
( te t,% .= 5:1f ;nzr. ( -- 94Y K AMENDNENT NO. 57 BRAIDWOOD - UNITS 1 & 2 3/4 4-21
l - -.- CTS INSERT (S)' Iff L d. SECTION 3.4 LCO 3.4.14 , : INSERT- 3.4 214 . (L7 )-' -. - . l l' CONDITION- REQUIRED ACTION COMPLETION TIME l: LA. One or more flow paths' .. -NOTE with-leakage from one Each valve used.to satisfy ;
.or more RCS PIVs not Required Action A.1 and-within'. limit. Required ~ Action A.2.must have been verified to meet-SR 3.4.14.1 and be in the' reactor coolant ]ressure boundary or the ligh pressure portion of the system.
A.1- Isolate the high 4 hours pressure portion:of a
'the affected system i t(
from the low pressure ' V)m portion by use of one closed manual, de-energized power operated, de-activated automatic ;or check valve. q, AND 5 A.2 ~ Isolate the high 72 hours
+ pressure portion of w the affected system from the low pressure 5 portion by use of a second closed manual.
l.' _de-energized power operated. de-activated
. - Ei. automatic, or check N valve, i.
W [.,). u . 10/1/98 Revision K f I i y-- 4 - , , . + +- ,, ,y + -- --
I h CTS INSERT (S)
- N-
-SECTION 3.4 K-)!
LCO 3'.4.14
-INSERT' 3.'4' 21B ' (An. )- i
"' JAdIONS 21.,SehrateConditionentryisallowe r each flow path.
e
- 2. Enter a)plicable Conditions and Required -Actions 'for systems made 6
Linopera]le by an inoperable PIV. r
-INSERT 3.4 21C .-(M4 )'
- CONDITION- REQUIRED ACTION- COMPLETION TIME LB. Residual ~. Heat Removal B.1 Isolate the affected 4 hours
- (RHR). System suction flow path by use of
.3 isolation valve one de-energized 5 interlock' function power operated valve. >
-+ inoperable.
k f l '. ' i, 7 i 8/22/98 Revision K u h l :-
i.C0 2A.G Lco 2F.N
~
LCC 2. ~.f
' REACTOR COOLANT SYSTEM, I
SURVEILLANCE RE0VIREMENTS
.SR 3. u . ls. I W.4.5.2.E Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:
fe. MUnitur iriy liis uuntsiaimaat simuapiscs u 9 5e0u3 and ; rti:Ul:tb; 4 ( erdierttivity eriter at leret :::: ::r 12 h:ur:;J
": nit: ring ti: r:::t;r ;;.ity :"-- di::h r; , :nd :::
t.
" f1::r dr:f :: di::h;r;; : d ' :: tery it h ::t :::: ;;r 12 5:u cen ;;.; 5 ]r n [Aj Ammd 1. Measurement of the CONTROLLED LEAKAGE to the reactor coolant pump T
; sm; 3.q 1 seals when the Reactor Coolant System pressure is 2235 i 20 psig at\i
; s,,boc4.c .- least once per 31 days with the modulating valve fully open. The
! s,g.w3.ss[ j@ 00E provisions 3 or 4;J of _ Specification 4.0.4 are not applicable for entry into)
[u scat 3.9 229) ,
- d. Performance of a Reactor Coolant System water inventory balancetat least once per 72 hours; and l 4
- "enitering th: ":::t:r "::d rhn; L:;teff ';yste; et h.et e.,ce p;r}
;24 5: r:.f g tco uly
^
't l
! f 4 f 4 =d 4 r k 4 , g
,I ;.4.6.2.2 bl: 2.01]shall _Each beReactor demonstrated Coolant System Pressure OPERABLElby Isolation veritying Valveto:;be withini' leakage o
' Tits limit: J d O , f b
d 3AaY ?,t 1::: ent: ;;r(18 months W4w:e Witk +ke IsT he yom
- b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 7 days or more and if leakage testing has not been performed in the previous 9 months, i
- . .arier t: returning the v: h : 0; ::rvie: folh&; -:inteaeace, rep":f r l t Or 7;ph;;;;nt r:rk :n th; 1:h:, :nd 7 i
! it . Within 24 hours following valve actuation due to automatic or manual action or flow throuch the valve;except for valves RH 8701 A and B) SR N.+e 3 + uand RH 8702 A and B. f (theprovisien;;fS;::ific; tin 4.0.4;r: n;t applicable fe entr., inte "00 i rw-+-J t L S R N.+e i An i b m.T 34 225] S R d +e .1 4 IAJ3 C P.T" 3.U 22 0 / ,j-l !O v BRAIDWOOD - UNITS 1 & 2 3/4 4-22 AMENDMENT N0. 74 t IZrv K
J lt ~ ! ;:,~q CTS INSERT (S) SECTION 3.4 (( LC0 3.4.13 s INSERT 3.4-22A -(y,)? SURVEILLANCE FREQUENCY L
@ )
4-.' SR: 3.4.13.1- --
- ... NOTE P' }; Not required-to be-performed until 12 hours
- after establishment of steady' state t j; operation.
- l Verify RCS operational. LEAKAGE 'is within 72 hours limits _by aerformance of RCS water inventory Jalance.
1 I l y;' Y 1 a l
- 4 .
8/22/98 Revision K
3.[ REACTOR COOLANT SYSTEM h(6 ) L bO EN./b l 1 ff,lhg .' t . 0. "' SPECIFIC ACTIVITY 1,
. LIMITING CONDITION FOR OPERATION QI f.4.lfo M The specific activity of the reactor coolant shall be(l_i-i:- A h g g,g7,fy .Less than or equal to 1 microcurie per gram DOSE EQUIVALENT N @. . 1.
Q I-131**, and SR Fl.I6.1 * **** "h*" ' **"** radioactivity.
* * "i*' "'i" P'" *" '""
APPLICABILITY: MODES 1, 2, C. 4. a oJ , goog 3 ujin RC6 avemgL k&peM+WL 1500'F] l ACTION: ! b' {Adb: Cort D A A)b TE: LLO $.0 N h nelOfpllca e IC D bo A , vita 3, ,,,,1,1, ,,eiyiey or en, ,,,,,,, ,,,1.,,
,,,,,,,e3 ,
RA A.'L - 1 microcurie per gram DOSE EoUIvALENT I-131** for more than 48 hours !
, ,during one continuous time interval /or exceeding the limit line shown on Figure 3.4-1, be in at least NOT STANDBY with c odo 6 than 500 r within 6 hoursi and T"" less ;
C0 4 C
-in with the specific activity of the reactor coolant greater than 100/E microcuries per gram, be in at least NOT STANDBY with T ag less than 500 F within 6 hours. i I
1
.t
( _ 1 t l l i e Llo 5.0.Ib AffI IcabOW l *With Tag greater than or equal to'500*F. k1 , i s
**For Unft I through cle 7, reacto coolant DOSE EpdIVALENT I-131 wpil be ;
'E l
I limit /d to 0.35 m recuries per am. / / / l s . BRAIDWOOD - UNITS 1 & 2 3/4 4-27 AMENDMENT NO. 92
\ lev Y--
- . . . . - . . . . - ---~ --..-~~~= - ~ . ~ . - - - - . . . . - - .
L f REACTOR CODIANT SYSTEM bC,0 3 'Y, /[0 ' w ' LIMITING CONDITION FOR OPERATION
~
ACTION fContinued1
-[
Mo0E E w& RC6 avuage, kupera bre ).WF l- ! MODES 1,:2, i' ceuo A . 4 -
, R A A .I With the specific: activity'of the reactor coolant' greater than L* 3 J microcurie. per gram DOSE EoUIvAI.ENT I-131.ri ;;;;::: t r.: :: /::
m r - r== _aa- cr==' perform the sampling and analysis requirement's
'of Item 4.a) of Table 4.4-4 until the' specific activity of the reactor l
' coolant is restored to within its limits. .
SURVEILLANCE REOUIREMENTS 1 A The specific activity of the reactor coolant shall be' determined to be k- withinthe limits by performance of the sampling and analysis program of l Table 4.4-4. , i , j e SR3. % l sa r.4.Ik.7, M SR 3 4 Ib.3-L 1
. /,
, I i
l 1 I L lt bt, l
*For Un limit
.1-through Cycle 7, reactor coolant to 0.'35 microc les per gram.
SE EQUIVALENT I-131 ill be I O I
-BRAIDWOOD - UNITS'l & 2 3/4 4-28 AMENDMENT NO. 92
+
pp y
.. . _ _ _ . . _ . . _ _ _ _ . . _ _ . . _ _ _ _ . - _ . . ._ _ __.. _ _._. _ _ _ . . - - _ - - - ~ _. _ __ _ _,
4 j LGo 34/6 5 _. l' I i i 1 a ~ t E 4 j m 6
- L i
O s
- s. 't j
t [250 6 i 4 E i $ ==I 'k h s E ', e s . l-4
@200
( -6 . _0 t
, & s UNACCEPTABLE i- O-m t OPERATION
. g s l- @. 'L s ' ' l z 150 (, t 4 i a.J - g {- O i
- O s
! O ,< l > g k I q l < i , 3 100 , i E i ! .A E
't i E ACCEPTABLE ?'
j ] OPERATION t
- g !
s 2 z !. Wt 50 t < 2 , i 3
- . O i W l W m
O 0-is o 1 20 30 40 50 60 70 80 90 100 1 !. PERCENT OF RATED THERMAL POWER j; FIGURE $3 4, lG -l) g j e- re:- =:==cy~-m pose EQUIVALENT 1131 RE ACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF / eesm 4 ld j
^7 ,.C0"_^3' 1 RATED THERMAL POWER % :-7=:: ="::=^,070" ---= 000: ","." 0:'~0"' 0 f,0*".':TY .
- * ^^!O
- l j
a b' +-- BRAIDWOOD - UNITS 1 & 2 3/4429 AMENDMENT NO. 92 P_e v IL (- 4
klO 3 N.O j k I'
' ,/
i - .e < ~ , l 9 20 J / f k /
; 3 <
/
s i3 T x f * ' l
! ]h x
[ ,
$200
< s
- j. ,
g s UNACCEPTABLE - g y OPERATION > 4 o s e 1 l W- s t t' G. / 3 M s , !- e 1s0 --- , i 5 s , i 8 . [ O L v
/
y n g % / N. ( 't
> x E 100 s , s
!O E T' - i, E, n / s ! L T > p 'L < \ , i g i / \
, w s
- 8
, 4 50 7i
> - L s
- 5 - % '
h
~~
[ACCEP, FABLE L x m -- - OPE, RATION f( *b' 0 ! i 20 30 40 50 80 70 80' 90 100 i \ i- PERCENT OF RATED THERMAL POWER FIGURE 3.4-2
- / ~
- UNIT 1 THROUGH CYCLE T f
i [ . DOSE EQUIVALENT l.131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VER US t: PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPEcl C ACTIVITY >0.35 pCi/ GRAM DOSE EQUIVALENT 1131 M i
. ... _ h a
[o i BRAIDWOOD - UNITS 1 & 2 3/4 4-29a H AMENDMENT NO. 92, v f ~~' i i ReJ Y-1 d
- = -
O Oi O E
~
TABLE 4.4-4
.. REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE u
~
~ _f AND ANALYSIS PROGRAM *
' TYPE OF MEASUREMENTL ' SAMPLE ANO~ ANALYSIS AND ANALYSIS ' MODES IN WICH SAMPLE =
' FREQUENCY Als ANALYSIS REQUIRED '!
^ 1. -- Gross Radioactivity.; 54 s.4.h,. I 4
. [At least once per 72 ". - d - 1,2,38
' Determination **' L [
(7g sc s.v.14. [
- 2. . Isotopic Analysis for?OOSE.EQUIVA- i{0nce . per ~ 14 , days - 1 j LENT I-131 Concentration "
Su.<.tL. 3 . - i
- 3. :
Radiochemical. for E Determination *** .On'ce per 6 months
- 1
- 4. Isotopic Analysis for Iodine af f0nce per 4 hours,.
-Including'.I-131, I-133, and-I-135- whenever the specific If,2f,3f[-4#,%]h j j
Lactivity exc 1- ' t
' coNP A K A 4.l + W
,fg_3c,__-.
3 3 ,,, i ., 3 sm,4 L l i (hys7-ef;Eh5EEE5NAvitv., t y and. o } i A7 ['One: sample between 2 and 6 hours following a - Il ?, ? ,E V 'f. JTHERMALPOWERchange 3 se,s.v.lG U exceeding 15% of W g. ftheRATEDTHERMAL' POWER within a 1-hour (period.- 61 r, Q
- C '
p! p BRAIDWOOD - UNITS i & 2 x
-3/4 4-30 AMEN 0 MENT NO. 69 . E I
. j
._._m_._ _ _ . - . . . _ . _ _...__..____.__.-~__,__._m__.. -
4 L LLO 3 4. te, TABLE 4.4-4 fContinued) TABII NOTATIONS
- . CoND A until the specific activity of the R.aetor coolant system is resterse
- - M A *I ,
within its limits. OldOV5
. Sample to be taken Miter a minimum of 2 ETPD and 20 days of POWER j
.6A. IM lfed OPERATION have elapsed since reactor was last suberitical for 48 N hours or longer. ,
(=* ross radioactivity analysis shall consist of the quantitati # ' meas at of the total specific activity of the rea coolant , except.for radio uclides with half-lives less o minutes and all j .- radiciodines. The specific activi* ...all be the sum of the i degassed beta-gasusa activi total of all identified gaseous activities in the sample , a2 after the sample is taken and extrapolated bac when the sample was t Determination of the , c'.ntribut a the gross specific activity shall sed upon those
-e peaks identifiable with a 954 confidence level. T est available_d4%A_may_be_ys_ed._f.or oure beta-mittina radianuelid==.
l _ j j [*** A radiochemical analysis for E shall consist of the quantitative measurement of the specific activity for each radionuclide, except for radionuclides with half-lives less than 10 minutes and all radio-iodines, which is identified in the reactor coolant. The specific activities for these individual radionuclides shall be used in the determination of E far the reactor coolant sample. Determination of 0- the contributors to E shall be based upon these energy peaks identifiable with a 954 confidence level. (**** For Uni through Cyc1 limited to .
, reactor coo 5 microcuries, r gram.j t DOSE EQUIVALEliT M '
ih;
. Admsed h nehan 1.1
\ Se.c Doc faseckm Io -
(-
. g . .
BRAIDWOOD - UNITS 1 & 2 3/4 4-31 AMENDMENT No.92 PJ J \L-
)
(1/AA-11 PRESSURE / TEMPERATURE
- LIMITS t.- C c 3 .4 3 Oh,\
REACTOR COOLANT SYSTEM 44 g LIMITING CONDITION FOR OPERt. TION ua.a 3.4.9.1 o The Reactor Coolant. System temperature _4Sd p Ar ssure ang haatup apt a 1 cpoldowD, rate _s_ shalle befinhintiine'd"G iccordance wit'h the Ti$Tts spectri~e'd " in2 he PTLR.' ~ ~J #s
'^
~-"^==~e = - - -
^ ~ ~ '
*1 l APPLICABILITY: At all times.
ACTION: b b With any of the limits exceeded, restore the tem rature and/or
'g R pressure to within the i w n 30 minutesJ perform syngineer LAu eval n to de ne the ef ts of the p .-of-li conditio_ n the etural i rity of t eactor Coo].afit Systearfdetergne that the Reactor.
R A A ;t Coolant System remains acceptable for continued coeration orfbe in at least HOT (STANDBY within the next 6 hours and reduce the RCS T ano pressure to less p6 t than200*Fand500psig,-respectively,withinthefoUowing30 hours. co4b c (_he.a 3.4-32Ak Q SURVEILLANCE REQUIREMENTS h n Maa t.t.0.1.1 The Reactor Coolant System temperature and pressure s_ hall be determined to be within the limits at least once per 30 minutesfduring system ) h3 ., ( heatup, cooldown, and inservice leak and hydrostatic testing operations.( - (~s
\ -
u.n -
'4.4.9.1.2 Thereactorvesselmaterialirradiationsurveillancespecimens\
shall be removed and examined, to determine changes in material properties, \ by 10 CFR Part 50, Appendix H, in accordance with the schedule inj g
=, / 4
. n~m%
Akes) L. %A. / Em Su. DOC &r kehs 6.D O m BRAIDWOOD - UNITS 1 & 2 3/4 4-32 d AMENDMENT N eu e
-. . . . -.- .. . ~. ,- . ~ . . . . . ~ . - . . . . . _ . . - . - - . . . - . - . . _ _ _ . . - - . . . -
R. 3 ft.S e 3.4 .l
.. REACTOR COOLANT SYSTEM j hEstTRITER *
{ f H iLTMTTINC NDTTION FOR OPERATTON L3.4.9.2..The ressuriser temperature shall be limited to: 1
- a. 'A max heatup off100*F in any 1-hour period, j 1
<, . b . :. A^ maximum _ ~1down of'200*F in any 1-hour- iod, and j
< c.- :ALanzimum spra water temperature diff ensial'of 320*F. ,
l I h APPLICABILI'"Y s 'At all times. . p
- i. ACIlQli8~ 'l jJ .
-With the pressuriser temperature it in excess of any of the above limits . ,
i ' restore the omr9erature ,to ' with the 1 te within 30 minutes; perform an
. engineering'avaluation to det ine the of cts of the out-of-limit condition l l
i- on the structural integrit f the pressurise determine that the pressuriser !. romains acceptable for e tinued operation or in at least' hot STANDBY within '; ; .the next 6 hours and ~ uce the pressuriser pressu to'less than 500 psig within the following 0 hours. , i j. l p sORvtTLLANCE OUTREMENTS
]
I !
.4.4.9. - The pressuriser teenparatures shall be determined to be withinyhe
.limi at-least.once per 30 minutes during system heatup or cooldown. The
.sp l l y water. temperature' differential shall.be determined to be within it at least once per 12 hours during auxiliary spray operation.
, 4 -- J. e
.(
)
l HfM Q
, smAzowooo - unzTs 1 r. 2 3/4 4-33 AntwoxxxTwo.893
- ._ s j '
wc 6 e+w., se r-. ..e~.w-_ _ . , - < . - - _. m -_ _ _ _ _ _ _ _ _ _ -
LCO J, J/, Q f OVERPRESSURE PROTECTION SYSTEMS (3.8/o12, LCkl TE M Mit ATU R6 ) SpecAke N0Y (I.Cn.(o LIMITING CONDTTION FOR OPERATION LLO 3.4.l2. C M d SA 34.12.3 .<---- ,(INS EAT 2 4- 14 A i J.4.12 d SEU:im At least two overpressure protection devices shall be OPERABLE, and i sach device shall be either 1
- a. A residual heat removal (RNR) suction relief valve with a lift
! setting of less than or equal to 450 peig, or 1 ! b. A power operated relief valve (PORV) with_a lift setpoint that varies with RCs temperature which does not exceed the limit established in the PTLR. i APPLICAEILITY: MODES 4, 5, and 6 with the reactor vessel head on. AC'" ION : INSEAT 3 4-348 M7
- CodD c /coNDD CotJO E g. with one of the two required overpressure protection devices inoperable in MODE 4, restore two overpressure protection devices to j
)
- OPERABLE status within 7 dayofor depressurise and vent the RCS ]
g4 , (through at least a 2 square inch vent within the next 8 hours. J t , I l 0 F pr. With one of the two required overpressure protection devices inoperable in MODES 5 or 6, restore two overpressure protection i devices to OPERABLE status within 24 houre for vent the RCS throughT l CDtJD 6 Mat least a 2 square inch vent within the next a hours. )
~
1 WD q /. With both of the required overpressure protection devices inoperable, ? ! . - depressurize and vent the RCS through at least a 2 square inch vent d j__ within a hours. j L C O 3.4.l?. g . (-ie. "- N'*5 :rt-4 --- "? e. t - c Averif y the vent pathway r2 T
- g 3,q g2..y at least once per 31 days when the pathway is provided by a valve (s)
- j. that is locked, sealed, or otherwise secured in the open position; j otherwise, verify the vent pathway every 12 hours.
P
@ Q he event either the PORVs, RHR suction relief valves, or th
- vente ed to sitigate an RCS pressure transient c a1
[ Report shall be ed and submitted to ission pursuant to
- - Specification 6.9.2 with a report shall describe the ,
, circumstances initiat transie effect of the PORVs, RHR l suction re ves, or RCS vents on the tra and any ' l ve action necessary to prevent recurrence. J i gg gpM The provisions of Specification 3.0.4 are not applicable. I p i 3 O BRAIDWoOD - UNITS 1 & 2 3/4 4-34 AMENDMENT NO. 09 h4K 9
'C 7
, ,. Tj '_ 6 l' '
i
. ? iL ,
, CTS'INSERTIS)^
@'~'V .
SECTION 3.4
;jj.: .
s ; 3 , [ :.. " ')
.LCO 3.4.12 -1
,;- a, . ,
; , lfINSERT33C34A l . (M 7 )L 0 j
~
B- i PLC0%3.4il2 TAntLTOP System shall be OPERABLE withi s> . l.s c ^ q
'a. . . . , l 1
L. -; .b; ... f ,
-c4 ;Each SIcaccumulator isolated, whose pressure:is .greaterj
- tLn or; equal to the maximum RCS pressure for the . 9 e _. : existing RCS cold ~ leg temperature allowed by the P/T 1 H * '
ilimit curves providediin~the PTLR, and'
- q
- d; ...
g -
..}}