ML20059N855
| ML20059N855 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 10/09/1990 |
| From: | GEORGIA POWER CO. |
| To: | |
| Shared Package | |
| ML20059N854 | List: |
| References | |
| NUDOCS 9010180199 | |
| Download: ML20059N855 (24) | |
Text
- _ -
5AFETY LIMIT 5 LIMITING 54rETY 5Y5 TEM LETTINE5 j
i 2.1. A.1.d. APRM Rod Block "rio settino i
This section de' eted.
{
i i
B i
+
' t 4
}
(
2.1.A.2.
Reactor Vessel Water low Level Scram Trio dettino (Level 3)
Reactor vessel' water low level scram trip setting (Level 3) shall be t 0.0 inches (narrow range scale).. l.
3.-
Turbine Stoo Valve C1'osure scram '
i Turbine rtop valve closure scram trip setting shall be 510 percent valve closure from full open. This scram is only effective when tur 1 bine steam flow is above that corres-panding to 30% of rated core thermal 4
. power, as measured by turbine first -
-stage pressure.
L E
5 l-HATCH - UNIT 1
.1.1-3 Proposed 15/03834/232-121 L
\\.
t
)
P t
i t
F. M.11MITING 5AFETY SY5 TEM 5ETTINE5 2.1.A.1.c.
APRM Flu 5: ram Tric Settinos (Run Mode) (Continued)
The APRM flow referetted sipolated thermal power monitor scram trip j
setting at f ull recitculation flow is adjustable up to 117% of rated power f or two-recirc station loop and single-recirculation loop operations.
This reduced flow re'erenced trip setpoint will result in an j
earlier scram during slow thersel transients such as the loss of 100'F
'l feedwater heating event, than would result wkth the 12D5 fixed high neutron flux scram trip. The lower. flow referenced scram setpoint j
therefore decreases the severity (6CPR) of a slow thermal transient 1
and allows lower Operating Limits if such a transient is the limiting i
abnornel operational transient during a certain exposure interval in 1
the cycle.
The APRM fixed high-high neutron flux scram trip, adjustable up to 120%
of rated power for two-recirculation loop and single-recirculation loop operations, does not incorporate the time constant, but responds directly to instantaneous neutron flux. This scram setpoint scrams the reactor during f ast power increase transients if credit is not taken -
fer a direct (position) scram, and also serves to scram the reactor if credit is not taken for the flow referenced scram.
l
}
L i
2.
Reactor Vessel Water Low Level Scram Trio Settina flevel 3)
The trip setting for low level scram is above the bottom of the separator skirt. Figure 2.1-1 This level is approximately 14 feet above the top; l
of the active fuel.- This level has been used in transient analyses i
dealing with coolant inventory decrease.. The results reported in FSAR Section 14.3 show that a scram at this level adequately protects the fuel and the pressure barrier. The designated scram trip setting is at least 22 inches below the bottom of the norsel operating range and is thus adequate to avoid spurious scrams.'
L HATCH - UNIT 1 1.1-13 Proposed TS/03834/232-141._
r l
I 1
e
'i i
~*
9007 NOTE: SCALEININCHES 1
A80VE VESSEL 2ERO l
t t'
t WATER LEVEL NOMENCLATURE f
HEIGHT ASOVE i
VES$EL 2ERO 800 - -
NO IINCHEli RE ADING INSTRUMEN T l
181 573.5 66.5 BARTON/ROSEMOUNT I
(7) 669 42 GE/MAC v
750 - -
I4) 549
+32 OE/MAC 3
_ 723.66-, VESSEL ggggg (3) 517 0.0 SARTON/ROttMOUNT (2) 470
- 47 SARTON/ROS8 MOUNT j
700 - -
til 404 113 SARTON/ROGEMOUNT lol 318
+302 SARTON/ROSEMOUNT 660=
<- 640 -
j g,gg 1
t
+
600-- -
i
- m...,,
.0--
.0.-
.0--
0--
4
- 669 876 101-64.5 17) 42H1 ALARM
$$0 -r 649(4)
HPCI& lol 32 LO ALARM i
- ORYE R $KIRT-- $17 INSTRUMENT 0 - 0(3)
~
0--
' 0(3)
- REACTOR SCRAld.
I IN CONTRIBUTE TO AOS LOW (LEVEL al FEED
$00 - -
- ~~
WATER
" 470 (2) CORE
- **#I b
Ib
- 406 ---* gpnay p
INITI ATE HPCl. RCIC, 460 - -
I
- 404 (1) 400 - -
- -113 LOW LUW LOW (LEVEL 11 -
3 M
- 367
- 150. - INtfl ATE RHR, C.S.,.
350-u 352.56 START DIESEL AND i
CONTRIBUTE TO A.O.S.
t 2/3 CORE CLOSE MSIV's
- 302 H E IGHT 315 (0)
PERMisslVE 300 -
(LEVELO)
ACTIVE
' FUEL 250
-317 - -
200
- 208 56
-178 56 Ol$CH ARGE -
)
RECIRC SUCTION - 161.5 NOZZLE NO22LE 150 -
100 - -
i i
g 50 - -
1^
0--
.)
e380 3 FIGURE 2.11 =
HATCH - UNIT 1 Proposed TS/03834/232-121 4
)
. ~ - -
Table 3.1-1 (Cont'd) x 2=N Scram Operable x
Number Source of Scram Trip Signal Channels Scram Trip Setting Source of Scram Signal is Required (a).
Required Per to be Operable Except as Indicated Trip Systee Below E
(b) 5 High Drywell Pressure 2
1 1.92 psig Not required to be operable when primary containment integrity is w
not required.
6 Reactor vessel Water Level 2
1 0.0 inches l
(Low) (Level 3) 7
-Scram Discharge Volume Permissible to bypass (initiates High High Level control rod block) in order to reset RPS when the Mode Switch is 4.
Float Switches 2
1 71 gallons in the REFUEL or SHUTDOWN position.
b.
Thermal Level Sensors 2
1 71 gallons 8
APRM Flow Referenced Simulated 2
5 1 0.58W+62% - 0.58 et See Specification 2.1.A.I.c(1) for Thermal Power Monitor-(Not to exceed 137%)
. definitions of W and L*f.
Tech Spec 2.1.A.I.c(I)
Fixed High High Neutron 2
5 1 120% Power Flux Tech Spec 2.1.A.I.c(2)
Inoperative 2
Mot Applicable An APRM is inoperable if there are 7
less than two LPRM inputs per level as or there are less than 11 LPRM
~
inputs to the APRM channel.
o O
U O
t#3 (D
O.
--4 (A
N Ow CD m
'...D N
N (47
'N eW-
- . A W._
,r
,. +
-e.
-..,v.
W:,
- l ~
= *-
Table 3.2-1 2:
}D INSIRUMENTATION WHIEH INITIAIES REACIOR VESSEL Ata) FRIMARY
,N' CONTAINPfMT ISOLATION 2
e-Required Ref._
Trip Operable Action to be taken if
~
HE No.
Instrument-Condition Channels
' Trip Setting number of thannels is.
(a)
Nomenciature per Trip not met _ f or both trip :
Remarks (d)
'Sustem ibl sustees (c1'
.w I
Reactor Vessel.
Low (Level 3) 2-2 9.9 inches Initiate an orderly Initiates Group 2 &'6 l~
teater Level:
Marrow Range.
shutdown and achieve isolation.
the Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or isolate' the shutdown tooling system.
tow Low 2
1-47 inches
. Initiate an orderly Starts the SGTS.
(tevel 2) shutdown and achieve ~
initiates Group 5
.the Cold Shutdown.
isolation, and Condition within 24 initiates hours-setondary <ontainment isolation.
Low Low tow
.27 1-113 inches Initiate'an orderly Initiates Group i
'(tevel 1)"
stutdown and achieve isolation.
the Cold Shutdown Con-W dition within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />st
^ J "N
., Low Permissive. -
e 2,
Reactor Vessel Steam 1 45 psig-Isolate shutdown Isolates.the shutdown 9
Dome Pressure (Shut-
- cooling;
. tooling section valves
--down Cooling Mode)-
of the RHR system.
3 ~
.Drywell Pressure-
- itigh -
'2:.
"11.92 psig' Initiate an orderiv Starts the standby shutdrno and. achieve gas treatment system.
'T
'c the f.old shutdown
-initiates Group 2 13 Coe;ition within 24 isolation and second =
..'8 hr urs.
ary. Containment isolction.
-w
-m
. C1. >
. -4 m.
N
.g-
.- w CD N:
N' W
..N,,'
p.
M '.l
- w d
.2 w
s 7-Table 3.2-4
.c
, ~j lt45TRtR1Lf4TATION WilCff INITIATES OR C0tJiR0t5 ADS
_.e p.
Ret.
Instrument Irip Required trip Setting Remarks No.
Condition operable (a)
Nomenclature Channels C.
per Trip 3
System thi
--4 l
- 1.
Reactor Vessel Water level-Low (Level 3) 1 10.0 inches
- Confirms low level. ADS permissive Reactor Vessel Water level tow tow tow 2
1-113 inches
. Permissive signal to ADS timer (Level 1) 2.
'Drywell Pressure High_
2 11.92 psig Permissive signal to ADS timer 3.
RttR Pump Dis (harge High 2
tilt ps69 P+ rmissive -signal to' ADS timer Pressure' 4.
C5 Pump Discharge High-
'2 1831 psig--
Permissive signal to ADS timer Pressure'
'1 3 minutes-
' Bypasses high drywell pressure Auto Depressurization 2
1
.~
Low Water level Timer permissive upon sustained Level:I
.. W
."ro
_ 'b.
Auto Depressurization-I.
120 t 12 seconds.
beith level 3 and tevel l and high Timer-
. drywell pressure and C5 or RHR pump
'o
- at pressure, timing seguence-
~
~ begins. - If the A05 timer is not -
reset it will initiate ADS.
7..
Automatic Blowdown Control-
.I~
tiot applicable Monitors availability e power to 1 tp-Power failure Monitor-u-
. logic _systee.-
m.
- oo g,
a.
The column entitled "Ref. No." is'enly for convenience so that a one-to-one relationship can be established between
-m;
.- items in Table 3.2-4 and items in Table *4.2-4.
a..
^
. -4 b.~
Whenever any CCCS subsystem is1 required to be operAle by Section 3.5.--there shall be two operable trip systems.
R' If the' required number of operable channels cannot be met for one of ' the trip systems, that system.shall be repaired
.o or the: reactor shall.be placed in the Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> af ter this trip systems:i_s made or. found to
.g be is. operable. _
5
^
'~
N
~ g.
N w._
'(.m 4
w-
-'- ~
N.-
~
N.
,s hh W
rg+1e#"r+ d M-pv ~4 =uem
+14Wd4.-
-"r
-g--F#.*E WM T w-Wd'*
$4f-%'"'M'
'4'WT*"**C"N' T
f'"f"'"'HF'-
t
""""Y9D 41 4
Vf Y' N W W "4 'S $ h '% ' Yt'-%
.W FF * '"pth 4 F 6"
- 2)'F"*O
'TW-7t 7"-e
evn=*'t+*v MF
't vagv &M
'+-'it=++
-*w e=u-T
EASES FOR LIMITING CONDITIONS FOR OPERATION
=i 3.2. PROTECTION INSTRUMENTATION In addition to the Reactor Protection System (RPS) instrumentation which In.
l It14tes a reactor scram, protective Instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond l
the operators ability to control, or terminates operator errors before they result in serious consequences. Thlt set of Specifications provides the lim-iting conditions for operation of the instrumentation:
(a) which. Initiates reactor vessel and primary containment isolation.
(b) which initiates or controls the core and containment cooling systems',
(c) which initiates control rod blocks. (d) which initiates protective action, (e) which monitors leakage into the drywell and (f) which provides survell.
lance information. The objectives of these specifications are (1) to assure the effectiveness of the protective instrumentation when required by preserv-ing its capability to tolerate a single failure of any component of such sys,
tems even during periods when portions of such-systems are out of service for maintenance, and (11) to prescribe the' trip settings reautred to assure ade-gutte performance. When necessary, one channel may be made inoperable for.'
F brief intervals to conduct required functional tests and calibrations.
A.
Instrumentation Which Initiates deactor Vessel and Primary Containment Isolation (Table 3.21)
Isolation valves are installed in those. lines which penetrate the primary con-
,tainment and must be. isolated during a loss of coolant accident so.that the-radiation dose limits are not exceeded during an accident condition..Actua-tion of these valves is initiated by protective instrumentation shown in. Table 3.2-1 which senses the conditions for which isolation is required. Such in-l*
strumentation must be available whenever primary containment integrity is re-quired. The objective is to isolate the primary containment so that the guidelines of 10 CFR 100 are not exceeded during an accident..The events when isolation 15 required are discussed in Appendix G of the FSAR. The Instrumentation which inttlates primary system isolation is connected in a dual bus arrangement.
1.
ReactorVesselWaterLevel1.ow(Level 35(NarrowRange) f The reactor water level instrumentation.ls set.to trip when reactor.
water level is approximately 14 feet above the top of the active fuel. This level is referred to as Level 3 in the Technical Speci-fications and corresponds to a reading of'0.0 inches on the Narrow l
.I Range scale. This trip initiates Group'2 and 6 1 solation but does not trip the recirculation pumps, b.
Reactor Vessel Water Level Low Low (Level 2)
The reactor water level instrumentation is set to trip when reactor water level is approximately 9 feet above.the top of the active.
fuel. This level is referred.to as Level 2 in the Technical Spect-fications and corresponds to a reading of'-47 inches, This trip initiates Group $ isolation starts the standby gas-r treatment system, ano initiates secondary containment isolation.
HATCH - UNIT 1 3.2-50
. Proposed TS/03884/232-121 t
l
.4
<4
o c
BASES FOR LIMITING CONDITIONS FOR OPERATION D.
Instrumentation which Initiates or Controls ADS (Table 3.2-4)
The ADS is a backup system to HPCI. In the event of failure by HPCI to maintain reactor water level, ADS w111 initiate depressurization of=
the reactor in time for LPCI and CS to adequately cool the core. Four 1
signals are recuired to initiate ADS: Low water level confirmed low water level, high drywell pressure, and either a RHR or Core Spray pump available. The simultaneous presence of these four signals will initiate a 120 second timer which will depressurire the reactor if not reset.
1.
Reactor Vessel Water Level low (Level 3)
The second reactor vessel low water level initiation setting
(+0.0 inches) 1s selected to confirm that water level in the l
vessel is in fact 10w. thus providing protection against inadvertent depressurtration in the event of an instrument line (water level) failure. Such a failure could produce a simultaneous high drywell pressure. A confirmed low level is
'one of four.stgnals. required to initiate ADS.
R.eactor vessel Water Level Low Low Low (Level 1) b.
e The reactor vessel low water level setting of -113 inches is selected to provid6 a permissive signal to open the relief valve.
and depressurtre the reactor. vessel in time-to allow adequate cooling of the fuel by the core spray and LPCI systems following a LOCA in which the other make up systems (RCIC and HPCI) fall to matntain vessel water level. This signal is one of four required to initiate ADS.
2.=-Drywell Pressure Hidh=
A primary containment high pressure of 2 2 psig indicates that a breach of the nuclear system process barrier has occurred inside the drywell. The signal is one of four required to initiate the ADS.
L l
l L
l.
l' l
HATCH - UNIT 1 3.2-58
. Proposed TS/0388Q/232-121 l
1 1
6 e
q
' o...
TABLE 2.2mi-1 REACTOR PROTECTI0ts SYSTEM INSTRUMENTATI0ta SETPOINTS g
' FUNCTIONAL t#dIT IRIP SETP0lN1 Alt 0WAEt E VAttJES
-4Q 1.
Intermediate Range Monitor, Neutron Flom-High i 120/125 divisions.
I 120/125 divisions (2C51-K601 A.B.C.D,E.F.G.H) of full scale of full scale e
c.
2.
Average Power Range Monitor:
(2C51-K605 A.B.C.D.E.F)
-4 a.
Neutron flum-Upscale, 15%
1 15/125 divisions i 20/125 divisions g
of full scale of full scale b.
Flow Referenced Simulated Thermal 1 (0.58 W + 59% - 0.58 p)**
.1 (0.58 W + 62% - 0.58 p)**
Po-er-Upicale with a maximum -
with a manimum i 113.5% of RATED 1 115.5% of RATED THERMAL POWER THERMAL POWER c.
Fixed Neutron Flum-Upscale. 118%
i 118% of RATED 1 120% of RATED THERMAt POWER THERMAL POWER 3.
Reactor Vessel Steam Dome Pressure - High 1 1954 psig i 1954 psig (2821-N678 A.B.C.D) l 4.
Reactor vessel Water tevel - Low (Level 3) 1 0 inches above 2 0 in(hes at,ove (2821-N680 A,8,C,0) instrument zero" instrument zero*
5.
Main Steam Line Isolation valve - Closure-1 10% closed i 10% (losed (NA) 6.
Main Steam Line Radiation - High 13m full-po-er i 3 = fu11-power ro (2011-K603A B.C.D) background ***
background ***
7.
Drywell Pressure - High 1 1.92 psig.
i 1.92 psig (2C71-le650A.B.C.D) ~
m
>oooy "See Bases figure B 3/4 3-I.
n.
. Rated loop recirculation flow is equal to 34.2
- W = Total loop recirculation flow rate in percent of rated.
y (n
MLB/hr.
N O-L y = Maximum measured difference between two-loop and single-loop drive flow f or the same core flow in percent
- g of rated recirculation flow for single-loop operation. The value is zero f or. two toop operation.
o y'
- Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to the planned start of the hydrogen injection test with.the reactor power at greater than 20% rated power,- the normal full-power radiation background level and associated trip w7 setpoints may be changed based on a calculated value of the radiation level espected do"ing the test.
oo The background radiation level and associated trip setpoints may be adjusted during the test based on P
either calculations or measurements of actual radiation levels resulting f rom hydrogen.njection. The background radiation level shall be determined and associated trip setpoints shall be set within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of re-establishing normal radiation levels af ter completion of hydrogen injection vnd prior to establishing reactor power levels below 20% rated power.
--.,. ~
o TABLE 3.3.2-2 g-ISOLATION ACTUATION IriSTRUMENTATION SETPOINTS
--4 n*
AttOWABLE
' TRIP FUNCTION TRIP SETPOINT VALUE C2:
l.
PRIMARY CONTAINPENT ISOLATIQN
-t a.
Reactor Vessel Wier Level 1.
Low (Level 3) 1 0 inches
- 1 0 inches
- l N
2.
Low Low (Level 2) 1 -47 inches *'
1 -47 inches
- 3.
Low Low tow (tevel 1) 1 -113 inches
- 1 -813 inches
- b.-Drywell Pressure - High i 1.92 psig i 1.92 psig c.
Radiation - High i 3'm full-power background **
i3a full-power background **
2.
Pressure - Low 2 825 psig 1 825 psig
- 3. ' Flow - High 1 138% rated flow-1 138% rated flow d.
Main Steam Line funnel Temperature - High i 194*f i 194*F w
e.
Condenser Vacuum - Low 1 7" Hg vacuum 1 7* Hg vacuum N
f.
Turbine Building Area Temp.-High 1 200*F 1 200*F wb g.
Drywell Radiation - High 1 138 R/hr 1.138 R/hr cn 2.
SECONDARY CONTAINMENT ISOLATION
- a.. Reactor Building Enhaust Radiation - High i 60 mr/hr i 60 er/hr b.
Drywell Pressure - High.-
I'1.92 psig i 1.92 psig
~
so c.
Reactor Vessel W ter
.Leve1~- Low Low (Level 2) 1 -47 inches
- 1 -47 inches
- 1 d.
Ref ueling floor Enhaust.
Radiation - High 1 20 mr/hr i 20 er/hr cn N
Ow
- See Bases Figure 8 3/4 3-1.
co E~
- Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to the planned start of the hydrogen injection test with the reactor power at greater than.20% rated power,-the normal full power radiation background level and associated trip setpoints N
"w may be changed based on a calculated value of the radiation level empected during the test. The background y
radiation level and associated trip setpoints may be adjusted during the test based on either calculations co or measurements of. actual radiation levels'resulting from hydrogen injection. The background radiation 03 level shall be determined and associated trip setpnints shall be set within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of re-establishing normal radiation levels af ter completion of hydrogen injection and prior to establishing reactor power levels below 20% rated power.
e-TABLE 3.3.2-2 iContinuedi
,T 150LATION ACTUATION INSTRUMENTATION SETP0ff4TS
-4 m*
Att04ABLE TRIP FtMCTION TRIP SETPOINT VAlti c
2 S.
REACTOR CORE ISOLATIQg 5
CQQLIE SYSTEM ISOLATION a.
RCIC Steam Line Flow - High i 307% of rated flow i 307% of rated flow b.
RCIC Steam Supply Pressure - Low 1 60 psig 1 60 psig c.
RCIC Turbine E=haust Diaphragm Pressure - High 1 20 psig i 20 psig d.
Emergency Area Cooler Temperature-High i 169'F i 169'F e.
Suppression Pool Area Ambient Temperature i 169*F i 169*F High f.
Suppression Pool Area di - High 1 42*F i 42*F g.
Suppression Pool Area Temperature Timer Relays
.AA NA h.
Drywell Pressure - High i 1.92 psig i 1.92 psig w
- i. Logic Power Monitor NA NA 5
6.
SIMIDOWN COOLIE SYSTEM ISOLATION a.
Reactor Vessel Water Level - Low 1 0 inches
- 1 0 inches
- l (Level.3) b.
Reactor Steam Dome Pressure - High 1 145 psig i 145 psig.
o t)
E X;
a
-4 '.
my
- See Bases _ Figure _B 3/4 3-1.
E' g
N
~
~_
. ~
,. ~.
_-4
TABLE 3.3.3-2 (Eantinuedi ENERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINL5
--4 m-
.1LLOWA8tE -
TRIP FUNCT1Dlg J11P SETPOINT VALUE C$
3.
HIGt PRESSURE C00: ANT INJECTION SYSTEM
--4 a.
Reactor Vessel Water Level - Low Low (Level 2) 1 -47 inches
- 1 --47 inches
- y b.
Drywell Pressure-High i 1.92 psig i 1.92 psig c.
Condensate Storage Tank Level - Low 1 0 inches **
1 0 inches **
d.
Suppression Chamber Water Level - High i 154.2 inches ***
.i 154.2 inches ***
e.
Logic Power Monitor NA NA f.
Reactor Vessel Water Level-High (level 8)*
1 56.5 inches 1 56.5 in(hes 4.
AUTOMATIC DEPRESSURIZATION SYSTEN a.
Drywell Pressure-High i 1.92 psig i 1.92 psig b.
Reactor Vessel Water Level - Low Low Low (Level 1) 1 -113 incnes*
1 -113 in(hes*
c.
A05 Timer i 120 seconds i 120 seconds
.d.
ADS Low Water Level Actuation Timer i 13 minutes i 13 minutes e.
Reactor Vessel Water Level - Low (Level 3) 1 0 inches
- 1 0 inches
- l f.
Core Spray Pump Discharge Pressure - High 1 137 psig 2.337 Psi 9 g.
RHR (LPCI N00E) Pump Discharge Pressure - High 1 112 psig 2 112 psig h.
Control Power Monitor NA NA 5.
LOW LOW SET S/RV SYSTEM 4
w a.
Reactor Steam Dome Pressure - High i 1954 psig i 1954 psig N
u)
- m o
E us fD O.
-4
- See Bases Figure B 3/4 3-1.
"w
- Equivalent to 10,000 gallons of water in the CST.
O
- Measured above torus invert.
g to U1 N
NW-N e
Q N
4 p..-
, ~
w-
., ~~
-m..,--,w
-r-,.,
a
~.
,-,,v.n w..
e--,
+,. -.
..w
.m.
e
'i o
i a'
9007
'O NOTE: SCALEININCHES ABOVE VEESE L ZERO -
-t
-j 1
WATER LEVEL NOMENCLATURE y
HEIGHT ABOVE
)
. VESSEL ZERO.
]
300 -
NO clNcHtSt RE AOING INST RUME NT j
""~
ts) 573.5
+ 56.5 BARTON/ ROSE MOUNT I
(7) 559
+42 GE/MAC 750 - -
VE$5EL I4) 549
+32 GE /MAC
. 722.75 - F L ANG E (3) 517 0.0 BARTON/ROSEMOUNT l
(21 470
-47 BARTON/ROSEMOUNT l
7g,
til 404
- 113 BARTON/AOBEMOUNT 40) 318 202 BARTON/ROSEMOUNT J}
650 -.- 640 MAIN $TE AM LINE
/
-1 600 - -
- f. 5 18)
'60 - - -.
+60.-
' + 60 - -
+ 60 - -
4 459 478 (8)- 56.5 (7) 42 Hf AL ARM
-i 550 -n 549(4)
~
- ORvEHEKIRT 517(3f ZERO g
~0 - 0(3)
~
0--
0(3F - RE ACTOR SCRAM CONTRIBUTE TO AOS Low (Lgygt 3)
" " ~ ~
500 -
waygn ""II 470 (2)
"* * #7 '
CORE
,, 466 -- gpgay INITI ATE HPCI, RCIC, 450 - -*
III 400
- -113 LOW LOW LOW (LEVEL II -
'l
- 367
-150.
. INITI ATE RHR, C.S.,
q START OIESEL AND
~ t 350 --.352 56 CONTRIBU.'E TO A.O S.
2/3 CORE CLOSE MSIV1s
- 202 HElGHT 315 (0) d P E RMIS$1V E 300 -
(LEVELD)
ACTIVE FUEL 250 - -
j u
-317 - -
200 - 2N "
RECIRC
-178 56 OISCH ARGE
- Recirculation Pump' Trip j RECT".C
+
[
l' SUCTION - 161.5 -
NOZZL E Analytical Limit is.58 inches L
NO2ztt 150 --
100 --
(-
i r
' 50 - -
-r 0.-_
9380 2 r
BASES FIGURE B 3/4 31 RE ACTOR VESSEL WATER LEVEL l
LHATCH - UNIT 2.
B 3/4 3-6
. Proposed.TS/0000q/232-121 i
',3-t t
.~
s
]
s.
ENCL,0SURE 4 t
EAS-13 0290 l
DRF A00 03737 March 1990 f
SAFETY EVALUATION FOR RELAXATION-0F, SCRAM WATER LEVEL FOR PLANT: HATCH UNITS 1 AND 2 i
t i
W.M. Wong l
L'.L. Chi i
?
Reviewed by:
. /T.
3-6-30 l
A. R. Smith, Sou".hern Region,
.?
Licensing Services Manager Licensing & Consulting Services l
t L
Approved by:
W l
intPef ngineering:
t v
1 l
-l
--(
t IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT i
I PLEASE READ CAREFULLY The only undertakings of General Electric Company respecting information in this document are contained in the contract between the Georgia Power Company, (GPC) and General Electric Company. as identified in the purchase order for 1
this report and nothing' contained '.in 'this document shall be construed as' changing the contract.
The use-of this information by anyone others than GPC or for any purpose other than that for which it.. is. ' intended, ' is ' not; authorized; and with respect to any unauthorized use, ~ General Electric:
j Company. makes no representation or warranty, and assumes no liability as to the completeness, accuracy, or usefulness of.the information contained in this, I
document, or that its use may not infringe privately owned rights.
l.
The information contained in this justification is-believed >by General
'[
Electric Company to be an accurate and.true representation of the facts known, j
obtained or provided to General. Electric at-. the time this s information was prepared.
t l
d e
3 l
4 i
...............w.--.vv.
rw~,.x-.-,.-.-
1.
1 3
l I
i TABLE OF CONTENTS i
- f. art l
i
1.0 INTRODUCTION
1 1.1 OBJECTIVE 1
i 1.2 ANALYSIS SCOPE 1
(
i 2.0 ANALYSIS OF LOSS OF FEEDWATER EVENTS 2
i 2.1 ANALYSIS FOR THE LOOSP EVENT-2 2.2' APPENDIX R ANALYSIS 3
i n
2.3 HELB ANALYSIS 5
e, 3.0 ANALYSIS OF LOSS OF COOLANT ACCIDENT EVENTS 6
1
4.0 CONCLUSION
7
5.0 REFERENCES
8 r
o-j
- A
. l t
P k
j
--.s n,
+.n,..-,,-
.----...+,..w---.,.-.,-
,,. ~..
~+.+r n.
9 i
a 3
r 1.0 INTRODUCT10N 1.1 OBJECTIVE
{
This study evaluates the impact on transient and accident analyses. with a
.l i
scram water level (L3) at 12 inches below the current setpoint for Plant Hatch Units 1 and 2.
[
1.2 ANALYSIS SCOPE-i Events which involve loss of feedwater flow and RCIC operation were previously evaluated for relaxation of RCIC performance requirements for Plant = Hatch i
i (Reference 5.1). These events are (a detailed discussion is also presented in l
Reference 5.1):
~
i a.
LossofOffsitePower(LOOSP),
I b.
Appendix R Events, and-c.
HighEnergyLineBreak(HELB) Events.
.j I
The previous studies showed that the RCIC'.is capable'of, performing its design function with relaxed RCIC delay time and rated flow rate.
These events are reevaluated herein, using consistent. assumptions and basis, to. demonstrate that the' low water'1evel (L1)-trip can be avoided.with a lowered, L3 setpoint-and relaxed RCIC performance requirements.
The limiting loss. of' coolant accident (LOCA) event is-the design basis
- accident (DBA) recirculation suction! line break: with al single failure of
[
battery (Reference 5.2).
For this event', two low pressure coolant : injection-l
-(LPCI) loops, two low pressure core spray (LPCS). loops-and the automatic depressurintion system-(ADS) are available for core cooling. The impact of a lowered L3 setpoint on this DBA is also assessed herein.
{
-la
-m-..m__--_______._ _ _____., ~ ____.,, _,,-,,,,.....
c.
,,..,...w,,_,.%4.
+-.~,,4.-.
..,,.c%,,
..w.-
,.y_.g.,.,
c.
1 i
l 2.0 ANALYSIS OF LOSS OF FEEDWATER EVENTS 2.1 ANALYSIS FOR THE Loss 0F 0FFSITE POWER EVENT i
1 2.1.1 Event Descrintion l
The LOOSP event is the worst case loss of feedwater event.
Once all 'offsite power supplies are lost, the Reactor Protection System (RPS) is doenergized, j
This causes the closure of the Main Steam Isolation Valves (MSIV) and a reactor scram.
Reactor feedwater is then lost and the Safety Relief Valves l
(SNV) cycle to control reactor pressure.
Assuming HPCI is unavailable, RCIC j
is actuated upon reaching reactor water Level 2 (L2). The RCIC flow provides sufficient makeup inventory to prevent reactor water, level from reaching L1.
l I
2.1.2 Performance Criteria l
l The RCIC is provided to be capable of preventing reactor water' level from f
reaching L1 following system initiation at. L2 (Reference 5.3).
Initiation of the low pressure ECCS at L1 for this analysis is designated to occur at -124.5
[
= inches with respect to instrument zero (this setpointis based' on the L1 f
analytical limit plus 28 inches to account for the bias due tof the high j
drywelltemperatureconcernsspecifiedinReference'5;4).
l 2.1.3 Assumotions and Initial Conditions i
The following assumptions or initial conditions are made within this' analysis:
- a i
a.
Reactor scram and MSIV closure are initiated at time zero for 1
- analysis, i
I b.
Initial water level is at the scram setpoint.(L3):which is assumed to
]
be at 42.71 ft (this is 12 inches below the current setpoint),-
L
- c.. Decay heat is based on 1979 ANS Standards, 2-
i
?
l
)
l d.
Relaxed RCIC performance requirements consistent with Reference 5.1 1
analysis.
I l
2.1.4 Result.s The SAFE code was utilized to model the 'LODSP loss of feedwater event. - For' l
inventory loss cases where the core remains covered.the SAFER and SAFE codes j
demonstrate similar results. The SAFE code was selected for these evaluations j
to enable direct comparison to previous study (References.5.5 and 5.6).
i The results of the evaluation show that the margin,to actuation of ~10w j
2 pressure ECCS at LI is 21 inches..That is, indicated water level remains. 21 l
inches above the 124.5 inches designated as L1 for this analysis. Therefore, j
with 'L3 set 12 inches below the current setpoint, and. with the relaxed performance requirements for the RCIC system, there remains some margin to L1 trip avoidance.
{
q 2.2 APPENDIX R ANALYSIS 2.2.1 Event Descriotion l
Plant Hatch utilizes one of three sets of safe. shutdown systems to proceed to-cold shutdown during an Appendix R fire event ;(Reference. 5.6).
These are j
Division 1 and Division 2 safe shutdown systems.and the alternate safe shutdown system.
Division 'l systems include RCIC, SRVs, lone RHR; pump for f
LPCI, another for Shutdown Cooling and the RHR Service Water system. Division 5
2 systems do not include RCIC; 'therefore, fire events requiring Division 2 q
. systems are not analyzed within this report. 'The: alternate safe'. shutdown
}
[
systems, which-include RCIC,. are utilized in ' case of a'. control : room' fire.
i!
However, as-shown in Reference. 5.6, for the' limiting control room-fire' event-i l
' RCIC can not. be initiated in time ~ to prevent - subsequent ~ actuation of ADS.
=
Therefore, only those fire events which require the' ude of Divi $ ion I safe.
j shutdown = systems;are analyzed within this report.
I 4
3..
-... _, ~ _,,
n.. _ a,. _
.. I
1 q
}
?'
j 1
When the Division 1 - safe shutdown systems are utilized, the. Appendix R fire i
p event is assumed to have a concurrent loss of offsite power, loss of all-automatic functions affected by the fire and ' spurious -operation of plant
(
equipment affected by the fire. The LOOSP conditions initiate a reactor water
{
i, l
1evel response that is essentially the same as that analyzed in Section 2.1.
The RCIC system is then used to maintain indicated reactor water level above l
L1 until the reactor is ' permitted to'be manually depressurized' to allow low
{
pressure.ECCS injection into the vessel.
{
2.2.2 Performance Criteria l
o
-In accordance with 10CFR50 Appendix R, safe shutdown is acheived if there is.
l no damage to the fuel cladding and. no ' rupture ' of the reactor vessel and.
This is assured if reactor inventory remains above the I
top of the' active fuel (TAF).
Momentary core uncovery is allowed if the safe shutdown operation involves ADS and a low pressure makeup-' system.- If ADS is l
automatically initiated before the operators gain' control of the. low pressure j
makeup system, then the rapid depressurization has a potential to cause' core uncovery and subsequent fuel heatup.
For this' reason, RCIC is. utilized to t
maintain reactor water level greater than L1 to prevent automatic : initiation-of ADS.
I l
2.2.3 &giggtfions and-Initial Conditions
[
t +
The Appendix R analysis performed previously-(Reference 5.6) ' assumes a LOOSP at time zero which represents the same transient. analyzed in' Section 2.1 of this report.
The performance criterion for each are identical;_- essentially the avoidance of Ll
'The assumptions utilized for the LOOSP analysis are the same as the Reference 5.6 Appendix'R analysis except for the following added conservatisms:
I 1-a.
Initial reactor water level for the the Section 2.1 analysis was a relaxed L3*.
For the' Reference 5.6 Appendix R analysis, water was:
initially at normal water level.
1
- Twelve inches below current level 4
I
.~.
.. 2...u
. o l
i b.
The Section 2.1 analysis used relaxed RCIC performance requirements..
l' The Reference 5.6 Appendix R analysis used RCIC design values.
i 2.2.4 Results RCIC performs the same function for Appendix R events as it does for the'LOOSP event ~under similar conditions.
Therefore, the results of Section 2.I' apply i
to the Appendix R analysis.
For an Appendix'R event, the' indicated reactor i
water level is ma'intained above the L1 setpoint.
Therefore, a lower L3 does 4
not affect the RCIC systems capability of providing sufficient makeup i
inventory to preclude the automatic actuation of ADS during a fire event.-
r 2.3 HELB ANALYSIS i
2.3.1 Backaround
?
Reference 5.5 demonstrated that RCIC ensures the core remains covered through.
(
out the following HELB events. The RWCU'line break is the most limiting case, with the lowest water level being 4.5 feet above TAF during the event.
l
?
a.
Main Steam Line Break Outside Containment, b.
HPCI Steam Supply Line Break Outside Containment,.
c.
Reactor Water Cleanup.(RWCU) Line Break'0utside Containment'.-
2.3.2 Results The results for the RWCU HELB~ event'showed that a reduction of less than 1.5
- t ft for the lowest water level from Reference 5.5 result.., Water level is expected to remain at least 3 feet above.TAF during the applicable HELB events.-
- h r
5-
- ~..
-,,,,m*--,,._-,,r....--y,-4,,,-_,,
4, 4
,,,y, y,
,. +
- e va,
, 7
__ ~
i e
?
..I 3.0 ANALYSIS OF LOSS OF COOLANT ACCIDENT EVENTS i
-i
3.1 Background
I The Plant Hatch LOCA analyses (Reference 5.2) were performed with a bounding l
maximum average planar linear. heat. generation rate (MAPLHGR) and conservative I
I values for some ECCS performanc.e requirementsL relative to the current technical specifications or expected equipment performance.
i Reforence 5.2 indicated that the-limiting LOCA event was a DBA recirculation k
~
suction line break ~with a battery failure. The calculated PCT.using Appendix K assumptions was 1504'F for Unit 1 and:1521*F for Unit-2.
i 3.2 Results A ' sensitivity study was ~ performed previously (Reference 5.2) with-revised water level setpoints including reducing: L3 by 12.5 inches.
The results i
demonstrated that there is no effect on the' calculated PCT for. the DBA event, f
~
e These studies. also showed that there was a small impact (less than 30*F) for i
small breaks but the PCTs would be still well below the' calculated PCT for i
DBA. This result applies to both Plant Hatch Units.
I l
j L
i o
?
t t
a
~_
_ _ _ _ _ ~ _
t
'0 4
L
4.0 CONCLUSION
)
Lowering L3 setpoint involves a Technical Specification change and should be i
examined as specified in 10CFR50.92.
It is concluded that no significant safety hazard is constituted as defined in 10CFR50.92 for : the following reasons:
a; The proposed L3 setpoint does not involve a significant increase in the
{
probability or consequences.of an accident previously evaluated in the j
FSAR Update. The analyses herein demonstrated that the circumstances -
initiating the transient event ~(loss'of feedwater) and LOCA event remain i
unaffected, and the' consequences are acceptable.
]
- b. The proposed L3 setpoint does not create:the possibility for an accident of a new or different_ type than any evaluated previously in the FSAR l
Update. This change affects only scram water level in response to-accident conditions. Therefore this modification does not affect-conditions initiating any' type of accident.,
- c. The proposed L3 setpoint, does-not involve a significant reduction'in the q
margin of safety as defined in the Bases of the Plant Hatch Unit 1 or 2 Technical Specifications. The analysis in this report 1 determined the.
6 l'
impact'of a 12' inch reduction from the current L3 setpoint-The results i
showed adequate margin to the safety limits. Therefore the margin of l i safety within the Bases are unaffected.
l-In summary, the evaluation presented in this report has demonstrated that lowering L3 by 12 inches from the current setpoint can be justified analytically.
Adequate L1' trip ' avoidance can be provided during: transient events.. For LOCA events, there is no effect to the DBA: and-a small impact to the small breaks.
Therefore the l change is also acceptable' from an' ECCS viewpoint.
r
-c 7
i
. ~
. L.......-
.. l
-~
l e
1
.r l
9 l
5.0 REFERENCES
i i
5.1 " Safety. Evaluation for Relaxation of RCIC Performance Requirements for i
Plant Hatch Units 1 and 2", EAS-410688, July 1988.
+
5.2 "Edwin I. Hatch Nuclear Plant Units 1 and 2 SAFER /GESTR LOCA Loss-Of-Coolant Accident' Analysis', NEDC-31376P, December 1986.
J t
5.3 Design Specification, " Reactor Core Isolation Cooling System".
j General Electric Company, 22A1354.
l 1
L L
5.4 "High Drywell Temperature Effect on Reactor Vessel Water Level Instrumentation", General Electric Company, SIL-299,. July 25,:
j l
1979.
l l
l 5.5 "High Energy Line Break' Evaluation For Edwin I. Hatch Nuclear:
Power Station", General Electric' Company, NED0 24873-02, I
September 1981.
5.6 L. L. Chi, " Safe Shutdown Appendix R Analyses for Edwin I.. Hatch Nuclear Power Station Units 1 and 2", General Electric ' Company,-
MDE-03-0186, December 1985.
i i
Y
)
i 9
l
'(
L i l L,
..