ML20235V685
| ML20235V685 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 09/30/1987 |
| From: | DAIRYLAND POWER COOPERATIVE |
| To: | |
| Shared Package | |
| ML20235V671 | List: |
| References | |
| NUDOCS 8710150271 | |
| Download: ML20235V685 (183) | |
Text
,. _ _ - _ _ _ _ _- -_ _ __ ____ .
i i4 J
ATTACHMENT 1 CURRENT AND Pit 0 POSED SPECIFICATIONS i
I h01y,h,k e t Pc4-20 1
____. -___-___ A
g 4
' PAG 8 27d' ClownDvT AVERAGE PLANAR EXPOSURE The AVERAGE PLANAR' EXPOSURE shall be applicable to a specific planar height-and is equal to the sum'of the exposure of all the fuel rods in the specified
' bundle at the specified height divided by the number of fuel rods in the fuel ;
bundle.
- AVERAGE PLANAR LINEAR HEAT GENERATION RATE
~
l
' The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) shall be applicable to a specific planar height and is equal to the sum of the LINEAR HEAT GENERATION RATES for all the fuel rods in the specified bundle at the -
specified height divided by'the number of fuel rods in the fuel bundle.
PR0f M Ep
. Delete. definitions for Average Planar Exposure and Average Planar Linear Heat
- Generation Rate, i
j JUSTIM ATION .
These definitions are not needed for~a permanently shutdown and defueled ;
- reactor vessel. The specifications in which these terms are used are no i longer applicable under current conditions and therefore are being deleted.
4
)
i; PC4-20 2 I
-_____ . i
r_ _
PAM 27e Cl0 n DVT i
CRITICAL POWER RATIO !
The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in a fuel assembly which is calculated by application of the XN-2 correlation to cause l some point in the' assembly to experience boiling transition, divided by the actual assembly operating power.
PnGPOSED
~
Delete definition for Critical Power Ratio.
Jb5TIFICATIM This definition is not _needed for a permanently shutdown and defueled i reactor vessel. The specification in which this term is used is no longer
-applicable under current conditions and is being deleted.
l l
PC4-30 3
J PAGE 27f C(RMr#T IDENTIFIED LEAKAGE j i
IDENTIFIED LEAKAGE shall be:
- a. Leakage into collection systems, such as pump seal or valve packing leaks, that'are captured and conducted to a sump or collecting. tank, or
- b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not be be PRESSURE <
BOUNDARY LEAKAGE. j PnCFCGED !
Delete definition for Identified . Leakage.
1 JTGTIF1CATIOV This definition is not needed for a permanently shutdown and.defueled reactor vessel. 'The specification in which this term is used is no longer applicable'under current conditions and is being deleted. ';
i l
1 l
l PC4-20 4
i.
.g PMar 271 CN 1
- LIMITING CONTROL ROD PATTERN r.
1 A LIMITING CONTROL ROD. PATTERN shall be a pattern which results in the core !
bein'g on a thermal hydraulic limit, i.e., operating on a limiting value for APLHGR, LHGR,-or MCPR.
LINEAR HEAT GENERATION RATE
' LINEAR HEAT GENERATION RATE (LHGR) shall be the power generation in an arbitrary. length of fuel rod, usually one foot. It is the integral of the heat'Plux over the heat transfer area associated with the unit length.
- MINIMIN CRITICAL POWER HATIO The MINIMIN CRITICAL POWER RATIO (MCPR) shall' be the smallest CPR which exists in the core.
PR0ft M D i- Delete these definitions.
[USTIFTCAT10N These definitions are not needed for a permanently shutdown and defueled reactor vessel. The specification in which these terms are used are no longer applicable under curre:d conditions and therefore are being deleted.
I i
PC4-20 5 l
_- __- _ - _ _ )
q -
n.
i ,
^ PAQr 27h '
C(RDINT i
PARTIAL SCRAM
' A PARTIAL SCRAM signal _ shall' cause.thn ' electric and hydraulic' scram motors
~'..
for 13' preselected control rod drive' mechanisms to be actuated for control rod insertion. . Full insertion of PARTIAL SCHAM control rods.during POWER :
OPERATION shall render the' reactor suberitical.
PHYSICS TESTS-PHYSICS TESTS shall be'those tests performed to measure the fundamental' nuclear' characteristics of the reactor core and related instrumentation and (1) described in Chapter 13 of.the Safeguards Report,- (2) authorized under.
the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.
PRESSURE BOUNDARY LEAKAGE PRESSURE BOUNDARY LEAKAGE nhall be leakage through a non-isolable fault in a .
' Reactor Coolant System component body, pipe wall or vessel wall.
RATED THERMAL POWER
- RATED THERMAL POWER shall be a total reactor core heat transfer rate to the i reactor coolant and reactor component of 165 MWt.
SHUTDOWN MARGIN SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor is suberitical or would be suberitical from its present condition assuming all control' rods'are fully inserted, except'for the single control rod of highest react'ivity worth which is assumed to be fully withdrawn, and the reactor is in the shutdown condition, cold, i.e. I 800F, and Xenon free.
PIDI M ED Delete these definitions.
JUSTIFICA TION These definitions are not needed for a permanently shutdown and defueled reactor vessel. The specifications in which these terms are used are no longer applicable under current conditions and t herefore are being deleted.
PC4-20b 6
__1_ ___1_1____ _ _ _ _ _ _ 1
\ p .>
V l x, Mar 2T1 i
N
- THERMAL POWER:
I c .
. THEINAL POWER shall be the total reactor core heat transfer rate to the reactor coolant and reactor components.. ,.
- d -
' UNIDENTIFIED LEAEAGE l . . 'l UNIDENTIFIED LEAKAGE shall be all leakage which is not 3DENTIFIED LEAKAGE. :
i
'O mamssa f Delete these definitions.
.TWTIFICATIOV - ,]
-1 These definitions 'are not needed for a pemanently shutdown and'defueled -
l- reactor vessel.. The specifications in which these terms are used are no .,
longer applicable under current conditions and therefore are being deleted. --!
.i s
'l u
i i
I PC4-20 7 l
1
1MGES 271-27x a
J CimRENT }
i i
l r
TIIIS PAGE INTENTIONALLY LEFT BLANE 1
'I (Pages 271 - 27x)
\
PRolGSED 1
TilIS PAGE INTENTIONALLY LEFT BLANK .
Deleted ,
4.0.2 4.1.1 4.1.2 (Pages 271 - 27rr)
JIGTTF[ CATION l:
i* Pages 27y - 27rr are also being deleted.
l-f.
I' i
PO4-20b 8
b: j PAM 27y CBBRENT 4.0.2' SAFETY LIMITS AND LIMITING SAFETY' SYSTEM SETTINGS' m
'4.0.2.1 SAFETY: LIMITS THERMAL' POWER (Low Pressure or Low Flow) 4.0.2.1.l' THERMAL POWER shall not exceed 15% of RATED THERMAL POWER with.the reactor vessel steam done pressure less than 615 psia or the core flow less -
than 10% of rated flow.
APPLICABILITY: OPERATIONAL CONDITIONS 1 and'2-
. ACTION:
With THERMAL POWER exceeding 15%.of RATED THERMAL POWER and the reactor ivessel steam done pressure less than 615 psia or core flow less than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
THERMAL POWER (High Pressure and High Flow)
'4.0.2.1.2 The MINIMIN CRITICAL POWER RATIO (MCPR) shall not be less than 1.32 with the reactor vessel steam done pressure greater than 615 psia and
-the core flow greater than 10% of rated flow, i.
APPLICABILITY OPERATIONAL' CONDITIONS 1 and 2 ACTION: .
With MCPR less than 1.32 and.the reactor vessel steam dome pressure greater than 615 psia ~and core flow greater than 10% of rated flow, be in at least HOT SHUTDOWN within'2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
11KlR1%2?
Delete these sections. .j l
l
.RGTIFICA TIM These limits were applicable only in Conditions 1 and 2.
i
)
PC4-20 9 a
~ ~ ~ ~
Q ,
..; ,i
[y '
c
-c J,: '
j y?
jom fys .~
g l'
CBRENT ,
~
.J
' - SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS'
^ REACTOR COOLANT SYSTEM PRESSURE '
~
' 4.0.2.1.3 'The reactor coolant system pressure at the top of the reactor ;
'L' vessel lshall not exceed 1540 psig.. .
APPLIC'BILITY: A OPERATIONAL CONDITIONS 1, 2, 3 and 4.- l 1
ACTION:- I With the reactor coolant, system pressure at the top of reactor vessel abovel N 1540 psig, be in at'least HOT SHUTDOWN with the reactor coolant. system-pressure 1 1540 psig within 2. hours.
REACTOR VESSEL WATER LEVEL 4.0.2.1;4 The' reactor vessel water level shall be above the. top.of the active-irradiated fuel.
. APPLICABILITY: ' OPERATIONAL CONDITIONS 1, 2, 3, 4 ' and 5.
ACTION:
) With the' reactor water level at or below the top of the active irradiated fuel, manually. initiate the ECCS to restore the reactor vessel water level, af,ter.depressurizing the reactor vessel, if required. .
i
- RWGSED Delete these' limits.
d
.nwrmcmow I With the plant pennanently shut down and reactor pressure vessel defueled, these limits are no longer applicable. These limits were
~
established to maintain two of the'three physical barriers which protect the l
environment from the fuel, the primary system and the fuel cladding. With no j fuel in the reactor pressure vessel, these limits therefore do not apply. 1 i
I I
l s
PC4-20 10 i
__--.:.w.---_a - _ _ . _ _ - - -
sw , ,
y ,
, I
~s 1
!j! g,y. ;: >
hg 4 .
P i'
- PAGW 27aa' 27ffI m, . CMWilBWT
[ l W
4.0.2.2 ? LIMITING SAFETY SYSTEM SETTINGS 1 REACTOR- SAFETY SYSTEM REACTOR SHUTDOWN INSTRUMENTATION SETPOULTS l4.0.2.2.1 The reactor safety system reactor shutdown' instrumentation
.setpoints shall be set: consistent withithe Trip Setpoint values shown'fn Table 4.0.2.2.1-1 and the LSSS~. curves of Figure. 4. 0. 2. 2.1-1.. ..
I APPLICABILITY: As required for each channel in Specification 4.2.F.
ACTION:
4 With a reactor safety system reactor shutdown instrumentation setprint less
- c conservative than the value shown in the Trip Setpoint column'of Table
,: .4.0.2.2.1-1,. declare-the channe1' inoperable and apply the applicable U ' requirement- of Specification 4.2.6 until'the channel is ' restored to OPERABLE j status.withiits trip setpoint consistent with the Trip Setpoint value. .
.i L
y
.)
PC4-20b 11
PAGE 27bb CIDDiEW TABLE 4.0.2.2.1-1 REACTOR SAFETY SYSTEM REACTOR SHUTDOWN INSTRUMENTATION SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT
- 1. Manual Reactor Trip Not applicable
- 2. Source Range, Neutron Flux Not applicable (Nuclear Channels 1 and 2),
- 3. Source Range, Reactor 2 7 seconds i Period-Short (Nuclear Channels 1 tmd 2)
- 4. Intermediate Range, Neutron Not applicable Flux (Nuclear Channels 3 and 4)
- 5. Intermediate Range, Reactor 1 3 seconds Period-Short (Nuclear Channels 3 and 4)
- 6. Power Range, Neutron Flux
- a. Reactor Power-High I 80% of selected scale
< 15 1 24 indicated power on Nuclear Channel 7 or 8 (Nuclear Channels 5 and 6)
I b. Reactor Power-High 1 120% of RATED THERMAL POWER
> 15 1 2% indicas.ed power on Nuclear Channel 7 or 8 l (Nuclear Channels 5, 6, 7 and 8 with Automatic Gain Control)
PC4-20 12
PAGE 27cc Cl0WENT TABLE 4.0.2.2.1-1 (cont'd)
REACMR SAFETY SYSTEM REACTOR SHUTDOWN INSTRUMENTATION SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT j
- 7. Reactor Pressure-High (Pressure 1 1325 psig Safety Channels 1 and 2)
- 8. Reactor Power-to-Forced Circulation P-(2/3)F 1 48.3
- Flow Abnormal, (Power-Flow Safety Channels 1 and 2)
(Fig. 4.0.2.2.1-1)
- 9. Reactor Coolant Flowrate-Low I 30% of rated forced (Power-Flow Safety Channels I circulation flow and 2) (Fig. 4.0.2.2.1-1)
- 10. Reactor Water Level-High (Water High: 1 19 inches above level safety channels 1 and 2) "0" inches indicated and Low (Water level safety water level channels 1, 2 and 3) Low: 1 12 inches below "0" inches indicated water level
- 11. Main Condenser Vacuum-Low 1 19 inches Hg vacuum (Main condenser Vacuum Switches 1 and 2)
- a. Containment Bldg. MSIV Not Fully Open (Valve Closure Relays 1 & 2) 1 90% full open travel
- b. Turbine Bldg. MSIV Not Fully Open 1 90% full open travel (Valve Closure Relays 1 and 2)
- c. Turbine Stop Valve Not Fully Open 2 90% full open travel (Valve Closure Limit Switch)
- P =
- of RATED THERMAL POWER, F =
- of rated forced circulation flow.
(See Fig. 4.0.2.2.1-1)
IT4-20 13
t 3
p c
pm 274g ?
a CLERENT TABLE 4.0.2.2.1-1 (cont'd)
REACTOR SAFETY SYSTEM REACTOR SHUTDOWN INSTRUMENTATION SETPOINTS
_ FUNCTIONAL UNIT- JRIP SETPOINT
- 13, Control Rod Drive Accumulators
- a. Oil Level-Low I 1.4' inches' piston travel-
'(Limit Switch). from fully charged position
- b. Gas Pressure-Low 2,2150 psig (Pressure Switch)
- 14. Bus Voltages
- a. 2400 v. Bus 1A-Low Voltage I 1.8 seconds for 100*.
(Undervoltage. Relays Phases voltage loss
,A and C)
- b. '2400 v. Bus IB-Low Voltage I 1.8 seconds for 100%
(Undervoltage Relays Phases voltage loss A and C):
- c. 2400 v. Bus 1A-Low Voltage 1 1.8 seconds for 100%
(Undervoltage Relay Phase A) voltage loss and 2400 v. Bus IB-Low Voltage (Undervoltage Relay Phase A) d .~ 2400 v. Bus 1A-Low Voltage I 1.8 seconds for 100%
(Undervoltage Relay Phase C) voltage loss and 2400 v. Bus IB-Low Voltage (Undervoltage Relay Phase C) c.- Containment Bldg. MCC-1A-Low I 1.8 seconds for 100% !
Voltage (Undervoltage Relays voltage loss Phases A and C)
- f. Turbine Bldg. MCC-1A Low I 1.8 seconds for 100%
Voltage (Undervoltage Relays voltage loss Phases A and C) l PC4-20 14 l
p i
Phar 27ee *
- came, i
TABLE 4.0.2.2.1-1 (cont'd) l
' REACTOR SAFETY SYSTEM REACTOR SHUTDOWN INSTRITt4ENTATION SETPOINTS-FUNCTIONAL UNIT- TRIP SETPOINT
- 15. Reactor' Scram Relays Not applicable
'16. Automatic' Scram Logic Not applicable
.i Ir4-20 15
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ 1
l\lljlll1:
kNQ -
0 u l qg lp' g!
- j. . . . -
L uL 2
- :' M Ll F. :.biu u q.
. 4 '- - f. r- ':
mi m
.. . :.: jr aI' J' ,
- L. Jg- 5 ._l "- a_ -
1 -
L;**J -F L ,, - - - -
T frF
',. j J' 4 -
- . A J ,*- - . .
-4 ! * : ' .
- ,.*. *,. A. -
!L f .
. - - .;i'. ~- . -
- L.' JAuI. u.;l.
n i~H"-
L. 2. i.,
._ L 2 l -
1-:-- i.
P .a " ;-lgg vt_ t. f.
o 'hM :iF' a.it:J ' -_ -
- i. #.'lii.- L6 '- r ii sL
- - - _ . i tt,
- -l.L. j- L.tL.AWg,! :iy' AL_ .. 31. A1
? , -
. tL . - / _ I. +- 81 -j-H -
j2
- - t n
._ :.- P- s. ._F di ; l o
2 i. -
L . ..7- L t __ pA j. P H L,- iL i
A'*2"' )wLi .-
'"L+'
.?I -
.i::
l 0 r I
+-Ll.
LJ
- . t}.t, ..
-. e 4. I' !- .
. - . nt - - 0 h/ H .
- F._I '
l
. L 1i ;, . . _ . A. '3.jg Coo r pi
. h .L 1 b S
N
- i. ' AL }- , tp'.: 'L;.t
- .-> i *_ i 1 i . u.L.p;,
.* f :
r_ 4i. - - . e -
O I
i l.
L.l- wp LE p, l tu;J3:F .{.p
- r. i - I LL! -
h'L_ . . ..- . L' L iM Lp! T
!; t 6 .
_ LL - cO *3 L' 8L- .. j :- ]. L_ hd. , . - 6 h ;.J' - _
_ - 3.- - - . - .
- i. L. :L* l * . l-.- .
l
- - - i t1
- n. . 0 A
. l I. i- I - ' ilJJr 8
- .
- A * -
I. 8 l F e l.
i 1 L L* 4H .
- t - - .
j " L- ,.' F .L.I - L)- r F E e L- . Ljj- it" i
- t .
iL. l'
- - JlL, L' ' .L :_ L . ra3 -- . - - '
x R
- u. . 0 j i e l? j1 l .4 [
1 et 'l_
- LU k' A!.L. L t - -
, AL . L l-wS3b.
} i:
L.. -'r1.L
- '.l.-;
j 5 W LsIl' - - . '-t - - . gy-3 -
- b. 1 l -
7 O
- h. .h . Lcah. o4 33 . i .
- I1. J j ,
- . - . i .. -
P y li L .
II1l!.I
- l]w f_ - . -
_.j,p. 8 01 F
. - - ' : -[ t, d <Li[t :. L!F .' '.p . _l_
t!I s4 --
LL.1.t .
L.1f.I--
-b .'u*]-
! . .F .
i.*
. -' . . u* - -i _1L - - . [.g, itneF ll
. - . Ip -
' Jj'l.L {- 2 p ]. N O
t' ,
a'. ]_..' A.JtL_ L.'J- A.H I_ .
A1 g,,t. a_3l 3 . - . N, 1' I_i' i. . -'3J -[... w p4.ti'- J.tL.-
t :
tS38 l I
.2 > -
l.
I O, o T i
. i- ! .d ,. g - '- _. g t.. e8 ,i *g i
mr2
/
8 I 'i -.i . - . .-
1 i j:
.p -t .
- . l A 1 Jt
- t. J
- J [; '.F .,t9 o- 8 G- - . . - -1 uL F .
F L -
- 1. ; aL' f .
f t.!. a. 8
. J
. I
-'l i
LfP Ei il. . -n
- g J r_ ._ . U 1 l J. * *LL 1 {, 3 Ri : j l C H .,l.
2+
.-I.L ,u t. - . ..
l .r I *it -
tL.4_L Ll.
- 5. J AI . - R
.L .
I' . i l 2
.o L. n-
- l ' 1*
_ p I. - .
. !. G s P _ :. } 0 u I
.i .
- 1*.
- . 'i.t-N. *i._: t '
- 6 F C 2 I'
J! tlt IJ .Jg :
E T. }L a w l.p I . . .
- a. L1 ' ' '
- 8. :LL ? Li*
- *L ,3 . * :L-
. :I
+ T. -l f R 0
- l{.. .t .ui. -+ .F
. l .
, - . . - IA. uo .; o
^ . .
F .g/. -
- . u.L.
- J* 8
?
- e
.-L_
f '
' ;L. :_L
'i
.+lill
- 7.,
i1:
(
i
. .TI
. . R E, 1.t . - rF cl :F -
- f. i O
E 4
.L LL' l.I_ - ).' .
L, ll P. ~,.L .i 5.I .. C e
- s. '. . s;'
1 .
r
. _._- R
_ i ?*..
i
. . . cw$ _i ,
L lLf - j..'..
gI.I.' O. AA:T . .
. 4 _>. , .
! t l E. R L L - eoS- i
.J . e O u L1~ Y' 8LU
.rr A.., a..-
t F g
- d. <f . _
i I.l.
L.A1.
t
- t_ U LI- a
-J J a. . . - i. .J ! ' S S3 .
- i st L .J .
1 1
, - - I.,-l.._ - JJ .- 0 R R F Il!
L8 .
.. r - . . .. E
- .L_ :..J' L_2 AI?
L. L L! 4 .
. L. +
-h :- Ii . , t {_
[
lF t. ._
4 w W I.' -
- . .u. ig:.
J.1_ _L' ; d. .L. s:.L: b wo< Lt
!_ f_1. :.
1 .:
!.4 A. L J .uI .F L.t.i.L .. o O
._. l..a f.
J a;
- j. .. l.l ;.J i I.
- i.
} ,. l P
!'; f_ L( . L ;_i- -
L.. ,' f A 8 : .'
<[Fl F l
L_ L ;. L -.A: t
. F 1. g..j .
I ,:
jlr i
L s1 j '
L_j_ Lp ,, iL[..l 7, .
t A
- tL.:.i H .-.L HJGtif. (. .
l_.. - t L j., .
L_1 t1.-' ;.! r3 . s.
- li i '
p .
n M
~ t L !._gL.IL ; e/L.a L. L Ll. as_[.g a R I,,
1_L. l' J.a iL? t...l u! o-w2 . ?. . l.p. E
- t. L _ i' LLLL*
L A }_ _ .
g n.L.A*;a_AF l.
j
_ p_F .
l o H e
L ,IL* .,__
- L1'.
PP . t. J ;. t. jpi.p. -rl };
. - ) I. . .
.__.._-. 0 o T
, - l. .p 2 C
- t. I. 4 r -t.m L.t u- -
L l.b. .
- . tLte ; .; . I L* 1i L_ i . G
. . _ - L... L .- .
r I. -f ;L
- i I - .t
.2_
Ja . -
.;< L_
- t. A -'. $ !-
.i i
- L_ N 1;ALLt s_ o i i
- LL. ..L . H .q" ; .
._ .__ ; rr _L' ;
3,
+.L.p, 4. ;t i. F -
I
. . .... L_._t t T
/. . ;. ; L i. u, . - i. -
4 fL. ' 1. ; L L .
j[. L.; . l .:._
i .:
. _ A 'I.J . .- c
. -t
.L L ;. _ I NlLN:
_; J l. i " 1J _
hn.._ p p l.F . . . a M
-2' F ].
- 2. .1.l.
e_i 11 .
- J
- . a-
.{ ;; _
}i ;j.
.t u1li_
p . - -
bL' -
l,._ g
_ - L :. alp ii
- a. R e I L
. a. ; J: L i: . :
. is i-
+-:hr_;;,a-l 1f 2~ '
,{U _. _. .L. L. : p.._ I' l-'E.i L_ i +.'u_ t_f
.; a -
2 3
'. m. 1d t _.1-s .
- t g , i ! ..
, . . . ..t. . u i ,b t
. .. .A .F j.p,_, r.
0 I'l '[
!. i g3 ih 0 0 0 o 2 0 4 1 1
= m 5"* .. -,or ] ]2 c ., 5M a e.=m2sF gzmm 2
' 3
- hiwO
- c il!(Il!l!llI ,l!lill- ll' I ,l ,,:!l l{1
- s;g- -
1 .
/
f I iDelete Section 4.0.2.2,! including Table 4.0.2.2.1-l'and Figure 4.~0.2.2.1-1.
'JIETIFICATIM '
/ (.~.L- . ,
Since t_he reactor is permanently shut down, reactor shutdown-(trip)
. instrumentation is'not necessary.
,i i . '!
1 g.
r
} ,
l 1
IT4-20b 17
o _ _ . _ . _ _ _ _ _ _
D , ,
7 ~
<1 4 PAGE 77&T/bb d I
. camur i 3
4.0.2.10' SAFETY-LIMITS 1 i
!BASRS FOR SBCTION 4,0.2.]' 1
).
iL L4.0.2.1 'The fuel 1 cladding,'-reactor pressure vessel and primary system piping. I m ;are.the princ%al barriers to the release of: radioactive materials to the
- environs, uSafety Limits are established to protect.the' integrity of these-g <
barriers during norma 1Lplant operations and anticipated transients. The Safety Limit is set such that. no fuel damage is calculated to occur if the-Elimitlis not' violated. - Because fuel damage is ,not directly observable, ' a -
step-back ' approach is used to establish a safety Limit such that the MINI 510M .
s : CRITICAL: POWER RATIO (MCPR) is not less than 1.32. :MCPR > 1.32 represents a: ;
conservative margin relative to the conditions required to maintain fuel cladding integrity, i.e., the conditions which would produce onset of transition boiling.
- '4.0.2.1.1 THERMAL POWER'(Low Pressure or Low Flow)-
The use of the XN-2 correlation is not valid for critica1' power calculations at, pressures.below 615 psia or core flow less.than 10% of rated flow,'G <
Therefore, the fuel cladding integrity Safety R
.08x108 lb/hr-ft2, Ref. 1.;
> Limit is' established by.other.means, specifically by setting a limiting' ' '!
condition-on THEINAL POWER with the following basis.. A review of applicable low pressure and low flow. data, Ref. 3 and 4, has shown the lowest-data point for transition boiling to have a heat' flux of 144,000 Btu /hr-ft*. To insure applicability.to the BWR fuel rod geometry and provide a margin, a factor of one-half has been used, giving a minimum critical heat flux of 72,000 L Blu/hr-f ta , . This is equivalent to the hottest spot of heat flux in the core with core' average powerLof 28 W t, 17% of RATED THERMAL POWER. The limiting
. THEINAL POWER for' low pressure and low flow conditions has been conservatively Lset to 15% of RATED THERMAL POWER, somewhat lower than 28 Nt.
This value-is applicable to' ambient pressure and no flow conditions; there exista increased margin for any greater pressure and flow conditions.
4.0.2.1.2 ' THEINAL F0WER (High Pressure and High Flow) {
.)4 The fuel cladding integrity Safety Limit'is set such that no fuel damage is calculated to occur if the limit is not violated. Since the parameters which result in: fuel damage are not directly obsertable during reactor operation, ;
the: thermal and hydraulic conditions resulting in.a departure from nucleate '
boiling huve been used to mark the beginning of' the region where fuel damage could occur. Although it is recognized that a departure from nucleate boiling would not necessarily result in damage to BWR fuel rods, the critical I power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the
)
PC4-20b 18
CURRENT SAFETY LIMITS BASES FOR SECTION 4.0.2.1 THERMAL POWER (High Pressure and High Flow) - (cont'd) fuel cladding integrity Safety Limit'is defined as the CRITICAL POWER RATIO in the limiting fuel assembly assuring a 99% probability of avoiding boiling transition with a 99% confidence level.
The Safety Limit MCPR is determined using the Exxon Nuclear XN-2 Critical Power Correlation data base, Ref. I and 2. The XN-2 Correlation has been evaluated based on CHF data with the following range of operating conditions:
Pressure 615 to 1500 psia Mass Velocity 0.08 to 2.4 x 106 lb/hr-ft2 Inlet Subcooling -45 to 450 Btu /lb Quality at CHF 10% to 77%
Bundle Length up to 12 ft Ceometry Square lattice, axial flow 4.0.2.1.3 REACTOR COOLANT SYSTEM PRE 6SURE The reactor pressure vessel and the reactor coolant system are the other two physical barriers which separate the radioactive material from the environs.
The Safety Limit for the Reactor Coolant System pressure has been selected
.such that it is below pressures at which it can be shown that the integrity of the system is not endangered. However, the pressure Safety Limit is set high enough such that no foreseeable circumstances can cause the system pressure to rise over the limit.
The pressure Safety Limit of 1540 psig is derived from the design pressure of the reactor pressure vessel and the coolant system piping. The design pressure is 1400 psig at 6500F. The pressure Safety Limit was determined by including the 10% over design pressure allowed by the ASME Boiler and Pressure Vessel Code (110% x 1400 = 1540 psig).
4.0.2.1.4 REACTOR VESSEL WATER LEVEL With irradiated fuel in the core during periods when the reactor is shut ,
down, consideration must be given to water level' requirements due to the j effect of decay heat. If the water level should drop below the top of the j active irradiated fuel during this time, the ability to remove decay heat is i reduced. This reduction in cooling capability could lead to elevated l cladding temperatures and clad perforation. The Safety Limit has been established at the top of the active irradiated fuel to provide a point which can be monitored and also provide adequate margin.
l I
PC4-20b 19
__=_
PR0luSED Delete Bases for Section 4.0.2.1.
JUSTIFICATION Since Section 4.0.2.1 is being deleted, bases should be deleted.
l 1
i
]
PC4-20b 20
_.-___a
W4 g. - ~- ~~ - ' - -
[-
<>x .m '
y p r
- 3. :
w -.
by n - v:-
f; ,,,
@ ,, ,a PAGES 2711-27qqI CERNDW
~
l I 4.Ol2.2. LIMITING SAFETY ' SYSTEM SETTINGS
' BASES'FOR SECTION 4.0.2.2 _,,
9' ,14. 0. 2. 2.1 ) REACTOR SAFETY SYSTEM REACTOR SHUTDOWN INSTRUMENTATION SETPOINTS '
- The Reactor SafdtyLSystem Reactor Shutdown Instrumentation Setpoints specified'in Table 4.0.2.2.1-1 are the' values ~at which;the' reactor' trips are set for each" parameter.1 The Trip'Setpointshave been' selected to ensure that Lthe1 reactor core and Reactor Coblant Systems' are prevented from ' exceeding their Safety Limits.:
' Manual Reactor Trip The Manual Reactor l Trip'is: a redundant channel toithe automatic instru-Luentation channels'and provides~ manual reactor trip capability.
' Intermediate and Source Range TheIIntermediate'and Source' Range Reactor Period Channels provide reactor J
core protection.during reactor.'startup, The. Source Range' Reactor Period channel provides a prohibit' signal to the
~
Coritrol Rod Withdrawa1' Interlock when the reactor period is less than -
7 seconds. It also provides'a scram signal to the Control Rod Test Interlock when the reactor period is less than 7 seconds.
The Intermediate Range Reactor Period scram provides redundant protection to the .' low range ' scrams of the Power Range, Neutron Flux Channels 5 and 6.
A worst possible cause of a'significant power rise during startup.is continuous control rod withdrawal. This accident has.been analyzed for.an extremely conservat.ive~ case; reactor initially suberitical and at-ambient temperature, ' all reactor period and reactor power level scrams inoperative.
I and two control rods being withdrawn, with a reactivity insertion' rate'of
. .18 cents /sec. -The analysis'shows that a minimum reactor period of 0.045 Yb seconds occurs 6.2 seconds after criticality is achieved and that reactor .#
1
. power will peak at 443 Mw, ~270% of RATED THERMAL POWER, shortly thereafter due to the Doppler effect and then decrease further as voids are created.
The analysis concluded that even without scrams fuel damage would not occur ]
.since t he maximum peak fuel t emperature would be about 16850 F. 1 i
l' , .PC4-20b 21 f f A_n -- _A--.'.'_.__
. ~ ._ . - - - _ _ .--r . _ _ - . - _. J
x i.
u -1.
r.
CGERElW LIMITING SAFETY SYSTEM SETTINGS .
BASES FOR SECTION 4.0.2.2 __
i
- ~
r Intermediate and Source R(6ge;- (cont'd)-
' Protection against thisL type' of startup accident :is provfied by the Intermediate Range Rrcctor period'and by the high reactor power scram of
' Power Range Channels:5 and'6'while on>1ow scales.
' The setpoint for Intermediate Range channels 3 and 4 is 3 seconds while the setpoint for Power RangeLchannels 5 and 6 is 80% of full scale. Channels.5
~
and 6 are multi-range, instruments which the reactor. operator keeps between
~ 20% and:' 65%Lof ful1~ scale by range-changing as. reactor power increases duringLthe'startup. 'For. transients, such as the continuous rod withdrawal:
. y above, initiated from reactor ~ power levels atothe high endiof the. range:of 9 either nuclear. instrument channels 3f(and 4);or 5 (and 6), the short period scram or the high reactor power scram will' terminate the transient before-significant heat generation occurs. For continuous control' rod withdrawal R transients initiated at the low end of the range, the short period scram or the<high reactoripower scram will terminate the transient well before the MCPR value decreases'to 1.32.
-For continuous control: rod withdrawal transients; initiated at power levels well below the ranges.of Intermediate Racge channels 3 and 4 or Power Range:
channels 5 and 6, the short period or' hlgh reactor-power scrams may not -
~
'7 iterminate the: transient. prior to the first power. peak.' However, the results a
'of the worst caso analysis _.presented above clearly-show that<the initial- 1
= power peak.does not contain sufficient energy to damage the. fuel. Damage' j could. occur if subsequent power rise was permitted. ' However, the short
. period or high reactor power ~ scrams would terminate the transient prior to this subsequent power rise to prevent damage. - ;
Credit was not taken for. operation of either the Intermediate or Scurce Range LChannels-in the accident analyses; however, their functional capability at the specified settings'is required by this specification to' enhance the overall reliability of the Reactor Safety System.
Power Range.-Neutron Flux The Power Range, Neutron Flux channel high setpoint provides reactor core protection against reactivity excursions which are too rapid to be protected by temperature and pressure protective circuitry.
I I
I l
PC4-20b 22 1
i
Q~
4 .
~
)_y y , cs _ g' x ,
N 7 9
'i y y w y LIMITING SAFETY SYSTEM SETTINGS-' .
H
.,. J , .
- l . BASES FOR~SECTION 4.0.2.2-Power Range.-Neutron' Flux m (cont'd)~
ci"'*
, :Below 15-1'2*'of indicated power,; reactor-core protectica is provided.on a' q one-of-two basis,by Power Range channels 5 and 6. The~ active rangc'of ,
ichannels 5 and 6fis-selectable,and provides a scram at'80% of full scale on.
- the. selected rr.4ge.: .This provides redendant protection in the' lower power
> range for:a. power excursion beginning from low power.
LUpon reaching 15't:2* of indicated power, reactor core' protection'is provided:
. 1 -on a 2-of-4-basis by Power Range channels 5, 6, 7 and 8
. The calibration of the four power range nuclear instrument channels is verified using heat balance data.taken during steady-state condicions. Each !
channel responds directly to'the neutron: flux' detected by its, ion chamber.
The:tamplifier gain for each channel-is continuously corrected so thatLthe Loutput corresponds to the power level determined by steam flow.because, changes in' the void content in the steam separators, located in .the periphery
. of= the. reactor vessel, vary neutron attenuation, which causes erroneous nuclear instrument = indication. The continuous correction ~of the nuclear:
', instrument amplifier gain is limited to a relatively slow rate;.therefore, the ion chambers cause,the power range nuclear instruments.to respond directly to rapid changes in neutron flux.
The setpoint for the high reactor power scram has been set at the lowest practical value consistent'with reactor operations and will prevent MCPR 1
. values less than 1.32'during operational transients.
-i Reactor Pressure-High. l
'High pressure in the reactor coolant could cause a rupture to the system process barrier resulting in-the release of fission products. A pressure increase while operating will also tend to increase the power of t'ne reactor by compressing voids, thus adding reactivity.
\
I 1
l
\
L 1 i
i pC4-20L, 23
- _ _ _ - - . i
gg
~ ~
]
w , i i -
;; .s 1 4 < r
$60 '
CUAmiWT j LIMITING SAFETY SYSTEM SETTINGS
. .. 1 JLASES FOR SECTION 4.0.2.2-
] j
[p Reactor Priscsure_-lii_gh - (cort'd)- 4 - Actuation of th'e high pressure scram followed by actuation of the safety valves will prevent the pressure Safety Limit from being exceeded even during
- the maximum pressure transient . Reliance:on the high pressure scram and s safety valves is.a'very conservative-approach because of the following.
considerations ; the high pressure scram is actually a backup;for the high 4 flux. scram and' additional protection is provided by one' or more of the following: ,the turbine stop valve closure partial scram,' main steam bypass Lyalve operation, and shutdown condenser operation. The main steam bypass valve is capable of bypassing at -least 100% of full load' steam flow directly to the main condenser and has a design maximum full stroke opening time of
- 1.5 seconds. The shutdown condenser has a calculated capacity of
- approximately 20% of full load steam flow.
The 6. cram will quickly reduce the neutron flux, counteracting the. pressure increase by decreasing heat generation. . The trip setting is higher than the
-eperating pressure to permit normal operation without spurious trips. The' setting provides for a wide margin to the maximum allowable' design pressure and takes into account the location of the pressure measurement compared to
- the highest. pressure that occurs in the system during a transient. This setpoint is effective at low power / flow conditions when the main steam stop valve closure. scram is bypassed. For a turbine trip under these conditions, the' transient analyses'show a considerable margin to the thermal hydraulic limit exists.
Reactor Power-to-Forced Circulation Flow Abnormal and Reactor Coolant Flowrate-Low i
,1 The reactor power-to-forced circulation flow setpoint is specified for abnormal flow conditions in Figure 4.0.L 2.1-1 of LSSS Specification 4.0.2.2.1.
Safe operating values have been established over a range of operating conditions for the reactor power ' level and forced circulation flow rate. The LSSS curves establish the minimum allowed forced circulation flow rate at a , given reactor power level or, alternatively, the maximum allowed power level at a given forced circulation flow rate. These LSSS curves are a } modification of the limit curve presented in Fig. 4.68 of the final b Safeguards Report. ! j l l I PC4-20b 24 , i 1 a _- ___- - . I
~
~" ~
'c C ' Q:;- L..if , ~ l]
h- N .
, 1 ii. ' 's I
,,',- ;. 3
~
m, g t (f,l , ,
; " a j cggigyf ,
w- m , ,
>l , iLIMITING SAFETY SYSTEM SETTINGS. <
y 1 P IBASES FOR SECTION 4.0.2,'2'
.].
l]Leactor ' Power-to-Forced Circulation Flow' Abnormal and Reactor Coolant Flowrate-Low -(cont'd) 1 >
.., .) % .The bases 6f the LSSS curve are the burnout and flow instability safety .
y margin. Thus'.;the'LSSS curve establishes a limiting relationship between.the i
;y '.j fueljelement? heat' flux'at'a given power level ~and.the forced cirediation flow i
'rather-than a. limiting relationship between actua1Lreactor power and forced.:
circulation flow...Under; steady-state operating conditions, the heat flux and-U reactor' power relations woald be equivalent. 'During transients, however, heat _ flux and power level are not equivalent,'due to the thermal' lags of the oxide fuel.-
- The LSSS curve-has not been extended to forced circulation flows of less than 30' percent of full. flow,-Which is the low flow scram set point and which is .
well'above the hydraulic. instability flow. l 4- Scram' circuitry initiates a reactor scram whenever the relationship between
', , .the power level:and the forced circulation flow rate corresponds: to the IESS acurve. The power-flow ' scram signal is represented b'y the equation:
a P - 2 P 1 48.3 3 1Where P :: reactor power level, .* of RATED THEINAL'IVWER
.F = recirculation flow rate, %'of full flow, 10.75 x 106 lb/hr Thus, a scram will occur whenever the value of (P - (2/3)F] exceeds 48.3.
.i
'A' low flow scram signal is set at 30 percent of. full' flow.
This low flow scram signal compela the operator to establish a minimum flow before reactor startups and it prevents. operation in a low flow range. 1 1 iT4-20b 25 l 1 _1_.n___.______1__ _ . _ )
. , , t ,
,:, i 1, u 5
[? , i fy;7 d ,
, ,.7 , t , egggyT
/ r 1
' JIMITINGSAFETYSYSTEMSETTINGS 9
BASES FOR SECTION 4.0.2.2 Reactor' Water Level-- Low'and High'
.During normal / reactor' operations the zero indicat'ed water level can be' set at any' point'from 2 feet 9 inches to'4 feet 6 inches above the' active core. ,
, .i athe setpoint for the low reactor water level scram 'is less than or, equal .to
~
- 12. inches below the zero indicated water level. . At this value, the water pl '
;1evel Tabove the top of the' active core will be sufficient to assure recirculation flow after scram.: The low reactor: water level scram point was 1 u '
chosen farienough below the normal operating level to' avoid spurious scrams but-high enough'above the fuel to assure that there is adequate. water to account for evaporation! losses and displacement.of' coolant following the most severe transients. The'setpointLfor the high water level' scram was selected far enough abov~e the normal operating level to ' avoid spurious scrans but low enough to prevent release of. excessive energy to~the containment in the event of a LOCA. Main Condenser Vacuum-Low? i .-To protect'the main condenserfagainst overpressure, a loss of'c'ondenser vacuum initiates' automatic closure.of.the Containment Building main steam isolation valve. To anticipate the transient and automatic scram resulting from the closure of the Containment Building main steam isolation. valve, -low
< condenser vacuum initiates a scram.
j .. Main Steam Isolation Valves The Containment Building main steam isolation valve (MSIV) automatically-
- closes at i 1000 psig steam pressure to limit the amount.of fission products released and to protect against rapid reactor depressurization and cooldown v .during a' steam-line break accident. Both the containment Building and
- Turbine Building' isolation valve closure scrams anticipate the pres =ure and )
flux transients.that occur during normal or inadvertent closure of either j valve. The 90% open setting will give the earliest positive indication of
-valve closure.
l l j l- ] l PC4-20b 26 i
1 ! CURRENT LIMITING SAFETY SYETFM SETTINGS BASES FOR SECTION 4.0.2.2 Main Steam Isolation Valves - (cont'd) Operation of the reactor at pressures lower than 1000 psig is permitted only
~ when the reactor'is starting up or shutting down with steam flow rates
< 100,000 lbs/hr. The low main steam pressure channels are bypassed during l this time to prevent automatic closure of the valve on low reactor pressure.
The containment Building main steam isolation valve is also automatically closed on either low condenser vacuum or low reactor water level. In any event, closure of the Containment Building or Turbine Building main stean
. isnintion valve will initiate a scram and will also initiate operation of the shatdown condenser. These actions anticipate the pressure and flux trancients that occur as a result of isolation valve closure.
Closure of the turbine stop valve initiates a PARTIAL SCHAM. This scram anticipates the pressure, neutron flux, and heat flux increase that would result from rapid closure of the turbine stop valve. Transient analyses indicate that the reactor power would go to a peak of 260% of RATED THERMAL POWER if no credit is taken for a period nor partial scram. Because of the Doppler shutdown effect and the time lag in heat flux, the fuel cladding Safety Limit would not be exceeded. The opening of the turbine bypass valve (for which credit is not taken) prevents the continued buildup of the reactor vessel pressure. In the event that the bypass valve does not open, the operation of the shutdown condenser is initiated at 1325 psig and thereby prevents further pressure increase, as would safety valve operation should pressure reach their settings. , l Control Rod Drive Accumulators j In the event of significant loss of oil or of gas pressure in any of the 29 control rod drive accumulators, the associated control rod will lose
. capability for scram. The safety system incorporates only a PARTIAL SCRAM !
for low oil level or low gas pressure in any accumulator, because the control rod will scram even at the inoperable accumulator scram setpoints. The capability for an electric motor-driven insertion will also be unaffected. l PC4-20b 27 i
_ = _- . , - - - - - . _ _ . ._ .- - u --- ac _ I / M, LIMITING SAFETY SYSTEM SETTINGS L, BASESFdRSECTION4.0.2.2 Bus Voltages. 1 a> .. The electrical distribution system is. fed from two 2400-volt main buses. Loss of- one bue causes the loss of various vital' systems ~ and equipment. ( - Since' the loads are divided ~ to' provide sufficientLequipment for the. resultant.
- -- HOT SHUTDOWN condition,'a PARTIAL SCRAM is initiated if one. bus is lost for a time longer.than.that' required for. automatic transfer to reservezfeed. Loss- I of both' buses' for that'same' length of time, however,1 causes
- a full scram. A time delay undervoltage' relay (to allow: transfer of feeder breakers) is connected to' each :2400-volt bus- to initiate the. PARTIAL SCRAM if either relay is de-energized. If neither bus has ' sufficient voltage, . the redundant full' scram relays in the rod control system are'also de-energized by--the same
. undervoltage relays.
Loss of. voltage on Containment. Building MCC-1A'or Turbine Building MCC-1A-means. loss of electrical power to various vital systems and equipment. Loss of voltage or undervoltage on either. motor. control center for a time longer than that required for the reserve feed breaker to operate initiates.a full scram. . Redundant time' delay undervoltage relays are designed so that-the two ] normally ' opened contacts.of each de-energize the full scram relay in.the ; associated full scram circuit.- The time delay.is set for approximately 1'second longer than the time.that is'normally required for the reserve feed
' breakers to operate automatically.
l t i l l PC4-20b 28 _ - ____- -- - _ a
- m. .s ,-
, ng 3 r r !
.a
& s.
+ W QV , p g. , '
> ) r <
1..'-, , 7 , y (;f ' ' i l < }gic; y' ' " ' 'l ' j ' f, s-n, , CMMNT d t
. ~.
% , , 1 iLIMITING SAFETY SYSTEM SETTINGS' I TBASES FOR SECTION 4.0.2!2'- -
l q m 5 i . ;.,
- yr .,
7J ' ' "g .,
REFERENCES:
i '_ t i- .. . . . , , . . f
' l 1. : Galbraith, L K.L and ~J. Jaech. ..'.'The XN-2 Critical Power Correlation,"
' .XN-75-34,'Rev.,1. August 1, 1975. ;f
,< 3 4
- 2.i L.iH.Esteves,. A.':M. ' Sutey and D. E. Fitzsiassons, "XN-1 Critical Heat [
. LFlux Correlation for Boiling Water Reactor Fuel," JN-72-18,
', ,. August: 1, 19722
, 3
- 3. 'E.'Janssen, "Multirod Burnout:at' Low Pressure," ASME Paper 62-HT-26,t .
-l August 1962.
. 4. ; K. M. Becker, "Burnr .t CorAitions for Flow of Boiling Water in. Vertical Rod Clusters," AE-74, Stockh91m, Sweden,lMay 1962.
, r 1: >
s e C y a ' Delete Bases for Section'4.0.2.2. 7 'JfBTIFICA TION
.r-Since Section 4.0.2.2 is being deleted, bases should be deleted.-
l. L i I .e l PC4-20b 29
)
- I : - ,it ,' i :); .,
.@[ [.)isw,.~i T ' ' . 9 ;i ,,
w , , . ,.. , , e ., , .
# s. 1 W,
~
' .J ; -N J ' n;;g a.i:< v ' , .
o, " s. f,:-_ , ' f;, ' _ . ,_ , i e 4- ,
- 9 .:4:
, I a.
4 ppu ,' , i
, i .. I 1 b ;. . 1 er, o j I f I Y' ,
h i ,"
' . <, hl '
+
A ( m.- i (( lt-'
-,,i,s
']j .t 9
( / l 1, M. 24
.. > .G; 'l s
.. n ,
+
y , .., .3 c. L
; oe i
;4 kj-S ;; s
,j s-
.,.. ?f
*4 ti it .
.' ' -( 1' I !
q)
.s <
\ j .i i y
; i,.
1 ,. 47.. "y <
'(TMs i page isiblank Intentionally.)-
.$ j .I '-) '
.. / gl
' 1
((' s o e > y,'
, ,4
.yc ,
, $i B _ Del _et ed i b , t
.. { 4 /1.1' '
' t y
L :4;1.2
.wt k I'. - _ i
; PRORMED ~
.s Delete this page.
JUSTIFICATION This page being blank. and the deleted sections will be incorporated into - l proposed.Page 271-27rr.' ' J i
.l l
i
)
P PC4-20b 30 1
. . 1 y 1 iws s
)
w -_ _ ---- _ - - - - - _ . - - _ - - _ . _ - _-- - - _ - - :J
m . .
'!l '
w
$', q , .
PAGE 28
' Cin9%NY 4.1.4 Whenever the reactor contains one or more fuel' elements, any-
. operations.from points outside the control room of equipment which'may' affect.
the reactor shn11 be" conducted .under the direction, or with the ' knowledge, of
'J .the' control' room operator.
14.1.5 XIf 'the; plant is operational during a tornado warnirig, the Shif t ' Su'pervisor. on'daty shall keepiinformed of the actual tornado activityL which may , approach tho plant. . 'In the everst that ' reports ' indicate ant imminent'
; tornado strike. atiorLnear the LACBWR plant, the Shift Supervisor sha11' reduce 2
reactor. power to a' level'which permits prompt reduction of power generation-to^ station load. However, the. Shift Supervisor shall be instructed to
- discontinue: plant operation if, fr. his judgment, this action is required f.o
, , ensure plant safety.
4.1.6 :If the plant.is in CONDITION 1,.2, or.3'and.the Mississippi River.
~ level adjacent to the plant reaches 639.2' and is predicted to exceed 640', j
. commence reactor. shutdown and be in CONDITION 4 prior to the river level exceeding'640'.
1 l-4.2.1.1. CONTAINMENT-INTEGRITY shall be maintained in Conditions 1, 2,3 and during;
- a. CORE' ALTERATIONS,
- b. Handling of irradiate.4 fuel, or I $.,
- c. There is fuel in the reactor and any control rod is withdrawn- .
}, IRNGSED Delete Sections 4.1.4, 4.1.5 and 4.1. 6. Modify Section 4.2.1.1 as follows: 4.2.1.1 CONTAINMENT INTEGRITY shall be maintained during handling of irradiated fuel. JUSTIFICATION The portions deleted no longer apply with the reactor permanently shut
'down and the reactor vessel defueled.
i
. PC4-20b 31 l
o
.=, ,
J
- n : t' , ,
q
. PAGW 29 { i to l
's ~ ' .
cgggg#r 14.2.1.9 The containment building shall be isolated whenever.the spent fuel
~
~
storage well contains. irradiated: fuel which has decayed ~1ess tinui 43* days after exposure in .9 critical' reactor and a shipping ' cask for irradiated-fuel is bef ug moved by; the crane on. the 701-foot level'or located within one cask 4 length of the top,of the spent fuel storage.well or is within the spent fuel
. storage well,' . ' During cask movement near or atL the FESW. the water level in:
'the FESW must be at: 1 east 16Lfeet above the t0p of the' fuel storage-rack (no J
1 ' core than 17 feet' below the top of. the ' FESW). - i
~
"* 43 days for off loading less than' ene half of the core, i.e., less than 36 fuel elements.
51 days for off loading moie than 36 fuel' elements. f Ph0R2SRD ; g Delete.this section. JUSTIFICATION ' The~irradiatsd fuel has already decayed for more than 51 days;.Section 2.12.5 states that water shall be at least 16 feet above any fuel stored in the spent fuel storage.rachs. Since fuel must be stored in the upper tier for.all fuel to be 6 accommodated, the last. sentence in Section 4.2.1.9 is redundant to Section 2.12.5 and therefore not necessary. 1 PC4-20b 32 lI
m ) PAGE 29e S(UNENT REACTOR BUILDING' CONTAliMENT VENTILATION DAMPERS LIMITING CONDITION FOR OPERATION 4.2.1.10 The containment ventilation inlet and outlet dampers shall be OPERABLE with isolation times of less than or equal to 10 seconds. APPLICABILITY: Whenever CONTAINMENT INTEGRITY (Specification 4.2.1.1) is
. required.
ACTION: With one or more of the above ventilation damper (s) inoperable: A. Operation may continue provided that at least one damper in each affected penetration is maintained OPERABIE and either:
- 1. The inoperable damper (s) is restored to OPERABLE status within 4 hours, or
- 2. Each affected penetration is isolated within 4 hours by use of at least one automatic damper secured in the isolation position, or a blank flange; OR B. Be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the following 30 hours unless the affected penetration is isolated.
C. The provisions of Specification 3.0.4 are not applicable if the affected penetration is isolated. SURVEILLANCE REQUIREMENTS __ l 5.2.1.10.1 The ventilation dampers shall be demonstrated OPERABLE prior to returning the damper to service after maintenance, repair or replacement work is performed on the damper or its associated actuator, control, or power circuit by performance of a cycling test, and verification of isolation time. 5.2.1.10.2 The isolation time of each above damper shall be determined to be within its limit when tested pursuant to Specification 5.2.2. 5.2.1.10.3 The seat rings of the ventilation inlet and outlet dampers shall be replaced at least once per 5 years. PC4-20b 33
y o e CORREWL. BASES The Nuclear Regulatory Commission requested that similar. Technical Specifica-
. tions per Generic Item B-24 ' and NUREG 0737 Item II.E.4.2 be submitted to help assure operability'of containment ventilation dampers. DPC had previously committed to' replace the containment ventilation inlet'and outlet' dampers' resilient sealing material.at least once each 5 years until such time as ;
> additional in-situ data an be accumulated to justify a longer interval. If in-situ data is accumulated which supports a longer seal replacement inter-vals, a change to Specification 5.2.1.10.3 may be requested. Specification 3.0.4 is not applicable .if the affected penetration is isolated.since the
. safety function of: the dampers is to close.
PRt HtSED
' REACTOR BUILDING CONTAINMENT VENTILATION DAMPERS
' LIMITING CONDITION FOR OPERATION
. i.
4.2.1'.10 The containment ventilation inlet'and outlet dampers shall be r OPERABLE with isolation times of less than or equal to 10 seconds. APPLICABILITY Whenever CONTAINMENT INTEGRITY (Specification 4.2.1.1) is required. ACTION: A. With one or more of the above ventilation damper (s) inoperable, suspend fuel handling or isolate the affected penetration with an automatic valve secured in its closed position, or with a blind flange, within 1 hour. B. The provisions of Specification 3.0.4 are not applicable if the affected penetration is isolated. SURVEILLANCE REQUIREMENTS 5.2.1.10.1 The ventilation dampers shall be demonstrated OPERABLE prior to returning the damper to service after maintenance, repair or replacement work is performed on the damper or its associated actuator, control, or power circuit by performance of a cycling test, and verification of isolation time. 5.2.].10.2 The isolation time of each above damper shall be determined to be within its limit when tested quarterly. 5.2.1.10.3 The seat rings of the ventilation inlet and outlet dampers shall be replaced at least once per 5 years. l l PC4-20b 34
0 .
.;)-
te a, - D 1 PRWtWED BASES The Nuclear Regulatory Commission requested that similar Technical Specifica-
' tions. per Generic' Iten B-24 and NUREG 0737 Item II.E.4.2 be submitted to help assure operability of containment ventilation dampers. The ACTION statement has been modified due,to the plant's permanent shutdown. DPC had previously committed to' replace the containment ventilation inlet and outlet dampers' resilient sealing material at'least'once"each 5 years until such time as additional in-situ data an be accumulated to. justify a longer interval. If in-situ data is accumulated which supports a longer sea 1' replacement inter-vals,.a change to Specification 5.2.1.10.3 may be requested.- Specification 3.0.4 is'not applicable if.the affected-penetration is isolated since the safety function of the dampers is to close.
JUSTIf1 CATION The current action statement relates only to plant operation and therefore does not apply. The test performed under Specification 5.2.2 is done only prior to cold startup.and so would not be performed. Therefore, the proposed change to the ACTION statement and testing frequency are more stringent and make sense under current conditions. l PC4-20b 35
p t. N' ,
.PAGE 296 CUh' RENT :
l'
- 4.2.2.2 The reactor coolant shall be light water and shall conform to the fellowing requirements.
CONDITION'l Normal Limit Maximum Limit l Chloride concentration .2 ppm .5 ppm p11 5.3 - 8.6 N/A Conductivity. 3 umho/cm 10 umho/cm The time spent above 3 umho/cm at 700F - 80oF and .2 ppm chloride should not exceed 72 hours per incident nor 2 weeks per year. If etcher time limit is exceeded, an orderly shutdown shall be initiated within 4 hours unless returned to within the limits. When the maximum conductivity or chloride limits are exceeded an crderly shutdown should be ititiated immediately. If the pH is outside the limits for a period of greater than 72 hours an orderly shutdown shall be initiated. 00NDJTIONS 2 & 3 Normal Limit Maximum Limit Ch'ioride concentration .1 ppm N/A pF 5.3 - 8.6 N/A 1 Conductivity 5 umho/cm N/A The time above 5 uhmo/cm at 700 F to 80o F and .1 ppm chloride concentration is restricted to 48 hours for any single occurrence during Conditsa:. 2. When this time limit in condition 2 is exceeded the reactor shell be brought to the hot shutdown condition (Condition 3) until the limits are restored. If the limits can not be restored in an additional 7 days, the reactor shall be taken to the cold shutdown condition '.00ndition 4). CONDITIONS 4 & 5 Normal Limit Chlorld.e concentration .5 ppm pH 5.3 - 8.6 Conductivity 10 umho/cm The primary system chemistry parameters defined in this section shall be determined ec seast once every 72 hours in Condition 1, 2 and 3 and at least once every 7 days in Conditions 4 and 5. PC4-20b 3G _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - 1
_, e > > b: i .U , 3 f'f P l
-l
,' I 4' t 'f, t i i Y. (
-..f. 4 1 :
t ' ' 5 , i
< -7'
- 't,'1, f ]'.
~:i'-
> t R.t 1
{ '. f , l , _
;p ,
..]
- 3. .
g y ' o i 4.2.2.2! The ' reactor coo $t shall': be light water.' and shall confone to the 4c -
< - following requirements.'
..n l' h '. ' ,,:
a 9 , , CONDITIONS'4 &'5- Normal Lin'it H i' + [ -jchloride. concentration
.5 ppm.
l , . PH 5.3 - 8.6 y., Conductivity' ;10 Iumho/cm ". r . ,
'The primary system chemistry parameters defined in this section shall be determined at least'once every 7 days.in Conditions 4 and 5.
t- ,
;y JISTIFICATICW
!r- -
. The. portions applying to Condition 1, 2 and 3 are deleted since they no longer apply. i
.c , <
f h i l PC4-20b 37
PAGES 30-30f
~ CURRENT l
REACTOR COOLANT SYSTEM
' PRESSURE / TEMPERATURE LIMITS I REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 4.2.2.4 The reactor coolant system pressure, measured at a level above the normal operating water level, and temperature above RTwor shall be limited in accordance with the limit lines shown on Figure 4.2.2.4-1 as Curve No. 1 for ,
inservice hydrostatic or leak testing; curve No. 2 for heatup and cooldown when the core is subcritical (except when the reactor vessel is vented); and Curve No. 3 for operation with a critical core (except for low power physics tests ur when the reactor vessel is vented) with:
- a. A maximum reactor vessel heatup rate of 1000F per hour,
- b. A maximum react >r vessel cooldown rate of 1500F per hour,
- c. A maximum difference between the reactor vessel flange temperature and the closure head flange temperature of 50oF,
- d. The forced circulation loop pressure:
- 1. Less than or equal to 280 psig unless forced circulation loop temperature is at least 130eF.
- 2. At atmospheric pressure unless loop temperature greater j than 70o F. j a
APPLICABILITY: At all times. ACTION: With any of the above limits exceeded, restore the temperature and/or i pressure to within the limits within 30 minutes; perform an engineering determination of the effects of the out-of-limit condition on the fracture toughness properties of the reactor coolant system; determine that the reactor coolant system remains acceptable for continued operations or be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the i following 24 hours. l SURVEILLANCE REQUIREMENTS 5.2.2.4.1 During system heutup, cooldown, and inservice leak and hydrostatic testing operations, at least once per 60 minutes determine: 1 1 pC4-20b 38 I l i
- 1 Li. . .J
w7;gg - O,s
; g. .
.t- ,,o ,
1 - + x- m 3 g:., < :., a Q9 i L
; 2,
, n y e (- <
. . . , ,; . , 9 , ,
i f) , ._ < ;4;
*;[ : ,
- ', ; n;..- .
i .f . ~ (
. ..kr I 4 * .I g .
I, ,'h 7 --:,.,.'a
~
4 p:a + c ' ;, . . c(Algg pr ' r.: ? ~ ~ ): ,$, ,::,..4 .' REACH 0RCOOLANTSYSTEM' I ? { 'l 03 , , .;.L' O. I i- ElmVETLLANCE REQUIREMENTS -'(cont'd) !
, 9 g i.f n ' a.' .The reactor' vessel'to.be within the limits of LCO 4.2.2.4.
. bi ' Thelforced circulation loop' temperature to be within the limits of:
. LCO 4. 2. 2. 4 '.-
1:
', :15. 2. ... 2 4 2 'The reactor material irradiation surveillance specimens shall'be' Y.y.y'
. removed andaexamined'to determine changes in material properties'at the y , intervals-shown';in Table 4.2.2.4-1. The results of:these examinations shall-
' be' used to update' Figure 4.2.2.4 -2, 'and .as a factor to predict vessel
' lifetime., .,
/,' _
t 2I: I
,s
?.\ ,
ia.. t t 4 ( l
' PC4-20h 39
--.m___m. .______2._.-ri___2._m
's
} -
_Z_E t r_ : - Z __ tF x' _ __ + x 2 r__ x:
- Z
_ Z_ i!y7
~
g1Rb_ T. MY_ _ ~
~ .~ _ . - _ _-
V ~ _ ~
- . _~nr ._ -
- . .~ l _ _-
ut -
- ~
- - --i - ? h. n* " . .~ _
i a.ip_- c.' .. _-
.L- - _
0 u _.ti_ - _x: _ s .i. l - _
,0 y {i L.T-rt **I ,tr.1 - b. .
T ! k ,r .-,..~ T - [ h : i _ fn l h r . r.t. C- :
- (( , ;
.l l
1 l" _ V
.:~
4.+4 t'~ 1 :- _ _. - - _ Rr _- E t1 : : - . - _ __ - - . o _: C mt.l & . _. - N l2T-:
~
~ uc ;
0 E _x r.x: ,_ '
~
ie ma _: 6 1 R E n R . - _ F s_Z_ _ Z _ _ - M i M r E R
; _Z_-
M__ _
- ,- M _
Z _ - f o _ _ Z T DE s: H V
- : J .
T' - 3 e 0T O
- /
r ZZ
- : B 4R
- - - _f _ . u . _-- 1 A W
i^ - o t 2 - I)_g_ZZ e ) N a ._ r
. x .
o v E R T D l
- Z - 2 -
. M .f
) . Z ~
e e v p n _ b U N T T I ; : : 1 - _ - _ s - - ~ r m
- - A Z . _ _ - _
- - - : - . . _. _ A R f
- f. /'
~
u e ,
~
- 0) R (
C T E 7 2F J l+ - . - _ - - ~ _.x;
. - ' . _ : . .t- : 3 : : - P E 1*
1_ Z . . Z _ Z
~
___ M R
- - - : i n (
_ - - . _ ~
- Z- . - ? _.
U
- i - ::M -
x_ Z - -
- i
_ Z
~
e E T T
- :: M N,x x. 2 e
- r u E R A
..Z x- _ - .xJ
_ . _ _ : _ ~ t L E x:f 2 _< x _ _ ~
- . - _ , _ _ _ ~ 0 a B P 0 r A M
- x: m.
1 e WE f- g__ j
- - _ - . p O T x ~
L m e L N T'1+T ~ t :
.J
.~ T AO I P
-y- _~~
x 1
- Z -
- _ e. UI MT
- .~ 0l
.~
Z l]:E
- - - : _ . 8b MS N 3:!i a I
__. / Y w N A l . Z j.I n x . . - _ - . - o I R i ;] !
- I l M T t1/:- i - - _-
x
- l g,
- E j : .
A .
/
l t" Z jh
- - . t h !
l a c x
. m E
S d i - : 0 u S
.k
~ n t --- ._ m_dn .
_.. . x Z
~
6 m E V V e
.kr ;:
V . . _ _ - _ _ i a t R r lI. R r+ . x r T: _ - . . n Z_ N u p e + ;,_3'- R
.d[.c r:N..
. i
,i m ss .
m 5 . r _ - M O rtU u b _. ti u . t . u t u +: p2: .
- __Z
_ _ b _ _. T n : b C m e _
- ,- .tI- . m a S : t. i j_ x :
i nP r . . in le . _ 6 . {3 _ 0 A E b . l r t4. : . b-.a4 el R i Mr o l . i M r t o .d n 2 .' _ 1 Q.T..
._ x _ 4
.h'x r
o c
~ _
- b c
x ZZ
- _ . . _ ~
f - : b : x: nn _ _ _ ~
- - f a _
i Z . e _ .. _ _ ~ b x e .Np.x. :. 1 2 e R __ r u i
. u r ,
- M. .
b b i
- Q _
ot . ot n :: ; . . . 0 1$2
<n:._
Na i _. _ _ _ _ r
..x e e su h _. r. g M N arowa __ x_:_:x. :, _ _
vpt I .v pleed- _ _ b .:, rms rmo- ._
. CTC- ueo-M.__
' uCTT e e r {4.d.M g: _.
_ _ .n
- [
_x b i
.- .x - _ -
_. .X _ i*_ :."Z .
- xb _$
h 0 0 0 0 0 0 0 5 0 5 0 5 5 2 0 7 5 1 1 1 - 2 3 - _ T E * ,,g,, ge[c
. . , r 3,m >i E.x . _ no e
- siao t 3
~
l
1
. CURRENT-t{0TE: 'PER SE DATED MARCH 22. 1985, THIS CURVE NOT APPROVED BEYOND 15 EFPY.
1
,1 400 I mr ' U.i,fg_
s:id.i. 8.n. . .i.l..i.i.i.i..ii.i.l.i ji.;i.i.s. ;j f. s.-d...ig gl.i.ii.p. :e.. . : i.i:!.!.p. . : . .:s o. .!!.i.. .ii.ii.iiii:::
. . .. r?}is: l p.".
- ;J1: : 't.7*- ~
*:
- p; !
2***j'r*
.a ..J.J . . 'u,*
~ . *I.d.g,IU e.* ....
', .:.:.s
.g". .'*{..l._J
_ ' ,t ",.:t .. . .
,. 1 f::
.p.. t. . . .m .- r .".'p . . -. 1.*.
.-. ...~.:l;p"7. ~D T. .* g.=L.;;;;
l DN J E-W WH:si.!!!: ::iG@= g i *, j .' ?!i f .@ i.y:' ut. 8
%]
'J g -.i y<e 'I#1'rh '
4 c:.
- ]9)5 1
h.p!i
. e : :. d. .fi.f. .ggw.
i
!!Ji@y.s h. p#1p m m.
~
.i.,
7 :m :m Ms#1pi}i
.pm= =
e :.. : ~.cs . jN.: : .x.n :ru 1 . t.p.t.rq:. a
- ru &,, errs .ne~::n. .,. .
,j:
a t . q: me"er..:;;
- a:u .3: rg i .: c
;r:;. 5: p:: :.: .a: m m--.r:::: 4.4 .n-njj,l
- .t i.u ;r.:
- 1 3; =: n.
~; .
- H.;. a :=!::- .r.rr t.. .
4 r: :3 nun!:r: asa. :.
.a=.= :. m- . .
- .. .
- :n :. n}; p,;'r .p:.: - -
p== *: "." I 200 ' .+. ".%r P. 'a. n.': ':.".P. Nl.
. . . . . .e. nc. . . . a.t. :. .
. =. 6. . .".in..
. :. a.7.. . - i y m . ---. ,.;
.:: .rJ r...; ...
c.:.. oJ;,p- n Ae r--, .:.. .-: g;r .4p, !h
; i 1 h--er- :.:. ... m- .. :n: --::.- t .p:S -- - . m- :. .::
-. .,:- -- %..g n< !.:j,ih;--.
'.T "- o L.r F.7 u:: . .-2 :n :
- u. -t --
e . r.i:,
. 7.s+:'.m=.*:
r.-:
-+
n::
"r :O. -*.6
. ,&s r - .,,, .
m;
;;.:: =.
+ me, -
9---" ;;rWC 7 **++'m-t-- * - - _]
. ! . . h Pressure Vessel Plate (b)
%=. .;*m ,:a r- f- *w+ w t-tv. ut+-' -- -
%.. n=* ni. o - -- -
- , , m m nin..... l
~it..-i.....n..... ~~- +..-"
. ~
o 2l: 100-.:apr.~ r.z,rg .g ==- =r :::g:;n ;,p:.;.; g..s]=e= u::::=a=:g.
... ,q , ;
m 1.~.4.
-- ,--~-r .-
. __ _.g :t Iu Qll
-.m .
. .. = , .
..,=.....je .u:==-....:,_.
-.t-__
- -r ::r.t:::;::::r.- --,_ ,j. c=. ::: .i::
80- . * * * -r ):* -
=r- =::n_%fr-L:: ... ;. : _:-.:::: a " '=:r- ' .-
.=:u: -- u.= := == We1d Meta 1 --
d- _::w--
.u.: :::r r :c;2 -;
n ::.: --
- .n:::: - - -
= -- -
= = F- - . - -
=&::
.: *. :.- .-___.._ , -.. - , . . :.n:::' =..._ :.-.__,,--- ..-%~- f. -
. .o. E:g --
60 - - - -- - --- ~ . .- = ;;";: .
- . :l.;.- ..
a - - -- - --- -
- ,. . MmsWsm th w W .* Mt~it -rs.es.v. n m= a ====. w - =-P.= N-sm-m I
!A W:E M fIE.:N3tr.N -115?'ENSh E d t -N.M'5 ~ :
1:~=IECM' :M-WEES@E $$ 05 5-E W=_.~fE:i:es E=f. ".2 1F == 3'- g ge- G z == ="4 s . ^~~ ~~7"-~~=: fiME2
== ..
c- _ - - . . - ,_ __.
- a. 40
==^ - . _ . .* = ' =
- m. =-:. : -
.. =.
-: - +
_ ,.. .} .- .u.- s- - m .- -:_._ .
- m. m.. -
-.4I
-j **n! -e:+2t::::
g +.-+ . . .. .. 4 - , --
==d t .= -* =*-. **
g .z:: :.e __.m
-s
- m::@. ::-b _--@ . ==
C U'?E:~ % iIL _..; M EE; Esp t.I .'.: .:ju.: 1*W., FEE r - ._=;; =um=_s,,y, : J!i: =: :RE.:=C:=--REEriMs"M@~.:i=:
- M.n=~S'S.;.. - - -.. _ e. m c
,.lp..:--
- t. E~-f@@2:r' . - :: .: ;
n-
.r. ::==.,.z.
.- : .j
.. .tu ---L=.=w.,.__.
_ _ 3. _._7.=.c:w.=--.~. A- , , . _ _ , .---
+::;:== .u.:.:_c _.=: r u. =- ...-. .F. . --===
. ~ r.. :q
_= .r_.
.: :p*:. . ,' 1 n. .:: ;w. :u m -m. e .=+ n .:__.m._r_._,. =- - -
.m . :
= -- . . . . . .- - -
ur:=.p = .:- _=n.w=::.===.::
; =: . - a.A: .: = up:::. .u :::
- 20. : . ..T "= t. i :.=. . .. . . . ,
=.... .:
= . . =- . .= .v.. _ 2.=...r_ . =_. "_ . _ ,_ .e.r. = = = .:::w t =u . _ "._ _" j -=: 4. . :=:.
. . _ . ==. . . "
1 2 3 5 I 10 20 3'O 50 Effective Full Power Years at 165 MW t (a) Weld Metal in pr, essure vessel belt-line region. (b) Material of pressure vess'el Plate NP-1056. (c) Minimum RT NDT for vessel pressurization is the j higher of the two. PREORTED REFERENCE TRANSITION TEMPERATURE VS PLANT OPERATIONAL. LIFE f!GURE 4.2.2.4-2 PC4-20b 41
.e l i CIDlWENT TABLE 4.2.2.4 REACTOR VESSEL MATERIAL IRRADIATION SURVEILLANCE SCHEDULE REMOVAL SCHEDULE SURVEILLANCE CAPSULES i Full Power Years <a) Typ.e Quantity l.4<b) A 1 B 1 2.5(c) A 2 ) I B 2 VW(*) 2 l 6 (d) A 2 B 2 10 A 1 B 1 l Spare a 2 B 1 , NOTES (n) One full power year equals 60,200 mwd <t> Withdrawals to be made during the nearest scheduled refueling outage. (b) Capsules removed during August, 1972 outage. (c) Capsules removed during May, 1975 outage. (d) Capsules removed during November, 1980 outage. (e) Vessel wall dosimeters. PC4-20b 42
U.k /4 $ f , '
'j j
b . s
% yf j;. ,
e3
. + BASES 7 14/5.2.2.4~ PRESSURE / TEMPERATURE LIMITS
?All components in the reactor coolant system are designed tol withstand the i effects of cyclic' loads due to system temperature and pressure changes.
~
These cyclic loads are; introduced by normal load transients, reactor trips,
-and startup and shutdown operations. .-The various categories of load cycles used for design purposes are.provided in Section 4 of the Safeguards Report.
During heatup and cooldown, the rates of temperature and pressure changes are o > limited so that'the maximum specified heatup and cooldown rates are i consistent with the design assumptions and satisfy the stress limits for' cyclic operation.
' ' :During heatup, the thermal. gradients in the reactor vessel wall produce , .
- thermal ' stresses which vary from compressive at the inner ' wall to tensile at the outer. wall. These thermal induced compressive stresses tend to alleviate-
' 1 the tensile stresses induced b'y the internal pressure. .Therefore, a pressure-temperature curve baqed on steady state conditions, i.e. ,' no thermal
. stresses,. represents a lower bound of all similar curvec for finite heatup rates when the inner wall'of 'be vessel is treated as the governing location, q The heatup analysis also covers the determination of pressure-temperature limitations for the case in which the' outer wall of the vessel (4ecomes the L
controlling location. The thermal gradients established during beatup L . produce tensile stresses at the . outer wall of the vessel. These etresses are i additive ~to the pressure induced tensile stresses which are already present. The thermal induced stresses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the time along the heatue ramp; therefore, a lower bound curve similar to that described for the heatur of the inner wall'cannot be defined. Consequently, for the cases in which the l outer wall of the vessel becomes the stress controlling location, each heatup i rate of interest must be analyzed on an individual basis.
.l The heatup and cooldown pressure-temperature limit, Curve #2 on Figure 4.2.2.4-1, is a composite curve. It was prepared by determining the most conservative case, with either the inside or outside wall controlling, for <
any heatup rate up to 1000F per hour or with the inside wall always- j controlling where the cooldown thermal gradients tend to produce tensile stresses while producing compressive stresses at the outside wall for any cooldown rate up to 1500F per hour. 1 l
..The reactor. vessel design specification required that the reactor vessel be designed for a maximum heatup and cooldown rate of 1500F per hour. The 1- reactor vessel analysis for thermal stresses and fatigue indicated that the ;
design heat up and cooldown rate of 1500F per hour would result in an I excessive fatigue usage factor for the closure head boltings. However, the analysis showed that the cumulative usage factor for the bolting is satisfactory when the following limits are imposed: heatup rate not exceeding 100o f per hour; cooldown rate not exceeding 150oF per hour; and difference in temperature between the vessel shell flange and the closure head finnge 1imited t o 500F. pC4-20b 43 1 1 j _____________-A
i CWlRENT . REACTOR COOLANT SYSTEM HASRS P_R_ ESSURE / TEMPERATURE LIMITS - (cont'd) The reactor vessel materials have been tested to determine their initial RTw o r . ' Reactor operation and resultant fast neutron, E greater than 1 Mev, irradiation will cause an increase in the RTwor. The pressure / temperature limit curves on Figure 4.2.2.4-1 are used with Figure 4.2.2.4-2, the predicted RTwor change over the operational life of the reactor pressure vessel. The vessel plate NP-1056 will control the primary system HTwor for more than 15 EFPY of operation because of its high initial value of RTwot. The actual shift in RTwor of the vessel material will be established periodically during operation by removing and evaluating, in accordance with ASTM E185-79, reactor vessel material irradiation surveillance specimens. These specimens are installed inside the thermal shield of the reactor vessel in the core area. The neutron spectra at the irradiation specimens is accelerated by a factor.of about 2 over the adjacent section of the reactor vessel. Figure 4.2.2.4-2 must be revised when the delta RTwor ' determined from the surveillance capsule is different from the predicted delta RTwor. The ' pressure-temperature limit lines shown on Figure 4.2.2.4-1 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50. The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in Table 4.2.2.4-1 to assure compliance with the requirements of Appendix R to 10 CFR Part 50. I l l PC4-20b 44
<!- , ,:s <
, t , , , w;
_ {: i it
'O:, ,,
, t :,.
- p
.Ph. llNMED n
' Delete Sections 4/5.2.2.4,' including Figures'4.2.2.4-1 and 4.2.2.4-2 and 'l Table.4.2.2.4-1.. ~
t. JEETIFICATION
+<
+ .Since the reactor.has been permanently shut'down and-the reactor vessel' Vdefueled, there is no need.for' vessel heatup, cooldown and temperature
. limits nor is there a need to continue to examine vessel specimens.to.
predict vessel lifetime based on : irradiation effects. j; .e .. t i PC4-20b- 45
7 , . . .. . . . .
'U i
t '; . { , 4 1
.' ) .- ggg y ;.
. E@MWT
- REACTOR COOLANT SYSTEM SAFETY VALVES-LIMITING CONDITION FOR OPERATION
- 4.2.2.5 At least two reactor coolant system safety valves shall b- OPERABLE A with lift settings within i 1% of the set pressure:
-I' Safety' valve with a set' pressure of 1390 psig, and' s -1' Safety valve.with a set pressure of 1390 or~1426 psig. {
E < APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3* . l ACTION: With one or both;of. the above required reactor coolant system safety valves
. inoperable, either restore'the inoperable valve to operable status,.or be in
. at'least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the' next
'24 hours. -1
*lixceptiduringperformanceofprimarysystemhydrostaticorleaktests.
1 SURVEILLANCE REQUIREMENTS 5.2.2.5 The reactor coolant safety valves shall'be demonstrated OPERABLE by
- verifying the set pressure in accordance with the schedule and requirements
. of Section XI of the ASME Boiler and Pressure Vessel Code, 1974 Edition, with Sunmier 1975 addenda.
PhDivsR11 Delete this section. JUSTIFICA TION . This specification was applicable only during Conditions 1, 2, and 3. i PC4-20b 46 a-___- -_
l l PAGE 30h l C(MRENT 1 REACTOR COOLANT SYSTEM BASES L 4.2.2.6 SAFETY VALVES 1 The safety valves are designed to meet the requirements of the ASME Boiler I and Pressure Vessel Code. The reactor primary system overpressurication j protection is sufficient to limit the pressure within'the pressure-retaining - L boundaries to less than 1540 psig, which is less than 110% of the vessel , design pressure of 1400 psig.
]!
The safety valves have a minimum stamped relieving capacity of 294,612 lb. per hour at a relief pressure of 1390 psig and 302,160 lb. per hour at a relief pressure of 1426 psig. Three safety valves are installed. The relieving capacity with one valve inoperable is sufficient to limit the primary system preasure to less than 110% of the vessel design pressure 4 during an abnormal transient with the highest pressure, which is the MSIV closure. A high pressure scram is initiated at 1325 psig and no credit is taken for the MSIV closure scram signal, the power-flow scram, the overpower l scram nor the pressure reduction due to autofoatic operation of the shutdown ; condenser heat sink. ' I During the postulated most limiting pressure transient, caused by a turbine j trip without bypass valve operat. ion with full scram on high power (120%), l reactor pressure would not reach 1390 psig, the lowest safety valve setpoint. ' The safety valve function is therefore not expected to be required under the most limiting operational transient. ; The testing frequency applicable to the safety valve function is provided to ! ensure operability and demonstrate reliability of the valves. The required i testing interval varies with observed valve failures. Setpoint drift within
'1 3% of the setpoint is not considered to be valve failure for the purposes '
of this test schedule. Setpoint drift > i 3% of the setpoint will be cause j to test additional valves in accordance with ASME Section XI test schedule, f The popping setpoints are significantly below the 110% primary system design ! pressure safety limit. Therefore, adequate margin exists between the setpoint and the safety limit of 1540 psig. For the purposes of establishing the test frequency, a valve shall be considered to have failed to function properly if the test relief pressure is i determined to be outside of the allowable setpoint tolerance specified in the ! ASME Code to which the valve was constructed. For the LACBWR spring-Icaded valves which are constructed to the ASME Code Sect ion VIII,1962, and Nucleer Code Case N-1271, 1962, and which must be removed from the primary steam j system to conduct the test, the al.lowable setpoint tolerance is 1 3% of the ! set pressure. Ilowever, when the safety valve relief pressure is set prior l to inst alling the valve on t he reactor system, the maximum deviation of the I test relief pressure from the specified set pressure shall not exceed 1 1.0% I of the required set pressure. l IT4-20b 47 1 ______._________.______________.____________________J
i PIMMMED Delete these bases. l
.R5TIFICA TION Section.4/5.2.2.5 is being deleted, so the bases should also be deleted.
4
\
a h i PC4-20b 4H ________-___A
~
f.. J i j D g c
- -e s .; ,. -
i 4 r Y , S , l, ? fMGE M1 t , 3, ,
'Cin@nrNT .
-'4.2.2.6;'At least one. forced circulation loop'shall be non-isolated,.i.e...
suction?snd discharge-valves open, any time the shutdown; condenser-is operatin i, lor the Reactor Building Main Steam Isolation. Valve .(MSIV) and the Turbine Huilding MSIV are~non-isolated. n . 4.2.2.7' Thy suction, discharge, and discha'rge bypass valves of'the forced-circulation puaps shall operate as described in Section'2.3.3.4. 148tMMim
. Delete.these specifications.
OL i l J751717 CAT 10W ) i With' the reactor ~ permanent ly shut down and reactor pressure, vessel ; defueled,' there is no need'for any requirements on the forced circulation-4 system. 'i
- l s
1 V
?
I
. PC4 -20b 49
m
? : ff. , .
f' ,- I y
. , i k {'I' ' o , ,
' d !) .
.'f.
f" u
, m, , . ,, ,
cggggyr x i , i .,
'f
,- 14.2.2.8L Operation of.the' forced' circulation' pumps shall"be asLdescribed in.
' U.
, ' Sect ion ' 2.3.4.'3. - ,
o . .
-4.2.2.9s . The ' reactor. shall not be operated above a power level Lof 82.5 L megawatts'thermalLwhen only.one forced circ,ulation loop is..in operati'n..- o i .
u., , 4.2.2.10 E Automatic) initiation of. shutdown condenser operation sisall cause
.the' shutdown. condenser tube. side. vent;controlival've't'o open and.then to'close- ~
automatically 'after a' maximum ~ of 2 minutes of coridenser operation. . This'
~ ~
s k ' ' action.shall'be subject.to remote manual override..
;t ..
;4.2.2.'11 -The return isolation. valve on tho decay heat' system shall!be locked-
~
M , f :-) , closed:except'as required during reactor shutdown.
, 4 ,
, t 4.2.2.12 L The : decay heat pum:p shall'not be' placed in service unless the
' , < ' reactor-is 'suberitical by' at least: 0.5% delta k/k by the criterin'of Sedtion c
" 4.2.4.6, or unless boron solutionJhas been injected into'the reactor.
L4.2.2.13 The purification system shall:not'be operated whenever.the presence: ' X > of' borori solution . is ' required -;in the reactor. y q
;. 5 .' 4.2.'2.14 - Deleted :. ,
l 4'.'2.2.15 Delethd'
' 4 '. '4'. 2. 2 l 16 ! LDeleted
'4.2,2.17 The boron injection system shall be.available for remote manual.
operation:whenever.the " control Power" key switch is in the "0N"' position or whenever all control rods are not fully inserted, except' during tests;. D -specified.in Sec.'5.2.9'and 5.2.24.1.
..N J 'RtMtWED P
l , , sDelete these' sections. l t
.175TIf1 CAT 10N These specifications do not apply to a reactor that has been permanently Lsinut down, with all ' fuel removed from- the pressure vessel. Regarding Section 4.2.2.10,.the shutdown condenser is not required when the reactor is shut l'
L down 'and the primary system depressurized, per Section 4.2.2.19. 1 l l 1 l 1
,f PC4-20b 50 s
4
- e. . . .
M/S 4 ;' , Ly' ,
.)*
1 ,~,.
~'
~
t , t .;
' ' . '- .' I s o e .
;,9j i b' j ,
.p , I f
\ '
, .<1 N ',f
'y
' z
' ' ~
, ; PAagg' [31a~32 m ?
p cgaggyr, 4, d , , f 4.2. 2.'18 Deleted. , t , ky2 , ,, s.4.2.2.19 - TneIshutdown condenser system shall' be available- f(r' automatic. operation except at: times when :the reactor is' shut- doni and the primary
.( isysten~depressurized to approximately' atmospheric. pressure. .Manural valves in-the' shutdown cendenser, system lshall .be locked in'.that position which will not 1
;g ; impair.l gystem capability. when ' automatic ' operability is ~ required. -
- 5. m gg ,,
4.2.2.20 Titeicondition of low accumulator pressure it thethydraulic ' valve--
'accumuistor system shall be'. corrected'within a'maximus time of one-hour'-
~
O. Y followingl annunciation in'the control room or the reactor shall be shut'down. W ,
.pEACTOR COOLANT SYSTEM y7 '- -OPERATIONAL LEAKAGE ,
4
,y -
t :.f
' LIMITING CONDITION"OR OPERATION n
w >' x ' 4.2.2.21 L. Reactor.: coolant. system Icakage shall be limited to:
., u.. No' PRESSURE BOUNDARY LEAKAGE.
7 1' b. .TDENTIFIED and UNIDENTIFIED LEARAGELas shown below: e. i p' .. f" Maxn Allowable ,' m t i
~ Area' Source -Leakage +~'GPM
- q. ,
, ;~ Lower Reactor . IDENTIFIED 0.04. s N -Cavity ~ ' UNIDENTIFIED -0.01 . 4 2.0
'* . Forced Circulation . IDENTIFIED
. Pump Cubicles UNIDENTIFIED 0.5 r"
Balance of- ' IDENTIFIED '4.0
' Containment Bldg. UNIDENTIFIED 1.0 E ~ c. 2 gpoi increase in UNIDENTIFIED 1.EAKAGE within any 24-hour
, period. j i
,-[ ATTLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3. !
!,' .\ t s , , i ' l ',
# 4 i
4 ) o i b PC4-20b 51 1
- je ..
.g. <
J
- p J' ..
l 1 n
?
i ., l l' REACTOR'C00LANT SYST @ . LIMITING CONDIJION FOR OPERATION._- (cont'd) -
)
ACTION: .j
~
- a. With any PRESSURE BOUNDARY LEAKAGE,' be in at Icast HOT SHUTDOWN
.within 12 hours and'in COLD SHUTDOWN within the next 24 hours.
- b. With any reactor coolant system' leakage greater than the limits'in b.;and/or c.'above, reduce the leakage rate to within the-limits within 8 hours or be in at least HOT SHUTDOWN within the' next 12 hours and .in COLD SHUTDOWN within the following 24 hours.
- c. With any reactor coolant" system leakage greater than the limit in c.
above, identify the source of leakage within 4 hours or be in at'
-least HOT SHUTDOWN within the next 12: hours and in COLD SHUTDOWN "
l within the following 24 hours. SURVEILLANCE REQUIREMENTS l
- 5. 02
. 2 : The reactor-coolant system leakagc shall be demonstrated to be 1 l within each of the above limits by monitoring' a. belew and at least one of b., or c. l q
- a. Monitoring the containment Building retention tank level at least .l once por 24 hours,
- b. Monitoring the Containment Building forced circulation pump cubicle and the reactor vessel cavity atmospheric particulate radioactivity j
4 at least once per 24 hours. I
- c. Monitoring the containment Building atmospheric gaseous 1
-2 radioactivity at least once per 24 hours. l I
i lPhOlOSEl1 D6lete Sections 4. 2. 2.19, 4. 2. 2. 20 and 4/5. 2. 2. 21. JEST 1 FICA TION These requirements are not applicable with the reacter permanently shut down. i 1 FC4-20h 52 l t I
w, , 7 -- 9
'O #
y < p' PAGES 32a-32e y0 . -
~' h ' CM RENT e
. REACTOR COOLANT SYSTEML r .. .- ,
T 4.2.2.22 ,-ACTIVITY LIMITING ' CONDITION FOR OPERATION 4.2.2.22 : The : activity * **' of the: 0 m
- a.1' Reactor coolant shall:be limited to:
1.',f 0.2 uCi/ gram DOSE EQUIVALENT I-131, 1 s .
+ ,
- 2. II 100/E.uci/ gram', and g .
3.-- 1 5.0 x;10-8:uci/ gram gross alpha activity c . L b. , Off-gas emission,L measured at the .150 cu. ft. off-gas holdup tank effluent monitor, shall be limited .to I 750 Ci/d = (P/Pr) ..+- Ar '
~
(1-0.018t) (P/Pr), where:-
- 1. J(P/Pr)'.=lfraction of RATED THERMAL POWER between 33 MWt ands
.m. , 165' MWt. ,.
1 2. : At ' tramp activity,.Ci/d,'not to exceed 1500,.at, THERMAL POWER (Pr)Cao determined upon initial power escalation'after each, refueling. i 3. t = days.after the-determination'of At with the limitation-c that til 50. ]
- 4. Activity' values shall be normalized for correlation with previous cycles due to' changes in monitoring, sampling and/or q analysis methods.
',;j -- ---
$
- Limits for Cycle 6 core configuration with reactor operation extended j
. beyond 15,000 MWD /MTU lead asserably average exposure in the remaining AC j
core II" fuel assemblies: ]
l 1-131 Dose Equivalent 0.082uCi/ gram .) 2
.Grosa Alpha Activity- 0.9 x 10-6 uCi/ gram Off-gas 571. Ci/ day .,
i
" off-gas activity limits for Cycle 7 reactor operation wit h maximum
( average fun 1' assembly exposure greater than 16,800 MWD /MTU shall be I det ermined by substit uting 571 Ci/ day for the 750 Ci/ day figure wherever
~
it appears in Section 4.2.2.22. 53 x .PC4-20h
3 W . ggg REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION - (cont'd) APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and 4.
' ACTION:
- a. In OPERATIONAL CONDITION 1, 2 or 3, with the activity of the reactor coolant:
[ 1. J)0.2 uCi/ gram DOSE EQUIVALENT I-131 but 14.0 uci/ gram, entry to OPERATIONAL CONDITION'1, 2, 3 or 4 is permitted and operation may continue for up to 48 hours provided that. operation under these conditions'shall not exceed 800 hours in any consecutive 12 -month period. Should the' total operating time atla primary coolant specific activity >0.2 cCi/ gram DOSE
' EQUIVALENT I-131 exceed 500 hours in any consecutive six-month period, the licensee shall report the number ~of hours of, operation above this limit-to the NRC within 30 days.-
- 2. >0.'2 uCi/ gram DOSE EQUIVALENT I-131 for more than 48 hours a during one continuous' time interval or.) 4.0 uCi/ gram be in at least-HOT SHUTDOWN'with the main steam line isolation valve
\ closed within 12 hours.
- 3. ') 100/5uCi/gramor)5.0x10-e uCi/ gram gross alpha activity,.
p be in at least HOT SHUTDOWN with the main steam line isolation valve closed within 12 hours and in COLD SHUTDOWN within the next 24 hours. l f I PC4-20b. 54 j \ a--.-___. -==.
- y. ,
j f"
< +
lcl , y}r , 8 y ' m3 , , y s (i , 2 T ,
.. y . ;. .[. .
. REACTOR COOLANT SYSTDI
' 4l LIMITING CONDITION _FOR OPERATION - (cont'd)
. )> .
. ACTION:: :(cont'd)
!b. sin OPERATIONAL CONDITION'1, with off-gas' activity.
1 (( 12 1 750 Ci/ day'(P/Pr)' + Ar (1-0.018t) (P/Pr) bu't i 1000 Ci/ day
~
d
.~ (P/Pr) Cars (1-0.018t)--{P/Pr) . ' POWER OPERATION may-continueL 1/;
.," for up to 48 hours provided that operation under these i Tconditionsishall'not exceed 800 hours in any consecuti've:
, 12-month periodi . Should the total operating time' at an off-gas.. c
^
w; activity" E 750 Ci/ day P/Pr ' + Ai (1-0.018t) ' (P/Pr ) exceed. 500 ' ' l' W ' hours in any consecutive six-month period,'the licensee shall report the number of hours of ~ operation above this limit to the u NRC within'30 days. y.
- 2. > 750, ci/dayf(P/Pr) + Ar '(1-0.0}8t)'(P/Pr) for more than 7 48 hours during one continuous time. interval or.).1000 ci/ day I be in'at'least HOT SHUTDOWN with the main steam line isolation valve' closed within.12 hoursJand in COLD SHUTDOWN within the?
next t 24 : hours. ?
# -c.- With the reactor required to be'~ shut. down.as.a result of ACTION a'3 .
i
+
> 'r b'.2 above. obtain Commission approval prior.to increasing' reactor o
coolanL' system temperature ab'ove'212oF. e" ~
. d. ' In OPERATIONAL CONDITION 1, 'with THERMAL POWER increased by more than 25% of RATED THEINAL POWER since the last performance of Table'4.2.2.22-1. Item 4, perform Table 4.2.2.22-1, Itema Ic and.4c.
- e. In OPERATIONAL CONDITION 1 or'2, with off-gas activity increased by
.more than 1000 uCi/see within one bour, perform Table 4.2.2.22-1,.
ltem Ic and 4c. Off-gas activity values shall be normalized for correlation wi.th previous cycles due to' changes in monitoring, :
' sampling and/or analysis methods. ,
'l
- f. In OPERATIONAL . CONDITION 1, '2, 3 OR 4 with the activity of the _ ,
primary coolant > 0.2 uCi/ gram DOSE EQUIVALENT I-131, or > 100/E j uCi/ gram,'or > 5.0 x 10-6 uCi/ gram gross alpha activity, or with the : q;;
- off-gns activity > '750 Ci/ day (P/Pr) 4 Ar (1-0.018t) (P/Pr) perform !
i4, ' the sampling .and analysis requirements of Items Ib, 4b and 6b of i
- Table'4.2.2.22-1. l
- 1 A Licensee Event Report shall be prepared and submitted to the a Cnamdssion, if the activity is indicative of fuel cladding failure.
- ,;. This determination will be based on additional sample results; other
~
previous, concurrent and subsequent activity levels; operating experience; and/or recent or ongoing plant evolutions. This report . shall contain the' results of the activity analyses and the time j duration when the activity exceeded each limit together with the following additional information. PC4-20b 55 h_m-______m m_m__ _ _ _ _ _ . _ _ _ _ _ _ _
m . , m - - q j {;.f M,
/< -
z..,
?r51 l Wj 1
., < \'i ./
Vt 3
- CMBilEW a L- , ' REACTOR COOLANT SYSTEM.
&a
' O' I,IMITING CONDITION FOR OPERATION'- (cont'd).
m1 ' ACTION:- (cont'd)i d
- g. (In OPERATIONAL CONDITION 1 or 2, with THERMAL POWER changed'~oy more than 15% of RATED THERMAL POWER within'one hour, perforra the sampling and analysis requirements of: Item Ic and 4c of Table 4.2.2.22-1. .
Prepare and submit ~to the Commission a Special Report at'least once j
-per.92. days containing.the!results of the' specific activity analysis-together with the below~ additional infondation = for each occurrence.
Additional Information j u.
- 1. Reactor' power history starting 48 hours prior to: -> ,
( I a) The first sampie in which the, limit was exceeded, and/op-b)' The THERMAL POWER or off-gas level change.
- 2. Fuel burnup by core region,.
- 3. Cleanup flow history starting 48 hours prior toi -
h.
;:1 t
.1:
The first sample in which the limit was exceeded, and/or.
-a) ib). The THERMAL POWER'or offagas level charge.
- 4. Off-gas level starting 48 hours prior to:
.l
.a) The first ample in which the limit was exceeded, and/or- q b) The THERMAL POWER or off-gas level change.
- 5. Gross alpha activity level starting with the last' sample taken 3 prior to: f a) The first sample in which the limit was exceeded, and/or b) The THERMAL POWER or off-gas level change.
-l SURVEILLANCE REQUIREMENTS _
5.2.16 The activity of the reactor coolant and off-gas emission shall be
. demonstrated to be within the limits by the performance of the sampling and ,
analysis program of Table 4.2.2.22-1. l t i 1
)
PC4-20b 56 I u 3
w . p.; t ~ A d.h , CMYENT A 1 k
~ TABLE 4.2.2.22-1 ACTIVITY SAMPLE AND ANALYSIS' PROGRAM OPERATIONAL CONDITIONS TYPE OF MEASUREMENT ~ SAMPLE AND ANALYSIS IN WHICH SAMPLE 4 AND~AMALYSIS FREQUENCY AND ANALYSIS REQUIRR4-4 1. ' Gross . Beta 'and a) At least once:per 72 hrs. 1, 2,' 3, 4 Gamma Activity- b) 'At least'once~per 4 hrs l#, 2#,.3#, 4#
l - Determination whenever the' activity exceeds n'specified~1imit as: required by ACTION f. i .c)Atricast one. sample. 1,2 between.2 and 6 hrs foilowing the' THERMAL POWER or.off-gas level change as required by. ACTION d, eior g.
- 2. . Isotopic' Analysis At least once per'31 days 1 for DOSE EQUIVALENT
'I-131 Concentration-
- 3. Isotop'ic Analysis < At least once per 6'mos.* 1 for:E .
1
- 4. Isotopic Analysis .a) At.least once.per- 1 for Iodine includ- 31 days.**
ing I-131, I-133 b) At least once per 4 hrs, 1#, 2#, 3#, 4# , and I-135 whenever the activity l exceeds a specified ' limit as required by' ACTION f. c) At least one' sample 1,2 between 2 and 6 hours following the THERMAL POWER or off-gas level change an required by ACTION d, e or g I i i l l \ 1 ! u PC4a20b 57 !
- Ahwa_-w--_._ ___ .-_.2.._. - . _ _ ..
Y= . p
' 'r j1:
,w
.,.p
!O > . . . [W N y :gLq _ TABLE 4.2.2.22-1 .(cont'd) fl --fCTIVITY SAMPLE AND ANALYSIS PRCGRAM OPERATIONAL CONDITIONS i TYPE OF MEASUREMENT SAMPLE AND ANALYSIS IN WHICH SAMPLE AND ANALYSIS FREQUENCY AND ANALYSIS REQUIRED
' 5. - Isotopic' Analysis- At least once per 31 days 1 of an Off-Gas . ,
. Sample including i Quantitative l Measurements for at least Xe-133, Xe-135-and Kr-88!
6 .~ Gross. alpha 'a);'At'least once per 7 days.- 1-
,, activity deter. .b)',At.least'one sample when- 1,2,3;4
/ mination ever the. activity' exceeds
,; 'a specified limit as.
- required by ACTION f.
'
- Smrple to be taken after a minimum of 2 EFPD and 20 days of POWER !
OPERATION havel elapsed sinc.e reactor was'last suberitical for 48 hours or longer.
'** Each isotopic iodine analysis shall be accompanied by a gross activity determination.-
# Until the specific activity of the primary coolant system is restored to .j within its limits. ;
1 PIKHY) SED j 1 Delete Section 4 2.2.22, Surveillance 5.2.16 and Table 4.2.2.22-1. 1 JUSTIFICAf]O_Q The purpose of these requirements was to detect fuel degradation and 1.imit operations when activity was high. Sampling of the primary system and measuring offgas activity for a defueled reactor vessel do not serve any function, i Pc4-20b 58
-_--____---_-2
)
1 PAGE 32f - i
- 1 f
CURRENT REACTOR COOLANT SYSTEM ;
'4.2.2.22 and 5.2.16 Activity The limitations on the specific activity of the primary coolant ensure that the 2-hour thyroid and whole body doses resulting from a main steam'line failure outside the containment during steady state operation will not exceed small fractions of the dose guidelines of 10 CFR 100. Permitting operation j to continue for limited time periods with higher specific activity levels j accommodates short term iodine spikes which may be associated with power level changes,and is based on the fact that a steam line failure during these short time periods is considerably less likely. Operation at the higher activity levels, therefore, is restricted to a small fraction of the unit's total operating time. The upper limit of coolant iodine concentration during e.hort term iodine spikes ensures that the thyroio dose from a steam linc ;
failure will not exceed 10 CFR Part 100 dose guidelines. Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analysis following power changes may be permissible if justified by the data obtained Closing the main steam line isolation valves prevents the release of 6ctivity to the environs should the steam line rupture occur. The' surveillance i requirements provide adequate assurance that excessive specific activity ! levels in the reactor coolant will be detected in sufficient ti:re to take corrective action. The limitation on off-gas emission and the limitation on gross alpha activity 1 are established to ensure fuel integrity. The numerical limits were selected using extensive data collected during fuel cycles 4 and 5. The off-gas activity limitation formula has been derived to relate off-gas production and a power level and conservatively precludes reduced power operation without also reducing the limiting condition for operation off-gas activity. The formula contains a factor Ar (1-0.018t) P/Pr which accounts for base level uranium 4 tramp activity in the coolant and on the core surfaces at the beginning of the cycle. This tramp activity term is reduced to 0.1 of the BOC valve within 50 days to account for cleanup of the system during operation. l The activit y of the off-gas emission (including the at Factor) is limited to I ensure that the potential radiological consequences of an off-gas system failure resulting in the release of the entire off-gas syst em volume wil'1 not exceed small fractions of the dose guidelines of 10 CFR 100. Operat ion at higher levels of off-gas emission for short periods is rettricted to a small frnction of the unit's tot al operating time. This upper litrit permits normal t ransients as might be expected during periods of power change, pC4 40b 59 l __ J
.)
l i 11 j l 1 L : 1 ;. j t . PmfuSED .i; Delete the bases for Sections 4.2.2.22 and 5.2.16. - JUSTIFICATION i These sections are being deleted, therefore bases should be deleted. i l PC4-20b 60 ! I i
o.
- PAGES 32r(1) - 32r(S}
CURRENT ELECTRICAL POWER SYSTEMS A.C. DISTRIBUTION - INSTRINENTATION LIMITING CONDITION FOR OPERATION 4.2.3.3.1 :The Essential Bus instrumentation channels of Table 4.2.3.3-1 'j shall be OPERABLE with their trip set points set consistent with the values i shown in the Trip Set Point Column of Table 4.2.3.3-2. l APPLICABILITY: As shown in Table 4.2.3.3-1. ACTION: As shown'in Table 4.2.3.3-1. j SURVEILLANCE REQUIREMENTS l 1 5.2.11.5 Each Essential Bus instrumentation channel shall be demonstrated- '
' OPERABLE by the performance of the CHANNEL FUNCTIONAL TEST at the frequency shown in Table 5.2.11.3-1.
Um current Tables 4. 2. 3. 3-1, 4. 2. 3. 3-2, and S. 2.11. 3-1 follow. PIMI M ED
' Delete Specification 4.2.3.3.1, including Tables 4.2.3.3-1 and 4.2.3.3-2 and Surveillance 5.2.11.5, including Table 5.2.11.3-1.
LUSTff1 CATION These specifications were applicable only in Condii. ions 1, 2, and 3. PC4-20b 61
@fE i 1 , :, ,< ? l jf ' l_ ' l ir Uh 4 s vn CNE@T 1-
~
TABLE 4.2.3.3-1
~ A.C. DISTRIBUTION-UNDERVOLTAG2 INSTRlNENTATION
; TOTAL. "MINIMIN APPLICABLE' TRIP' FUNCTION AND NO. OF ' CHANNEIS CHANNELS OPERATIONAL FUNCTIONAL UNITS CHANNEIS TO TRIP OPERABLE CONDITIONS
- ACTION 4
Loss'of' power. / a 480v Essential 2/ Bus 1/ Bus .2/ Bus 1, 2, 3 1 Bus.Undervoltage'- (Undervoltage relays phases A and C)
- b. 480v. Essential 3/ Bus ** 2/ Bus 2/ Bus 1, 2, 3 2 Bus Undervoltage (Undervoltage relays phases-
.A,.B and C)
*. Required when engineered safety equipment is required to be operable.
** To be~ effective only after unit is modified to have 3 channels per bus.
I
)
i l l l 1 i l'C4-20h 62
tw +-. .
' ' ' ' ~
,O ^
'r b_Gl y ). .;;.l, _
o e), m . . .u , , 1
- 1. ' ,k i'
}^ h y h, l
, g
. n- 1 I ' '
) l [ t .j ' _,
W i^ ', ,
. 7'
;f ,
paggyy W, 4 , I,,l-'
.. J ,
. . - . -Li s'
'1 TABLE 4.2.3.3 (cont'd)
T,,1 ja y . .
+ ACTION STATEMENTS <
J ACTION 1 -- 'WiththenumbeEof'channelsOPERABLE-'oneles'sthanrequiredby the Minimum Channels OPERABLE requirement, be in at least HOT
' SHUTDOWN within 12 hours;iexcept that one channel may be L bypassed up' to two hours for surveillance testing or maintenance -
.. -provided that:.
u , A , 6
- a. The other channel'is OPERABLE, and 1- '. . , .
- b. The time during which' redundant protection is not provided
, shall not' exceed 24 hours;in any 30-day interval.
+ .
.y ACTION 2' -- . With:t'he number of OPERABLE Channels one less than the; Total-Number of Channels, operation may proceed until perforinance'of
' the next CHANNEL FUNCTIONAL TEST provided the inoperable
! . channel is placed in the tripped condition within~one hour.
.r. .
t t , ,
'i
. o ,
I
+
5 .:14 , PC4-20b G3
. , f 1 ,
MF :u: a .s v c m,+ yg , s >3 y, :, ...;,i . . '- ! "Ar o;. 3
' l ', .i ; '- 1
~
's r {'
s ;;g > , , . s
.id:4 131-;
. i t '.]
; .( ~m ,
e m i'r 6 4 ', '
' TABLE 4.2.3.3-2 A.C.' DISTRIBUTION--UNDERVOLTAGE' INSTRUMENTATION' TRIP VALUES ~
? 'l . /
9 : cTRIP FUNCTION AND
' FUNCTIONAL UNITS TRIP SETPOINT Loss . of ' Power .
e ,.9
- a. .480v Essential Bus. 88.3.f-93 1 97.6')
Volts Undervoltage (Undera
~ voltage relays'phasbs. >
1.91312.)-) Seconds, Time Delay
.A and C) 480v Essential Bus.** Volts
~
n r ., ; b'. 95 1 100 1 105 ) Undervoltage (Under-voltage relays phases' 8.1 f 9 I 9.9 ) Seconds,LTime Delay A, B and C)' f') n- .= _. - _ _ . _ . - _ , _ _ i- ,
- To be effective only,after unit is modified to have 3 channels per bus.
I L 1: I l PC4-20b 64 L . . . .. . . . . . . . _ _ _ _ _ _ _
g
- ?
.y
,) 5 Vl' .
CfA BW N T m
- a -
TABLE:5.2.11.3-1.
.A.C.! DISTRIBUTION--UNDERVOLTAGE INSTRlNENTATION SURVEILLANCE REQUIREMENTS i OPERATIONAL-CONDITIONS'
,. CHANNEL FOR WHICH.
TRIP FUNCTION AND' CHANNEL FUNCTIONAL. CHANNEL SURVEILLANCE FUNCTIONAL UNITS . EIECK TEST ' CALIBRATION REQUIRED -: ., iy
- Loss of. Power i
- n. .480v Essential' Bus N/A 'R N/A 1,2,3
.Undervoltage (Under-
' voltage' relays
' phases:A'and C) i
- b. 4S0v Esseatial Bus.** N/A M. R '1, 2, 3'
- Undervoltege.(Under-voltage relays, phases A, 0'and C) i
I
** To be effective only after unit is modified to have 3 channels per bus.
l 1 1 IT4-20b 65 1 mil!___. ._._. _ . _ .
n .
.{:) Ii ') .,
) ..n.
t e i + s
*1?
fPAGM 32t--32H
.c.
+ '
q 4.2.4.2 _P_O,WER DISTRIBUTION-LIMITS .L., IAVERAGE PLANAR LINEAR HEAT GENERATION RATE
..k*i1 TING' CONDITION FOR OPERATION -
g, t y
/4.2.4.2.1 All' AVERAGE PLANAR LINEAR HEAT GENERATION: RATES-(APLHGRs)ifor eachi ,
. type of fuel as.a' function of AVERAGE PLANAR EXPOSURE shall not exceed the.-
. limit shown in~ Figure'4.2.'4.2.1-llor 4.2.4.2.1-2.
APPLICABILITY:. OPERATIONAL CONDITION 1.
- ACTION:
With an APIllGR exceeding' the limits of Figure 4.2.4.2.1-1 or 4.2.4.2.1-2
-initiate: adjustment, within 30 minutes.so that APLHGR is within the limit ~
.within 2 hours or reduce THERMAL POWER to less than 25% of RATED THEIPiAL POWER within the'next'4 hours.
SURVEILLANCE REQUIREMENTS
-5.2.17.'l All APLHGR's shall.b'e determined to be equal to or less than the applicable ' limit'by verifying. that each control rod .is within the control rod pattern and. withdrawal sequence requirements during operation at'l 25% RATED.
THERMAL POWER:
- a. At least once per 24 hours,'and
- b. Whenever THERMAL POWER has been increased by at least 15% of RATED <
THERMAL POWER and steady state operating conditions have been established.
- 125% of RATED THEINAL POWER b
l 1 PC4-20b 66 m .. , , 3
;yl 3 # I D' , I f. '
g u .. '/& i in)[ ,, a_ ~m. j- a.. m 3 ., e b 1 - f:'
, j;b 3
.\; 1 n i, ,f ,
- wr ,
t ., 1 i'
')
cgggyf <
. .' 2 K . - '
'320 -
. .o :t t
,Et's '. ' r g
' t.
s ea
.l t . 3
; k
; i ,. a. t '..
1, s v, . i ..' . j!.t.43...
's> -
. J.[ } /
'j..I! i sil . . . .. . *
).. . . l. . ! Le . . . .! .,i '
.!. s. !.& ,..r .
s ! . .'. l ' 2 4 A r w u ,. m :-v : .g .e. t.- ... m n' r Jli n- .
!.l't .
'j-.l1:
' ~ , .
1
- c 'r -
.. 4 . 1 ;.
....I*
o . ;
.- .s g..?.s ! , * -
(i
, .t 8. r .. .
. i ;. l' .i . i
4 ia. m-g . j. . , : ,1
.l .g q ,f, 8 -
i ! 2 .' .' 4..:; ' '
**. . .q .. . !
- .., s . . ., . 3 is I' *4 ,l l.r..'
,1
.f
- .t . .
- p. .
,-..,.t.,.. .., .
- .f s.. : p 1
, . . . i.
.I
.. .a
. . ..t . , s s. - .; ..._ t.
t f ..L- . s. .* -;.. 4. . t. ~ ,.
. .... . J.
g
.a e s . . . .f . 1 t
. .r;ye ......,s.
.. .- e .
.s . . ...,.
-l 8-j ] a : f v'* 7l- . * . I e1l : . *.
'\, .
. La .J. , . .J .:
l
. . . . . . i/.
.. j .1.1.t. L' ' ; ! / . . .. ..
. .: r . .: ; .6 i1.i
. . s. , ,53 N-
'[i
!y t
..jt..j. ,.... f.
- ..l!: . , l.. . . ;, .'
.. : : .."i,, .'
.l.. v.. {
.L .l..., .
.. ....t . .
,i .ia . ga . . .
m
. . l.., .1.. 't. ... j ; . , ; . . . !.., c.
.- l b. .. 3.t{y -l .'j
- C s i '. : 4 f ., ,: .j *
;, gw
. ,.i . ! .. . *A'll] }.',
; r.;. i.; r. ' * *I- ,'d
- 4. .
3 ..-.4 .
..2 O am.>,
g,.; , .
; .i ;3 .:.,
., a 4 a . .:
. a
, y om-
- n. 7 ,
Ql - .
- , - . . i
. .J.; . ..!t .
.. N *-N k ,
.J ' ;,..; l, '. ;....g... .
4 ., . ...i ; : :. *
. ,j . W .
m .: . 94 , . g.
. . . . .......'i .
. ~ , .. . ; .... >t w .-N.'-
.!a ! g .!. .. q .1
. . . . . . ! t.1 . : .i .)
. . .. 3 a. n: a. ,
.J .
1
. ..1; , . .
.: i!j .
l,
; : i
.4 i.'.t. '
.l .: ; . .
c< m= y . 1.i .1.1.:. - . .'.,; . :-t . . . .
- l. . L t :. . .1. ; : ' a 1. J; t .
O< v
. .T.. .. g .; .y .; ; j. .. ...- 4 6 i i i .. .- i..l. N. .
. . * * * * * ** ' ** i1 o. i : e
.j . . . .; .7. .;.. ;. .. .
. , .f * .* .* * - l' j* *. i; t. . .j.i - w.
X Q. i
., J....,. . . 7..... .
3 .
..,. g_ .. ; . . 1. } ~
y e;
. . l. . . . . . . . . .I g a. .
, 1I. a. LJe:, - 5.
.' g J ; , 2 .t.
4 s
. .g ... . . .
. 1 . . ., ,
.i .
...{. ,. .... r-t g ,.
. . .; ,* n,,...
;..!. .j . . .
, =4
- ...; .e ;..j ...s Z .C.
.t ,
. . sq' W g'
,t {;
4 .a .. ...Ie iI . I 1 .g g
.1 i ;. .;.
. . q g;..;.j ", . . . ; 3 i .
; .jij.;.i..s! . y .. ;.
=
o .. . g .. -. .; . .; j lij;I n< J> i;.3m. x .g. .
. ., .: . ; . . .a
.. . . l -
.i . : . ., . .:. . :
- 4. .
. .. 3i..3. ....... . . ,
s
; . . .i. .t, 3 7.pg : : .. ..
g , cm
.. e2 .t , .a ::
.! . .q.y.:. ;.] : i .l : . ; w,y..
i.;i s .. .i;!. o :, . . .. .. ..
, . . t. .; .: : p. 3 ..tt
, g i, e. '
g
. .] . .. . . .. . ..i a . t op : . -
.a a ul :.l. l : .. ;' .1 !.ti.l ;
a g3 . . . a
; ~1'. }.yi.Q .
.l. . . . :. . ...... i;:... . . f. . ;.,i , ,, ;.i,..
. . i pt7 mm ..
. l'.:. ..g;..1 n.
g
. .l .
.g .,
J . . . pi s ;. +, l . 1 s a..: . ;t,. .. ..i
.t. g .; . y
. .. 7 ..
. g. 4
, i.....
.a t 1.
- m o . .. ....l. .
..e . l ..
- i. g.
, 1 .. . ..g ."it.. .: .j
. , . t .,J.,.t.
.. .g .. . .a .. i
,,e s gu ..
- 4. . .. 4 .
. . . :- .. . 3. .i
}. ii 1,.t'. ..i I f. .
i1. .'2. f a g .a .1
, . 1 .. l o ;!
!. 3. ; . . .
!. .! . i. { . _ . m
- l. !;j.Lj...'a ('* W s j ill..!;
. . .. - .t. . .
. g, I '.e .' l;;e t.t
.j* ..i, ; .lt. 4.
. 1.. ;.. . . . .. ' ,. . ....
i i : je ' i t*..1*. . l i If ji. .j {.!. 1 4 .
.t 4 *
. ..:..ti,; ,, .
. . .Ij, 1 1.i -
g
. i . .l.,.I '. '. t t1. g. ,,
8
.e..lt j
- ij 4 I I!
... : 4.1 ,. j
. .i : i ,. *
.. .,i... i i., . ..
.i.' . . .
, .. . . . .. . . . .i ... .
,,a m
m c so m N/4) (EDHidWI)31V8 IM1Vd3NL ly3H W3HM
~
WW1d 3Drd3hV wnwpeu ! L l
.PC4-~20 67 J :- '
L i___:_____..._____._____ _ _ _ _ _ . _ _m____ . _
x :
- ay8 t
.~.J %
b
, - . . . . . . . . s . . .
.94i
. - , - i . -
. . . . . . - .1 4 .t - . .
. , g. . .- . .4
- 4 . - .
1_1 . 0
. , * . . . ,. . . - ;i . . 3 ;f _
t
- s .
. , .
- j. ,y ..; . '
- i i.
. +
..r
. . r ,} . '
/ . . . ,
- c. a' . . . . . . . - .t . . : f. . ,j
. ; * . . .. . . . . ! . . i R
- .~ . y. i
)
_. . . . - 6: .; .i * . . . 5 O
.\
? 2 F g
. . i - .
. . . . ' j- ,; . ' . . {
- z
. . . * . . T . '. .' g.I i .! . P . . .j )
... y . . * . . . - . - 4 . . .I . ' . . .4 U E
?
I . . * - ; . . . . ' 3 5 T R
- M U L
/ S E D O U
. . i . . ' a . . . . ' - . ,.
. . . ?, . . ;
. .; . . . - . ii . . . ' . - ,, W P F G X '
- p. * .! .a . . . .I 11 . . , 3 . i t li I .
;. .i 0 ( E )
i . . , . . rt > ,. . .; '4, 3j .l I 4 .
, .j = 2 C
b E R -
, 4 ; . . . . . .j ,. .
R a . . ;. . . . - . . .' . . ' . 1 . . ! - . . . 9 A A(
, , * . . , . . . .> i,. ' ; . . . . . ' - . . l U N S A .
,. . t . . . .le~ . ' -l4l l g, 1 i i-O L I P P I a
! i. .- . - ,s . .i t i9.l .l. i:. . . .l X
_ . . ! g. e , . . . ; 4 .( ' l ,: i- .. ,
. , jj. . . . . , . : .
. i. g' i;. - i . .
E E E
,. .j 5 G P j; jg,
_j: . , (- .t; .i.' . gy' g. !i. . . s t} 1 R AY
- m. li IIg . i g,. t . : . . e II 1<
. i;i A R T N E ^
. m.
Il i l.s &j.t . t ? . i ll A
? g Ih.
. I .' i 6 . .i g 4 - ; , i4_. ,g,g V D N L AN aOI P A
- y. Et ; . ;: . . -
. .t
- ..a . j 45.i
. t. l!. tI _1 1 ,
. - . . s r
- .. - - il
. .l.j E S I h;I{. . i: ,; : ,jt ; .' .
GN 4l G V _ hi ' . ii g. i EO .r *4
.- - . RI - _. 0 A E 8 . i j, . 'l;6 . . .
8.ti . . . . , t j T . . I
- . 1 R RP
* . I i. I .; . .
, s . e . . .{ I EA . . . , E G Y
* . ta ,
. . . 7 _ 'e n .l . .
j LR . ..g VH T
- a . . - j. . . . ,
7..i . - . . l j BE . . . 4. AL IP . P A SO .
, . ;% tc I. . . 4 . . .i j .j 11 .
?. . * . . . -
S M j y9Sh
. . 4. t j .
IF
. . . * . a . , 2 . .
4
- h. . p,4 p .
- MO
_ . 4 . . } ,. g .5 1J S R l 4 ' . - E
- . , . . . .- .. .1 .
ds.4
. . a ' . . . P , . .}
_ . . * . . _ . .i . .i . . t. _ * ; i
- 1 . . . . . .
tI 1. l i!
- ' !. ;. , , . e . .! . t . . .
- g. .I . . .
. I . .
.t4
. , r .
,v .
I1
.6s 44iI ..
.i.}
.ii
__ w
.
- s . . 1
,i ,, . . . .i , . _ . . . . . ti '
. i- . 9 f.
7
.4uN3X-
-..cO=1cg-wen zOHE4Mwzto gr= *- a %6 az s a,
- m <z 5 n. c&cE0>< $gget 4
.d4 n $ 3 m0
h I ClHUMAT POWER DISTRIBUTION LIMITS THERMAL POWER-RECIRCULATION FLOW RELATIONSHIP LIMITING CONDITION FOR OPERATION i 4.2.4.2.2 The combination of THERMAL POWER and recirculation flow for steady-state operation shall be within the operating region of Figure 4 ~ 0.2.2.1-) of Specification 4.0.2.2.1. APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. ACTION: l With the combination of THERMAL POWER and recirculation flow outside the 4 operating' region of Figure 4.0.2.2.1-1, initiate THERMAL POWER or i recirculation flow adjustments within 30 minutes so that the combination of THERMAL POWER and recirculation flow is within the operating. region. SURVEILLANCE REQUIREMENTS s 5.2.17.2 The combination of THERMAL POWER and recirculation flow shall be verified to_be within the operating region of Figure 4.0.2.2.1-1;
- a. At least once per 24 hours during steady state operation, and
- b. Each time THERMAL POWER has been increased by at least 15% of RATED THER3AL POWER and steady state operating conditions have been ;
established. l PC4-20h 69 _______ _ a
~ :
- [> s y ,
, , );
ht 5 t gy.
,It
I . ,, f Lig. , t! i , cgmHrXT V i iPOWER DISTRIBUTION LIMITS: l- :c ' 9,: . MINIMIN CRITICAL POWER RATIO . < M, . ,. . , ,
.I L: >
' LIMITING CONDITION 10R OPERATION :p
- t. .
412.4.2.3 '.' The' MINIMUM CRITICAL POWER. RATIO-(MCPR); as a function of core ' -
- flow, shall:beiequal to or: greater than the limit shown in~ Figure k .
!: m ". e4.2.4.2.3-1. , > n ' APPLICABILITY: ; CONDITION .1 ' *. : + , , ACTION: With MCPR less than the applicable limit, initiate adjustments lwithin . , , 1
.30' minutes so-that MCPR is equal to or greater than the applicable limit' j
.within-2 hours'or reduce TIIERMAL POWER to less than 15% of RATED THERMAL POWER within the next 4 hours.
SURVEILLANCE REQUIREMENT i 5.2.17.3 MCPR shall be determined to be equal to or. greater than the limit
' determined from Figure 4.2.4.2.3-1 by verifying that each control rod is
- within.the control rod pattern and withdrawal sequence requirements during.
Toperation;at b l5% of RATED THERMAL' POWER:
- a. At'leasti once per 24 hours, and -!
- b.: Each' time power level of the reactor has been increased by at least
'15% of RATED TIERMAL POWER and steady state operating conditions
'have been established. ';
l
- 1 15% RATED TIEINAL POWER j
\ )
PC4-20b 70
4 4Jl!li' 1i .: ij lslli Jil]l ,I111
,ii R ,;;_; ~
r .i-
~
8 yw3'
..y 9
; . 5 y d n1 0
?
,. . - ~ . , . -
I I 2 T; . . a,. . g' 2 1 II m p-_E rt - ' 7 . ._ i
- i i
t ' , . L' *-
.,. . ,'Li.',
.'I
- '3- .
II mu-'-
- t_ . I _
. i
.i9 *I i ' . ii -
l' ' . , '
. .[ . .
;- e pe p1L
;i
. e
;+ !ii
' :,t t
_ d '- -,_ _'gI
- i t i-i_, L' l' yp_ . .
._; L , .
aI_. Ty_.
;I;4 i ._
4_ p._ hg,- z T
" lii +Lr ii+
. _- t. I~~ LiL. '- . .- 1
.L 0 J g-0
[.. ij -.* ,. F -
.. 1
_~ jI: 4- .i i.- i
;Ii dp:e .r .
.h - ;. J. . +-
~
; - ., 1 I 1_ . .. . ..
y-f;L!u.*m. :i ii h_H-- ,4 3*-
;* . ' ~ .
g r - L.aH g I iL l!! yT T ,k,1 .f, . ._ . -
}^IL ' :!
3_ .} - I
~
n
,3.,
.L
*o'
. . )
L.L L _ 4._ ._ - a_. :
- .g. t . .j- ,* ' .1 F
i ;
; oii. s. ; I +i i . ._.
. .T tt . - 4 - . .
i~g" . . _. . 0 tud
. - ;:- j . . .
1-ii 4 13 .
;_* [.
2 LP e* 3 t.-_it . . j g,4i - _ %_ 0 7 R '
;g,. -
T 9 5
. - g._;9L.1 I'
. .i
= g ,-
,7U.
,.r'i ! H!. p1 i a- .
t Q. . .. - . - ii *in+ L .- 7 H ;~ . { .t
- J a t.
' .i -
,L. T ._ b '
- ; ..g. 1 1
- t ' .:l I r
. 8 a '- i 4hr- -.!..
~-
y4 1. :
. .i' h
i4iii'L.t& rt - I IL i 1
- t. L. . . '
._ / E i L.J x.
et
.L t h :,
,,s t qfHl' . 2_ii O I. .ti T
i' 4 i p ; q4t t..y. Y b A _ 4i . 0 l y i.
.:: .Y 8 R
]'
i$
* * .i i .. .
L _ - b . 6 1 H l !! ii i} - l_ q 1 i . jF 0 W
- 4. 3li it_ i I
__ ;9 . . .. i 1 O . x i1 .i g. _ . . L 3 F 5 . pI' _ _ _. ', i {- i L.j. .. , i... . .? 7 2
.: l . . . ..
+
E Rj ,..! :t. J
=1 .g l
;3 .
_. - P
- iU'l
- 4. .!-
i ; :i ij R
- j. tl . 0 4
-!EL. 'ti._-
-I .
. .- 4 l .
t {. . - ' 0 1 O .
, p.e+. .
_- - . }il _ I. }iI iL .:.; ! }e . .
. i
. .i 7 ( C 2 i4 i 3j'
._.}- w . 4
.f
.ti G'd..
- +
' n . 'i
.. ~ - b_.
i4 jL !. o S L.a. aI-i'I . . o_ 3 N
. . :1 l V e I. I F r k .,
_ - _ - ... L1 .f 4:= ,. , t' I {.l S u {4.I.tt!
- I _
j~ e g. 2psH e p p.
. . ..\ 1%. ! . . .1 aei
- e T g
- y. h .- eP
; - i_' :a 3-N-
; '_1 '%-
tlt. .: ;l r I i . o
-i.i -l t .
- 3 .. 0 M F jI F 'yT.
'T Y , .v .4 ,i
. i j .I - I 6 C I.
e
T
- _ .3p. . :l 1!].. ,.l. L.p. 'I I u
1 i' . d h 74
. ,. j .
e
- . , I
._ }. . .: jh 1{,
R
,._ :. 4-.I q*!L. a -c '
. . . . _ i. . .
t P go.' a C oy!{y,a
. _l .
X- - . 1 1I 4 g:
.4!
t:{ 4 R M i l t- li sF._ W' . ii h
- d. %
Kii
;. . t. I L :.
} ,;
g t
- ..i i:
ps _. . jp.
.t!
rI i 0 f o (p- - A[v-
- u. h. i
- . .. ,7 ;Y 5
j l_ .p ..p
'. , Lp'- .
j 1 _. _. _. v{~ _ _ . _.t_ _. t .
-- { _ . . _ . _ b. _ }. ,b. .- n t IT
. 9 t . .
- t gT e
. i!. q. t L.
.- _ L. _%
g {.,41
. ' c
} i1
-ph
[ t_ d{.1.I.. j_ i_ !iI i.
.i .
LI i
- r e
P i.+.i s% __.t- NT1 . - j 1 j_ tj_.7i t j' R j
.- .I
~ . .
.' l ; 3_
. b.
-m ; . 1
} ,1 : i.:. }
- 9. t.i .
H .
- 3. tI .
0
. 4 1}+ r
! ~ ! ; t j.a .t
_. . a. :t.. :.; g 1
. . . yi i L. i _ L. - - {
- l. .!
.~ . - .
. _I
." %:._..' . % 4.L-P p._!' t i. _. -
i1 4 - 1 L. - . .
. 4. i. ! 4_
a t {_f' . : p_.- . . . _ i
.1 !-
i 4.f_ - i!ili i 3i 4 t _.. . . . . - -! L . L_ " _ rE1ILM . - . i' i T-> At- . 4 ! 1
. ~
p,1h4i-i'Q.t - .
' > i} '- _ * - L 1 i !
- u. L. _ ._I u u. . .
L ;4l1' :i!- , g
.i i # ' t+ 4 - . -i!
-I - _. -'
. 0 3
5 4 3 2 1 0 9 B 7 g 5
- 2 2
2 2 1 l 1 1 2 2 acoE o~O.e3o - < 2a to yi.-
!l Le itI;
J . F yd ; fpy .. /' j 5. { ^ f " CDRREWT (,, 4 POWER DISTRIBUTION RATES'
.. LINEAR HEAT' GENERATION RATE' LIMITING CONDITION FOR OPERATION n
L i
. f. 7 . . ,
~
- c. m 4.2.4.2.4; The LINEAR HEAT GENERATION RATES.(LHGR's) shall.not exceed the ~
idesign LHGRLof 11'.94 kw/ft for Type I or Type II'(A-C) fuel ~ rods and 11.52 kw/ft~for Type III!(ENC) fuel rods?during steady-state operation..
. APPLICABILITY:' OPERATIONAL CONDITION 1.*
s
' ACTION:
- With the'LHGR'of any! fuel. rod exceeding the limit, initiate adjustment within 30 minutes.so'that LHGR is below the-limit within 2 hours or reduce THERMAL'
' POWER to less-than 25% of RATED THERMAL POWER within the'next'4 hours.
, SURVEILLANCE REQUIREMENTS'
.5.2.17,4 LHGR's shall be determined to'be equal to or less than the limit by verifying that each control-rod is within the control rod' pattern and-withdrawal sequence ' requirements during. operation -at 125% RATED THERMAL .
POWER:
- a. At least once per 24 hours during steady state operation, and
- b. Each time power level of the reactor has been increased by at least 15% of RATED. THERMAL POWER and steady. state ~ operating conditions
, -have been' established.
- 1 25% of RATED THERMAL POWER i
1 i n.~ l I 1 PC4-20b 72 l __2=___-
- a. ,
- j 3- q o ,
'1 l ;, ', k CURREW l
, s 4
; POWER DISTRIBUTION LIMITSJ
, ' MAXIMUM AVERAGE FUEL ASSEMBLY EXPOSURE' LIMITING CONDITION FOR OPERATION
~
4.2.4.2.Sc.The, maximum. average exposure of any fuel assembly not on.the- .-1 1 periphery of.the core shall be limited'to.18,000 MWD /MTU.' t APPLICABILITY: OPERATIONAL. CONDITION 1.
'i ACTION:
With the maximum average fuel assembly exposure of any nonperipheral assembly greater than 18,000 MWD /MTU, be in. at least HOT SHUTDOWN with the main steam line isolation valve closed within 12 hours and in COLD SHUTDOWN within the next- 24 hours. SURVEILLANCE REQUIREMENT i 5.2.17.5 The maximum average exposure'of each fuel assembly not on the periphery of the core shall be determined to be less than 18,000 MWD /MTU by calculation at least once per 31-EFPD. ! PC4-20b 73 t
a
' H a
7 , q
..O f-t y.
,f k I ^ I-i
- l cgggwr q i
.g I4.2.4.2LPOWERDISTRIBUTIONLIMITS BASES FOR SECTIONS 4.2.4.2 AND 5.2.17 j
4.2(4.2.1'and.5.2.17.1.AVERAGEPLANARLINEARHEATGENERATIONRATE
'The. specifications of these sections assure that the peak'claddingL . ,
- temperature following.the postulated design basis loss-of-coolant accident; -]
(LOCA)-will'not exceed 23000F in compliance with the limits established by- ; the Interim Acceptance Criteria, June 1971,' as applied to LACBWR stainless-steel clad fuel, Reference 1,'2, 3, 4 and 7. The peak cladding temperature (PCT) following a postulated loss-of-coolant: accident-is primarily a function of the averag: heat generation rate of-all the rods of a fuel sasembly at any axial location and is dependent only-
-secondarily on the rod to rod power distribution within an assembly. The PCT (
' is conservatively calculated assuming the reactor is operating at 102% full-rated power and'the axial times radial peaking factor is at a maximum value ,
of' 2.43 from 0-24 GWD/MTU maximum average planar exposure for any Type I, . Type II or Type III fuel-assembly. The factor is then reduced to 1.75 at :l 30 GWD/MTU exposure to prevent calculated failure of internal' rods. Operation with peaking factors below these values at RATED THERMAL POWER will ensure that peak ' cladding temperature during a LOCA will not exceed the limit of 23000F for' stainless steel fuel. The corresponding limiting Maximum-l AVERAGE PLANAR LINEAR imAT GENERATION RATE (MAPLHGR) value for Type III is shown in Figure 4.2.4.2.1-1 and that for Type I and Type II fuel-in Figure
-4.2.4.2.1-2.
The. calculational procedures used to establish the MAPLHGR limits shown in these two figures are based on the loss-of-coolant accident analyses performed by Gulf United Nuclear Corporation and Exxon Nuclear Company using approved calculatiuhal models consistent with the requirements of ECCS criteria as it applies to the LACBWR fuel. .The effects of the following parameters were included: (1) Radial conduction within each fuel rod, (2) Rod and canister convection, (3) Rod thermal radiation among the rods and canister, (4) Gap conductance with exposure and densification, (5) Rod ballooning, and other mechanical and neutronic parameters. The peak cladding. temperatures achieved during a postulated LOCA for Type I, Type II and Type III fuel are shown in Bases Figure 4.2.4.2.1-1. Operation at RATED THERMAL POWER with fuel assembly AVERAGE PLANAR HEAT GENERATION HATES below
- the MAPLHGR limits of Figures 4.2.4.2.1-1 and 4.2.4.2.1-2 will ensure that the PCT's will not exceed the 2300oF limit.
A list of the significant input parameters to the loss-of-coolant accident analysis is prm,ented in Reference 4. PC4-20b 74
_ dg9 c
- 0
. . , ~
3 d1 ) - \ ' n1 C \ a A \ e( p I y
\
'e T
\
p \ . y \ 2
- T \- -
E1
. 5 R .
g ~ 2 U2 S . O4
- P X2 E
- 4 R
) AD
- U NN T AA
. - , 0 M 2 /D L
P1 V E . G G1 ( A . l R2 e .
. E E u .
. R V4 F U A .
S 2 l
, O .
'e P S4
. e . 5 X V 1 E S S
t
" R EE RR s A UU l
s e - N A L TG AI RF n P E
~
i - l E PF MO a l ) t i C G E S - eE N 0 AR TS T 1 r p( . E GI f o - T y V A NM I I DL D i t m .
- AR LG CH i
L L KP
.. AA
. . 5 EM P
T _ A 0 0 0 0 s N 0 0 0 0 0 0
~
0 0 0 0 0 0
~
r 4 3 2 1 0 9 8 7 2 2 2 2 2 1 1 1 C *ru , E 3
- e&U p "" - x;a N g,o ym
k'.J , J e ,. N ' 4 h , x ' ctme wr ' y l -. Y , POWER' DISTRIBUTION LIMITS:- :i f. r-BASES FOR SECTIONS 4.2.4.2'AND'5.2.17 \t ' D1;
~ '
, ~ The! daily. requirement:for surveillance of the core APlHGR above 25%.of- RATED '
THERMAL' POWER is' sufficient lsince power distribution shifts areivery! slow}
'when there have not been significant power or. control: rod changes.:.The:
surveillance of core LHGR after power iincreases') .154.of_. RATED THERMAL POWER '
'will' assure that significant increases in APLHGR are determined.
4.1.4.1.1 and 5.2.17.2 THERMAL POWER-RECIRCULATION FLOW RELATIONSHIP-
. The THERMAL POWER-recirculation flow limiting. condition for steady-state.
z
.' operation has.been conservatively. set.below-the-LSSS: curve presented in .n
- Figure'4;0.2.241-1 of. Specification 4.0.2.2.1.- This specification also require's that the reactor not: be operated in a steady-stete condition 'above
- the RATED THERMAL POWER, 165 MWth, authorized in the NRC license for the-facility.; The limiting minimum flow setpoint of'30% of rated recirculation
.- , flow is substantially'above the natural circulation flow and the flow at. 1
- L which hydraulic. instability occurs. ..A ratio of > 1.6 exists between the ,
low-flow scram setpoints and the instability-natural circulation flow. 1 Therefore, adequate protection of the core against. flow and core instability exists over the full power range of anticipated reactor operation as' limited by; the THERMAL. POWER-recirculation flow operatirig' region. The low flow limitation requires the establishment of a minimum flow of 30% of. rated recirculation' flow before reactor startup. Operation in a steady state.cbndition { 68.3% of RATED THERMAL POWER at-the minimum 30%'of RATED
' RECIRCULATION FLOW assures that the CPR remains above the minimum allowable
.value of 1.32 during an abnormal-reactor transient (recirculation flow speedup is the'most limiting). The steady state limiting CPR values-z corresponding to other operating conditions bound by the Power-Flow LSSS curve reported in Fig. 4.0.2.2.1-1 are defined in Fig. 4.2.4.2.3-1.
- g. .
The daily requirement for surveillance.of the power to' recirculation flow
- relation is sufficient since this relation' shifts very slowly when there have-not been significant power, flow or control rod changes. The surveillance of this relation after power increases 1 15% of RATED THERMAL POWER will' assure that significant changes in the relation are determined. l 1
4.2.4.2.3 and 5.2.17.3 MINIMUM CRITICAL POWER RATIO The required operating limit MINIMUM CRITICAL POWER RATIO (MCPR) at steady-state operating conditions as established in Specification 4.2.4.2.3 is derived from an analysis of abnormal operational transients with the transient CRITICAL POWER RATIO 1 1.32. The CPR criterion of 1.32 was j 1 PC4-20b 76 j
j 1 CURRENT POWER DISTRIBUTION LIMITS - BASES FOR SECTIONS _4.2.4.2 AND 5.2.17 MINIMUM CRITICAL POWER RATIO (cont'd) established, Reference 5, based on the XN-2 predicted power to the measured
~
critical power, assuring better than 99% confidence, a 99% probability of < avoiding boiling transition. For any abnormal operating transient analysis, evaluation with the initial condition of the reactor being at the steady-state operating limit, it is required that the resulting MCPR does not decrease below the MCPR limit of 1.32 at any time during the transient assuming Limiting Safety System Settings given in Specification 4.0.2.2. To assure that MCPR limit.of 1.32 is not exceeded during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO. The type of transients evaluated with pressure increase, moderator temperature decrease, core coolant flow increase and decrease, and positive reactivity insertion. The limiting transient which determines the required steady-state limit during operation at RATED THERMAL POWER is the rod withdrawal transient. A minimum transient CPR of 1.32 is caused by steady state CPR's of 1.73, 1.49 and 1.50 for Type I, II, and III fuel, respectively. A steady state MCPR of 1.73 has been accordingly established for Type I fuel at RATED THERMAL POWER and RATED RECIRCULATION FLOW conditions to ensure no penetration of the minimum allowed CPR value of 1.32. The corresponding steady state MCPR for Types II and III have been maintained at the cycle 5 value of 1.59 for conservatism and continuity, as Cycle 6 maximum peaking factors are lower than these for Cycle 5. For core flow less than rated recirculation flow, the most limiting transient is the recirculation two-pump speedup caused by a failure of the motor-speed control system. The MCPR limit for steady state operation at flows less than rated recirculation flow are shown in Figure 4.2.4.2.3-1. MCPR values were calculated using a flow control line that corresponds to the l Limiting Power-Flow line of Figure 4.0.2.2.1-1 which intersects 116.7% power ! at a maximum of 110% rated recirculation flow. The recirculation pumps are operated in the manual mode only, and the maximum core flow is limited by pump scoop tube travel and pump capacity. [ j l 1 l l 1 77 ! ( PC4-20h
U "g 3p - ; < > -
^
- b. _,
,e i;) y ,}.
', hrV r qN ... ; ; .
;. , 'w g-1
'I ? 1
+4 ccvoammer - '1 4
es
' POWER' DISTRIBUTION LIMITS v .
a s l h/ y ?
, ', BA'SES FOR SECTIONS'4.2.4.2 AND 5.2.17-
$l ~'l f
]
- c. m u MINIMUM CRITICAL POWER RATI0 f (cont'd) i i .Thellimiting bundle: relative power was'.ad,justedluntillthe'MCPR was equal to
'.1.32 at maximum' flow.. Forl additional conservatism, the transient ~was assumed.
g i to terainate ' at 120*' of RATED -THERMAL' POWER rather than ' expected (116.7% power q
;y ; intercept point. .Using/this relative b'ndle u power as'a basis, the MCPR's
.g were calculated at different flows. The calculated MCPR'a'fornType'I, Type, 4~, LIItand Type.III-fuel were then used to establish the curves:shown in Figure
.4.2.4.2.3-l'. :These.curvesTrepresentLthe minimum allowable operating MCPR for ithe most limiting' fuel assemblies. over the fullt core flow ^ range of permissible. operation.
I ~The: nomina 1' expected flow control line falls below thc Lim ~ ting Power-Flow line and intersects the 100% 'of RATED THERMAL POWER -- 100% of rated ' recirculation flow intercept' point.- The termi~nal.MCPR during'a postulated transient from a' normalelow flow starting. condition, ' therefore, would result in a HCPR ').1.32. In addition, san automat'ic reactor trip would be' expected , L+ to' occur due to power-flow and/or.120% overpower' trip signals during the j transient, but these were conservatively ignored for the analyses' The MCPR. limit. curves ~shown in~FigureL4.2.4.2.3-l'are> conservative and operation with g fgreater MCPR values will' assure that the MCPR limits will not be. penetrated j ' ifor.the most severe operational transient. X.
.At. THERMAL POWER less,than or equal to 15% of'the' RATED THERMAL' POWER, the-moderator . void. content will be small even at minimum recirculation flow.- For all' designated control rod patterns under these conditions, the resulting MCPR value is in excess of requirements by a' considerable margin. With the low void content,' any inadvertent core-flow increase would only place operation'in~a more conservative mode. relative to MCPR.
The daily requirement for surveillance of the core.MCPR -is sufficient since power; distribution shifts are very slow when there have not been significant
. power-or control rod changes. The surveillance of'ecre MCPR af ter power- .!
increases'1: 15% of RATED THERMAL. POWER will assure 'that significant reductions in MCPR.are determined. 4.2.4.2.4 and S.2.17.4 LINEAR HEAT GENERATION RATE 1 The LIbEAR HEAT GENERATION RATE (LHGR) specifications for Type I, Type II and ! Type III fuel assure that during steady-state operation, the peak LINEAR HEAT GENERATION RATE in any fuel rod is less than the design LINEAR HEAT GENERATION RATE. .The specifications also assure sufficient margin to accommodate maximum centerline fuel temperatures less than the melting point during operational transients. I PC4-20b 78 j a
~ a:
'g" 44. ' f 4i bm{ ,
, z.q .
; , c i.
,. ..e
;- Q
- i.
yJr Jc ggggg,y y s e> ; _ , g f E WER-DISTRIBUTION LIMITE
,7 , ,
,4 W
, iBASES FOR SECTION' 4.2'.4.2 AND' 5.2.17
.f iLINEARHkATGENERATIONRATEm(cont'd) w LFor-Type I and Type II (A-C). fuel, the original. design LINEAR HEAT GENERATION
, RATE'specified by the fuel manufacturer was conservatively reduced tol11.94 kw/ft to; account for the' effects of densification, power spikes and -
' manufacturing factors. For. Type III.(ENC) fuel,Jthe design LINEAR HEAT GENERATION RATE of 11.52:kw/ft is 'also calculated with design conservatism. .
1
~
ithatlare larger than'the' calculated axial densification~ effects plus manufacturing' tolerances.and power spike effects, References 6 and 7.
~
The daily Lrequirement for surveillance of; the core LEGR abbve 25% of RATED y THERMAL' POWER is sufficient since power distribution shifts ere.very slow when there have not been significant ' power or control rod changes. The
' surveillance of core LHGR after power increases 2.15% of RATED THEINAL POWER willisssure that significant increases in LHGR are' determined.
4.2.4.2.5 'and 5.2.17.5 MAXIMUM AVERAGE FUEL ASSEMBLY: EXPOSURE 4
References:
8, 9, 10,'11.and 12 Fuel cladding. integrity is a function of many parameters including-fuel exposure, pellet clad interaction, THERMAL p0WER, rate of change.in power density, coolant chemistry,:etc. Therefore, limiting fuel. exposure to 18,000' j MWD /MTU in the non-peripheral fuel assemblies which experience higher than ' average power densities and rate of change of power will give additional-assurance that the condition of the fuel during operation will be satisfactory. It is not necessary to limit exposure in the peripheral core ..;
' locations.since operating experience at LACBWR has shown that.the 28 peripheral fuel assemblies have a much lower rate of failure than the 44 interior fuel assemblies. This trend has been attributed to the lower power
' density at these locetions, and the minimal effects'of control rod movements
- which cause local power peaking in the fuel rods near the tips of the control-rods. The outer control rods are fully withdrawn at the beginning of cycle
'(BOC) and remain withdrawn during normal cycle operations. ,
i PC4-20b 79 r .l
, i 11 i , .
m -) ,' t 3 st..; 1
. . C$M l .y , POWER DISTRIBUTION LIMITS N '
k
. BASES FOR ECTIONS 4.224'2~ ann _5.2.17
);::.. ,
- Maximum Average Fuel Assembly Exposure - (cont'd) ,-l H
. Psliet-clad interaction. is 'a well known and documented contributing . factor. to -
fuel. rod failures; 1The preeence of pellet cladding ir.teraction has been r ". identified:in: post-irradiation examinations of, fuel rods' removed from LACBWR
-fuelicssemblies. . Fuel rods removed from fuel assemblies with! average m; exposure lup to 14',700 MWD /MTU.have been examined. The strength," ductility, ,
JJ, , . and. condition of the cladding in these rods was' found to be adequate as -
' determined by mechanical tests. The examination' further confirined that power
.histor/ of.the rods.:is of prime importance, though not-the only factor in
. contributing lto fuel rod' failure. During' operation the rate of withdrawal of U controt rods,,when the THERMAL POWER is above 25% of RATED THERMAL' POWER, is:
reduced from.that experienced during operation prior to Cycle 5 which also significantly reduces the stresses in the fuel clad.> .
- Minor cled~ defects that may occur would be expected to develop very slowly and the~ consequences of such failur;e would be minimal. Surveillance and. l
' limitations on coolant and off-gas esctivity will assure that' operation does not continue with significant. quantities of. failed fuel.
References:
- 1. '? Technical Evaluation Adequacy of La Crosse-Boiling Water Reactor Emergency Core Cocling System," Report SS-942, Gulf United Nuclear Corporation, May 31, 1972. 4
- 2. " Review of Densification Effects in La Crosse Boiling Water Reactor,"
Report SS-1085, Gulf United Nuclear Corporation,-.May 15, 1973. I
- 3. . NRC' Safety Evaluation Report, Letter, Reid to Madgett, dated August 12, 1976.
1 4, "ECCS Analysis for Type u and Type III Fuels for the La Crosse Boiling 'l Water Reactor," Exxon Nuclear Company, Inc., XN-NF-77-7, March 19#7. / l
- 5. " Transient Analysis for LACBWR Reload Fuel," Response to Question 4, !
Nuclear Energy Services, Inc.,' Report 81A0025, February 18, 1977. I i i i i PC4-20b 80
. . - __ _u_ _ _ _ _ _ _
j
j{T ' .
;fl{ ' .'. i,{ J j .4 4 , m !
.wlfu; .e4 f, 1
+- , %! ' cugyyyy l
!? ' i POWER DISTRIBUTION LIMITS' ' i
' U BASES FOR SECTIONS 4.2.4.2 AND 5.2.17-3 W. R_oferences .'- (cont'd) - .
.1
'j
- -4.. L6. " Description of Exxon' Type III Nuclear Fuel' for Batch 1 Reload'.in the H h LACBWR," Dairyland Power Cooperative, LAC-3929,_May 17, 1976.
' 3, ' l{
..l '
.71 . Exxon' Nuclear Co. Letter, J; A. White to C. W.' Angle,
Subject:
.MAPLUGR H>> ! Limits:for Typei.I~(Allis-Chalmers) Fuel,. dated June 22, 1977. -r 4
, J 8f~DPC Letter, LAC-6846, Linder to'Ziemann,: dated April If 1980. <
4
- 9. ' DPC Lhtter,'. LAC-7572, Linder to Crutchfield; dated June' 1,1981. 7 I
. 10, ~DPC' Letter,' LAC-8109, Linder to Crutchfield,. dated February. 23,.1982. J i.
- 11. " DPC Letter. . LAC-8131, Linder to Crutchfield,' dated March '4', .1982. .
- )
~
- 12. DPC Letter, LAC-ll318,. Taylor to Zwolinski,. dated December 12, 1985.-
.i
'1> i e a 1 i t.- (Next page is page 33) ,
. '{
.8 i
l l 1 1 PC4-20b f51 i j 1 _____ _ __ __ _ l
m ,, , 1
';t 'f l;^ 1 (;o c, l ;' h_'
,)
1 { ,s
; -t ,
." , n :-
i. i n a i 4-
'2 5 i
PntnRnER21
'aA \
Delete Section .:4.2.4.2,' Power Distribution Limits, ' and' its- bases.
., j . < . JfBTIFICATIO\W
,^
< , 1
. i .; .
f- These, specifications-are not' applicable when the reactor is shut'down.
1
-I p ,
s
-1 4
f, 7, , i d 4 l i i
'i 1 t:
l a ( l o
- 4.\
'I i.
-1 1
-(
)
! l l l ( ] l l 1 l
)
PC4-20b 82 l [' l: l .-
- p. I l ': )
t. t. t
# j W ,
)
PAGE 33 ,
;CURR8WT )
l 4.2.4.3 l Deleted.. 1
' 4.2.4.4- The reactor shall not, be operated at a power level above 1 Mwt with o less than 72 fuel assemblies installed in the reactor., .{c t ,
l- ,
'4.2.4.5 .The' reactor shall have'a negative' temperature coefficient for all-moderator. temperatures from ambient' to 5570F or shall be preheated to obtairi;
~
j a negative temperature coefficient' prior to withdrawal of.contr;o1 rods. s H 4.2.4.6 The core s ah ll be suberitical' by a least. 0.5% delta k/k in the cold clean condition with~t.he' rod'of maximum worth in itsLfully. withdrawn position and with all'uther rods.-fully inserte'd.
.y
- 4. 2. 4. 7 ' The reactor may be operated with control rod. drives inoperable- A
.provided that the associated. rods are secured in the fully inserted or fully .=
withdrawn' position by electrically' disconnecting.the clutch and drive member i and provided that all other requirements of these specifications are met.
~
4.2.4.8 .If an unexplainable change of greater than 0.6 percent'in' core . j reactivity is observed in a core which has undergone no physical change, the l] reactor shalllbe shut down. Except as required to investigate the reactivity change, the reactor shall not be operated until a satisfactory reason for the . change has been found. 4.2.4.9 Following ~ core reloading and/or core rearrangement,' operation wil'l'- be discontinued if the difference between the observed and predicted value.
~o f core reactivity exceeds 2 percent. If this difference'is exceeded, the '
effect of the difference on further operation will be evaluated before operation is continued. PIDI M ED k l Delete these specifications. q JUSTIFICATION Since the reactor is permanently shut down and defueled, these requirements do not apply. { l i i PC4-20b 83 i
mm ; . . f
~
r ng>( . , t i y-l. ,( , L
.{-* 3 i
}. .? l k 3 PAGE M
.M ,
CID D DVY 0 t , 1
- .J '4.2.51. Reactor Contioli tt:
I,; , 4.2.5.1 ;' fha tothl scram time of each control rod as measured by the rod '
. scram 41mer shall not. exceed 3.0.sec.
4.2.5.2 -The control rod configurations.used'during reactor'. operation shall . L , be such that the maximum increase in. reactivity; thet would be caused by? the .
, complete withdrawalof any.one control rod.from a critical core would'not
~
exceed-.2.5%' delta k/k.
~
4.2.5.3'. ' The maximum ' rate n't which reactivity $ hall' be incrSased by movement ' J of control. rods when k,rr of the reactor exceedsc0.99 shall not exceed
- 18' cents /sec. ,
4.2.5.4.;The maximum rate at which forced circulation flow rate may.be-rincreased by changes in forced circulation pump. speed shall be.1200.gpm/sec. f
- 4.2.5.5L-The reactor shall not'be' operated above:10-5 .of full power unless at leant one: forced circulation pump is operating at'or above its set minimum. !
speed (34' percent.~of fu'11: rated speed.)
.l
. . l
.4.2.54GL The. speed of.the forced cliculation pumps shall be controlled 3 manually orf automatica11y' by the pressure control system.
' 4.2.5.7 'Except during plant startup and shutdown, the' main steam bypass fj c
^
. valve. control shall be set to begin opening the bypass valve at an indicated 'J pressure corresponding to~a turbine inlet pressure not more than 15 psi above l its nominal value.
4.2.5,8[The'ronpositionIndicationsystems,.i.e.,thesynchropairs',.the 1
,rced switches, and the limit switches which indicate full rod? insertion or i
l withdrawelv shall be available for indication of individual rod positions,. O a --' 'except that lone of these systems may be removed for maintenance for a time .] (, period not exceeding 24 hours. i 4.2.6. Safety Instrumentation 4.2.6.1 The safety instrumentation shall provide scram, is61ation action, 'i and other safety actions as specified in Table 1 of these specifications. 1 NYMYENI) j Delete the entire Section 4.2.5. Modify Section 4<2.6.1 as follows: 4.2.G Safety Ins t rumen tat igil
'4.2.6.1 The safety instrumentation shall provide isointion action and other safety actions as specified in Table 1 of these specifications, pC4-20b 84 L = _- - .
i i
\
l l I i 1 JUS 1'TFICATION } Since the reactor is permanently shut down and defueled, the reactor control requirements do not apply, nor is there any need for setto functions. l i I d 1 l l 1 i l ! i l- I t 1 l l l l PC4-20b 85
1 i PAGW 35 t f}FRENT~ l r Safety instrumentation full scram protection shall always be through ' redundant channels, except that one channel may be bypassed, as permitted by Table 1, provided that' the time during which redundant protection from any one-out-of-two or one-out-of-three system is not provided shall not exceed 24 hours in any 30-day interval. 4.2.6.2 The setpoints for the safety instrumentation shall be as specified-in Table 4.0.2.2.1-1 and Table 1. 4.2.6.3 Key switchea shall permit' operational, maintenance, and test bypass ; of the safety instrumentation only with the approval of the Shift Supervisor, i ar.d in accordance with the provisions of Table 1 and the following ! conditions:
- 1. The requirement of a full scram signal to permit operation of the boron injection system may be bypassed to permit testing of that system.
- 2. The rod withdrawal prohibit action of both source range nuclear channels may be bypassed only at power levels above the normal range of these channels.
- 3. The rod withdrawal prohibit action of both in'armediate range nuclear channels may be bypassed only when reactor power exceeds 3 Mwt.
- 4. The " building locks secured" ir.terlock of the reactor start circuit may be bypassed when there is no fuel in the reactor or when testing or repairing this interlock function, provided the requirements of Sec. 4.2.1.1 are met.
- 5. The rod withdrawal permit circuit niay be bypassed to permit test i withdrawal of a control rod in accordance with Sec. 2.8.6.
4.2.6.4 A control rod shall not be withdrawn unless the conditions of Sec. 2.10.2.3 or Sec. 2.8.6 are met. 4.2.6.5 It shall not be possible to seal in the teactor start permit circuit unless the conditiens of Sec. 2.10.2.4 ere met. 4.2.6.6 During refueling or other changes in core configuration involving fuel element or control rod cemoval or insertion, the core shall be monitored by a minimum of two neutron count detectors.
/*fWPOSNil Delete these requirements, with t he exceptions of Sections 4.2.6.2 and 4.2.6.3, which should be modified as follows:
l l l 1T4-20h 86 l i -- - =__--- __- _ k
- s .:
5, , s 4.2.6.2' The setpoints for the safety _ instrumentation shall'be'as specified in Table 1.
~
4.2.6.3l Key stfitches'shal1 permit operational, maintenance, and test bypass .j
- of the ' safety instrumentation only.,with the' approval of, the Shift.: Supervisor.
mrmem, ,
'Since the~ reactor is permanently.' shut down and defueled,' there is.'no need for scram, rod withdrawal.. 'or reactor start protective ~ circuits..
Reference to Table 4.0.2.2.1-1 is deleted from Section 4.2.6.2'since that-
- . table is being deleted. . Since the reactor. pressure _ vessel is'defueled, Section 4.2.6.6 does not apply.
a f tl l i
'f I
l l-i. fl7 PC4-20b l 1 a
7 1-
, 7
'^
[
't ! .>;
, , , t.
g, ,
; i L_ e, ,
)
;g,, i i
t
,f.,- 4 d' c ' PAGE 36 - -
gy ^ ' ' ' R 7 .Cf4NDW !
;g-4.2.6.7lDuringoperationwiththenuclear'instrumentationchannels.in2of'4; trip l logic, at . least = three of' nuclear channels 5, 6,L 7 and 8, including their.
- automatic gain control subsystem channels, shall- be OPERABU3.
~
If nuclearL 1 channel 5.or) 6.is; inoperable, its scraafcontacts:shall be' placed in the trip 2 position.LIf power,is de-escalated. the. tripped channel's. output'shall-be-a lbyphased prior toLenteringil-of-2: logic, eubject to Section 4.2.6.1
. requirements. .If nuclear channel'7 or 8 is' inoperable.in a manner affecting.
; ' the operability of its corresponding power-flow channel the power-flow
, channel shallcbe bypassed, pursuant to the time' limitations ~ of Section-4.2.6.1,~and the scram contacts of the nuclear channel-shalllbe placed in the. -;
trip position' . b, .. . .
' 4.2.6.8.~ Safety channels directly backed up. by: an identical channel or - .
.l Lchannels may be bypassed for maintenance =or testing. Safety channels inithe; partin1Lscram circuit may be bypassed for. maintenance or testing for ur to
'24 hours.
4;2.6'.9' Both reactor forced circulation pumps shall be automatically shut down.by a high reactor pressure signal or by a low reactor water level- q signal.-
- Bases - The RPTS is a diverse and independent backup except for common current ' sensing loons to the: normal scram system for rapid shutdown of the
. reactor. To protect the primary system from an ATWS event in which either
~
'MSIV closes at pcwer,.thun eliminating the main. condenser as a heat sink,.the:
i- recirculation purps must be shut down -to prevent damage to th6 primary system =
~
i 'due to high pressure. A rapid shutdown of the recirculation pumps has the-
< effect of causing an increase in the moderator voids in the reactor core. A
{ substantial negative reactivity results and. the power and pressure surges < that might otherwise occur in the most limiting transient (MSIV. closure) are j substantially . reduced. With the recirculation pumps shut down, the' reactor 4 power will be reduced to ,a . steady, state' power level of less than 20% (based ] on natural circulation through the core). 4.2.7 - Deleted - ' l PhufDSED Delete Sc3ctions 4.2.6.7 and 4.2.6.9. Modify Section 4.2.6.8 as follows:
-4.2.6.8 Safety channels directly backed up by an identical channel or channels may be bypassed for maintenance or testing.
JUSTIFICATION j Section 4.2.6.7 does not apply to a shutdown reactor. The partial scram circuit is not needed fer & permanently shutdown reactor, nor is an ATWS circuit. pC4-20b 88
. _L-..'_sJ.J.-.___
t
/
iPAGE 37 4
?
.j 4.2.8.4 Irradiated' fuel elements'shall have decayed for at least 72 hours .i i
(-: .
, prior-to placing them in the' spent fuel storage well. . ,;
150fDSED , Delete Section 4.2.8.4. 3 4
.' JUSTIFICATION i
~
iL .Since the reactor has been permanently shut down and defueled, this requirement-is not applicable. 4 i I PC4 20b 89
o r.
,y;< ' :[
i,-
- PAGES 37n-37q CUhmWT
~ PLANT SYSTEMSJ 4/5.2.21 ' EMERGENCY SERVICE WATER SUPPLY SYSTEM
, LIMITING' CONDITION FOR OPERATION ,
4.2.21.1) The Emergency Service. Water Supply System shall be OPERABLE,' with:
- a. At:least-three' portable gasoline engine-driven pumps connected'in parallel"and-capable of delivering a combined flow of 900.gpm
'against a pressure 'of 50 psig within:the reactor vessel'.
- b. An operable flow path from the' Mississippi River,. consisting of pump' suction and discharge hoses, a- three-way ball valve distributor,. and Erelay hoses capable of transferring water to a valved inlet manifold into'the High' Pressure Service Water piping system.
APPLICABILITY:=-Operational-Conditions 1, 2 and 3
, ACTION:
- n. With only:2 pumps OPERABLE,' restore the inoperable equipment to L OPERABLE status within 24 hours.
'b. With the ESWSS otherwise not' OPERABLE:
- 1. Verify OPERABLE status of the High Pressure Service Water Fire Suppression Water System per Sections 5.2.18.1.1.a and 5.2.18.1.2.a within 24 hours.
- 2. Restore the ESWSS to OPERABLE status with at least-three pumps within seven (7) days or be in at least HOT SHUTDOWN within the next twelve (12) hours and. in COLD SHUTDOWN within the following twenty-four (24) hours,
- 3. Submit a Licenseo Event Report. ,
4.- If 1. above cannot be satisfied, place the reactor in at least 1 HOT SHUTDOWN within the next twelve (12) hours and in COLD SHUTDOWN within the following twenty-four (24) hours. L
)
i l-PC4-20b 90 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - - -2
_ . _ _ , _ . ~ _ _ _ _ .- _ CUM /ENT PLANT SYSTEMS SURVEILLANCE REQUIREMENTS 5.2.21.1.1 The Emergency Service Water Supply System shall be demonstrated OPERABLE:
- a. At least once per 31 days by verifying that each manual or automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in its correct position.
- b. At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel,
- c. At leati, once per 18 months, during reactor shutdown, by performing an integrated system test which directs coolant flow through the Alternate Core Spray System test loop and back to the river.
- 1. Verifying that the combined pump output develops 900 gpm at a minimum system test pressure of 90 psig,
- d. At least once per 18 months, by performing a drill which assembles '
the system under simulated emergency conditions. 5.2.21.1.2 Each Emergency Service Water Supply System gasoline engine driven pump shall be demonstrated OPERABLE:
- a. At least once per 7 days by verifying an acceptable built-in battery hydrometer indication, and voltage ? 12 volts,
- b. At least once per 31 days by verifying that the fuel tank contains f at least 4.0 U.S. gallons of gasoline, and that the crankcase oil is at a safe level.
- c. At 1 cast once per 6 months on a STAGGERED TEST BASIS by starting an engine / pump unit from ambient conditions and operating at least 20 minutes with pump load.
I
- d. At least once per 6 months by verifying that the fuel tank has been i drained and replenished with fresh fuel. Whenever fuel replacement {
is performed, each engine shall be started and run for sufficient I time to ensure that all old fuel remaining in the carburetors and ! fuel line is consumed and replaced with fresh fuel,
- e. At least once per 6 months, by performing inspection and maintenance 1 in accordance wit h writt en procedures, j
- f. At least once per 18 months, by performing inspection and maintenance in necordance with written procedures:
I PC4-20b 91
- J
r 1 3 . ov ',
.c . ': , '
..h 8 ..
-ij f c( ,
y .. _g >> - , t
,,,, -p.-
, :Q 3 i[ ,
f t
)
j .%
;c a'.
y y
! ] ..; u .. , .
, -q :
. PLANT SYSTEMS D f k
SURVEILLANCE REQUIREMENTS- [, iu ir . , . , h 5.2.21.1.3'. Each Emergency Service Water Supply. System pump suction hose
, shall be' demonstrated. OPERABLE:
.I a.; , At!. least,l once perg12 months,.'.by tes ting She.' pump suction. hoses ~and
~; .
couplings;at a pressure 1 50 psig, and visually. inspecting for damage 1and' abnormal' wear.
-5.2.21.1.4 3 Each Emergency Service Water Supply Systems' Relay' hose, pump-
-discharge hose, and'.the 3-way' distributor shall be demonstrated t OPERABLE:; '
- n. LAt leastL once per -12 months, by. testing ths 5" relay hoses and-4
. couplings at a pressure 1 225 psig and visually inspecting for damage and abnomal' wear.
- b. . At ,least once per 12 months, by- testing the pump 3" discharge hoses -
- and couplings at a pressure 1 225 psig and visually inspecting for.
damage and. abnormal-wear.
' c. At least"once per 12 months, by testing the 3-way distributor and' a- couplings alt a pressure 1-225 psig and vis~ u ally inspecting for damage,and abnormal wear.
- d. At least onces per 12 months, by verifying that each 3-way
'distributo'r ball l valve has been cycled through at least one complete cycle of full travel, i
L-PC4 -20b 92 x_-_--_______-
Pb .' i 3. K
- 2. l CuhntNT
- ' , ' PLANT SYSTEMS' !
1 BASES 4/5.2.21 EMERGENCY SERVICE WATER SUPPLY SYSTEM'~ The Emergency Service Water Supply System is designed to supply an additional' l alternate cooling source to the reactor. This backup water source is i
. necessary because of the review of site' liquefaction potential during seismic
~
.l events.- The; system enhances the' safety factor.in a seismic eventJwhere free i field ' liquefaction at~ the Safe Shutdown Earthquake' may, according to the results of this review, cause. failure _of the Crib House structure'and'the underground high pressure' service water ~ piping _ connecting the Crib House to
- j the Turbine Building.
The system consists of isolation valves to truncate.the potentially damaged systems inside the Turbine Building, thus closing the pathway to seismically'
- damaged' piping, portable pumps' capable of drawing a suction from the -
Mississippi River _and'the"necessary. hoses:and hardware to connect the pumps to the Turbine Building piping . either inside the' building or from the outside. The system is' capable of supplying 900 gpm flow rate against. a 50 psig reactor pressure. 'The system installed piping can' supply water to the High Pressure Core Spray System, the' Alternate Core Spray System, the Fire > Suppression Systems, the Shutdown condenser and the Low Pressure Service Water' System-(which in'the ultimate heat sink for decay heat). j j-
/WONSRD q
Delete Section 4/5.2.21. JUST1FICATTON This specification was applicable only during Conditions 1, 2, and 3. 'i l TC4-20b 93 l 1
s
. 3yiL_ '
q:; ':' .
, vy m,s ,
h, <'
"jp p ,
,, . ' PAGl' 37rI capisFr
- g 9
.. . . r ,
. PLANT SYSTEMS-i 4/5.2.22.. PRIMARY COOLANT SYSTEM PRESSURE ISOLATION VALVES M ,ITING CONDITIONS FOR OPERATION _ _
_ = . 4.2.22 The leakage through the following pressure isolation valves shall not j 1' exceed 5 gpm;(a, b).
- a. Alternate Core Spray Valve 38-26-001
- b. Alternate Core Spray Valve 38-26-002 0' APPLICABILITY: ' OPERATIONAL CONDITIONS 1, 2 and 3 I
ACTION:o
- a. If seat leakage through one of the valves listed above exceeds the limit, the valve must be repaired or replaced prior to entering Operational Condition 1. Otherwise beLin cold shutdown within ]
'24 hours, i.
. SURVEILLANCE REQUIREMENTS 1
.i I.
W 5.2.22~ Prior'to entering Operational Condition 1 the seat Icakage through' each valve shall be individually verified to be equal to or less than 5 gpm:
- n. After each refueling' outage,
- b. After each cold shutdown of 72 hours or more if not done within past 9 months, land
- c. -Pr. tor to returning to service after maintenance, repair or replacement.
BASES 1 1 4/5.2.22 PRIMARY COOLANT SYSTEM PRESSURE ISOLATION VALVES
-l, To increase the reliability of Primary Coolant System pressure isolation valves by establishing seat leakage limits and surveillance requirements, o
thereby reducing the potential of an inter-system loss of coolant accident. l i Foot not c: ,
.'(a) 1. Leakage rat es less thnn or equal to 1.0 gpm are considered
! acceptable. 1 FC4-20b 94 u1 _ _ _ _ _ _ - - _ _ _ _
'%,p ;4 l;r
' +; .,
; , ;^ ^"
t
;;c
; , '[gg . b 1 ,
.2h: Leakage'ratesgreaterthan1.0gpmbutlessthanorequalto5.0gpm; EH
, ~
are considered acceptable-if the latest measured rate has not exceeded'the rate determined by the previous-test by an amount that' reduces the margin between measured leakage rate end the maximum
. permissible rate, of 5.0' gpm by 50% or greater. ,
- 3. Leakage rates greater than:1.0 gpm but:less than or' equal to,5.0,gpm
' are considered unacceptable.if the latest. measured rate exceeded the.
rate determined by-the previous. test by an amount that reduces the margin between measured leakage rate and the maximum' permissible rate of 5.0 gpm by 504 or greater.
'4... Leakage rates greater than 5.0 gpm are considered unacceptable.
5.'- To satisfy ALARA requirements, leakage may be measured indirectly.
- (as from;the performance of pressure indicators) if-accomplished,in e s., accordance with approved procedures'and supported by computations-1 showing that'the' method is capable'of. demonstrating valve compliance- '
-with the leakage criteria, j
'(b) Minimum test ~ differential pressure shall not be less than'150 psid.
PROMSED .
' Delete SectionL4/5.2.22. l
'l n' .
.l JUSTIFICATION \
This section was applicable only during Conditions 1, 2, and 3. i i 1 1 1 l 4 li ' i PC4-20b 95 l
~ ~ ~
r K. r ; ;,. o,4 , , 9
' f'
Ls' l .; gy ,
;p g yjg n, 3 n x p ;
C(ABBrNT J K . PLANT SYSTEMS. 4 '
. c.
;4/5.2.23. DEMINERALIZED VIRGIN WATER TANK m
- P ' i LIMITING CONDITION FOR OPERATION ~
% .i.
;4.2.23-lThe Demineralized Virgin Water Tank shall be OPERABLE with a minimum . 1:
P 't waterJ1evel'of.1 foot ~.; APPLICABILITY! OPERATIONAL CONDITIONS.1, 2 and 3. i . ' ACTION:
. jy, 6 I' With.the Demineralized' Virgin Water Tank' inoperable, restore it to OPERABLE status within 72' hours, or be in'at least: HOT SHUTDOWN within'the next
'12: hours and in COLD; SHUTDOWN within'the fo11owing.30 hours.
.). '
!)
SURVEILLANCE REQUIREMENTS
' 5.2.23 The Demineralized Virgin Water Tank .shail. be demonstrated OPERABLE by -
verifyingLthe minimum water level in the' tank at...least .once.per 7 days.
.l BISIS FOR' DEMINERALIZED VIRGIN WATER TANK t
A minimum water level .in the Demineralized Virgin Water Tank is specified to assure availability of a short term supply of coolinf; water for;the shutdown .:
' condenser in the event of a loss of the Ultimate-Heat Sink, the Mississippi
~
1
' River. The specified minimum of I foot .of' water in the tank will provide
-sufficient time to allow set up.of an alternate water supply.-
'11)OIDSED
' Delete Section 4/5.2.23. j 4 i JUSTIFICATION This section was applicable only during Conditions 1, 2, and 3.
I 1 l PC4-20b~ 96 I 1 _ _ _ _ _ - - _ _ _ _ . _ _ _ _ _ _ l
b l PAGES 37t-37v. CORREWT il
. EMERGENCY CORE COOLING SYSTEMS
- 4/5.2.24 HIGI! PRESSURE CORE SPRAY SYSTEM
- LIMITING CONDITION POR OPERATION
- i. .
4.2.24.1 The high pressure. core spray (HPCS) systen shall be operable with:
- a. For the high pressure core spray mode:
1.. Two OPERABLE high pressure core spray pumps, and-
- 2. An OPERABLE flow path capable of taking suction from the overhead storage tank and transferring the water'through the-core spray header to.the reactor pressure vessel.
~
.15. : For the low pressure' core spray mode:
- 1. An OPERABLE flow path' capable of transferring water l from .the 7;," '
overhead storage tank to the. reactor pressure vessel by gravi'". q l APPLICABILITY: {
'l
- a. OPERATIONAL CONDITIONS 1, 2 and 3 for the high pressure core' spray ~ l mode, except with the boron injection system operating.
e b.' OPERATIONAL CONDITIONS 1, 2, 3 and 4 for the low pressure core spray mode. t i l i l l PC4--20b 97 j l
.1
__ .jl n *
#. 1 i
' CURMTNT ' )
. ACTION:..
.a. For the high. pressure core spray mode:.
l '. With one of the above required high pressure core spray pumps-inoperable, POWER OPERATION may continue provided-the manual'
' depressurization system and.the alternate core spray-' system are" i
' OPERABLE; restore.two pumps to OPERABLE status within 72 hours: .!
-or be.in at least HOT SHUTDOWN within the next'12 hours and.in l
COLD SHUTDOWN within the following 24 hours. i i
- b. For.the low pressure core spray mode:
- 1. In OPERATIONAL CONDITION 1, 2 and 3, with the Iow pressure core 1
. spray' mode-inoperable, POWER OPERATION may continue provided i-that. the alternate core spray. system is OPERABLE; restore the
. low pressure core' spray mode to OPERABLE status within 72 hours l or..be in at least HOT SHUTDOWN within the next 12 hours and in ;
COLD SHUTDOWN within the following 24 hours. J
- 2. In' OPERATIONAL CONDITION 4, with the low pressure' core spray l mode' inoperable, suspend all operations that have a potential- j '
for. draining the reactor vessel.' i t I l I t i f f PC4-20b 98
,L 7 i
;j O w ,
.l vy ; pgmy
- 4 . EMERGENCY CORE COOLING SYSTEM SURVEILLANCE REQUIREMENTS- 7 5.2.24.1 -The high pressure. core spray system shall be demonstrated OPERABLE:'
- a. At least once per 24 hours by verifying the valve actuation nitrogen d, supply. pressure from the regulator to.be 30 1 10 psig,
- b. At least'once per 31 days by verifying the valve actuation nitrogen supply bottle pressure to be greater than or equal to 100 psig.
a-
- c. For.the high pressure core spray mode: . .
'l. Each COLD SHUTDOWN, if not performed'within the previous -
3 months,'by cycling each' power operated or automatic valve in- , the flow path through at least one complete. cycle of full. I travel.-
- d. At least' once per 18 months, during shutdown by performing a system- ;
functionally test which includes simulated automatic actuation of the i system throughout its emergency operating sequence, and:
- 1. Verifying that each automatic valve in the flow path actuates to its' correct position on.a:
, (a) Release of a boron injection actuation signal for the high pressure core spray mode, and (b) Low pressure core spray mode actuation signal.
- 2. Verifying that both HpCS pumps start automatically upon receipt of a high pressure core spray mode actuation signal.
- 3. Verifying that the valve actuation nitrogen supply pressure regulators operate to control valve actuation pressure at 30 1 10 psig when cycling the associated valves. j
- 4. Verifying that each pump runs when started manually if and only 1 if a full scram signal exists.
- 5. Verifying during a test that each pump operates at a flow rate 1 50 gpm.
- e. Prior to startup during each refueling shutdown by verifying that i each valve, manual or automatic, in the flow path that is not ;
locked, scaled, or ot herwise secured in position, is in its correct position. PC4-20b 99 __ l
pW..in vi -- -
,,g. n s . . ,-;
g:.....,,.,j#, , , I. i;
; ,, ' . 'j -l
{... . ![l';- i is!. ' 4 A ' i j .
., j <
i ; g(. gW. t R5 3 ., 1
.l' Nt _'. *'$*$, , '/
J p!l;ff a 9;_ . . . . . . . f , ,, , Delete Section 4/5.2.24.1L b !: , u-
- p .. .
,y . p ..a.
., 1:
g JEBTiflCMIM
. <. m. . . ,.
. . ,1-Sinceithe reactor is permanently-shut'~down.and defueled,>there,is-no use O ,.'or.need for'an emergency core cooling' system. o n .1'
- , g-
, -+tf-1
-{4 '
t, A. < I f
) ';; . -.
oy 3- , . g .- .4, i ... 4 2 t j t j,l , f i i i I 1 i 1 i
,( -
I
...i
'.. I' ; f
-1 i
,. I ,
I h I 1 1 PC4-20h 100
q @ ; r bi ' i "j y +c' < , -
% i g 'PMiE' 37w CIRBMWT EMERGENCY CORE COOLING SYSTEM-MANUAL DEPRESSURIZATION SYSTEM ~ -l J
LIMITING' CONDITION F0J OPERATION , 4.2.24.2,- The manualdepressurization system (MDS) shall' be OPERABLE with:-
- a. TwoOPERABLEshutdowncondensersteaminl$tvalves,and; 1
- b. 'Beo OPERABLE shutdown condenser condensate.line reactor vent valves. '1
. APPLICABILITY: 0PERATIONAL CONDITIONS 1, 2.and 3. j
> 1 EACTION:
o With one of_the above required steam inlet valves and/or reactor vent
~ '
- valves inoperable, POWER OPERATION may continue provided the high Pressure core' spray system is OPERABLE; restore the inoperable' valve (s)1 to.0PERABLE status within 72 hours or be in at'least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within'the following 24 hours.
u SURVEILLANCE REQUIREMENTS a 5.2.24.2 The manual depressurization system shall be demonstrated'0PERABLE: 1
- n. At.least once per 24 hours by verifying the valve actuation nitrogen j supply pressure'from the regulator.to be 35 i 5 psig. J I
- b. . Each COLD SHUTDOWN, if not performed within- the previous 3 months, .
by verifying that each steam inlet _ valve and each reactor vent valve is manually OPERABLE from the control room by cycling each valve through at least one complete cycle of full travel.
/Yo/usWW j Delete Section 4/5.2.24.2.
I 1 3 Ji/STTFTCA TION l This specification was applicable only during Conditions 1, 2, and 3. l' With the reactor permanently shut down, there is no need for a depressurir.nt ion syst em. PC4-20b 101 C -.__.____m._._
) PAGES 37r-377 ) l CURRENT l j EMERGENCY,SORF APRAY SYSTEM
~
ALTERNATE CORR SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 4.2.24.3 The. alternate core spray (ACS) system shall be OPERABLE with: .j a .' ' Two OPERABLE diesel driven ACS pumps, each with a separate fuel storage tank containing a minimum of 270 gallons of. fuel for pump 1A l and 108 gallons for pump 1B. [
- b. OPERABLE redundant control valves, and 4.
- c. An OPERABLE flow path capable of taking suction from the Mississippi River and transferring the water to the reactor pressure vessel.
APPi,ICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3. ACTION:
- a. With one of the above required diesel driven ACS pumps and/or l redundant control valves inoperable, POWER OPERATION may continue i provided that the high pressure core spray system is OPERABLE; restore two pumps and both redundant control' valves to OPERABLE status within'72 hours or be in at least HOT SHUTDOWN within the l next 12 hours and in COLD SHUTDOWN within the following 24 hours. f
- b. In the event the ACS system is actuated and injects water into the reactor coolant system, a Special Report shall be prepared and j submitted to the Commission pursuant to Specification 6.9.2 within j 90 days describing the circumstances of the actuation and the total ;
accumulated actuation cycles to date. l SURVEILLANCE REQUIREMENTS f l 5.2.24.3 The alternate core spray system shall be demonstrated OPERABLE: a, Each COLD SHUTDOWN, if not performed within the previous 3 months, by cycling each power-operated or automatic valve in the flow path through at least one complete cycle of full travel. i 102 l PC4-20b _A
k
. c.
n d' ' h/ i, l RMERGENCY' CORE COOLING t SURVEILLANCE REQUIREMENTS " (cont'd) f b; ~At-least.once per 18 months,'during shutdown, b'::
~
y -) o .
~\
- 1. Performing a. system functional test which excludes actual
~
injection of. coolant-into the reactor vessel, but which includes simulated automatic actuation of the system throughout its emergency operating sequence, and: .
-(a) LVerifying that'each:
!(1)' Automatic valve in the flow path. actuates.to its:
correct. position.upon. actuation of a low reactor water level signal coincident'with a high containment-pressure' signal,-
'(2): Automatic' valve closes upon'deactuation of the low reactor water level ' signal, and (3) : Automatic. valve reopens upon reactuation of the low-reactor' water level signal.
(b) -Verifying that each. diesel driven'ACS starts automatically upon1 receipt of a high containment pressure signe1. 1
'l
- 2. Verifying that each diesel driven ACS pump operates for greater than or. equal' to 20 . minutes with a pressure greater than or equal to 90'psig, as measured by PI-38-35-801, at a flow rate greater than or equal to 900 gpm. j 1
- c. Prior to startup during each refueling shutdown by verifying that-each valve, manual or automatic, in the flow path that is not locked, sealed or~otherwise secured in position, is in its correct
. position, s)
. IVKWYMiRD i Delete sections 4/5.2.24.3. l 1
JTGTLFICA TION ; l This specification was required only during conditions 1, 2, and 3. There is no use or need for a core spray system for a permanently shutdown und defueled reactor. l PC4-20b 103 l l l - _ - - _ - - _ _ _ _ _ _ _ _ - _ _ _ _ . - .0
T i g
; Q
'+ '
4., , (f M i 16 ; .J
- o- , 4 us: ./M 1,
{u i t; ( e . yl L' ..
' j(sPAM Ms ~ ..
1 ;6, , N '~' y:; i
, s . pg ' ;
- w. ,
- g. . jp{RHGENCY' CORE COOLING SYSTEM-
'y. .: C/"
, , , i
; ii. ' '. . . . ,,
' ' , , 1 OVERHEAD: STORAGE TANK-L i I
@ l LIMITING CONDITION FOR OPERATION 4,
* $4.2.24.'4.J Thefoyerhead storage tank shall be OPERABLE with:
L
, , s .
a.. A minimum contained water volume of 15,000 gallons,' equivalent;tosa level of 40 inches. 16 o q APPLICABILITY:-- OPERATIONAL CONDITIONS 1, 2, 3, and 4.: t' i 5 -i
-ACTIONi:
a., :With the' overhead storage tank inoperable:
, i s .I E , l'. JIn OPERATIONAL. CONDITION'1,-2 or 3[ declare the HPCS system .
high pressure core spray mode: inoperable and be in atLleast HOT 4 -E JSHUTDOWN within 12 hours and in COLD SHUTDOWN within.the-next'
, :- 30 ; hours. -
- 2. fin ' OPERATIONAL CONDITION 4,' declare ~ the HPCS system low - l
'T pressure core spray mode inoperable and suspend all. operations that.have a potential for draining the' reactor' vessel. .]
'V, SURVEILLANCE REQUIREMENTS i
l5.2.24.4 The overhead storage tank shall be demonstrated OPERABLE by: a.-iAt least once per 7 days, verifying the minimum contained water ' volume'in the tank. q
- b. At least once per 18 months, verifying that the demineralized water
- makeup valve opens when tank level is:
- 1. Greater than or equal.to 80 inches with the makeup valve L control switch in the open position, and i
- 2. Greater .than or equal to 50 inches with the makeup valve ;
control switch in the closed position. i i
)
PC4-20li 104 __ _m.._m-___m J
g-
..f i r- 5"'" ,N,'
44. o..' S l' 4, I.[( ; r. L
;g IwO NtSED i
v Delete.Section. 4/5.2.24.4. ~ F JUSTIFICATIDW ' The 0verhead Storage' Tank was required to be operable.to supply water-fortthe High Pressure Core Spray System. Since the reactor is permanently: shut.down and defueled, there.is no use:or'need for a core spray system and' 4
'l "7'
, . therefore' the overhead storage tank should not be required to be operable. j
,t.
q '. J 1
.i ; c' I
l f,
..e l
1
)
1 l j l l l I l PC4-20b 105
=, -
g 3
&,i;ypl .6'L ~\ l? l \
,(, '
;,; bV
s >
' ' }J l '
'.) ll 1 L ~!
f', . W. < > i
'.[ [ h ,
,I lj.;,A [ ^ iPAGES 7/an77/bb . ,
y. 7 u , .,, 33 [
, m.
l, i '
?BMERGENC[ CORE'COOLINGSYSTEMS , . <
-( ,
{ '
5ASES -
n . ON '4/5.2.24iEMERGENCY CORE COOLING SYSTEMS
. The OPERABILITY of two: independent' ECCSisystems : the' high pressure core: spray -
i(HPCS) system and :the alternate core spray (ACS)' system.with the' manual
- h. :depressurization system (MDS) ensures that sufficient emergency, core cooling capability will be available~in the event of a. loss-of-coolant accident- '
- c. < y _ o (LOCA)' assuming: the~ loss' of one' ECCS system through 'any single : failure -
'r l Consideration.' Either system-is' capable of, supplyingisufficient core .
cooling;to limit' the peak cladding: temperatures withinL acceptable limits for. allipostulated break sizes ranging from the ' double-ended. break of the largest -
+ \ reactor coolant'. system cold leg pipe downward.
- 4/5.2.24.1 - HIGH PRESSURE CORE SPRAY' SYSTEM s .The~liigh~ pressure core spray,(HPCS) system high pressure core spray mode is providedLto a'sure'thattthe s reactor core is' adequately cooled.toilimit' fuel y '
. clad" temperature lin the eventL ofl a smal1~ break in the: reactor coolant system and a lossfofccoolant which does not result intrapid depressurization-of.the-l
- reactor vessel. The HPCS. system high pressure core l spray mode permits the--
- reactoreto be shut down while maintaining sufficient' reactor' vessel . water level. inventory until the vessel.is'depressurized. The HPCS system high-pressure core sprayfmode conti_nueslto operate until reactor vessel; pressure
~
', is below the pressure at which alternate core spray system operation' maintains core cooling. - The HPCS system high pressure core spray mode consists of two 4
' N.K pumps,'and associated valves and piping. The pumps each have the capacity to.
deliver.50 gallons per minute to the reactor at a. pressure i'n-excess of-
~
reactor operating pressure. .The system is actuated by automatically starting the pumps ~ on .a' signal from' either one of two reactor water level sensing
~
' channels.
' A function of the HPCS system is the low pressure core spray mode which
.provides, by gravity feed wLich bypasses the HPCS pumps, water from the OHST
' to the reactor vessel when the reactor is at low pressur e or the reactor vessel head'is removed, to provide a source for flooding of the core in case oflnecidental~ draining. The liPCS system low pressure core opray mode is 1
actuated on a coincidence signal from the reactor pressure and reactor water W , 16 vel sensing channels. (, i PC4-20b 106 { _a____:______ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ . _ - . . _ _ _ _ _ . . __ _ _ _ _ _ __ _ _ __
Np i 1 : j tac
.( <
s ,
, ' h 7,
L4 , n: gf y cgggrpr cc A EMERGENCY CORE COOLING SYSTEMS sv: . i 4 I
@ ? BASES- -
l 6> 1 b2b, (4/5.2.24.2L MANUAL DEPRSSSURIZATION SYSTEM and. j 1
- 4/5.2.24.3 ALTERNATE CORE SPRAY SYSTEM:
' j< ,
LAlong with the HPCS system,' adequate ccre cooling.is assured by the f " demonstrated OPERABILITY of the manual depressurization system (MDS) and the. 1
'% elternate core spray (ACS) system.
i 9' u- -The MDS.is-manually initiated.- 'It serves Eo reduce reactor pressure rapidly , so that the'ACS-system can perform its function. The.MDS provides the ACS l
- l system with ' the capability of performing its ' function -in' both long-term and '
short-term cooling. modes. y r The 'ACS system is provideil to assure that the. core is adequately cooled following.a loss-of-coolant accident. The system is comprised of-two fdiesel-driven pumps and' associated. valves and piping. The water supply for- '3
'the ACS., system is the. Mississippi. River. The ACS system is capable of .
,providing 900 gpm of cooling water to the reactor when reactor pressure drops j to"approximately 50 psig. ']
Two Containment Building pressure sensors and two reactor water level'sensore ]; provide the signals 'to actuate operation of the ACS system. A containment
' pressure of.5'psig will cause a sensor to generate a signal to actuate and g automatically start its' respective' pump.- A motor-operated valve associated 1, with each pump will be' opened on a low reactor water level signal from either of two reactor water level sensors coincident with high containment Building pressure, Similarly the second pump will start and the valve will open when
<m the respective set of instruments generate the required signals. The motor operator-for one valve is. supplied with,a-c power from an essential power bus; the motor operator for'the other valve is supplied with d-c power. The W: ,
ACS system is capable of' remote manual start from the control room. -j The surveillance requirements provide adequate assurance that MDS will be OPERABLE when required. A complete functional test results in reactor i blowdown and therefore is only performed during shutdown. ; The' surveillance requirements provide adequate assurance that the ACS system i will be OPERABLE when required. All active components are not testable and a , full functional- test requires reactor shutdown. L 4/5.2.24.4 OVERHEAD STORAGE TANK 3 lC The OPERABILITY of the Cont ainment Building overhead storage tank (OUST) as part of the ECCS ensures that a sufficient supply of water is available for injection by the HPCS system in the event of a LOCA. Demineralized water for the high pressure core spray system is supplied from the 42,000-gn11on overhead storage tank, located in t he dome of the Containment Building. The PC4-20b 107 1
p)f c
.,h (
'hi gh pressure core spray cooling-system is connected near the bottom of the-tank. The Containment Building spray system is connected to a standpipe-within the. tank'with the top of the standpipe located so.that 15,000 gallons are reserved for?the high pressure core spray system.- The demineralized water system replenishes the'OHST for long-term cooling. The contained water-volume. limit includes an allcwance for water'not usable because of tank'
' discharge line location or other physical characteristics.
PIKN M ED I Delete the bases for Section 4/5.2.24. JUSTIFICATION The Section 4/5.2.24 is being deleted in its entirety, so the bases should-also be deleted. i I l PC4-20h 108
'I PAGES 43 and 44 .
1
- CWERENT TABLE 4.3.2.1 i
RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION ^ l ! MINIMUM CHANNELS APPLICABLE f s INSTRUMENT OPERABLE' CONDITIONS ' ACTION ] 1
- 1. Reactor Containment Building Ventilation Monitor System a.- Particulate Activity 1
- B Monitor
- b. Gaseous Activity Monitor 1
- B
- c. Sampler Flow Rate 'i '
Measuring Device. 1
- C
- 2. Stack Monitor System
- a. Noble Gas Activity Monitor 1 ** D
- b. Iodine Sampler 1 ** E
- c. Particulate Sampler 1 ** d
- d. . Sampler Flow Rate 1 ** C Measuring Device
- 3. Air Ejector Offgas Monstor 1 *** E,F e
*
- When Containment Building Ventilation System is in operation.
** At all times, unless alternate monitoring is available.
*** When air ejectors are in operation. 9 s
A. For post-accident instrumentation, refer to Section 4.5.1. B. With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases through this pathway I may continue for up to 24 hours, as long as stack monitors are OPERABLE. j C. With the number of channels CPERABLE less than required by the Minimum l Channels OPERABLE requirement, effluent releuaes via this pathway may 1 cont.inue provided the flow rate is estimated at least once per 4 hours, j 1 D. With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may { continue provided alternate monitoring is available or grab samples are j taken at least once per 12 hours and these samples are analyzed for i gross activity within 24 hours. l E. With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releasem via this pathway may continue provided alternate monitoring is available meeting the requirements of Table 5.3.2.1 or continuous collection of samples with nuxiliary sampling equipment is initiat ed wit hin 1 hour. F. See Specificat ion 4.2.2.22. PC4-20b 109
,- j i CURR&VT ; TABLE 5.3.2.1 .j RADI0ACTIVF GASEOUS EFFLUENT MONITORING INSTRUMENTATION j SURVEILLANCE REQUIREMENTS CHANNEL (4) SURVEILLANCE 4 CHANNEL SOURCE FUNCTIONAL CHANNEL REQUIREMENT 'f INSTRINENT CHECK CHECK _ TEST CALIBRATION CONDITIONS j i
- 1. Reactor Containment Building Ventilation Monitor System
- a. Particu' late Activity Monitor D M Q(2) R *
- b. Gaseous Activity l Monitor D M Q(1) R *
- c. Sampler Flow Rate Measuring Device D N/A Q(3) R *
- 2. Stack Monitor System
- a. Noble Gas Activity i Monitor D M Q(2) R *
- b. Iodine Sampler D M Q(2) n *
- c. Particulate Sampler D N/A Q(2) R *
- d. Sampler Flow Rate Measuring Device D N/A Q(3) R *
- 3. Air Ejector Offgas Monitor D N/A Qcaa) n *
- During applicable conditions per Table 4.3.2.1.
(1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exist:
- a. Instrument indicates measured levels at or above the alarm setpoint,
- b. Instrument indicates a downscale failure (provides control room annunciation alarm only). i
- c. Instrument indicates a circuit failure (provides control room l annunciation alarm only).
(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exist: J
- n. Instrument indicates measured level above the alarm setpoint on one j channel.
- b. Instrument indicates a failure by a Low Flow and Low Count Rate signal,
- c. Instrument controls in Maintenance mode.
(3) The CHANNEL FUNCTIONAL TEST shall also demonstrate that the control room ' local alarm occurs if t he flow instrument indicates measured levels below the minimum and/or above the maximum alarm setpoint. l (4) The CHANNEL CALIBRATION shn11 be conducted in accordance with plant procedures. 1 PC4-20b 110
, u
,,ft,
,;t!Ql' t- 3j ,
h Y e
\ 3 W r
-3n 8
i
+ . - ,
g 5 ' : ;. }.? W-1 ' x
- 2. '
b 7.;p.;
- y. (.i4 FJ :: MED:
-n
% . TABLE 4.3.2.1 ,
. A . , , l
. _ l gc RADIOACTIVE GASEOUS EFFLUENT MONITORING-INSTRUMENTATION ^~ j
' [f '
. MINIMUM CHANNELS; . APPLICABLE
[T' p. INSTRUMENT. ,
' OPERABLE' . CONDITIONS- ACTION' 'L
.. .. . s . , . ..
l4 :1. Reactor Containment Building.
- Ventilation' Monitor, System ggn Ja? Particulate Activity' 1 ,
*r B <
o- 4 E- Monitor ' <. ib. Gaseous fActivity: Monitor 1. *- B! n ,t c. ' Sampler Flow; Rate-
' Measuring Device-1
- C i2. ' Stack 1 Monitor System: .,
t, J
- a. ' Noble Gas Activity' Monitor 1 ** D- i!
- b. 16 dine Sampler 1 - ** E '
** M
- c. Particulate ~ Sampler. 1 E 1 id. . Sampler Flow Rate 1 ** C j
- 4. -
.. Measuring Device
- 13. Deleted.
-)
..i.- q
- When containment Building ventilation System'is 'in operation.
<** At all times, unless alternate monitoring is available.
A. - For post-accident instrtimentation, refer to Section 4.5.2.
- B. With the number of channels OPERABLE less ihm required by the Minimum Channels OPERABLE requirement, effluent' releases through this pathway may' continue for-up to 24 hours', as long as stack monitors are OPERABLE.
C. :With the number of. channels OPERABLE less than required by the Minimum 'I Channels OPERABLE requirement,~ effluent releases via this pathway may-continue .provided the flow' rate is estimated ~at least, once per 4 hours. I
.D. With the number of channels OPERABLE less than required by the Minimum ,
.Channeln OPERABLE requirement, effluent releases via this pathway may l
, continue provided alternate monitoring is 'available or grab samples ore !
taken at least once per 12 hours and these samples are analyzed for gross activity within 24 hours. o , i E. With the number of ebannels OPERABLE Iess than required by the Minimmn Channels OPERAPLH requirement, effluent releases via this pathway may t continue provided alternate monitoring is available meeting the : requirements of Toble 5.3.2,1 or cont.inuous collection of samples with auxiliar;y sampling equipment is initiated within 1 hour. F. Deleted. PC4~20b 111 f
=. :_ : _ __. . _ . _ _ _ - _ _ _ _ _ _ _ _ __ _______ _ _____ _ __
, , ,n ,, _ ,
1
)
q
\
g e' +u.:'g 1 b - i PROIDSRD 1 TABLE 5.3.2.1 j 1 l RADIOACTIVE CASSOUS EFFLUENT MONITORING INSTRUMENTATION
. SURVEILLANCE REQUIREMENTS ]
. 'i CHANNEL .. (4) SURVEILLANCE CHANNEL SOURCE FUGCTIONAL CHANNEL REQUIREMENT- .
CALIBRATION CONDITIONS'
- INSTRINENT CHECK _ CHECK _ TEST
- 1. Reactor Containment - 1
- Duilding Ventilation ']'
Monitor System a.-Particulate Activity Monitor D M Q(1) R
- I
-b.- Gaseous. Activity.
Monitor! . D M Q(1) R *
- c. Sampler: Flow Flate -
' Measuring Device D. N/A Q<a) R *-
U
' 2. Stack Monitor Fystem-
- a. Noble Gas-Activity :
Monitor D- M Oc2) p -*. l
- b. Iodine. Sampler- . .D ' M. y2> a
- D ,1 c. Particulate' Sampler D N/A. Q(2) n *- !
- d. SamplerbFlow; Rate
,D R A Measuring < Device N/A ~ Q(3)
- 3. Deleted. .f
- During applicable conditions per Table 4.3.P.I. q l 1'
-(1) The CHANNEL FONCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarrrl annunciation occurs if ;
any of the following conditions exist:
- a. Instrument indicates measured levels at or above the alarm setpoint,
- u. Instrument indicates a downscale failure (provides control room annunciatioii alarm only)e ,
Instrument indicates a circuit failure (provides control room j
- c. 1 annunciation alarm only). 1
(?) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room ) alarm annunciation occurs if any of the following conditions exist: j
- a. Instrument indicates raeacared level above the . alarm setpoint on one !
channel. 1
- b. Instrument indicates a failure by n Low Flow and Low Count Rate signal. I
- c. Instrument controls in Maintenance mode. j (3) The CHANNEL FUNCTIONAL TEST shall also demonstrate that the control room local alarm occurs if the flow instrument indicates measured Icvels below the minimum and/or above the reaximum alarm setpoint.
(4) The CIIANNEI, CALIBRATION shall be conduct ed in accordance wi th plant procedures. PC4-20b 112 , l 5
- _ _--____--_-___-___O
.RISTIFICA TION Since the reactor is permanently shut down and defueled, the Air Ejector Offgas Monitor is not needed and so it is deleted. The reference to Section 4.5.1 in Table 4.3.2.1, Footnote A, is being corrected to 4.5.2. i i
PC4--20h 113
=' -
'z~3 7 ,
u; , ;;.- _7;y%, , . c , e -t.
. w n
w
- g. .
I 8 ,s
,I*
i 7 6 * ( t: , t.-
) ., . fj -
V PAGR '499 i
'i t ;. , [l c' ,
a' .
. .i . ,.
1
> RADIOACTIVE GASEOUS EFFLUENTS: ;-l
, s .a
' GASE0!!S' RADWASTE SYSTEM, '!
T-t 9 4 ' LIMITIfIG CONDITION FOR ' OPERATION u - - ' J L' . . . . . y , 4.3.2.5 'A GASEOUS RADWASTE' SYSTEM shall be in operation. L , APPLICABILITY:' 5Whenever the main condenser . air ejector. system. is in- ""' . operation. 4
, -J 4
ir [ ,
; ACTION:
L f
- a. 'With tbeGASEOUS RADWASTE SYSTEM' inoperable for more'than 7: days, 1 prepare and submitLto:the Commission within 30 days, pursuant
's
, Lto' Specification 6.9.2,.a-Special Report whi'h c includes'the
- following information:' a
'I'. Identification: of the inoperable equipment or subsystems'. and:
- the reas'on for nonoperability.
i , 2. Action (s) taken to' restore the: inoperable equipment to OPERABLE- :l
. status. ' l L :.
3
- 3. Surmnary description.of< action (s) taken'to prevent a recurrence,
- b. - The provisions of. Specifications 3.0.3, 3.0.4 and 6.9.1 are not f applicable. j i
SURVEILLANCE REQUIREMENTS
.1
'5.3.2.5 ~ Cumulative doses due to gaseous releases to areas beyond EFFLUENT =l RELEASE BOUNDARY will be calculated at'least once per 31 days in accordance f
'with methodology and parameters in the ODCM. d I
Ph0f M ED ! Delete'Section 4/S.3.2.5. 1 j
. JUSTIFICATION Since the reactor is permanently shut down and defueled, offgas is not being generated in the reactor and the main condenser air ejector system will i not be receiving any offgns. Therefore, the gaseous rndwaste system is not ,
necessary. l l l PC4-20b 114
.u--_.______ _ . _ . _ _ . . _
m ,
. , r {, /
i
',T i ,;
c g' , Eh ' i i j '.,j,, t r py1 m , , py; -f ,
-e
\ . ;p: ' [;* l ] .
. pAgg SS; l
! m: l CURHWT L ,
. y
% '4/5.3.2.5' GASEOUS _P OW4STE SYSTEM
.- [ , The ' OPERABILITY of the gaseous radwaste system er.aiures that the system will be ^-
available for use whenever; gaseous ' effluents require treatment prior to X-? release to the environment.', This specification implements the requirements ;
'" 4 of '10 CFR Pr"ti 50.36a and General. Design Criterion 60 of Appendix /.. to'10 CFR j'
Part;5G. , , o L t
'.pgfqqgg~~ '
.1 ,
.{
. Delete t'he baces for Section 4/5.3.2.-5.
l
' JUSTIFICA TIM i
Since Section 4/5.3.2.5 is being deleted, its bases should also be.
. deleted. 1 1
- l i
i l l 3 1 l
)
PC4-20b 115
- - )
a .. W
~
g rv
; ' M M]? '" ,
1 W 'r p f., r , ,,f-)*.' ' i Y ,Mp ;.y, * ' ., y. . . ,Mf' A M, b Jh. ;
~
4
-7 i i ,
e' ; ,
. p3gg Sg .
(d;' .
}L m
, CMNRM' D
%p y ,
- i i
~4/5:5) L ACCIDENT MONITORING INSTRINENTATION 1
j4/5.5.1 . CONTAINMENT PRESSURE. WATER LEVEL & HYDROGEN CONCENTRATION
'j-J POST-ACCIDENT IPSTRINENTATION y
,i , ,
-LIMITING CONDITION FOR OPERATION . J.
;v -
.' 4.5.1' The'two channels
- of the containment pressure, water level and
,;' . hydrogen concentration indicating systems.shall be. OPERABLE: 'q s
* (the.two
~
containment water levels utilize a' common readout) ] APPLICABILITY: , OPERATIONAL CONDITIONS 1 and 2 ACTION': ! h- . .
' t. a. With one or more post-accident' monitoring instrumentation channels;
- inoperable,.' restore.the' inoperable channel (s) to OPERABLE! status.
L .within 30 days or be in at least HOT SHUTDOWN. 3 b.' The provisions of. Specification' 3.0.4 are not applicable. . ' i n ;
; SURVEILLANCE REQUIREMENTS' .'.; i
}
.i 5 5.1 Each of the above required post-accident monitoring instrumentation channels shall'be demonstrated OPERABLE by performance of a: '
.. j
- a. CHANNEL CHECK at least once per 31 days.
- - b' . CHANNEL CALIBRATION at least once per 18 months.
PROIDSED } i
" Delete Section 4/5.5.~1. .,
l JUSTIFICATION This section was applicable only during conditions 1 and 2. 1 i l PC4-20b 116
., , c au
- . ;- - 7 l15 3-p .W 3 ir~ .,
[; _ 3 _ e - n , ; , y , , , PAGE 5-1
;1 fggWBr/MP: >
~ 5.1.1;;Mainten'anceoperations,includingrefueling'operationsand: routine j\
~
tests, L ahall be performed'in.conformance with these' specifications.. i
; 5.1.2". Maintenance operations on the, reactor coolant system, reactor . -l
- components,'orfemergency; systems shall be-performed ~as. authorized by the t
; Shift. Supervisor.whenever.fue1~is in the reactor.- Maintenance' involving the- :l s , ,
; opening of systems 1containing radioactive materials'shall be conducted under the surveillance lof'a Health Physics representative ~.
( 5.1.3, Maintenance' operations which involve the. opening. of the reacto'r j
' prima'ry coolant system shall be performed with the reactor' shut'down and !
Ldepressurized.to atmospheric pressure.. The reactor primary coolant auxiliary j Esystems may be serviced during reactor' operation if the' system is first-t l C .is'olated,fdepressurized totatmospheric' pressure, and cooled.to-1500F. The. l 111mitationstof this par'agraph do not npply to the' supervised collection of :l
' steam or water samples,f and supervised venting of process instruments. ;
LPM 2R SEW .; i Delete Section 5.1.3. Modify Sections 5.1.1 and 5.1.2'as follows: 15 .1.~ 1 Maintenance operations and routine tests shall be performed-in ! conformance with these specifications. ;
}
.5.1.2 . Maintenance operations ~shall be performed as authorized by the Shift
- Supervisor. . Maintenance involving the' opening of systems containing-
. radioactive materials shall be conducted under the surveillance of a Health :
Physica representative. i JUSUFICATION Since the reactor is permanently shut down, Section 5.1.3 does not apply. Since the. reactor..is permanently defueled, no refueling operations will be'taking place; therefore, reference to refueling can be deleted from Section 5.1.1, Additionally, without fuel in the reactor, Section 5.1.2 as currently worded does not apply. The proposed revision would require maintenance still be performed as authorized by the Shift Supervisor. 1 PC4-20b 117 P ______E______I_.
t _ / il. , i ,. I > T
;f f..'
= 5.1. 6. , All , annual' tests requiring the reactor.'to be shut down may .be performed atLthe nearest scheduled shutdown, provided that the interval between tests'does'not exceed 16' months. The interval between semi-annual tests shall1not' exceed 8 months.-
, 5.1~.7 Definitions applicable'to Section 5.2.15 are as follows:
Channel Check: .ALqualitative; determination of acceptable operability by at Observation-of. channel' behavior during operation. This determination shall.
" include, where possible, comparison of. the channel with other independent
. channels measuring the same variabley uW - Channel Test:J Injection of a simulated signal into the channel,to verify its
! proper response including,-where applicable, alarm and/or trip initiating-
' action.
Channelfcalibration: : Adjustment of channel output suchlthat it responds, ~ with acceptable. range,and ac' curacy, to known values of..the' parameter which-the. channel-measures. Calibration shall encompass theLentireLchannel, including equipment actuation . alarm, or-trip.- 's a
- PMfMED Delete Sections.5.1.6'and 5.1.7.
L-- J751717CA170W l l .Since the' reactor:is permanently shut down, Section 5.1.6 does not l l apply.- :The definitions contained in Section 5.1.7 are redundant to those in
.Section 4.0.1', Definit ions, and therefore are not needed.
H l? h l f-l l PC4-20b 118 l l j
e---- .. t , ; H .1 s L! . j :.6 , t g U , ,
; pygg 5_4; :b .
m casarer
-(e)' Test-Frequency: Type B tests (except for air locks:and electrical penetrations) shall be performed at intervals no greater:than 2 years. Air.
- locks shall be tested at 4-month intervals.4 The freight' door shall be. tested -i
. followingleach closure prior to_ plant startup. Electrial-penetrations'shall be tested at; intervals-no' greater than l' year. Type C! tests shall'be-
. performed during'each reactor shutdown for refueling, but in no case at
. intervals. greater than 2 years.
1%lGFOSED i
.. (e)' Test Frequency: Type B; tests.(except-for air locks and' electrical:
- penetrations); shall. be performed st intervals no1 greater than 2 years.. Air.
locks shall be' tested at 4-month. intervals. LThe freight' door shall bb tested: followingieachiclosure.- Electrical' penetrations shall be tested at. intervals
;no greater than 1 year. Type C. tests shall be performed at intervals-no
. greater than. 2. years.
1
> JEBTIFICATION
;Since the plant has been permanently shut down, the current requirement' nin Section 5.2.1.2(e).to test'the freight door following each closure prior
. to~ plant startup would not require ~ testing to ever be performed again. '-The-proposed version'is more stringent,' requiring testing ~of the freight door following'each' closure. . Also, the words "during each reactor shutdown for refueling, but. in no case" were eliminated 'trom the Type C test frequency,
=which was'left to be at' intervals no greater than 2 years.
, i s i
\
i (L t PC4-20b 119 p. L_'
pr < .o . . 1~ ,
- r. 4
, [ ,, .I; 5 3 ,A .
j-
.. h .
r,
;)
1
. pgS 5 3
. CalBRDVT
.q. m '
- 5.2.2.l' ' The reactor . building isolation system will,be tested for proper.
tr
- operation prior l to every cold startup, but this' test will not be required - 1
.more often than at 30-day intervals. ]
1 m u so ,
'512.2.1 .~.The reactor building isolation system will be tested.fr proper operation at.least,once per 18. months.
I JUSTIFICATIORf ; . Since the plant-lis perinanently shut down, the reactor building isolation system would never have to be tested again per the current requirement. The '- proposed version would. require testing at least once per 18 months and so'is
~
,]
more' stringent. t q l y
.J 1
L 4 I c o
'.L a
l l l i PC4-20h 120 4 s _.--m.___.m .___. __,,_m. _ _ _ _ _ , _ _ _ _ _ . _ _ _ _ , _ _ _ _ , , _ _ _ _ _ _ _ _ _ , , _ _ _ _ _ _
.c , ,
9 x 11 a
, . e I t )
PAM 5-6 CIMENT. 512.4 The reactor. vessel'shall be hydrostatically tested at 1400 psig after. any of its'gasketed joints have been opened and resealed. All hydrostatic tests shall beiperformed with the' vessel at,a temperature no lower than that. < I specified,in Section 4.2.2.4;
. 5.2.5 The forced circulation system controls and automatically-operated valves shall be tested for proper operation at each refueling shutdown with' ,
testiintervals not'to. exceed 18 months.
' 5.2.6, - The . shutdown condenser system. contro1 ' valves shall be tested at least quarterly to demonstrate their operability. The integrated system shall'be tested for proper operation'at each refueling shutdown with test intervals not to exceed 18 months. In addition,,tho' condenser tube bundle shall be pressurized to greater than 1250 psig and tested for leakage at each refueling shutdown.
5.2.9 -The. boron-injection l system controls and the remotely-operated valves shall be- tested for proper operation during cold shutdowns but not required more often than every 92 days. PRGiuMI)
- Delete Sections 5.2.4, 5.2.5, 5.2.6, and 5.2.9.
JUSTIFICA TION Since the reactor is permanent 1y' shut down and defueled, there is no need.to hydrostatically test the reactor pressure vessel. Since the forced ; circulation system,. shutdown condenser system and boron injection system are not required to be operable,.there is no need to test them. q i I i i PC4 -20b 121 ~ _ -___ -_--___ _ --_ _ ___- - _
a f l 1 pygg - w
'n l 5.2.12' Control: rod scram time tests shall be performed and the total scram insertion time shall be demonstrated:to be within the limit specified in Section 4.2.5.1',
-(a) .for all rods prior to each cold startup after,the reactor vessel head has been removed,- but at least annually unless 50 percent of the rods have been tested'during the previous 365 days as described in (c) below, .
.( b) for specifically.affected individual control. rods following I maintenance on or modifications to the control rod 'or control rod drive mechanism which could' affect the scram. insertion time of those epecific control rods, and (c) - for 30 percent of the individual control rods' (3 rods) on a {
rotating basis prior to resuming power operation following each reactor shutdown, hot or cold, but not required more-often than every 30 days. 1 5'.2.13 Each control rod drive mechanism shall be exercised by moving each partially or. fully withdrawn control rod at least one-half inch in any one direction, at least once per 31 days. 5.2.14 Proper operation of bothL control solenoids for each of the hydraulic scram' valves will be determined in conjunction with each scram time test required by Section 5.2.12.a. PROI M ED Delete Sections 5.2.12, 5.2.13 and 5.2.14. JUST1FICATION Since the reactor is permanently shut down and defueled, there is no need to test the control rod scram times or control rod drive mechanisms. PC4-20b 122 1 I J
f . %
} (
, < lg* '. 46 I
, , J ;l g- ,
' ' l i
5 , . l' pjgg5-72 Sajo
.4
, CnmW 1
- ?w <
MINIMUM FREQUENCIES., FOR TESTING, CALIBRATING, AND/OR CHECKING OF. INSTRUMENTATION
]
1 CHANNELS : ACTION l- MINIMUM FREQUENCY : l D r 1 l _
- .- l
!1.iReactor Water Level l' Calibration ..At each refueling shutdcwn..- .
;:' L: :: l l
! Test *' l Monthly when in service.and prior ~.:
. 1- l .l to each reactor startup'if test 1 I' l'has not been performed within 30 l
- ! ' :. I days. -l
.+ i . .
l'. ! Check Daily. 1: : : :
- . l l .
l2. Reactor Pressure ll Calibration l At each refueling shutdown. l
- 1. l Test * ! Monthly when in service and prior
! ! : to each reactor startup if test
- : has not been performed within 30 l l~ ! ' days.
- i= ! Check .l Daily.
l t l- _ l l- l l l
- '3. Reactor Power - Flow' Calibration : At each refueling shutdown.-
l l-Test
- 1 Monthly when in service and prior l l : to each reactor startup if test l l l l has not been performed within 30- l-
! ! days. !
- : ! l
-l l Check l Daily. !
- : : l l- :. l l
!4 Reactor Coolant Flow: Calibration At each refueling shutdown. l l' Rate Low l l 1 : Test
- l Monthly when in service and prior :
1 l l to each reactor startup if test ! l l l has not been performed within 30 l l days. ; l
- :~ : l l
. Check : Daily. l l ! l l t l l l l5. Source Range Test : Prior to each reactor startup if :
! (Channels 1 and 2) : (60 cycles : test has not been performed within:
- per second) 30 days. l l
l l l
; Check l Once per shift when in service.
,a l, l l l PC4-20b 123
i ,
,a COMENT i MINIMUM FREQUENCIES-FOR TESTING,. CALIBRATING, AND/OR' CHECKING 0F INSTRUMENTATION -(cont'd)
'~
'l CHANNELS ACTION : MINIHUM FREQUENCY !
l l l l l6 Intermediate Range. l Test (10-10
~
- g. : Prior to each' reactor lstartup if l-
-l ' (Channels 3 and 4). l Land-10-5 amps;: test'has not been performed withinl-30 days. i q l ~ period *) -l
! !. l. . .
l l- l Check l Once per shift when in service.. :
?!- ! l 't
!' l .
l l
-l7. Wide Range and Power! Check.by heat : Monthly.when in service. .:
- Range-(Channels 5, -balance l l
.I G, 7, and 8)- l l- l
~ l l l l
- A. Nuclear Instru- l Test * ! Monthly when in service and prior i
' mentation and ! : to each reactor.startup if test l l- Automatic Gain l l has not been performed within 30c !
l Control Sub-System! !. days. l
, l -l :-
!. -B. Nuclear Instru- 1 Check Once per shift'when in service. 'l
't mentation tmd l l l 1 Automatic Gain l l l l' Control Suis-System! . l l l- l l 2
l C. Automatic Gain l Calibration ! At each refueling outage. l l Control Sub-System! l l l l l~ l l NOTE: . Testing of.the Nuclear Instrumentation and Automatic Gain Control l l Sub-System shall be done concurrently. :
; l- l l l8. Full Scram Circuits : Test for hot : Once a month. l
! l short by means! l l l of built-in l .-
l l test switch. l l l ! : :
!9. Area Radiation : Calibration ! At least once per 18 months. l l Monitors : ! !
! ! Test : Quarterly. l l l l l l Check l Daily. l-l l l PC4-20b 124
d 4l 3; n __ . CIAIRFNT MINIMLN FREQUENCIES ' FOR TESTING, CALIBRATING, AND/0H CHECKING OF INSTRUMENTATION - (cont'd) ; L- CHANNELS : ACTION -l MINIMUM FREQUENCY- l 1 l10,'11, 12- !. l l
- DELETED' :l l l-1 l ,
l
} .
-l l
'!13. Portable Radiation -l Calibration l Semi-annuolly. :
J
- Detectors l l l
-l l Check l Every two weeks.
l l l- l l l 114. Main Condenser l Calibration l At each refueling shutdown. !
.: , Vacuum 1. .l !
- Test
- l Prior'to each plant startup if : .,
.! l l test has not been performed within l l : the last 30 days._ l l' l l :.
! .! l l
- l15. Reactor Building l Calibration l At each refueling shutdown. :
Pressure l l l
.: : l l
': l l l-l16. Low Main Steam l Calibration -l At each refueling shutdown. i.
Pressure l l l
': l l l
! ! l !
il7. Reactor Building -- l Test
- l Prior to each plant startup if .!
Main Steam' Isolation: ! test has not been performed within: l' Valve : 30 days. l
! : ! l i
! l l l !
.!18. Turbine Building i Test
- l Prior to each plant startup if : i' l Main Steam Isolation: l test has not been performed within:
': Valve l 30 days. l l l l l
~l l i19. Reactor. Building MCC: Test
- l At each refueling shutdown. l l -1A Under-Voltage l l l Relays l l l l l :
l20.2400 V Busses lA and: Test
- l At each refueling shutdown.
l' IB Under-Voltage l l l l Relays ': : ; l PC4-20b 125
,.E'}[~ g , f /. ' r 4 q , CIABEEW .f i MINIMUM FREQUENCIES' . FOR TESTING, CALIBRATING, AND/OR CHECKING OF INSTRUMENTATION - (cont'd)- i l- CHANNELS : ACTION ~l MINIMUM FREQUENCY ! l l . . l'
-' 21.CRD Accumulators Low! Test -l Prior to each plant startup if: l-
- . . Oil Level Scren :: .: test has not been performed within:
.- Relay .:'30 days. l l l .l' : i l -l .: .
}'
l22.CRD Accumulators Low! Test- .l Prior to each plant startup if' !
.: J . Gas Pressure Scram- ! test has not.been performed within:
!- Relay.. lJ l 30 days. l l !. l l ~ Check Pressure: Weekly. l l l' Indication ~!' l l l :
l l l 123. Turbine Stop Valve. : Test : Prior to each plant startup if
.: test has not been performed within!.
j ! : 30 days. I'
-l : ! :
.'t . l l l l24.Reector Pressure -l Calibration : At each refueling shutdown. ::
l (RPTS) l l
-l ,
l l 1 l l l l25. Reactor Water Level : . Calibration l At each refueling shutdown. l
'l (RPTS) .l l l'
! l l :
- l26. Reactor Safety Valve: Check l Monthly. !
l Position Indication l- l l l- l Calibration !-At each refueling shutdowr.. l l l !
- Test shall include tripping of the scram relays K-ll4.
PIDIYA9ED The table entitled " Minimum Frequencies for Testing, Calibrating, and/or Checking of Instrumentation" should be modified as follows: PC4-20b 126 J
m - - - -
+
.y s
. t. .
h, $. .y, , ,
* " MINIMUM FREQUENCIES FOR TESTING, CALIBRATING, AND/OR CHECKING OF INSTRUMENTATION w
l CHANNELS ~ ! ACTION : MINIMUM FREQUENCY- ! 11.~ Reactor = Water Level :. Calibration l At least once per-18 months. .': l l' !. -:
!" l Test
- l Monthly when in service. !
-l- .: Check Daily. l
!. l : ;
!2. Area Radiation l Calibration : At least once per 18 months. !
' !- Monitors ! ! ! .;
! Test : Quarterly. ! !
! -l^ !
'l' Check : Daily.
l- l
- 13. Portable Radiation l Calibration l Semi-annually.
Detectors -l l l l l Check l Every two weeks.
- ! l l-l l l l l4. Reactor Building l Calibration At least once per 18 months. l l l Pressure l l l
- - l l l i
JUSTIFICATION The instrumentation eliminated from this table is not. required to be operable for a permanently shutdown, defueled reactor. The instrumentation functions were to shut down the reactor, initiate the Shutdown condenser, reduce power in the event of an ATWS, or to provide position indication for the reactor safety valves. The surveillance frequency for reactor water level and reactor building pressure was changed from "at each refueling shutdown" to "at least once per 18 months" so that some calibration would still be required of these channels at this time. Also, the water level channel test frequency was_ changed to be just monthly when in service, eliminating the portion dealing with prior to reactor startup. l , PC4-20b 127
g _ u q.;
)
. PAGE S-10 ' $
1 camar q ] , 5.2.'16. Corrosion' test coupons shall.be inserted.in the forced circulation-U loop to evaluate. the corrosion ' deterioration of. chromium-molybdenum piping; and that piping shall be replaced if the reduction in' pipe wall. thickness,' as
- indicated by. weight -loss and metallographic evaluation ~ of the . test ' coupons, is greater.than 0.190. inches. . The' replacement. piping shall be stainless.
. steel or! shall be clad internally with stainless steel and shall meet the design' requirements of Section 2.3.2' l: 1 JYDIMED
,e
' Delete Section 5.2.16.
JilSTIFICA TION . Since the reactor has been permanently ~ shut-down and'defueled, monitoring of forced circulation loop piping is not necessary. I PC4-20h 128
I a n edoMl on ,na S net ef o m I m 'r a nsc3 n i r V l on asal ndt . l on HO ef o h aesaeag ef o CR - n i . cprdhsrn n i _- TP ~-
~
ndt g yi ecsbi ndt aeag : I _ aean h' bne ait WS hsri t oeh xnyae e c e pl s h srn SS csbt csbi YA e ais BbweOb ct ait EP n e pl e . - e pl s
)'
KY o nyat ) nyae
.B N Obc' -
1 2 Ob ct T r s 's _ e t r ae s ee .G n l p fl nm e f a d I a OV n E - o n t r mC f otf ne a oieoed rns tl . m m m m cweeat ene a a a S ot rl uriH r r r .d auiOua c c c lt rst stt N S S S l ueond onn O - uh pl enl oe FSOCV aCCV I l l l T l l l C u u u )) ) ) A F 12 3 F ' F 4
~
1 'l 1 S 1 1 1 T . . .
' I 2 2 2 M g 1 I 2 2 2 L . i E T 0 0 0 s L G N p B N I 4 4 4 A I O 5 T T P e e e 2 A l l l 3
, R T b b b 1 E E a a a P S T T T 1 O t r n e e w c o r 4 p e d . p r rnsl - o aail 5 R e ~ u f 3 O l 7l f i< S c e l y N u , vf r6 s e t2 E n6 eo a i r n e S r ,l t er l ol w e n a f r ao R u5 rn cu5 eo h S O o ee vp C l' f swc n e e L l or ll l r rl E f epe re l a ue sn N onnf p enru e N h nef l sn A oai5 t aw c ea H wh ih of u rh C TC82 eCpo N PC e r 8 d u s 1 r o e i s 5- w r e o e r
- P P P 1 r r r 1 N o o o T O t t t t 5 V I ch cr ch T ag ao ag
_ S E E I ei eh ei _ n D RH RS RH _ G R N A U O . . . P C C 1 2 3 l !
e -,an a S m .a m:r a m p o b I r r : 3 r V l on l on l on u l yon
- O efo ef o ef o -
el afo nem HR n i n i n .i o i CP ndt . aeag ndt aeag ndt aeag
. t nn anA eag dt . ~
T IS hsrn h srn hsrn l h a7 srn WS csbi csbi csbi e ch _ sbi SA ait ait - ait u C mait YP e pl s e pl s e pl s f -e e pl s EY nyae nyae nyae nrt yae RB b ct Ob ct Ob ct e O oIb ct . O. L ! h - t d - . hg r n r l e o a o~ a ' v i t thv o .H c e cgo B b a f e amv eal aum eoe I
- or os mnC pfS - sfTt Rea - RhrR t V -
d n 9
'2 ao rieu mo goa nso ne a
A a ct rPenipewH l .. m m m a a a m S au rcit yro . r r r o i syudaB ud yet euen
^
c c c r l t sasll swavrssdi N S S S f li eroiosoocl aeona O u n r pl ustll eatil or I 'l l l y FIPSCBI ICBDVSDCCD T l l l r ' C u u u a )) ) ) ) A F F F v 12 3 4 5 o t d l 1 1 1 e a
- - - t n 1 1 1 t il
. . . i me 2 2 2 m ov r ne 2 2 2 e l p w-T 0 0 0 od N e l e I 4 4 4 b et O b a P e e e l l
c l l l i T b b b a "2 d E a a a h l n S T T T s {_ i l e v l e) . n ro ei t t y aa y R y y t wre t e O t t e f S e e f d p N f2 f2 a2 eo a2 E a - a S t S
) S S r S r r ad r d o o l o rn l o
' R w w e ua e t O o 1' o1 v1 t ap v1 e
n l l e o L Fl Fl Ll su Ll c E - e - e e t e ( N rn rn rn d a rn N en en en ee en
- A H
wa oh wa oh t a ah dh i t a ah S C PC PC WC or WC T vo I nt M w uc l e I o l a L w l e d e v o F v er L e G l e t ag N F t L I - n cn r T rl a r ii e A ea l e d r nu t a R wm o t W E or o a id P P o n C W r O w l l rb ro r ae o
- N oA oL o nu t c
O t t t i f T V 1 I ce ce ch ag m oe aw eo T at at E E I RR ea RR ea ei RH Nh RL I L D ( t R B N U A O . . . . C T C 4 5 6 7
I S foo. pana l un edo pau at an pspar I ~ 3 y n i ylh yeyl o V l b n ndt . bps btb p
.O el o aeag .
HR neei h srnegd egeguo CP nnbt ag
. csbib nn bnb ntdi T an hayrn "e ait pl syria i
yryrau
' I S nyaeaup WS chabi auauth SA C mit Ob ~- ctMd u .MdMdL e
YP EY nrAae l s
)
.)' .-
) )
KB - p Oo7 ct ' l 2 L 1 2
~^
a n
- g. m u h g
r - er o
~
- i l i - P
~ i .at in 'e e i t s n
t - ,t n tt kv H c : e
~. aa amv e nd enroVsne -- es nl aa f e or eal d n
V aeC - p ne sve H 'e v f TV - os - R tVea o ft mf alfdl one mnC p mfS mC - oeaogaona og ao m ao r go n a rns l D fV oV eir rieu
._ en ef eC rt a ct rP ceni cwe rIt rOrr nunh S au S rit S ot u eu:eumrsec lt sa isy ud a .
lt r d a snl std sauot s l sll
+. N oot onaoetl ei li er l oio l ue O l i ul eel t eCRD unrp ul us uhp
- I CtOC mHCSR FIPS FCBI FSO T .
) -
C ) ) ) 0 )) )) )) A 7 8 9 1 12 12 12 _ ~ d e l
'. t e a v c a i t r
. d nT T
wn oI e cn N l g re ep I el h O B a PO P nl ie "9 0l T E "2 1 oe mv 1 9l u C S 3NL 1 1F ^ 2 e2 r v - o l r y .ao
.V R t 1 O e g 1 S f s nn N a e i os E S h di y
) S c l t a d l t i al R e i ul e
't O v3 w B oR n e S s o L Ll rI e c E e m o r
( N rn u t mu N en u cas
- A t a c aeo I
I ah a etl S C WC V RSC T I M n I l e L e p v g O G e nny N L I r i ol T r r e dil A e e s l t u R t ) t n i aF E a d a ew ul P W e u Wwo d o nL B ot so O r n rL o rI N
- N o i o C mu o O t t tl t me T 1 I c n ce nu cav N T aw o c
av ee i c aa ael E
? E I eo MV RSV et a 1 L D RL (
R. L li B N U A O . A . . C T C 7 7 8 9
I~ yeyl o yey yeylio pl np nr I . l n t yaiy io V btbp n btbr a btb ow: a bctb ow O pw eeh t an- l t an egee s HR egeguo egevnt ege- rieu rieu g CP b nbntd 'bi nbei b nbrd gm b ntbrd T i:i rt nbs i oh sru i oh sru IS yryrau yryers yryitd ac yrd yi t d a c WS auauth auah ue auariohc a u n a ri o h c SA MdMd ss M d. M w t l MdMpwrca Md aMpwrca YP -
~
EY
) ) ) ) ) ) ) )
KB
- 1 2 1 2 1 2 l 2
' '~ ~ _
i ~ r e s n
- _e d
n o mC m m m a a a a rns r r r cwe c c c S ot S S S d a - lt r l l' l N l ue a a a O uhp i i i I FSO t t t T . r r r C A
)) a a a 12 P P P l 1 1 l l -
u 1 1 1 F . . . 2 2 2 t' nl 2 2 2 ee T cv 0 0 0 N ra I er 4 4 4 O PT P e e e 0 n l l l T 9 e b b b E p a a a S > O T T T e u mr au es t o
, Sl R
g C s O e S ne h N i v . s c E dl 2 e t
) S l a h h i d iV r c c w R u o t t S
't O B'n i i n o1 w w e o L ei S S r c E nt s u
( N i ay t t s N bl a i i s
- A rol m m e H use i i r S C TI R L L P T
I M n I e I L g v r r nny l n o e o G a I dt rdt N i ol V n e oa uoa I dil l Rl sRl T l t u pp e u s u A i aF oO vl m el m R E ul B ot t S y eou L rc rou P rc P so l t c t c O eI N el l nA snA N n i me nu iF i o ao
- OC e GC e T O b avn b v v 1 I rel e rt wyi w yi N T ut ap uo onr onr E
7 E I TSVO TN LAD LAD L D M D B N . . . . A O 0 1 2 3 C T C 1 1 1 1
~
lg l jf l
r S sp
- I V
su at
~ O pr HR ya .
CP bt n T IS e
.swo WS 5 btd SA e e e e.
yaunt
- YP n n n n o
EY o o o alh 3 ~ KB N N. N N Mps 1 .
~L
= ~
= _
~-
r o t - % cJ e
, amv m'
a eal R ea r ; - t V c m m m f S n S a a a o
= -
r r r .go l c c c eni
. N a S S S rit O i -
ud a _ I t _l l l sll T r l l l oio C a u u u l us A P F F F CBI
~ .
1 1 1 l' 1 1 1 1 2 2 2 2 - 2 2 2 2 g i T 0 0 0 0 s N p I 4 4 4 4 O - 0 _ P e e e e 0 l l l l 0 T b b b b 1 E a a a a ' S T T T T >_ r r o o 2 2 - - . t t r- r rd r rd r o o e R er er ene d ad ene d ad M - A M - A r u d o d o O n n n- n n n gl gl s S U1 U1 U1 U1 U2U2 n - n s N :
- . ir ir e E A y B y A yB y A yB y d e - d e r
) S l a 1 a I aI a I al a l t2 l t2 P d l l -l i n in r l
s le's e l R se - se sese uer uer me
't O uRR uR uR uRR uR uR BC o B C. o at n D OB B B OB B et o L e e e e e e rl1 el 1 ti c E V g V g V gV g
.a a V gV g oo _ no S m
( N a a a a t ry iry s N A 0t 0l 0t 0l 0t0t 0l0l 0t0t 0l0l ct a anl bt a rnl nn i a H 4 o 4 o 4 o4 o 4 o4 o eoe uoe ar S C 2V 2V 2V2V 2V2V RCR TCR MT T I M I e L a t a G rn r N I fh oa e p) m T ( t oy a A rd l e R eroeol t E gef et a S P ag F c O t nd si ne l oeert ir
- N ol rvea au T O V irk m Ms 1 I eueao s N
E T wmqset we or E I oieeru W L D Lt rRB a LP i
}
1 B N . . [ A O 4 5 C T C 1 1
S - I V
- O HR
;, CP
._ T -
IS WS SA 'e YP n EY o u. KB N m s ' ' ( p
- e - -
r nt
's m u . wa
~
hg ? - r l e t e ae gv os o- a t a
- i P d n thv he cgo gc W H; - ~
l fl et e e n f a OV k vud aum eoe ii d fereff r I nlh n aaS o R rR Hv r e o n t r>TV z oo- ~ o "t C pf C e .h t ee
.t i
r C pn f nef f .vftt f nS vf ni o moemoie oed oneorl o a oe l oel i euit u tl t ;ma og~ ea ne mea enrV enr ma t rPt aPeat seneeiresV ewH auian: rl urriHrt arn- ro - riu rie i s yi r y 2 i Oeua- unh uenud yeuasruanest av N t sat eart pst t secsdi s w 1: l st sesti vonel O i erit rond monnot sonaoocl onet onmll ot a nrpnl pl enal oel eil orl l eal oral oeaCCSV_ I T IPSI ASCV aDCCVCRDCCDCBDVCCPWCCDV) C ) - ) : ) -
) ) ) ) ) ) ;~
A 1 2 3 4 5 6 7 8 9
~ -
e T - N g I i O s~ P p T , 5 E S 1 e r u s s e
- r P2 R
O gr S N no
)
E S d 'l l i d i r R ue
'tn O Bt
- t o L ri
( c E N t s om N cn
- A aa H er S C RT
' T -
I M I L G g N n I _ i T dh l g A R ii E uH P _ B O - e _ rr
- N ou T O t s N
1 I cs T ae W E I er L D R B N R. P U A O 6 C T C 1 jl! !f L ;
I V O HR CP I I' S WS SA e e e e YP n n n n EY o o o o KB N N N N
. e s s . mr g y t r ae ge a d a t e ee gv at ry n r ng l p fl ra oa a p ea nm f a h wyt r r - s ur l o i a d
ov p agt l sp s osef t enoe ft n t r incd ovi r f s f otf ne di eae ml gao e e oieoed tl ma wrer f oih o f' vem anpc . nh eat ene ol d rc o ue gawvos e ot rl urih l . df pt i uioua geo a . N O so rt s
. st ond onn st t n i el oot e l
nt i at ere n n :n I e k l enl oe nvf mk z nrrvet T C vsn i aa CvaC cv)
)
el paorao onz eeait s ppt rl y A Dgt 1 2 O vtft n Oosd as 2 2 o "2 y st 1 t e 3 l d i h ed i . e v t 4 vn l t i eo d - e a t o n l pf n nl v c cht o soo i aae d i ac i ne nl nwd id t orst gi aon T sh nf a i rt a i md l i N uwooc t oic . s r- e ge I o p i acmi2 ppt i bl G P esssf i if a sp" a . sl eti dhli 2 0 wc nl aeric gvrme ac c 3 oi 2i e T riee3 - ld 51 mv E S eoip hhp . 5 en oe Il cl S Iwt S4 2bi I1nL d n a 1 r1 e r rt l e eae s t y t Wn R r r tt t n O o o i e gi ra S t t mf nmoh N i i sa2 i st C E n n nS d nc
) S o o a r l aay d M M rl o i ret
' R T e uTR e t O n n v1 B f n
e L o o ee ed a i i rLl rrnS u e oua c E t _ t ( N a __ rs srn t s l N i i sen cs2 e2 v
- A d d et a ae H a a rah errer S C R R PWC RP oL o T
I M I t l - L k s e ny u v h r G at ga wee gg e N Ti nh orL i n t I v i x L u Hi a T pi dE s r d W A R ut l sse sl d d c i n uet uinw E l A uo ora ouao P o Bi ePW eB L O Ht t n n e n ra arw arrr
- 'p se ol t on t ouo 0_ au t i l t L l t st T 1 j gl h ct uc ucsc A
E E T 7 ff g ffi an ee mad i en maea i ere W L D RV a I U B N O. E H . S. R S. R P R A O 7 8 9 0 C 1 C 1 _ 1 1 2
!!!,! ,h
I eb eu ra
.. S v nhn .e I i nesosr V t ah ii
- O ccwst y HR e CP T~
t n 'd ilant ooeruif oe _. IS risoct a WS st ras - SA ht i c a c i. r - e x, YP enpacer n EY huyeepo o
- Kl !
Tfb rrof 5 ~-
- L ? l h
,, t n
- s s .
e ,
. _ d i
cy
. cl
- ns A n n
h or _ o t i e t . otk sn baa oo - - l e Pi f ur t
~ -
N O orcb - ac I pi p ei T i cm nd
- C reu on
- A -
Trp Ni
- =
-wd oe
.~, rea
.lt ob c
- i
- a. gsd ien e T shi s N pc: o
' I nl l O 9ia . C P 5 nl - . 3Gi e n T 13 e vr e E oe p S I1n1 o r
o n o e . R r se eey O u S sl h rt N se ch e E ev ttf
) S res i a d PL r weS R .
e Sh
-._ 't O rrt t e -.- n oot n c o L t t i ofi ccm c E i ovs
( N aas t re N een i h ev
% A RRa scsl H r oana S C 33T PEiV T
I M I s
- L e v
G l d N w ae I o V s T L r o A r e yl R ort t C E t oa e c ae W P f y O al err Sl
- N R uo u T O stl mF 1 I hgeev s c- e a N .
T et E 7 E I i ree t o D L D
} B N H. P R L S. N U A O 1 2 C T C 2 2 y.,. ,
I aci c S m t
- rs r - [- 1 I V l oe 5
O eft HP n
;_ _9 ~^_; _
CP T ndd aen
~
IS hsa
=
WS cs
' SA :epo an- _
YP E EY n yi _ KB Obt 2
.-: - 'e -
p nt i . wa a g = ;=j7 - t r o : thv r o l a os d n l i n i : at n' i e m u P - e cJ e:cgo t e t ;! t n tt amvsaum aa
.~
- eal seoe ReaaRrRp ud h n S o~
nd sne - es-enroV V aeCp : eHne k v nl aa
.~ __ ':'
m
= . t V, fS .yfTt o- gon b. o h -
a f Ce; vft mf alfdlf - orl oeaogaonaone sv .ev TV ne ea lD fV oV og =- enisewH ritt udaiud yeuenu ro <: -rn~ esV rIt en rOrr eu ef eC eumrunh eir nrt a
~ "
sll swcvsdisnl stdsausec N oiodoocl onaoot onaoet ot s
. O l usnll eal orli ul eelt el ei I CBI aCBDVCCDCt OCmHCSRCRD T . .
C ) ) ) ) ) ) .)
. A 1 2 3 4 -
5 6 .7 l a n~
- S i l T
I me ov
.Q l'
I M nel L - w - E T od L G N l e B N I et A I O ba
; T T P .c A " i n.-
R T 2d E E l n P S O L fi _u y R t O e S f' N a2 E S S r l o R e O v1
~
L e Ll E e N rn N A en t a H ah
+
C WC l e v e L r e t a W r o N y t f g O c I aw T eo I RL D
=
m _ N O C 1 e
,l[' t
I S I
- O V HR CP
- T IS WS - e e _ Sv Y n n R?
- o o EB N N
l
'p : e .
nt r -~ wa nm s- m r l e - s - - ae u os o a - t i a t r i gv thv he a aee ee
- t. . -
fl P ' e.tden cgo gc W tt v l p i e
&s r
f a kvud aum eoe ii nSl o a nm OV nl h n Hv d i a Vt e aaS o .RrR _r - e CgV d
-.ep t r TV -f Ce h .t ee~ t z n- n fi e f otf fl mf n e f ' n 'e o r l onaoed o ID ma : og ea vfTt o
ne a f nS vf ni oe- l oel mea ma ott aa oieoe tl eL eneeiresV ewH enrV enr ees eat e rnt riHrt ro yeuasruaneu riu - rie rH n rl ur uoeua unh a rue n 'n u d e uioua 7 . sil stt secsdi swavst sest ivstd st N ott onnot sonaoocl onet onml onn ond o O l aul oel eil orll eal oral oeal eo l enl I T ClOCCVCRDCCDCBDVCCPWCCDVC mC
- C_ v C ) ) ) ) ) ) ) ) )
A 1 2 3 4-5 6 7 8 1 o st l . ed vn eo~ l pf n so ne orst T' irt N g t oic I i acmi2 O sp i i P dhli2 ac T 5 riee3 E hhp S 1 fwtS4 e r u s s e r ~- P2 s R r
~ O gr o S
N i no t i b E n d1
) S l o d i r M
' n ae t O Dt n n t o o L ri i c E N
om t ( t s a N cn i
- A aa d - H er a S C FT R 3 T I
_ M _. I t _. L s u G g ga N n nh I i i x T dh dE A l g l R ii i n E uH uo P B Bi e t lO_ rr ra D - N ou ol O t s t i E 1 I cs ct S T ae an O P E I er ee L D RP RV O B N R A O . . P / C 2 3
c, n 0: s t r
-) c J i, '
~
\
Jb5TIF7CA TlYNV , l The instrumentation l channels and actions deleted from. Table l'are not. l necessary for a perannently shutdown, defueled' reactor. The channels and actions fun'ctioned to shut down. the~ reactor, operate the shutdown condenser, initiateicore: spray,: initiate ' diesel-generators for?the purpose of. supplying 4 F'i the'high. pressure core spr6y pumps, isolate the.steamLline on low main' condenser vacuum,'or low main steam line pressure, divert offgas-flow, trip.
-forced ~ circulation pumps to reduce-reactor power, indicate safety valve-
<poettion. or to. isolate. containment ventilation on highJreactor pressure; The containment isolations on' low, reactor water level and high reactor-
; building pressure are remaining at this time, since some of the isolation valves have'~no-other closure' signals.
1 I l
^
s'l f i I p. I l. o 1' I pC4-20b 139
]
L__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ . . _ . . . _ _ _ _ . _ . _ _ _ _
m . , ., . m ,. - . l' l ' ,. . I 1
' j% ; ij ( ,
t
.g ,
,- 'O ' .p'. ;;, .
- , ..., :m
; .u . , 7. -
\
;pq!" y Fi ,
f ,, ,,( glt 'i. 4
# I
i J
;l
~
,' PA Gl 6.~0 - ,
,. ,1 : ,
'l ,. k!
[' I' . '. [ 7
..o.
.ig t f ;' . -
i - 6.2.2. The unit; organization shall'. be as sho'm;on' Figure ' 6.2.'2-1 ' and: . 1 -)a
!}l .' , s - , >
;i 1 .j
.n.,. . . ...
d. a : c.':At!1 east-two licensed Operators shall:be'presentiin the Control; Room" l
'1
'during reactor. startup',' scheduled creactor shutdown and during -
4
, ; recovery from reactor,. trips...
s v ., . ( >, .. p)gvRggy ';;
'. l l
3
- Delete Section 6!2.2.c.
.o oI p, . . ' i
'l e
t
s ]
-:L
~
,.JU5TIFICATIW !
Since the reactor'is permanently. shut down, this- specification does not~
- apply.
> 11, '
.l L . . I j
i, ; ,
.] '
s
'l
;If 4
.'g' l
o
'l .
I- l 1 r < {
\
l i PC4 20b 140 l s
-]
'1
.. .. i
l 1 PAGE 6-1 0 CURRES f, A Fire Brigade of at least 5 members shall be maintained on site at
]
3 all times.* The Fire Brigade shall not include the two LACBWR Plant Operators necessary for safe shutdown of the unit or any other i personnel required for other essential functions during a fire emergency.
- g. At all times when the reactor is critical, or when its contro b are being manipulated with fuel in the reactor, the control Room shall ,
be attended by a minimum of two persons, one of whom shall have a I valid Operators License and shall have full responsibility for operation of the facility.
- h. The working hours of the Reactor Control Operator, Turbine Generator Control Board Operator, the Duty Shift Supervisor, Nechanical I
Maintenance and Instrument & Electrical Technicians and Auxiliary Operator when performing duties which may affect nuclear safety., and Health Physics Technicians, when performing radiation protection .) duties.which may affect the safety of the public shall be limited. In the event overtime must be used, the following restrictions shall be followed:
- 1. The specified personnel shall not be permitted to work more than 16 hours straight, excluding shift turnover time.
- 2. The specified personnel shall not be permitted to work more than 16 hours in any 24-hour period, more than 24 hours in any 48-hour period, nor more than 72 hours in any 7-day period.
- 3. A break of at least 8 hours shall be allowed following overtime before the next scheduled shift for the specified personnel, if the above limits are exceeded.
In the event overtirne must be used in excess of the above .j restrictions, the Plant Superintendent or his designate, must l authorice the deviation and the cause must be documented. l l G.2.3 Shift Techniccl Advisor The Shift Technical Advisor shall serve in an advisory capacity to the Shift Supervisor on matters pertaining to the engineering aspect:s assuring safe l operat. ion of the unit. l
- Fire BriJade composition may be less t han the minimum requirement s for n period of Lime not to exceed two hours in order to accommodate unexpected absence of Fire Brigade members provided immediat e action is taken to restore the fire Brigade to within the minimum requirements. This provision does not permit ar.y Fire Brigade position to be unmanned upon shift change due to an oncoming brigade member being late or absent.
PC4 -20t> 341
_ r
-7 , "f;' , '> S; "!D , ,
q~ ' LiN h ,10 ~
}lf '
-p >
l
- n<w 1 h' ,
V ~
-- PNCFC6ED; p -,
JE, "\ f. : A Fire: Brigade oflat .least 5 meubers shal1~ be maintained on site at i b
- Lall times.*; The Fire Brigade.shall.not include the two LACBWR Plant'
,? 4 Operators necessary for safe shutdown of the. unit-lor any other - f
*personnelJrequired'for otheriessential' functions'during a. fire.
emergency. J . g .1 Deleted.-
!N
.h. LThe.workin'g hours of.the' Operators, the Duty Shift. supervisor, l
; Mechanical Maintenance and Instrument & Electrical Technicians when performing duties which may. affect nuclear. safety, and
-Health PhysicsiTechnicians, when performing radiation protection C . duties .which'may affectithe safety of. the public shall be limited.
In'the-event overtime must'be used, the following restrictions.shall b ' : be ;followed:
- 1. .The specified personnel.sha11 not be' permitted to work more than- .
16 hours straight,' excluding' shift turnover time. I t
- ' 2. The specified' personnel shall not be permitted-to work more than.
.16 hours'in any 24-hour. period, more thar. 24 hours in any, 48-hour _ period, nor'more than'72 hours in'any 7-day period.
3.5 'A' break of at least 8 hours shall' befallowed followin'g overtime before the next scheduledLahift for the specified personnel, if the above limits are. exceeded. 5
- c '
i
- Inithe. event' overtime must be used in excess of the above.
in restrictions, the Plant Superintendent or his designate, must authorize the' deviation and the'cause must be documented. j
. . I 6.2.3.. Deleted.
i
- Fire Brigade composition may be less than the. minimum requirements for a period of time not' to exceed two hours in order to accommodate unexpected absence of Fire Brigade members provided immediate action is taken to
' restore the Fire. Brigade to within the minimum requirements.- This~
provision does not permit any Fire Brigade position to be unmanned upon _!
~s hif t change due to an oncoming brigade member being late or absent. i 4
l i i l PC4-20b 142
.g ,~ -
.i
.RISTIFICA TION
. Since the plant permanently is shut down, Section 6.2.2.g does riot
. apply. With the reactorishut .down, there is not a separate Reactor Control and Turbine' Generator Control Board Operator, so the. term " Operators" :is !
being substit'uted into the overtime specification '(6.2.2.h) for.the
' individually named operator duty' assigninents. Section 6.2.3 is being deleted i
+ .since Shift Technical' Advisors were required only'during Conditions 1, 2' l
-and 3.
> .-i I l l l l
'1 i
1 l 1 1 PC4 20b 143 j
xG
, ,-r-"
4. PAGE 6-4 CURMTNT !/5: 4 > LTable 6.2.2-1 1 MINIMUM SHIFT CREW COMPOSITION NIMBER l Position ' ' Number'of Individuals Required to Fill Position- l'
- . -Conditions-1,'2, & 3 : Condition 4 : Condition 51:
'! l l_ l' l
- SS: l 1 -l l' -l 1- .:
4 L l SRO l None -l None' l 1 ': ROD l- l' -l .1 ; l- l
! A0 l 2. -l- 1- l 1- ;
!,STA' 1 l- None -l None l SS - Shift Supervisor with a Senior Reactor Operator's' License SRO'- Individual with a . Senior Reactor Operator. License or .a Senior -
Reactor Operator: Limited to Fuel Handling -that is supervising core alterations Ro - Individual.with'a Reactor Operator's License AO - Auxiliary Operator STA - Shift Technical Advisor PROIM ED Table 6.2.2-1 MINIMUM SilIFT CREW COMPOSITION NUMBER , l Position l Number of Individuals l
-l l Required to Fill Position : -l l =l Condition 4 ;!
- l l l l l SS l 1 l l SR0 l None -l l R0 l' 1 l l A0 l 1 l ]'
i SS - Shift. Supervisor with a Senior Reactor Operator's License c, SRO - Individual with a Senior Reactor Operator License or a Senior j Reactor Operator Limited to Fuel Handling that is supervising core l 1 alterations R0 - Individual with a Heactor Operator's License I' AO - Auxiliary Operator L-
- j. JUSTIFICA TION STA is being deleted from the table since t he STA was required only during Conditions l' 2 and 3.
, Also, coltunns for conditions 1, 2, 3 and 5 are i
[ being deleted since they don't apply. ] 1 l.
-PC4-20b 144
I 1 l ,c !' j l l 1 a a I
- a ATTACINENT 2 i
PROPOSED TECIINICAL SPECIFICATION PAGES i l I s
r i h
- 4. J0pERATING LIMITATIONS 4.0.1 Definitions L The following terms are defined so that uniform interpretation of these
- specifications may be achieved. When these terms appear in capitalized type, the following definitions apply in these Technical Specifications.
' ACTION
" ACTION shall be'thet part of a Specification which-prescribes remedial measures required'under designated conditions.
CHANNEL CALIBRATION A Cl!ANNEL CALIBRATION shall be the adjustment, as necessary of the channel Loutput such that it responds with the necessary range and accuracy to known-values of the parameter which the channel monitors. - The CHANNEL CALIBRATION shall encompass the' entire channel. including the sensor and alarm'and/or trip functions, and'shall include.the CHANNEL FUNCTIONAL. TEST. The CHANNEL
' CALIBRATION may be, performed by an series of sequential, overlapping or total channel steps such that the entire channel is calibrated..
CilANNEL' CHECK A CllANNEL CHECK shall be the qualitative assessment of' channel behavior
.during operation by. observation. This determination shall include, where
.possible, comparison of the channel indication and/or; status with other
-indications and/ortstatus derived from independent instrument channels measuring the same parameter.
i 1 l 4 1 4 l 27d
- e. , , m -
9.. ,
; .: f !
f4. 0PERATING;LIMITATI'ONS' 4.0.1 Definitions '(cont'd)
,t'.
-CHANNEL FUNCTIONAL' TEST'
. A CHANNEL FUNCTIONAL TEST shall be:
r ~ L a. : Analog channels - the injection ofEa simulated signal into the-channel as close to the sensor as practicable'to verify OPERABILITY including alann and/ori trip functions and channel-failure trips.
". b. .Distable channels. 'the injection of a real.or simulated signal into.-
the. sensor- to verify OPERABILITY including alann .and/or trip functions.' CONTAINMENT INTEGRITYJ CONTAINMENTIINTEGRITY shall' exist when:
- a. All penetrations required to be isolated during acciden't' conditions are'either:
- l. Chpable of being closed by an OPERABLE containment automatic' isolation valve system, or ,
.<c 2 .~ . Closed-by at least one manual valve, blind flange,.or deactivated-automatic. valve secured in its closed position.
- b. The freight' door.is closed,
- c. . Each air lock is OPERAELE,
'd . The containment leakage rates are within the limit, and e .- .The sealing mechanism associated with.each penetration (e.g.,
welds, bellow, o-rings) is OPERABLE. CORE ALTERATION CORE ALTERATION shall be t he addition, removal, relocalic>n or movement of fuel, sources, incore instrumentation or reactivity controls within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe conservative position. 27e
1
- 4. OPERATING LIMITATIONS ,
i 4.0.1 Definitions - (cont'd) l DOSE EQUIVALENT I-131 l l DOSE EQUIVALENT I-131 shall be that concentration of I-131 uCi/ gram, which alone would produce the same thyroid dose as the quantity and isotopic. mixture of I-131, I-152, I-133, I-134, and I-135 a~tually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of' TID-14844, " Calculations of Distance Factors for Power and Test Heactor Sites." 5'- AVERAGE DISINTEGRATION ENERGY E shall be the average, weighted in proportion to the concentration of each radionuclides in the reactor coolant at the time of sampling, of the sum-of the average beta and gamma energies per disintegration, in MeV, for isotopes other than iodines with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant. ; EFFLUENT RELEASE BOUNDARY The Dairyland Power Cooperative property line within the 1109 ft. radius Exclusion Area is the EFFLUENT RELEASE BOUNDARY. See Figure 4/5.3. FREQUENCY NOTATION The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table of Surveillance Frequency Notation. GASEOUS RADWASTE SYSTEM A GASEOUS RADWASTE SYSTEM is a system designed to reduce radioactive gaseous effluents by collecting primary coolant offgases from the primary system and providing for delay, holdup or filtering for the purpose of reducing the total radioactivity prior to release to the environment. 1 27f
_ . - _ _ _ - _ _ _ , _ _ _ _ .,-m,, , . _ 7-E: y<l A; ,, l 1 :. , ;
;..- r
' - yG .4.1 OPERATi.NG LIMITATIONS -{
' 4.0;1- Definitions -~(cont'd)-
iMEMBER(S) 0F THE PUBLIC MEMBEH(S) 0F THE PUBLIC shall include all persons who are not-occupationally
> . associated with.the utility. 'This category does not include employees of the
' utility, .'its contractors or vendors. Also excluded from this category are
. persons who. enter the. sitello service equipment or make deliveries. This category does include persons who use portions of the site for recreational < l
.or other purposes not associated with'the utility.
OFFSITE DOSE CALCULATION MANUAL ii m. An OFFSITE DOSE CALCULAT. ION MANUAL (ODCM) shall be a manual containing the
. methodology and parameters to be used for the calculation of offsite doses A due.to radioactive gaseous and liquid effluents and for the calculation of gaseous and liquid effluent monitoring instrumentation alarm / trip ~setpoints.
- It shall describe the radiological; environmental monitoring program.
' OPERABLE-OPERABILITY' i:
i A system,c subsystem,. train, component or device shall'be OPERABLE or have OPERABILITY when it is capableLof performing.its'specified function (s)~and when all necessary attendant instrumentation, controls, a normal and an emergency. electrical power source, cooling or seal water, lubrication.or other auxiliary equipment that are required for the system, subsystem, train, component ' or device' to perform its function (s). are also' capable of perf orming their related support function (s). OPERATIONAL CONDITION - CONDITION An OPERATIONAliCONDITION, i.e. CONDITION, shall correspond to any one inclusive combination of power level and average reactor coolant temperature specified in Tnble of OPERATIONAL CONDITIONS. lt
- rg :
1 I
T ! .' , s , lib b 0
'4.. OPERATING LIMITATIONS
- . J, i .
m . P i .f. .+
. ' 4.0.1 Definitions - (cont'd)'
PROCESS' CONTROL PROGRAM (PCP) M
" The PROCESS' CONTROL PROGRAM shall contain the current formula,' sampling, analyses, tests and determinations;to be made to ensure that the processing L and ' packaging of solid ' radioactive wastes based onzdemonstrated processing of '
': actual or simulated solid wastes will be accomplished in such a way as to "g assure compliance with)l0.CFR Part 20, 10 CFR Part 71 and Federal and State regulations and other requirements. governing the. disposal of the radioactive.
- i. . waste. l l.
REPORTABLE EVENT A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50. . A Licensee Event Report shall be submitted for M REPORTABLE. EVENTS. RESTRICTED AREA r - A RESTRICTED AREA shall be any area within the exclusion boundacy, access to
? .which is controlled by the licensee for' purposes of protection of individuals from exposure to radiation and radioactive materials. See Figure 4/5.3.
& i 1l j. u I e j i 27h j l i u.li__.--_-__ _ ._
-- ~ -- - -- - ---
rgr:77 g q r , a- , , Lc . c 4 1 4 , I
/ 4 } 1 . . . .
' i f; ) , . < ,
,4 .v :
o
; g7- l /
. , , m .
-]
py; , ' , l'4q ' OPERATING LIMITATIONS. , s, ,
'J '
^ 4.0. I' ' Definitions ~ - (cont'd)
,9 5 4 .n-E,f SOLIDIFICATION , i
,, -u 1
) . SOLIDIFICATION shall be the conversion of wet wastes into a form that meets-e g, , Jshipping and burial ground requirements, Wi i ?; SOURCE CHECK' 1
, ^: , '.A SOURCE CHECK shall'be the qualitative assessment of channel response when
, .the channel sensor,is exposed to'a radioactive source.. .
P ; STAGGERED TEST BASIS
\
. A' STAGGERED TEST BASIS shall consist of:
4
' le , '
<a. .'A test schedule for,n systems, subsystems, trains or other:
designated components obtained by dividing the specified test } 4 4 interval into n equal subintervals. T 3 b, ! The . testing of one system, subsystem, train or other designated 1 J' component at'the beginning of each subinterval..
. IINRESTRICTED AREA
, - An UNRESTRICTED AREA.shall.be any area not controlled by the licensee'for.' ,
j fpurposes of protection of MEMBERS OF THE PUBLIC from exposure to radiation O and radioactive materials. VENTILATION EXIIAUST TREATMENT SYSTEM c' A . VENTILATION EXHAUST TREA'INENT SYSTEM is any system designed and installed to reduce' radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through liigh Efficiency Particulate filters
'for the purpose of removing particulate from the gaseous exhaust stream
. prior to the release to the environment. Such a system is not considered to L have any effect on noble gas. effluents. i i
f. l, 271
.)
1 r l l 1 I i TIIIS PAGE INTENTIONALLY LEFT BLANK peleted 4.0.2 'l 4.1.1 4.1.2 1 (Pages 271 - 27rr) i l l - . - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
= ;- . - - - - - - - --
7y. , .; ,
^'
h- j 'l > r
, q '
.y i g !. , t
- ,l
[',~ t ; - i D- i +J < . p ,
+
p,
.2j j
f,- GENERAL q l' J.4.1.
' i g 4.1. 3; ' Delet'ed. . ,
L4.1.,4i Deleted. <
^
1 ,. 4 '.1. 5. Deleted.- I I' > 1 4 . 11. 6 Delbted. , 9I L4.2-: OPERATING LIMITS, i 4.2. IL ^ Resetor Building i
; 4.2.1,1. CONTAINMENT INTEGRITY shall be maintained during handling of,
.' irradiated fuel.
4.2.1;2, 'Gasketed closures and ventilation l system closures which have been (subjectedito maintenance .repairJor'other operations which might affect:their!
- performance shall, before any, subsequent operation. for.which containment 1 l Integrity:is--required, be tested; f or leak: tightness using the. soap-bubble, techniqueL(orjother method of. equivalent sensitivity). This' test shall be-Lperformed using a pressure differential no less than 0.5 psi and the results
~
q ahall be used usLa guide in evaluating leakage. ! sj- , L
.l S.
l I l 28 i
)
li l 4.2.1.3 The number of containment vessel electrical penetrations may be E o- varied, as may the containment vessel piping penetrations, provided the new l L penetrations'are equivalent in design to the existing penetrations. New !' penetrations are defined as conduit or piping requiring attachment to the containment vessel shell. .After'any such penetration changes, an integrated l l l reactor building leak test shall be performed at approximately 52 psig. This l l test must demonstrate that the containment vessel meets the leak rate
.specified in Sec. 5.2.1.
4.2.1.4 Existing containment vessel penetrations may be removed from or placed in service, provided the containment vessel is not affected and the closures are equivalent in strength and tightness to those previously installed. After any such change the penetration, exclusive of the containment shell connection, shall.be tested for leak tightness at a pressure no less than 52 psig using the soap-bubble technique (or other i method of equivalent sensitivity) or by determining the rate of pressure loss of a test chamber. The penetration leakage rate shall not exceed 1.0 percent ' of the containment vessel leak rate Lp specified in Sec. 5.2.1. L j 4.2.1.5 The reactor buildings vacuum breakers shall be set to relieve at a ! differential (external-over-internal) pressure not exceeding 0.5 psi. j 4.2.1.6 The reactor building ventilation system isolation dampers shall close upon loss of control air or loss of electrical power supply, and they shall be automatically closed by abnormal conditions as specified in Table 1. 4.2.1.7 The main stetun line, the reactor building vent header, and the decay
- heat system blowdown linelshall be isolated automatically by abnormal i
conditions as specified in Table 1. l' 4.2.1.8 The containment building steel shell temperature shall be greater
. than 00F during reactor operation or during any integrated leak rate test performed with a pressure exceeding 10.4 psig.
4.2.1.9 Deleted, f 4 I 1 1 l l L l 29
- l. .. . - - _ _ _ . _ _ _ _ _
~ ~ ~ ' '
7 T' ' m . . p;c t
$ y i
.cQ, $ ,
- v. iq
'a
?'
f ;. . ' , um - c '
..i: . u , .
i; 4, s
' REACTOR BUILDING, 1 q y
')
<' CONTAllNENT VENTILATION DAMPERS
~
e LIMITING CONDITION'FOR OPERATION g , 1 g b f.2.1'.10l;Th'e containment.ventilat! ion inlet;and outlet dampers shall.be OPERABLE with Lisolationitimes of less than 'or equal' to,10 seconds, L m r u. E',,, = APPLICABILITY:: Whenever CONTAINMENT INTEGRITY (Specification'4.2.1.1) is , required. ACTION: s r A. . With one or more'of the above. ventilation' damper (s) inoperable, d suspendifuel handling orLisolate the affected penetration with an automatic valvs secured in its closed position, or.with a. blind
. flange,-within L hour.
y
' B.; ..The provisions:of' Specification' 3.0.4 are not applicable if the
' affected penetration is isolated.
s, SURVEILLANCE EE00IREMENTS
' 5. 2 ; 1.10.' l The ventilation dampers shall be. demonstrated. OPERABLE priorIto
.returnin'g;the damper ~to service after maintenance,. repair or? replacement work-is. performed on the damper or its. associated actuator, control, Eor power circuit by performance of, a cycling test, and verification of isolation time. ,4 5.2.1.10.2 The isolation time nf each above damper shall be ' determined to be ( ,
within its limit when tested quarterly. 5.2.1.10.3 The seat rings of the ventilation inlet and outlet dampers shall be. replaced at least once per 5 years. BASES
, +
The. Nuclear Regulatory Commission requested that similar Technical Specifica-tions per Generic Item B-24 and NUREG 0737 Item II.E.4.2 be submitted to help
- assure operability of containment ventilation daml i ers. The ACTION statement g
has been modified due to the p? ant's permanent shutdown. DpC had previously committed to replace the containment ventilation inlet and outlet dampers' resilient sealing' material at least once each 5 years until such time as
" additional in-situ data an be accumulated.to Justify a longer interval. If in-situ data is accumulated which supports a longer seal replacement inter-vals,'a change to Specification 5.2.1.10.3 may be requested. Specification 3.0.4 is'not applicable if the affected penetration is isolated since the safety function of the dampers is to close.
29a l
.. i
p , 4.2.2 React'ro Vessel, Coolant, and Auxiliary Systems 4.2.2.1 Additional penetrations to the systems containing reactor coolant
-shall be designed, manufactured, and tested according to the provisions of
.the ASME Boiler and Pressure Vessel Code and the ASA Code for Pressure Piping applicable as of June 1962. These additional penetrations shall be limited to instrument connections and piping connections, the latter being no larger than 1-inch inside diameter.
4.2.2.2 The reactor coolant shall be light water and shall. conform to the following requirements. CONDITIONS 4 & 5 Normal Limit Chloride concentration .5 ppm pli 5.3 - 8.6 Conductivity 10 umho/cm The primary system chemistry parameters defined in this section shall be determined at least once every 7 days in Conditions 4 and 5. l 4.2.2.3 Deleted 29b
l
-1 1
l i
.l 1
1 THIS PAGE INTENTIONAL.LY LEFT BLANK Deleted' . 4/5.2.2,4 4/5.2,2.5 4.2.2.6 4.2.2.7 4.2.2.8 4.2.2.9 4.2.2.10 4.2.2.11 i 4.2.2.12
- 4.'2. 2.13 i 4.2.2.14 i 4.2.2.15 4.2.2.]6 4.2.2.17 4.2.2.18 4.2.2.19 4.2.2.20 4 4/5.2.2.21 -!
4.2.2.22 5.2.16 i r i ( hft((('S dl) 220) i 1 e
gr : l in :'t
.v
- l. .
,; , , 1 L ..i n-.
,j, p ,
4
- t ,
I i ,i.j l 1 i s THIS PAGE INTENTIONALLY LEFT tlLANK 1. I j, Deleted
,4.2.2.22-and 5.2.16 BASES 1
I i (Page 32f') ,
l' i
. k 7..
.l 8
I g .. i l l s
-)
4 1 4 r j 1 l n THIS PAGE INTENTIONALLY LEFT BLANK , i Deleted 4.2.3.3.1 5.2.11.5 [ 1 i i 1 I l 1 l (Pages 32r(1) - 32r(5)]
i- I p< r 4> y j I .,' , i: '! u- / 'l i I I 1 THIS PAGE INTENTIONALLY LEFT BLANK i i
,i Deleted 4.2.4.2
'4.2.4.3-4.2.4.4' 4'2.4.5 1
4.2.4.6
- 4.2.4.7 l 4.2.4.8 l 4.2.4.9 i
i i i l f (Pages 321 -- 33)
4 ,
-i a
- 4. 2.~ 5 . : pe_leted ' [-
' l 4.2.6: Safety Instrumentation ,
.j
'b
'4.2.6.1 The safety instrumentation shall provide isolation action and other ')
l
-safety actions as specified in Table 1 of these specifications.
'; l
-j 4.2.6.2 : The setpoints for the safety instrumentation shall be as specified 9 Lin Table'1. l ,
'4.2.6.3 . Key. switches shall permit operational, main'tenance, and test bypass ;l of the safety instrumentation only with the approval' of the Shift Supervisor. .l.
,4 '. 2. 6. 4 Deleted. .
l 4.2.6.5 Deleted. l ; 4.2.6.6 Deleted.'- '! l-4.2.6.7 Deleted.. l 4.2.6.8 Safety channels,directly backed up by an identical channel or ) channels may be bypassed for maintenance or testing. l.
- l 4 4.2.6.9 Deleted. l 14.2.7 Deleted. l l
u q i j a f1 l l 1 34 l l
,3a. , p;; ,1
- iI s c s
.. .I '
I. +
, i
.]
l
.'. l l
'i i
(
)
1 1
-a l
TIIIS PAGE INTENTIONALLY LEFT BLANK .! 4
-1 54 l
1 s i
.v a
. .>1
, \
i
]:4 l
.i i
1 i
)
.i A
j 1 I l l l l 1 (Page.s 35 - 30)
' - - - "-----_m.__ ___ ___, _ "'"--"-- -----____ _
g _ - -_ ._..-- _-- - . ,. _ --
, 1
,f ': 17- ' t
,m,z ,
~
y . ,
.l g >
q- , , 14.2.8' ' Spent Fuel Storage and Handling
- 4. 2. 8. I ' Fuci elements and control rods shall be. inserted or. removed from'the-reactor. vessel one et a' time.. 1 A, s;i.
. 4.'2. 8. 2 Irradiated fuel elements shall be stored underwater in . spent fuel .
i storage racks 'that are positioned on the bottom of the spent fuel storage l w -well, or-in an approved shipping cock.-
]
J. E , 4.2.8.3 .During the' handling of irradiated fuel elements that have been operated at' power levels greater than 1. Mwt 'the' depth of water in the reactor .) upper cavity and/or the' spent fuel storage well 'shall be at least 2 feet 1 above the active fuel. ]
, ' 4. 2. 8. 4 Deleted. .( .
4.2.8.5 With the exception of.~a spent' fuel shipping cask, the core spray bundle, the . transfer canal shield plug and the. other components and fixtures , that 'are normally located and used within the spent fuel storage well, no !, objects heavier than a fuel as'sembly shall be hatulled over the spent fuel l storage well.. r
~ s.
h 1 s 37
- j
-l,
.I
'l
'3 s E F i THIS PAGE INTENTIONALLY LEFT BLANK 1 Deleted 4/5.2.21 4/5.2.22 4/5.2.23 4/5.2.24 4
.. r - l
'I l
1 e l \ \ , l i l~ i l i e (Pages 37n - 37bb)
)
o 1 , TABLE-4.3.2.1
- - RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRINENTATIONA
.a,
;. -MINIMUM CHANNELS APPLICABLE.
INSTRUMENT OPERABLE . CONDITIONS' . ACTION
- 1. Reactor Containment Building Ventilation Monitor' System-
- a. Par2 ticulate- Activity -1 *- B
. Monitor a
'b. Gaseous Activity Monitor 1
- B.
- c. Sampler Flow Rate
,ea Measuring Device 'l
- C' ,j q
'2. Stack Monitor System H
- a. Noble Gas Activity Monitor 1 ** D. 'i
- b. . Iodine Sampler 1. -E
- c. . Particulate Sampler 1 ** E
- d.-. Sampler-Flow Rate 1 ** C Measuring Device i
- 3. ' Deleted. [ q
- When Containment' Building Ventilation System is in operation. l l
** At all times, unless alternate monitoring is available.-
,A. For post-accident instrumentation, refer to Section 4.5.2.
B. With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases through this rathway may continue for up to 24 hours, as long as stack monitors are OPERABLE. i C. With the number of channels OI'ERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may I continue provided the flow rate is estimated at least once per 4 hours. D. With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may ., continue provided alternate monitoring is available or grab samples are taken at least once per 12 hours and these samples are analyzed for gross activity within 24 hours. -( E. With the number of channels OPERABLE less than required by the Minimum , Channels OPERABLE requirement, effluent releases via this pathway may f continue provided alternate monitoring ,is availablo meeting the requirements of Table 5.3.2.1 or continuous collection of samples with auxiliary sampling equipment is initiated within 1 hour. F. Deleted. l 43 j ___E_.____.____
, --r; '- ; ,
- = 'w y - - =~
9}g MQ M~] , 1 ij.i et= .o ( '
- i. 9 .;s a.', ,
' g, ., ,
~
u _ . , j
; ,1,-
O ,
) >
N [ X '
<< .a tBLE 5.3.2.l' ]
'4
'i '
T . RADI0 ACTIVE' GASEOUS EFFLUENT MONITORING INSTRI_ MENTATION SURVEILLANCE REQUIREMENTS < > Mp ' < .y i 'w '
! CHANNEL (4)- S' SURVEILLANCE D 1 : CHANNEL SOURCE FUNCTIONAL . CHANNEL- ' REQUIREMENT *)
-{c 3 INSTRINENTT ;; CHECK? CHECK TEST- CALIBRATION CONDITIONS '
'j o )
, 51'. t ' Reactor' Containment j LBuilding Ventilation- <
g Monitor System
*j ,
- a. Particulate _
-0 ]"
. ' Activity Mon! tor. 'D. M Q( 2 h R
- j w:. i;b. Gaseous: Activity d
~ Monitor D- M. =O(1)- .
R .* . < l
'c.1 Sampler Flow Rate ;]
Measuring Device 'D :N/A Q(3) R '* j
- s. ,
L2. Stack Monitor System
.:al Noble Gas Activit'y~
' Monitor. '
D M Q(2); a * , J
- b. Iodine: Sampler D 1 M Q(2) 'R-c/ Particulate' Sampler D N/A- Q<2) LR-
- 1 id.J SamplerfFlow Rate ~
g h~ Measuring' Device. D N/A Q(3) R *. j sc )j u 3 .~ ' Delete'd'. l.
- m
} a
;* During applicable conditions per Table 4.3.2.l'.
lq
.i ^ (1) The CilANNEL FUNCTIONAL TEST shcIl also demonstrate Orat automatic j isolation' of this ' pathway and control' room alarm annunciation occurs 11 - j any of the following conditions exist: j
- a. Instrument indicates' measured levels at or abov'e the alarm setpoint.
- b. Instrument indicates a downscale failure (pt ovides control room
~
)
annunciation alarm only). . l
- c. . Instrument. indicates a circuit failure (provides control room annunciation alarm only). -{
-(2)- The CilANNEL FUNCTIONAL TEST shall also demonstrate that control room !
alarm annunciation occurs. if any of the following conditions exist: a.. Instrument indicates: measured level above the alarm setpoint on one channel. 4
- b. - Instrument indicates a failure by a Low Flow. and Low Count Rate' j a
signal. .
- c. Instrument controls in Maintenance mode.
- (3) The CIIANNEL FUNCTIONAL TEST shall also demonstrate that the control room 1 o local nlarm occurs if the flow instrument indicates measured levels below l the minimum and/or above the maximum alarm setpoint. )
(4) The CHANNEL CALIBRATION shall be conducted in accordance with plant procedures. l l 44 4 j
=
t ' , j i ci i i TI!IS PAGE INTENTIONALLY LEFT BLAgg Deletegi 4/5.3.2.5 1 49
i RADIOACTIVE EFFLUENTS BASES 4/5.3.2.4 DOSE, RADIONUCLIDES OTHER THAN NOBLE GASE3 1 This specification. is provided to implement the requirements of Sections II.C., III.A, IV.'A and Annex of Appendix I,10 CFR Part SG. The ODCM calculational' methods specified in the. surveillance requirements implement the requirements'in .Section. III./. of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the' actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. l 4/5.3.2.6 Deleted. l J
'4/5.3.2.6 VENTILATION EXRAUST TREATMENT SYSTEM (CONTAINMENT BUILDING)
The OPERABILITY of the venti ~1ation exhaust treatment system (Containment , Building) ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. This. specification. implements the requirements. ofl10 CPR Part 50, and the.desigu 6bjectives given in Section II.D of Appendix I to 10 CFR Part 50. 4/5.3.3 SOLID RADIOACTIVE WASTE The OPERABILITY of the solid'radwaste system ensures that the system will be available for ase whenever solid radwastes require processing and packaging prior to being shipped offsi te. This specification implements the requirements of 10 CFR Part 50.3Ga and General Design Criterion 60 of Appendix A to 10 CFR Part 50. 4/5..?.4 TOTAL DOSE This specification is provided to root the dose limitations of 40 CPR 190. The specification requires the preparation and submittal of a Special Report whenever the calculated doces from plant radioactive effluents exceed twice the desigri objective doses of Appendix 1. The Special Report will describe a cour se of action which sht,uld result in the limitation of dose to a real . individual for 12 consecutive months to within the 40 CFR 190 limits, j 1 i l l L 1 l 1 I 55 L. __ . _ _ _ _ _ _ _ _ - _ - _ _ _
y my y,--- y
,y,, -- .
. , , - , - = , , , -,
,' .' -:t'.I I I' i k.
I L 7
,)'. .,'
f
,1
' p;$ ' ,
! 5 -
i I 1 i g v f
? I -
l
,j
( s 1 .j 1
. 1. .
Tf!IS PAGE INTENTIDMALLY LEFT BLANK <
.i I
Deleted' 1
'4/5.5.1 l i
'.)
L b h 4 1 l , i 58 < 1 E___-_--__ _ _ _ _ _ .
ry , ' 4, 3 -. --,;--- -
;Y .' ' -
r lr:) d. . : ,.l l l4 [ [.h-' i ,.:, '. ,! b ' R :- , e.s '; o.
^,}_ {j i s
.5.. MAINTENANCE:
t 1 o. n ..
- 5.1 . GENERAL j
,i ?
y' ;5.1.1 Maintenance. operations and routine tests shall be performed in, l :{ conformanceLwith these specifications. ,
, W . , ;,
5.1.2 Maintenance operations shall be performed'as authorized by the Shift. Supervisor. . Maintenance involving the opening of systems containing .,
-radioactive materials shall be conducted under the r;u'veiliacce r of a Health
! i'. , Physics representative.. < ' 5.1.3 ' Deleted. l 5.1.4 Deleted. ! I
'l'
{ l
)
j u v. t
.)
m j l! i h i I k i t-l 1
)
l 5-1 j..
l 1 5 .' 1. 5 Components which have been repaired, replaced, or otherwise subjected - to. temporary or permanent modification shall be tested in accordance with l procedures which are appropriate in view of the nature of the repair, - replacement or modification, and in view of the condition of the system. 5.1.6 Deleted. 5.1.7 Deleted. I 1 5.2 TESTING i 5.2.1 Containment Testing l I 5.2.1.1 Containment Integrated Leakage Rate Test (Type A Tests): ! (a) The integrated leakage rate test shall be performed at a pressure f of at least 52 psig without any preliminary leak detection surveys and repairs except as necessary to correct any evidence of structural deterioration which may affect either the containment's structural integrity ; or leak tightness. Such structural deterioration and. corrective actions taken shall be reported as part of the Type A test report. Closure of containment isolation valves shall be accomplished by normal mode of actuation and without any preliminary exercises or adjustments. If valve closure malfunction is detected which requires corrective action before. the test, this information shall be included in the report submitted to the Commission as required under Section 5.2.1.5. The test duration shall be for a sufficient period of time to obtain ! ' meaningful leakage rate results. In addition, a controlled leakage rate test shall be included to verify the test accuracy. i l l Reference 5 5-2 affects this page.
y
' the electrical penetrations . . . .
the. reactor building spray valve shaft penetration, s the freight' door, .;
- the main steam'line penetration,
~
1 the feedwater line penetration, the heat W steam line and condensate return penetration,
- the containment building airlocks, and _
the flenges.of the' ventilation system inlet and exhaust ducts. 'l
' containme'ntEcomponente other than mentioned above, which develop' leaks ~
requiring repairs during.the' performance of Type A test, shall be. included in ! subsequent Type B Tests.
- Component Leak Surveillance System: A leak surveillance system (i.e.,
continuous pressurization of individual containment components) that l maintains a pressure not less than 52 psig at individual test' chambers of ! containment penetrations 'and seals during normal reactor operation shall be
. acceptable in If eu of Type B tests for the components under such leak l surveillance. 1 (b) Type.C Tests: Containment isolation valve leak detection testing !
shall be conducted at a pressure of 52 psig. 1 (c),-Acceptance Criteria: The combined leakage rate for all penetrations and valves subject to Type B'and C tests shall be less than 60 l percent. of the maximum allowable Type A test leakage rate. 1 (d) Corrective Actions: Leaks which cause the acceptance criteria of
~ (c) to be exceeded shall be repaired and retested until the criteria is met.
. Repairs of : lesser leaks are optional. l q
lc) Test Frequency: . Type B. tests (except for air' locks and electrical 1 penetrations). shall be performed at intervals no greater than two years. Air ! locks shall be tested at 4-month intervals. The freight door shall be. tested l following each closure. Electrical penetrations shall be tested at intervals 1 i no greater than one year. Type C tests shall be performed at intervals no greater than two years. , 5.2.1.3 Deleted. l 1 5,2.1.4 Permissible Periods for Testing: The performance of Type A tests shall be limited to periods when the plant facility is non-operational and secured in the shutdown condition under administrative control and safety procedures. 5.2.1.5 R1e. port of Test _Results: The leakage rate results of Type A, B, and C tests that meet the acceptance criteria shall be reported in the applicable LACBWR operating report. Leakage test results of Type A, B, and C tests that fail to meet the acceptance criteria shall be reported in a separate summary that includes an analysis and interpretation of the test data, the least squares fit analysis of the test data, the instrumentation error analysis, and the structurul conditions of the containment or components, if any, which contributed to the failure in meeting the acceptance criteria. Results and ann. lyses of the supplemental verificat lon test employed to demonstrate the validity of the leukage rate test measurements also shall be included. < Reference 5 5-4 affects this page.
e n: , i ,
' ' il 4.if ",. 5
- lJ
.s t , ,
..c .t - !"
y, + . ,
,o
.l '
T .- '
, ,_ _lY ,
_ , . .__Y
,f, , , __
Mi yf4 5.2.1'.6 CONTAINMENT' VENTILATION ISOLATION VALVE LEAKAGE TESTS p 4
, 0 ,
- q. , . . ... .
,. ~ . . . . .
- o. .. . .
"% < 5
' 1(a): -Tests: LThe containment ventilation. system' dampers shall be'
' ' _ subjected to leakage: tests, in addition- to the. tests: required by 4 .
vp Section'5.2.1.2.. The tests'shallibe conducted at an initial F .; pressure'of at.least'52.psig.- i ' f(b);iAcceptance Criteria: . Excessive degradationiis determined not to lexist..a'd.the isolationtv'alve(s):is considered operable if
~
, + n
,. ; pressure decreases by no gEeater than 10 psi in a ten:minutei
{ "
?~ : test period.'
L(c) . Corrective Action: Ifl excessive: degradation exists, the leakage. 3 . path must-_be-repaired or-isolated,;and retested until the- ! c , ; criteria is-met.l A passing. leakage: test mustLbe achieved'within'
~
224 hours crithe~ reactor'must be placed:in' hot shutdown within 'y ,
<the.next 24 hours,1and! cold shutdown within the'following 300 J hours.unless a passing' test.is; achieved. ,
-(d)L[Tes't. Frequency: :The leakage tests of the. containment ventilation isolation. dampers shall be. conducted at'le st
~ '
i% : quarterly'. . ,
.g
' 5.2.2 l REACTOR BUILDING ISOLATION AND VACUIN RELIEF VALVE OPERABILITY TESTSy 1
J 5.2.'2.1; The reactor building' isolation system will be tested for proper ' : , j operation'at least once per 18 months. l l
.y . ... .
. 5.2.2.2' The. reactor building vacut.a relief valves .will be -tested for:properm operation during each reactor shutdown for~ refueling but in no case at l : intervals greater than two years.
e i l i i. i / S-5
._ _ = _:
l w
.- , l
?: 5.2;3 The exterior surfaces of the LACBWR ventilation stack and the smoke _ ; stack of;the conventional steam power generating station,. Genoa-3, adjacent l
- to the LACBWR plant shall be inspected for structural integrity at an
~
l
' interval no' longer than five years followis.g the initial construction inspection, and at subsequent intervals no longer than five' years apart.
5.2.4 Deleted. l 5.2.5- Deleted.~ l 5.2.6 -Deleted. l 5.2.7 Deleted. .; 5.2.8 Deleted. 5.2.9' Deleted. l . 5.2.10 The door seals on the containment personnel and emergency airlocks will be visually inspected for degradation every 72 hours. 5.2.11 The door seals on the containment personnel and emergency airlocks l will be replaced periodically in accordance with manufacturers recommendations. 5.2.12 Dele t'ed. - 5.2.13 Deleted. 5.2.14 Deleted. 5.2.15 Instrumentation shall be checked, tested and calibrated as indicated . in the following chart. f I 1 1 5-6 l l
- ,-----7 --~-_- ._.m o .,
4 i TIIIS PAGE. INTENTIONALLY LEFT BLANK-1 i
)
I
, I l
4 4 1 . I 5-Ga
i MINIMLN FREQllENCIES FOR' fESTING, CALIBRATING, AND/OR CHECKING OF ' INSTRUMENTATION '
'; -CHANNELS ,
ACTION- MINIMUM FREQUENCY : l l l
-l1. Reactor Water Level : Calibration :.At least once per 18 months. :
l l l l
.' : Test : Monthly when in service. l 1 l l 1 : Check : Daily. :
- D.~ Area Radiation : Calibration l At least once per 18 months. :
; Monitors : : :
1 Test l Quarterly. l l l
- : Check : Daily. :
.l l l- l : l
- 3. Port able Radiation : Calibrat ion : Semi-annually. l- i Detectors l l : ';
l l Check : Every two weeks. : l ; l4. Reactor Building : Calibration l At'least once per 18 months. l l Pressure ! ; l I l i l i i i j
\
l l 1 l fi 7 I
l l' i 1 1 l THIS PAGE' INTENTIONALLY 18FT BLANK i Deleted s 5.2.16 i i 1 l l 1 l (Pages 5- 8 5-10)
i qli r - a r eb bi l g . N O yan I aci S m t I V rs l oe O eft HR n CP ndd T aen IS hsa WS cs SA epo an YP EY nyi KB Obt e p nt - - r o r o l a wa os l a i n g n m u t thv dn i at i e P c e cgo t e t t n tt e amvsanm eal seoe ud hn ndsne enroV aa es kv nl ReaaRrR' S o V aeCp eHne aa t V p yfTt h Ce sv ev TV fS f vft mfalfdlf o go nb o a orl oeaogaonaone og ne aa lD fV oV enisewH etV en ef eC eir ritt ro rn rIt rOrr nrt a udaiudyeuenu eu eumrunh _ sll swavsdi snl stdsausec N oiodoocl onaoot oaaoet ot s O l usnll eal orli ul eelt el ei I CBI aCBDVCCDCtOCmHCSRCRD T C ) ) ) ) ) ) ) A 1 2 3 4 5 6 " S l T a I n M i l 1 I L me ov E nel L G B N w A I T od T T N l e A I et R O ba E P c P i O T "2 d - E 1 n S 1i y R t O e S f N a2 E S S r l o R e O v1 e L Ll E e N rn - N en _ A t a - H ah C WC l e v e L r e t a W N t r o _ O c I T aw eo _ I RL D N 0,
. 1 YZ ll l llu
~
I W V ' O O c. HR CP T -
-f =
I S WS ? SA e e YP n n EY o ~ o KB N N 1' :e - p nt r - . wa nm s j s m r l e c s
- - _ ae u os o_- a t i a - t r ae
~ ~ i gv P d n th v he a aee , ee gv t
n& s fl f a et e k vud cgo aum gc ii W t t v nSl l p nm fl f a e r OV nlh n eoe Hv d o .a i a ov Vt e aaS o:RrR ' r - e CgV - d ~ -
- ep t' r~ TV - C eh - t ee t z n in ~- t r s
1 fl mfnef - f . v f T t' f' nS vf ni fi e f otf ne onaoed o n 'e og o rea l o nea oe l oel ot t oieoed ma mea ma aa
- ma tl e n e 'e i r e s V ewH e I D.
enrV enr ees eat ene uoeua . u n h u e ro: r n't riHrt arn
'n u riu d yeuasruaneu rie- rH n e
rl urih uioua stt sil stt secsdi swavst sesti vstd st N ot t onnot sonaoocl onet onml onn ond onn O l aul oel eil orll eal oral oeal eo l enl oe I Cl OCCVCRDCCDCBDVCCPWCCDVC mC CvaC cv T
~
~
1 '
~
C ) ) ) ) ) ) ) ) )' ) A 1 2 3 4 5 6 7 8 1 2 y o st l ed vn eo-l pf n
. soo ne i orst T i rt a N g t oic .
I i acmi2 O s i if P p dhli2 ac c T 5' riee3 hhp E S f < wt S4 - e r u s s e r . P2 s R r O gr o S no t N i' i
~
E d1 n
) S l ~ o d i r M
' R ue n t O Bt
, - n t o o .
ri i oms ( c Lt N cn t i a
-- A aa d a
H er R S C RT T I M t I L s u G g ga N n nh I i i x _ T dh dE A I g l R ii i n l uo E uI P B Bi t O e ra rr _ - N ou t i ol _ O t s ct I cs an 1_ T ae ee E I er RV L D RP B N A O . _ T C 2 3
~
T'C
I f q
.)
l' ) l l l) > .I
~
I L i ' i TI;tS PAGE INTENTIONALLY LEFT BLANK l
/Pages 5 13 5 18)
G. ADMINISTRATIVE CONTROLS 6.1 RESP 0NSIBILITY G.I.1 The Plant Superintendent shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence. G.1.2 The Shift Supervisor (or during his absence from the control Room, a designated individual) shall be responsible for the Control Room command function. A management directive to this effect, signed by the (highest level of corporate management) shall be reissued to all station personnel on an annual basis. 6.2 ORGANIZATION OFFSITE 6.2.1 The offsite organization for unit management and technical support shall be as shown on Figure 6.2.1-1. l UNIT STAFF 6.2.2 The unit organization shall be as shown on Figure 6.2.2-1 and:
- a. Each on duty shift shall be composed of at least the minimum shif t crew composition shown in Table G.2.2 -1.
- b. At least one licensed Operator shall be in the Control Room when fuel is in the reactor,
- c. Deleted. l
- d. An individual qualified in radiation prot ection procedures shall be on site when fuel is in'the reactor.
- e. All CORE ALTERATIONS shall be directly supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent resporaibilities during this cperation.
l s I l 1 1 G0 I
4 , . < - - d-i,5 j l r 1
; il.
y;$ ,
; },; O i, .;jt g- & ' '
, i
/* f ,,.
1
%d
~
v gr.f J . > ., , ~ ,~ - .
- f. ?A" Fire Brig. .ade of 'at- least: 5 members shall.be maintained on . site at' ~
%.S' all times.*" The Fire l Brigade shall,not include the two LACBWR plant
' Operators necessary:for safe shutdown of the unit or any.other Ut .
s 4 , ? personnel ^ required for other essential functions during a fire. i i! : , Lemergency.a j
- g. IDelete '
Y h :The working hours of the operators,Lthe Duty Shift Supervisor,r Mechanical Maintenance and Instrument & Electrical Technleians
,when performing, duties which may. affect' nuclear 1 safety, and' V: , . Health Physics Technicians,.when performing' radiation protection
. duties which may affect the safety of the:public'shall be limited.
),
- d- '
Intheevent'.'overtimemuNtbeused,'the'fh11owingrestrictionsshalll be!followed:. Ll . The specified personnel-shall not be permitted to work more than 16 hours straight, excluding shift turnover. time.
- 2. The specified personnel.shall not'be permitted to work more than
, .16 hours in any. 24-hour period, more than 24 hours in any 48-hour period,. nor more than 72 hours in any 7-day period.
- 3. A break of at 'least 8 hours shall be allowed following overtime -
before the next scheduled shift for. the specified personnel, if-
~
the above limits are exceeded. In the event overtime must be.used in excess of the above restrictions, the Plant Superintendent or his. designate, must authorize the deviation and the cause must be documented.
. G.2.3 Deleted. l
*' Fire Brigade composition may be less thun the. minimum requirements- for a
. period of time not to exceed two hours in order to accommodate unexpected 4 absence of Fire Brigade members provided immediate action is taken to restore the Fire Brigade to'within the minimum requirements. -This provision does not permit any Fire Brigade posit. ion to be unmanned upon shif t change due 'to an oncoming brigade member being late or absent. l Reference G f 61 affects this page. l
i l Table 6.2.2-1 ] i MINIMUM SHIFT CREW COMPOSITION *- .] . I l Position l Number.of Individuals ! l l Required to Fill Position : J l l Condition 4 l- j l l l :j l SS l 1 l . SR0 : None l l RO l 1 . ! l AO. 1 l l l l
'SS - Shi f t Superviso with a Senior Reactor Operator's License l SRO. Individual with a Senior Reactor Operator License or a Se.iior Reactor Operator Lirited to Fuel Hand)ing that is supervising core al t era t. ions j RO - Individual with a Heactor Operator's License -
A0 - Auxiliary Operator l Except for the Shift Supervisor and the Reactor Operalor, the shift crew - composition may be one less than the minimum requirements of Table G.2.2-1 for a period of time not to exceed two hours in order to accoimnodate unexpected absence of on duty shift crew members provided immediate action is j taken to restore the shift crew composition to within the minimum require-- ' ments of Table G.2.2-1. This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent. During any absence of the Shift Supervisor from the Control Room while t he unit is in Condition 1, 2, 3, 4, and .5, an individual (other than the Shift Technical Advisor) with .a valid R0 license shall be designated t o assume the Control Room command function.
- Shi f t crew compos i t ion may l>e less than the minimum requirements for a period of time not to exceed t wo hours in order to accommodate unexpect ed absence of on duty shif t crew members provided immediat e act jon is, taken to restore t he shif t crew compor.0 ion to wi thin the minimum requi rements of Tab le G. 2.2 - 1.
l l CO L--}}