ML20082Q755
| ML20082Q755 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 08/30/1991 |
| From: | Matthews D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20082Q759 | List: |
| References | |
| NUDOCS 9109130053 | |
| Download: ML20082Q755 (19) | |
Text
.
.~._.
s'o Il 5 j 3
??
UNITED STATES b
,i
- NUCLEAR REGULATORY COMMISSION WASHINGTON. D C,20866
$,c
_4 *... * -
GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-321-EDWIN 1. HATCH NUCLEAR PLANT, UNIT-1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment-No.:173 License No. DPR-57 1.
The Nuclear-Regulatory Commission (the Commission) has-found that:
The app (lication for amendment to the Edwin 1. Hatch Nuclear Plant,the facil A.
Unit 1-Georgia Power Company, acting for itself, Oglethorpe Power Corporation, Municipal Elcctric Authority of Georgia, and City of Dalton, Georgia-(the= licensees) dated October 9,_1990, complies with the standards and requirements of the: Atomic Energy Act of 1954, as-amended-(the Act),land the Commission's rules and_ regulations set
-forth in 10 CFR Chapter I;
~B.
The facility will-operate in conformity with the application,'the
-provisions-of the Act, and the rules and regulations of the Commission; L
.C._
There:is reasonable assurance (i) that the-activitie:; authorized by this amendment can be conducted without'. endangering the health and safety of.the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations-set forth in 10 CF*v 1 Chapter I; D.
The issuance-of this' license amendment will not be inimical to the common defense and security or to the-health and safety _of the public; and E
'E.
The issuance of this amendment is in accordance with 10 CFR Part 51 l
of the_ Commission's-regulations and all applicable requirements have been satisfied.
9109130053 910830 PDR ADOCK 05000321 P
[
e-
,-,e
-m
,+w y
4+-e-r.i.-e r
,,4..
~,
ww.
. 2.
Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-57 is hereby amended to read as follows:
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.173, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION hYA. $
David B. Matthews, Director Project Directorate 11-3 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance:
August 30, 1991
l
. g* "'c y%
w I
+
$.T UNITED STATES
- ' !s,(f' %VE NUCLEAR REGULATORY COMMISSION f
WASHINGTON. D.C. 20066 ogv /
GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-366
- EDWIN 1. HATCH NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 113 License No. NPF-5 1.
The Nuclear Regulatory Commission (the Commission) has found that:
The app (lication for amendment to the Edwin 1. Hatch Nuclear Plant,the facility) Faci 5
Unit 2 Georgia Power Company, acting for itself, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the licensees) dated October 9, 1990, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The f acility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.
The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
J
2-1 2.
Accordingly,-the license is'hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment
- and paragraph 2.C.(2) of Facility Operating License No. NPF-5 is hereby amended to' read as follows:
Technical-Specifications The' Technical Specifications contained in Appendices A and B, as revised through Amendment No.'113, are hereby incorporated in the license. The licensee shall operate the facility in accordance with-the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
.I
[ dd.Al b d h
David B. Matthews, Director Project Directorate 11-3 Division of Reactor. Projects - 1/11 Office of Nuclear Reactor Regulation
Attachment:
- Technical Specification Changes Date of Issuance: August 30, 1991 i-
+
e--
%-.e-
-.4
--,---q wi
,, - + -
g-p-
,,sw--,
+-
-+p,,
, ATTACHMENT TO LICENSE AMENDMENT NO.173 FACILITY OPERATING LICENSE NO. DPR-57 DOCKET NO. 50-321 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the areas of change.
Remove Pages Insert Pages 1.1-3 1.1-3 1.1-13 1.1-13 Figure 2.1-1 Figure 2.1-1 3.1-4 3.1-4 3.2-2 3.2-2 3.2-10 3.2-10 3.2-50 3.2-50 3.2-58 3.2-58 l
l-l I
1
l -.
l '.
SAF EYY L IMIT5 tlMlYING 5AFEYY SYSYEM SEYYlNGS 2.1.A.I.d.
APRM Rod Block Irip 5etting This section deleted.
s 2.1.A.2.
Reactor Vessel Water low level Scram.
Trio Settina (t evel 3) 9.
Reactor vessel water Isw level scram trip setting (Level 3) shall be
> 0.0 inches (narrow range scale).
l 3.
Turbine Stop Valve Closure scram Turbine stop valve closure scram trip setting shall be s 10 percent valve closure f rom full open. This scram is only effective when tur-bine steam flow is above that corres-ponding to 30% of rated core thermal power, as measured by turbit.e first stage pressure.
HATCH - UNIT 1 1.1-3 Amendment rio.173
l BASES E Rj lNITING 5AF[ly SYSTEM 5[T11NGS l
2.1.A,1.c.
APRM Flun Scram Trip Settinas Qun Mode) ((ontinued)
The APRM flow ref erenced simulated thermal power monitor scram trip setting at fell recirculation flow is adjustable up to 117% of rated power for two-recirculation loop and single-recirculettbn loop ophretiens.
This reduced flow referenced trip setpoirt will result in an earlier scram during slow thermal transients, such as the loss of 100-F f eedwater heating event, than would result with the 120% fix0d high neutron flux scram trip. The lower flow referenced scram setpoint therefore decreases the severity (ACPR) of a slow thermal transient and allows lower Operating Limits if such a transient is the limit)hg abnorn.a1 operational transient during a certain exposure iri*rval in the cycle.
The APRM fixed high-high neutron flux scram trip, adjustable up to 120%
of rated nower f or two-recirculation loop and single-recirculation loop operations, does not incorporate the time constant, but responds directly to instantaneous neutron flux. This scrcm setpoint scrams the reactor during f ast power increase transients if credit is net taken f e a direct (position) scram, and also serves to scram the reactor if credit is nnt taken for the flow referenced scram.
2.
Reactor Vessel Water tow level Scram TriD Setting (level 3)
The trip setting 'or low level scram is above the bottom of the separator skirt, figure 2.1-1.
Thi'. level is approximately 14 feet above the top l
of the active fuel. This level has been useo in transient analyses dealing with coolent inventory decrease. The results reported in FSAR Section 14.3 show that a scram at this level adequately protects the fuel and the pressere bc rier. The jesignated scram trip setting is at least 22 intnes below the bottom of the normal operating range and is thus adequate to avoid spurious scrams.
HATCH - UNIT 1 1.1-13 DuMtM M f
900 "T' l
NOTE.
SCALE IN INCHES ABOVE VESSE L ZE RO i
WATER LEVEL NOMENCLATURE HEIGHT ABOVE VESSE L ZERO 800 -- -
g teNCHist RE ADING INST R UM E NT to) 573.5
+ 56.5 B A RT ON/ROSEMOUN T 750 - -
(7) 559
- 42 GE/MAC VESSEL tai 549
+32 CE /MAC 23 56 F L ANGE 13) b17 0.0 BARTON/ROSEMOUNT 9
(2) 470 47 BARTOt;/ROSEMOUNT 700 -
til 404
- 113 BARTON/ROSEMOUNT tot 315
-202 DARTON!ROSEMOUNT 650 - -'
/~
600 --
. 577
- 60 -
+ 60 - -
+60 -
56.5 573,5 (81
+60 --
(g3-47 - 42 HI ALARM N
- 559 (7) 550 -r $4944) g
,g
} ' 32 LO AL ARM RCIC T RIPS BOYTOM OF ST E AM -
~
- ORY E R SKIRT-- $17.
(
- 0(3) 0--
0(3)-
_ FEED - 4835 NT 4 UTE TO AD3 LOW (LEVEL 31 5
- 47 LOW LOW (LEVEL 2)
WATER
- 470 (2)
~
spray 450 --
"- 404 (1)
- -113 LOW LUW LOW (LEVE L 11 INITt ATE RHR. C S.,
- 367
- 150 --
MM DM AW 352 56 350 -
CONT RIBUT E TO A C.S.
CLOSE MSIV's 2QCORE
- 202 NElGHT 315 (0) 300 --
P E RMISSIV E ACTIVE (LEVEL 01 FUEL 25'i
-317 --
200
- 108 56 RECtRC d
-na 56 -OisCH ARot )
RECinC NOZ2LE SUCTION - 1615 NO Z2L E 13 100 50 --
l O~
9380 2 FIGURE 2.11 RE ACTOR VESSEL WATER LEVEL iNuendment ilo 173 HATCH - UNIT 1
~ - - _
.- ~._ _ _..
~
..m.
.-m..m...--
...w Table 3.I-I (Coni'd)
~
- :t
>:7j Scram Operable at Nur6*r Source of ' 5 cram Trip Signal Channels-Scram Trip Set ting.
Source of Scram Signal is Required (a)
Required Per to be Operable Except as Indicated Trip System Below th) 5 High Drywell Pressure
'2 1 1.92 psig s.
Not required to be operable when primary containment integrity is 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 liigh tevel (ontrol rod blo(k) in order"to reset RPS when the Mode Switch is a.
Float Switrhes 2
1 71 gallons Lin the REFUEL or SHUTDOWN position, b.
Thermal level sensors 2'
i 71 gallons 8
APRH flow Referenced Simulated 2
-5 1 0.58W+62% - 0.58 oW
'See Specifisation 2.1.A.I.c(1) for Thermal Power Monitor (Not to exceed ll7%)
definitions of W and (W.
Tech Spec 2.12A.1.c(1)
Fixed High High Neutron 2
5 5 120% Power Flux Tech Spec'2.1.A 1.c(2) w Inoperative 2
Not Applicoble An AP2H is inoperable if there are less than two LPRM i.iputs per level 45 or there are less than Il LPRM inputs to the APRH channel.
?
ma a
3r+
M N
OJ 3
1 1,
latale 3.2-1 Af4a FkitM W IN5f kutet tJI Allur4 WHI(ft irJI TI Al[5 kf ACIOR VI $5f t 1
COfflAINMENT ISOLATI0f4 n'
~
x Required Attion to be taken if Irip Cperable riu.ber of t hannels i s n
Condition Channel' Trip 5etting for both trip ko1 arks {d)
Ret.
not met E
No.
Instrument Nomenclature per Trip (a)
Sy112lM
,__Jnith hl_ _
f
~
Initiotes Group 2 8. O I
Reactor vessel tow (level 3) 2 2 0.0 inches Initiate an orderly isolation.
shutdown and achieve Water tevel Narrow Range the Cold Shutdown Condition within 24 Fours or i stri at e t f.e shutdown tooling system.
Initiate an orderly 5 tarts the 5615, initiates broup 5 2
A-47 inthes shutdown + d asheeve-tow Low isolation, and (tevel 2) the Coht Shuicawn C ornii t : on weth.
/1 initiates set oridar y (ont ainment hours.
isolation.
lnitiat* dn ord s)v Ini t lates Grwp 1 g_}]j i n g.he ',
shutaawn sud **Liew" tow Low low (tevel li t i.e l u l.1 S h u t tbswh I O" -
d e t it;n i t hin 24 t-our s.
i sol at es t he stmtdown GJ Isolet e shut do o (ouling suition valves 8
/
Ecat t or Venel Steam to Permissive 1
<145 psig e
tooling.
tv o'
the NHR syston.
Dome Pressure (Shut-down Cooling Mode) or atee l y Starts the stendby 2
51.92 psig I,+ e t i o t e on gas t reatenent m yst + m.
!t t yh shutdo=+i and e(hieve initta'.es Group 2 1
Drywell Pressure the Cold Shutdown Condition within /4 isolation and secoral-ary (ori t a i nn.en; hour 3 esoleticn.
i5 O
C1
~3 O
O M
N LJ
f
. e I
7 lable 3./-4:
1 T-b if15f kUtiltJTAl!0t1 MilOf ItalTIAIES OK COf41kot $ AD5
-4 i
j Q
Ref.
Instrument Irip
' Required-Irip 5etting kemarks No.
- Condition Operable i
8
{a) tiomenclature.
Channels j
b:
_., _, _ ~ _
$n1O!L (12) c
. per Trip j
4
--4 1
l.
Low (level 3) 1 10.0 im bes -
Confirms low level, ADS permissive f
i Reattor Venel Water Level.
Low 4. ow L ow '
2
' f-113 inthe s Perminive siyrial to AD5 timer (Level 1)
,',j 2.
Drywell Pre nure Ifigh 2
11.92 psig Perminive sitful t o AD5 t imer k
i j
i.
WtiH Punip Distharge
' lingh 2
a ll/ psitj Pe-e nes n i we s t yna l to AD5 tsees j
Prenure 1
4.
C5 Pump biulian ge High 2
2Is/ psig ee. i nive,sige.al to Aus timer Pressure I
5.
Autn Depreneiriratior-2 1 14 ma n.t a-s ttypo.ses b reth tie ywe l l pre neeee
}
tow Wales-level Timer perminive upon suht oined t evel I
.u ro 6.
Auto Depre uurization I
liti 1/ seconds wi t h t ewel' t er><f l evel i and high j
f.,
T i sr.e r dry-ell pre % ure and CS or RilR pump h
o at pressure, timing sequente p
begins.
If the ADS timer is e,ot i
reset it will initiate A05.
1.
Automatit Blowdown Control I
' Jut opplit eble f49nitors availobility of power tu Power failure tionitor logic system' 2
3 L
1t m
-1he column ent itled '* Ret. tio." is onl y f or convenience so tha, a one-to-tme relat ionship (en t>e established betweer.
- s a.
.y
}
items in Table 3.2-4 and item in Table 4.2-4.
[
i m
t b.
Whenever any (CCS sutnystem is required to be operable by Sect ion 3.5, there shal! t,e two operable trip systems.
I If the rew rd nuarber of operable thannels cannot tre niet for one et the trip systems, that siistem shall be repaired a
4 or the rem %.
thall be placed in the Cold Shutdown Condition within 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> af ter this trip s pines is made or found tu j
be i riope rob b.
a
'r i
N W
i 1
r
?
v b
nn..e..,,,.i,.,,
n.
~.
a+
-e,
,~,+n a,,,,,,,
i BAIT TdQ}pjj iL C[ MIf6N 3.2 PDOTECT10N lNSTRUMENiaTIO3 In addition to the Reactor Protecticn System (kPS) instrumentation which in-itiates a reactor scram, crotective instrumentation has teen provided 'hich w
initiates action to mitigate the consequences of accidents which are beyond the operatots ability to control, or terminates operator errors tiefore they result in serious consequences, This set of Specifications provides the lim-iting conditions for operation vf 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 block 5, td) whicn initiates protective action.
(e) which monitors leaA6ge irto the drywell and (f) which provides survell-lance information. The objectives of these specifications are (1) to esture the effectiveness of the crotective instrumentation when required by preserv-ing its capattlity to tolerate a single failure of any co.*po"ent 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 required to assure ade-Quate performooce When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.
A.
Instrumentation Whtch Initiates ieactor Vessel and Primary Containment
___._I a b E___3_. 2 1 )_
i s ol a t_i on
(
lsolation valves are installed in those lines which penetrate the primary con-tainment and most be isolated during a loss of coolant accident so that the radiation dose limits are not enteeded during an accident condition, Actua-tion of these valves is initiated by protective instrumentation snowa in Iatle 3.2-1 which senses the conditions for which isolation is required. Such in-strumentation must be available whenever primary containment integrity is re-Quire, The Ctjective ti to isolate the primary Containment so that the guicelines of 10 CFR 100 are not exceeded during an acctcent. The events when isolation is required are discussed in Appendin G of the FSAR. Tne instrumentation which initiates primary system isolation is connected in a dual bus arrangement.
1.
Reactor Vessel Water Level Low (Levei 3) (Narrow Rangel The reactor water level instrumentation is set to trip when reactor aater level is approximately 14 feet above the tcp of the active fuel. This level is referred to as Level 3 in the Technical Spect-fications and corresponds to a reading of 0.0 inches on tne Narrow l
Range scale. This trip initiates Group 2 and 6 isolation 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 Speci-fications and corresponds to a reading of ~47 inches, This trip initiates Group 5 isolation, starts the standby gas treatment system, and initiates secondary containment isolation, HATCH - UNIT 1 3 2-50 Ameninent ilo.173
4 BASES F0i' LIMITING CONDlI[ONS 60R OPERATION l
0.
Instrumentation Which Initiates or Controls ADS ( Table 3.2-41 The A05 is a Dackup system to HPCI. In the event of failure cy HPC!
to maintain reactor water level, ADS will initiate cepressurization of the reactor in time for LPCI and C5 to adequately cool the core, Four signals are requireo to initiate ADS: Low water level, confirmed low water level, high drywell pressure, and either a RHR or Core Spray pump available. The s'multaneous presence of these four signals will initiate a 120 second ticer which will depressurize the reactor it not reset.
1 Peactor Vessel Water Leve) a.
Reactor Vessel hater level Low (Level 3)
The second reactor vessel low water level initiation $rtting
(+0.0 inches) is selectea to confirm that water level in the l
vessel is in fact low, thus providing protection against inadvertent depressurization in the event of an instrument line (water level) failure, Such a failure could produce a simultaneous high drywell pressure. A conf'rmed low level is one Of four signals reagired to initiate ADS.
t.
Peactor Vessel hater level tow Low Lo. (l eve l 1)
The reactor vessel low water level setting of 113 inches is selected to provice a permissive signal to open the relief valve and depressurite the reactor vessel in time to allow acequate cooling of the fuel by the core soray and LPCI systems follo.ing a LOCA in which the otner make uo systems (RCIC and HPCI) fall to maintain vessel water level, ibis signal is one of four recuired to initiate ADS.
2.
Drvwell Pressure High A primary containment hign Dfessure of 2 2 psig indicates that a breacn of the nuclear system process carrier has occurred inside the crywell. The signal is one of four requitec to initiate the AD5.
HATCH --UNIT 1 3.2 58 Anendment ilo.173 a
.._...______j
ATTACHMENT TO LICENSE AMENDMENT NO.113 FACILITY OPERATING LICENSE NO. NPF-5 POCKET No. 50-366 Replace the following pages of the Appendix "A" Technical Specifications with The revised pages are identified by Amendment number and the enclosed pages.
contain vertical lines indicating the areas of change, Insert Pages Remove Pages 2-4 2-4 3/4 3-16 3/4 3-16 3/4 3-18 3/4 3-18 3/4 3-29 3/4 3-29 8 3/4 3-6 B 3/4 3-6
I/4LLLL1-1 BLMiHR EEuMLDiLSYS101BS19UMENTAiICLLSL12 Ult 1D y
LWM10!L4L1: fill 1811'_1Ut91ril atteggottyetuts g
1.
Intermediate Range Monitor Neutron flu =-High
$ 120/125 divisions i 120/125 divisions
-4
( 2C514601 A.B.C,D E,f,G H) of full state of full stale e
c--
2.
Average Power Range Monitor:
(2C51-K605 A,8,C.D.E,F)
Neutron fluN-Upscale, 15%
4 15/125 divisions 1 20/1/5 divisions
-4 of full scaie of full scale a.
b.
Flos Rei renced Simulated Thermal 1 (0.58 W + 59% - 0.58 p ) -
1 (0.58 w + 62% - 0.58tw) -
with a manimum with a manoeum Power-tIpsc al e i 113.5% of RATED 5 115.5% of RATED THERMAL POWER IHfRMAt F OWE 5i f ixed Neutron f lus-tJpscale, 118%
i 118% of RATED
$ 120% of kATED THERMAL POWER THERMAL 00wER c.
3, Reactor Vessel Steam D0me Presst,re - High i 1054 psig s 1054 psig (i'821-N618 A,8,C,0) f 4.
Reactor Vessel Water level - Low (tevel 3) 2 0 irut.es above f 0 inthes obuve instrument zero" instrument zero-(2821-Nb80 A,B,C,D) 5.
Main Steam Line Isolation valve - Closure i 10% closed i 104 t.lused (NA) 6 Main Steam t ir Radiation - liigh 1 3s full-power s1 = full-puer (2D11-K603A,B C,0) background ***
bad yround*-*
re f
7.
Drywell Pressure - High 1 1.92 psig s 1.91 psig (2C71-H650A,0,C,0)
-T6 45ee Bases figure B 3/4 3-1.
- W :. Total loop re(irculation flow rate in percent of rated.
Roted loop retirculation flow is equal to 4<1.2 MLB/hr.
C)
AW. Ma=imum measured difference between two-loop and single-luep drive flow for the so w (cre flow in penent
=
of rated retirculation tion for single-loop operation. The value is zero tur t.o-loop operation.
b
"*Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ( rior to the planned start of the tydrogen injection test with the reo tor power at the normal full-power radiation background level and associated trip greater tr:an 20% rated po.er, setpoints may be changed based on a calculated value of the radiation level espected during the test.
The background radiation level and associated trip setpoints may be adjusted during the test based on injectico. The either calculations or measurements of actual radiation levels resulting from hydrogen background radiation level snall 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 after completion of hydrogen injection and prior to establishing reactor power levels below 20% rated power.
l 7
dnu s
o td r
nns g
un 6
iooo t
pring a
t gt onr b
t ek aiie w
ascl th w e
lo auaso r
r pb cii p m
ei l dl w
f
- s o
u wreaabr se p
d u
s oth crao a eh g
e c
g e
p T
tt E ;
sh c i
l gt a
r r
i h
r d
rdsc L
t cn s
l ia v
h h
s c
h re enea B(
h ni p
u sr
/
/
p n
/
ot
.h u-e AU ci f
p f
g f
R r.
r t
WI n
1 2
H n
2 i
m (iatt oer sir r OA i11 9
x 58 4
0 8
9 7
aceeg g
l V 41 1
3 81 1
7" l
e-23 9
0 1
0 1
4 0
eot k f n 3
2 6
2 rs nc oi A
seoa h
222 i
i 2i i
2 1
1 i i 2
i eah b ss h
td ri t d e e ul ngsh ob hanaTha t
i b t
d 4s il r
n weut 2e u
vdsn l
o t e eono lu r
sl dtii t g
e' e
th o
k tdt ect r P
c ncheio T
a nuetjwi E
b oop n r
S w
irs gi t p r
o t gen e
4 e
l ck i nsd Q
w f
m ecl re n
I
- s o
u jaeugea T
se p
d u
s nb vdob A
- eh 2 n u
ecn ig e
c g
e i
e r
n i
l sh c l
gt a
r r
is c
h no e yl i h
r nl ddl o s
l ia v
h h
t l
h ni p
u sr
/
/
p n
/
einthat 2 H 0
ci f
f 9
f R
r i
e o ai u m s e p%
r gt os hc 4
u P
n 3
2 a
H m
2 3
R T
i71 9
x 58 4
0 8
9 ritjo j
7 T
E 41 23 9
7" 0
3 0
1 4
0 ddad rsn 3_
S S
0 - -
l 3
81 2
1 6
2 y ai a f ti 1
N h rd n
L I
P 222 i
2i 1
2 1
i 1
1 2
1 a e gi n L
I errb noe B
n R
h e ipg A o T
t w e yt t o I
l.
oh al er l
f pt musd a
o s
y u
lf s e ph t
tl ot ri c
ru n
rf A
af eist o h
t uol pp ig sll ped n aat veo l
H dmveeti i
er sl at M-
- p D
2 nnepncl 9
)
nod ie l
0 D e
)
m l
a ti oop 5
0 v
1 e
h i
t l
l earism 1
1 e
h T
g M
s h
e t
p h. l tt so 1
l l
g l
w u
g v
s 8
u aac
) e i e
o a i 0
a e
u e
( di A
H L
r 2 v H
h nh L
e 5
h iH rL a
h l eddr.
0 e
e 1
t l t ig ng r
1 x
e(
h l
ow t rat aner w
ui A
L Eh t
xh ecarat e 1
a) e(
H o
l H
n g
aw
[
w e
t m
g o i gi e Wo ig i
f LW3v t
3 t oacl dao ew r
e-h e-u n i i inl r
L rH p
oae p
i it llL o u
n g
ee(l s
i n i
n u i t
u l
o 4
r nsuns E
seww e
i el i e c
d a
1 d -
s ew o-
/
od ostil d f
sv s
L o i
l r a
iu a
T uo r
e o
B paed eea l
1 l
s so l
3 irtd arvt e
acmee r
u V
id A
i i
n e
sL f n ll eL oo r
mt u-mt A
V(tl P
aa s
aa r
B R
N Bi P
V -
gi rsnf tl r 1
ei sw er e
O t
nt e
s aaoe 3 rwww l
t d eo t e s
e l
C ra l
rl i
0 oooo l
S a rl S p n
n l
oi l
oe l ia r
r%b dn%
u u0 l s o0 C
ttLt e
R PF m
e i
e Y
t d e
t v ed i
h evnbt g
o2det ei2 N L c
w n
ne d
b w
R ca w
ce ua O
a y
R e...
r i
il n
r y
A aR y
aL f R f
ngee aw a.
a o
u r
0 e
r e
e 4 anl ml I
i o T
A R123 D
M1 23 H
C T
D 4
R D
R R
s 2ha el dl C
i Q
e t h nraae t
N i
C s
n couhrb U R g
a b
c d
B h eet a l s L
a i r i ss F
E a
b c
d e
f t tbael al P
e i a i meme I
e W e yd vrv R
S
- raareoe I
l 2
- gmrol nl o % *> c. s c
,O$ E3$ c 5*a$[
gg,x ' C::_e L
3 a an
_c u
~ _. _, _ _. - ~ -_
m.
4 LABLE 3.3 2-2 f fnatinuestl g
I SOL A T 10t1 AC 10A T10!LitJS T Rt; MEN T A T I QU _$LILV1tjl$
w.
n At10WABLC 8
1 RIP FUNLl]QH 1 RIP _SLIEQ1111
._YALUL__
' c$
5.
BLACIOR CORE 1501A110N y
(QQLING SYSTEM ISOL ATION a.
RCIC Steam Line flow - High 1 307% of. rated flo.,
.1 3071. of rated flow b.
,RCIC Steam Supply Pressure - Low 1 60 psig 2 60 psig RCIC Torbine t.mhaust Diaphragm c.
Pressure - High 1 20 psig i 20 psig.
d.
Emergency Area Cooler Temperature-Hig" i 169'T i 169'F' e.
Suppression Pool Area Ambient Temperature 1 169"I i I t>9
- F High f.
Suppression Pool Area AT - High I'42*f 1 4/*f 9
Suppe 'ssion Pool Area Tempe rature limer Relays i4A NA N
h.
Drywell Pressure - High 1 1.92 psig i 1.92 psig b
u
- i. Logic Power Monitor NA NA 5
6.
$hijlDQWN COOLING S1$1LtL15QLAllDN a.
Reactor Vessel Water tevel - tow
.2 0 inches
- 1 0 inches" l-(Level 3) b.
Reactor Stease Dome Pressure - High 1 145 psig 1 145 psig
=
[
't o=
r+
=
C' "See Bases figure B 3/4 3-1.
M N
4 y
a.
Reactor Vessel Water Level - Low Low (Level 2) 1 -47 inches
- 1 -47 inches
- n3 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 1 154.2 inches ***
i 154.2 inchev e.
Logic Power Monitnr NA NA f.
Reactor Vessel kater Level-High (level 8)*
I 56.b inches i $6.5 inches 4.
AW0tB11LDLf2115W12AL101Uili1LB a.
Drywell Pressure-High i 1.92 psig i 1.92 psig b.
Reactor Vessel Water level - Low Low Low (level 1) 1 -813 inches
- 2 -113 inc hest c.
AD5 limer i 120 seconds i 120 seconds d.
ADS Low Water Level Actuation Timer i 13 minutes 1 13 niinutes e.
Reactor Vessel Water Level - low (Level 3) 2 0 inches
- 2 0 inche,*
f.
Core Spray Pump Discharge Pressure - High 1 137 psig 2 137 psig g.
RHR (LPCI MODE) Pump Discharge Pressure - High 1 112 psig 2 112 psig h.
Control Power Monitor NA NA hd 5.
LQu_LQu_1ll_SLRY_ilSILD I
.c-Reactor Steam Dome Dressure - High i 1054 psig i 1054 psig o
a.
N ua I
P
$s CL 3o3
- See Bases figure B 3/4 3-1.
jy
- Equivalent to 10.000 gallons of water in the CSI.
- Measured above torus invert.
a W
I nM
- 9^
1-
+
.' 4' x L
I_ABtE 3.3.3 { ((gntingsui]
' -EMERGENCY CORE CQQLlHEJi1SIEM ACILL4110*2 INSIEUMLNTAllu!L3EltQ1HIS
--t n
- 3:
ALLOWABLE 1RifLLUNC1 Ion IRIIL3EIfninI vAtUE c$
3.
ti1GH PREiiUBE EDOL ANT IMJEC110N SYSTEM -
M a.
Reactor Vessel Water level r Low low (Level 2)
' l -47 inches *
.2 ~47 inches
- y b
Drywell Pressure-High 1 I.92 psig i 1.92 psig c.
Condensate Storage Tank tevel - Low 10 inches" 2 0 inches **
d.
Suppression Chamber Water Level - High 1 154.2 inches ***
1 154.2 inches"*
e.
Logic Power Monitor NA NA f.
Reactor Vessel Water Level-High (Level 8)*
i 56.5 inches 1 56.5 inches 4
. AUIONATIC DEPRESSURIZA,110!L3yi][tj I
a.
Drywell Pressure-High i 1.92 psig i 1.92 psig b.
Reactor Vessel Water Level - Low low Low (level 1) 1 -113 inches
- 2 -133 ieuhes*
c.
ADS Timer i 120 seconds i 120 seconds d.
ADS tow Water Level At tuat ion T imer i 13 minutes i 13 minutes e.
Reactor Vessel Water Level - Low'(Level 3) 1 0 inches
- 2.O inthes*
, l '
f.
Core Spray Pump Discharge Pressure - High.
l'137 psig 2 137 psig q.
RHR (LPCI MODE) Pomp Discharge Pressure - High 1 112 psi 9 2 112 psig
[
h.
Control Power Monitor NA NA
^
}
5.
LON_10LSELS/RV SYSILM 4
y a.
Reattor Steam Dome Pressure - High 1 1054 psig s 1054 psig E
v t e
?
8 l
=
Rs 3
3
- See Bases Figure B 3/4 3-1.
g
" Equivalent to 10,000 gallons of water in the CSI.
Measured above torus invert, m
.==9 3
t P
"9
- Ml-O' A
T w
3 s
900 7 -
NOTE-SCALE IN INCHES ABOVE VESSEL zERO 2
e WATER LEVEL NOMENCL ATURE
' HEIGHT ADOVE 800 --
VESSEL zERO g
flNCHf S RE ADING INST R UM E NT tal 573.5
- + 56.5 B A RTON/R OSEMOUNT 750 --
41) 559
+32 GE/MAC.
- 122.7F F L ANGE'"
(3) 517 0.0 BARTON/ROSEMOUNT 12) 470 47 BARTON/ROSEMOUNT 700 -
III 404 113 BARTO;4/ROSEMOUNT (0) 315
-202 BARTON/ROSEMOUNT 650
UNE f
600 - -
73.5 186 60 --
+60 -
56 5
- $$9 471 (C)~
(7) >42HiALARM 549(4) 550 - r HPCI R.
(
-ORYEH SKlRT -
$17(3lN$ " MENT
'-* 0 - 0(3) 0+
0(3F.RE ACTOR SCHAM ztHO 500 -
CONTRIBUTE TO ADS LOW (LEVEL 3)
-.4835 470(2)
- 47 LOW LOW (LE VEL 21' wayrn CORE l
5
~~404 (1)
- -113 LOW LUW LOW (L EVEL 11 f
- 367
-150..
int il AT E RH R, C.S.,
352 56 SI ART OtLSEL AND 350-CONTRIDUTE TO A D.S.
W3 CORE CLOSE MSIV%
- 202 H E tGHT 315101 P E R M'SSIV E 300 --
(LEVEL 01 ACTIVE F UE L.
250 --
317 4-200 4#N RECtRC j
-178 56 -04SCH ARGE YN*# "" D RtCinC SUCT ION - 1615 NOzzt t
^"*'Y"**' L i'"i' . 58 inches Nozzi.E 150 --
100 --
53 +-
[
w z, 9380-2 i
BASES FIGURE B 3/4 31 REACTOR VESSEL WATER LEVEL HATCH - UNIT 2 B 3/4 3-6 Amendment No.113