ML20070T941
| ML20070T941 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 02/02/1983 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Georgia Power Co, Oglethorpe Power Corp, Municipal Electric Authority of Georgia, City of Dalton, GA |
| Shared Package | |
| ML20070T943 | List: |
| References | |
| TAC 49318, DPR-57-A-092, NPF-05-A-030 NUDOCS 8302100051 | |
| Download: ML20070T941 (23) | |
Text
.
a traoq hh UNITED STATES
[
g NUCLEAR REGULATORY COMMISSION s
~j WASHINGTON, D. C. 20555
%,*...*s$
o GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORIlY 0TTEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-321 EDWIN I. HATCH NUCLEAR PLANT. UNIT N0.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 92 License No. DPR-57 1.
The Nuclear Regulatory Commission (the Commissian) has found that:
A.
The applications for amendment by Georgia Power Company, et al.,
(the licensee) dated March 10,1982, April 5,1032, and January 3, 1983, conply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Connission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the applications, 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; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the publ.ic;
,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.
2.
Accordingly, the license is amended by changes to the Technical Spec-ifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License Nc. DPR-57 is hereby amended to read as follows:
8302100051 830202 PDR ADOCK 05000321 P
. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.92, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specificaticns.
3.
Within 90 days after the effective date of this amendment, or such
'later time as the Commission may specify, the Licensee shall satisfy any applicable requirement of P. L.97-425 related to pursuing an agreement with the Secretary of Energy for the disposal of high-level radioactive waste and spent nuclear fuel.
4.
This license amendment is effective as of the date of its issuance.
FOR THE NUCf.lAR REGULATORY COMMISSION Johti F. Stolz, Chief Ope ating Reactors Branch #4 ision of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: February 2,1983 i
?
ATTACHMENT TO LICENSE AMENDMENT NO.95 FACILITY OPERATING LICENSE NO. DPR-57 DOCKET NO. 50-321 Replace the following pages Jf the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain a vertical line indicating the area of change.
Remove
' Insert 3.2-11 3.2-11 3.2-16 3.2-16 3.2-22 3.2-22 3.2-61 3.2-61 3.6-9a 3.6-9a 3.7-21 3.7-21 3.7-22 3.7-22*
- 0verleaf page provided for document completeness.
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TABLE 3.2-5 5
INSTRUMENTATION h11IGI INITIATES OR CONTROLS 'lllE LPCI MODE OF RIIR Required Operable kRef.
Trip Channels No Condition Per Trip h(a).
Instrument Nomenclature System (b)
Trip Setting Remarks
?a i
1 Reactor Water Level (Yarway)
Low Low 2
146.5 inches Initiates LPCI mode of Rm g
Low (LL3) 2 Drywell High 2
1 2 psig Initiates LPCI mode of R$
3 Reactor Pressure High (Shutdown 1
i 135 psig With primary containment isola-Cooling Mode) tion signal, closes RM (LPCI) w L
inboard motor operated injection L
valves w
Low 2
1 335 psig Pemissive to close Recirculation Discharge Valve and Bypass Valve Low 2
1 500 psig Permissive to open LPCI injection valves 4
'ileactor Water Level 1
> - 203.5 inches Acts as pemissive to divert j
(Shroud Level Indicator) some LPCI flow to containment spray 5
LPCI Cross Connect N/A 1
Valve not closed Initites annunciator when valve is Valve Open Annunciator not closed e
a
e f
g il Table 3.2-7 (Continued) l 5
Required
=
Operable Ref.
Trip Chamels No.
Condition Per Trip
.g (a)
Instrunent Nomenclature System (b)
Trip Setting Remarks 3
APRM Downscale 2(e) t 3/125 of full scale Not required while performing low
,g power physics test at atmospheric pressure during or after refueling at power levels not to exceed 5 MWt.
12% Flux 2(e)
< 12/125 of full scale This function is bypassed when the Mode Switch is placed in the RUN position.
High Flux 2(e)
< 0.66W + 42%
W is the loop recirculation flow rate in percent of rated.
Trip
?
level setting is in percent of g
rated power.
Not required while performing low power physics tests at atmospheric pressure during or after refueling at power levels not to exceed 5 MWt.
4 RBM Inoperative 1(e)(f)
Not applicable Inoperative trip produced by switch (g)(h) mt in operate, circuit boards not in circuit, fails to null, less than required number of LPRM inputs for rod selected.
Downscale 1(e)(f)
>3/125 of full scale (g)(h) 2 High Flux 1(e)(f)
,c 0.66W + 41%
W is the loop recirculation flow
~
r Wot to exceed 107%
rate in percent of rated.
Trip 1evel setting is in percent of rated thermal power.
e e
TABLE 3.2-11 g
INSTRUMErlTATI0fl WHICH PROVIDES SURVEILLANCE INFORMATION E
Required
?o
?+ Ref. "
Instrument Instrument Type and Operable No.
- g (a)
(b)
=
Channels Range Action Remarks
.k1 Reactor Water Level (CE/MAC 1
Recorder (c)
(d) k 2
Indicator 0 to 60" (c)
(d) 2 Shroud Water Level 1
Recorder (c)
(d) l 8
1 Indicator -317" to -17" (c)
(d)I 3
Reactor Pressure 1
Recorder (c)
(d) 2 Indicator 0 to 1200 psig (c)
(d) 4 Orywell Pressure 2
Recorder -5 to +80 psig (c)
(d) 5 Orywell Tegerature 2
Recorder 0 to 5000F (c)
(d) 6 Suppression Chamber Air Temperature 2
Recorder 0 to 5000F (c)
(d) 7' 7 Suppression Chamber Water Temperature 2
Recorder 0 to 2500F (c)
(d) id'8 Suppression Chamber Water Level 2
Indicator 0 to 300" (c)
(d) 2 Reporder 0 to 30" (c)(e)
(d)
)
9 Suppression Chamber, Pressure 2
Recorder -5 to +80 psig (c)
(d) 10 Rod Position,Information System (RPIS) 1 28 Volt Indicating Lights (c)
(d) 11 Hydrogen and Oxygen Analyzer 1
Recorder 0 to 52 (c)
(d) 12 Post LOCA Radiation Monitoring System l-Recorder (c)
(d)
Indicator 1 to 106 R/hr (c)
(d) l 13 Orywell/ Suppression Chamber Olfferential Pressure 2.
Recorder -0.5 to + 2.5 psid (c)(e)
(d) i 14 a) Safety / Relief Valve Position Primary 1
Pressure Switch 4-100 psig (f)
Indicator b) Safety / Relief Valve Position Secondary 1
Temperature element 0-600DF (f)
Indicator 4
)
3.2.E.3. Reactor Pressure Low (Continued) jection valves.
The valves do not open, however, until reactor pressure falls below the discharge head of LPCI.
- 4. Reactor Water Level (Shroud Level Indicator)
A reactor water level > 203.5 inches below instrument zero is I indicative that LPCI has made progress in refboding tha core.
A simultaneous high drywell pressure trip indicates the need for containment cooling.
The >
203.5 inch setpoint acts as al permissive for manual diversion for some of the. LPCI flow to containment spray.
- b. LPCI Cross Connect Valve Open Annunciatoi-With the modified LPCI arrangement, the cross connect valve status was changed from nomally open to normally closed.
. Inadvertent opening of this valve could negate the LPCI system injection when needed.
The annunciator will alarm when the LPCI cross connect valve is not fully closed.
A pressure > 100 psig on the RHR pump discharge indicates that the pump has st'arted successfully.
The setpoint provides a pemissive signal to ADS which allows ADS initiati~on if ether requirements am met.
- 7. RHR (LPCI) Pumo Flow (op Switch) Low A flow switch is provided downstream of each RH1 pump to Jndicate the condition of each pump.
To protect the pumps from overheating at low flow rates a minimum flow bypass line, which routes water from the pump discharge to the suppres:; ion chamber, is provided for each pair of pumps.
A single motor-operated valve controls the condition of each bypass line.
The minimum flow bypass valve automatically opens upon sensing low flow in the discharge lines from both pumps of the associated pump pair.
The valve automatically closes whenever the flow from either of the associated main system pumps is above the low flow setting.
- 8. R_H1 (LPCI) Pump Start Timers If normal AC power is available, four punps automatically start without delay.
If nomal AC power is not available, one pump starts without delay as soon as power becomes available from the standoy sources.
The other three pumps start after a 10-second delay.
The timer provides correct sequencing of the loads to the diesel generator.
Amendment 110. g,FI, 92 3.2-61
? I'~ITI:G CD:CCIIC::S MR CfE.:AT1014 SLTvEIL'A'CE FH.CIF.DET'S 4.6.H.
Relief / Safety Valves (Continued) 5.' ' operability /of.: Tail Pi~pe Pressure'Swit~ches a.
Functional 'Ibst:
2.
At each scheduled outage greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during which entry is made into the primary contain-cent, if not perfe red within the previous 31 days.
b.
Calibration and verifying
- - the setpoint to be 85,+15,-5 psig at least.
cree per 15 r.:nths.
I.
Jet Ptr=s I. Jet P,ros Wnenever the reactor is in the Whenever both recirculating pumps Start & Hot Standby, or Run FExie are operating with the reactor in with both recirculating pumps the Start & Hot Standby or Run operating, all jet pumps shall be Mode, jet pump operability shall cperable. If it is determined that be checked daily -by verifying a je: p.=p is in g rable, an that the following conditions do crderly shu:do.n snall be initiated not occur simultaneously.
and the reactor shall be in the Cold Shutd:.e. Ccnditions within 1.
Tne two recirculation loops 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
have a flow imbalance of 15%
or more when the pumps are operated at the sa.ne speed.
l 3.6-9a
- Amendmtet flo. M,25I92
~.-
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Table 3.7-2 Testable Penetrations with Double O-Ring Seals Penetration No.
Penetration Description Notes X-1 A&B Equipment Hatch (1) (2) (4) (6)
X-2 Personnel Lock (1) (4) (7) (8)
X-4 Head Access Hatch (1) (2) (4) (6)
X-6 CRD Removal Hatch (1) (2) (4) (6)
X-35-A through TIP System (1) (2) (4) (6)
X35-E X-43 Drywell Test (1) (2) (4) (6)
X-200 A,B,&C Suppression Chamber (1) (2) (4). (6)
Access Manhole X-218 A&B Construction Drain (1) (2) (4) (6)
O e
t i
3.7-21 t
i Amendment No. 92 4
,-w.
.---.-----,-r,w--
c---,.,
. ~ ~ - -
Table 3.7-3 Testable Penetrations with Testable Bellows Penetration Number Penetration Description Notes X-7A Primary Steamline 'A '
(1) (2) (4) (6)
X-7B Primary Steamline 'B' (1) (2) (4) (6)
X-7C Primary Steamline 'C' (1) (2) (4) (6)
X-7D Primary Steamline 'D' (1) (2) (4) (6)
X-8 Steamline Condensate Drain (1) (2) (4) (6)
X-9A Feedwater Line 'A' (1) (2) (4) (6)
X-9B Feedwater Line 'B' (1) (2) (4) (6)
X-10 Steam to RCIC Turbine (1) (2) (4) (6)
X-11 Steam Line to HPCI Turbine (1) (2) (4) (6)
X-12 RHRS Shutdown Cooling Suction (1) (2) (4) (6)
X-13A RHR LPCI to Reactor (1) -(2) (4) (6)
X-13B RHR LPCI to Reactor (1) (2) (4) (6)
X-14 Reactor Water Cleanup Line (1) (2) (4) (6) 7-16A Core Spray to Reactor.
(1) (2) (4) (6)
X-16B Core Spray to Reactor (1) (2) (4) (6)
X-17 RPV Head Spray (1) (2) (4) (6)
X-201A through Drywell Suppression Chamber (1) (2) (4) (6)
X-201H to Vent Line v
3,7_ 22 l
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s.o trog'o UNITED STATES
^g
[
g NUCLEAR REGULATORY COMMISSION g
/j WASHINGTON, D. C. 20555
~s.,
/
GEORGIA POWER COMPANY OGLETHORPE POWER CORPURTION MUNICIPAL ELECTRIC AUTH0hlTY OF SEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-366 EDWIN I. HATCH NUCLEAR PLANT. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 30 License No. NPF-5 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Georgia Power Company, et al.,
(the licensee) dated March 10, 1982, complies with the standards and requiremats 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 facility 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; D.
The issuance of this 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.
2.
Accordingly, the license is amended by changes to the Technical Spec-ifications 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:
..,n.
, ~,.
O
. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.30, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
Within 90 days after the effective date of this amendment, or such later' time as the Commission may specify, the Licensee shall satisfy any applicable requirement of P. L.97-425 related to pursuing an agreement with the Secretary of Energy for the disposal of high-level radioactive waste and spent nuclear fuel.
4.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION e
1 John F. Stolz, Chief g erating Reactors Branch #4 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: February 2,1983 i
l 1
{
i
ATTACHMENT TO LICENSE AMENDMENT NO. 30 FACILITY OPERATING LICENSE NO. flPF-5 DOCKET NO. 50-366 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain a vertical line indicating the area of change. The overleaf pages are provided to maintain document completeness.
' Insert Remove 3/4 3-28 3/4 3-28 B3/4 3-6 B3/4 3-6 D
4 w
,l n
l TABLE 3.3.3-1 (Continued) k EERGENCY CORE COOLING SYSTEM ACTUATION INSTRtMENTATION 5!
MINIMUM NUMBER APPLICABLE c.
z.!
OPERABLE CHANNELS OPERATIONAL i
TRIP FUNCTION PER TRIP SYSTEM CONDITIONS #
H ro 3.
HIGH PRESSURE COOLANT INJECTION SYSTEM i
a.
Reactor Vessel Water Level - Low Low (2B21-h031 A,B,C,D) 2 1,2,3 j
b.
Drywell Pressure - High (2E11-N011 A,B,C,D) 2 1, 2, 3 c.
Condentate Storage Tank Level-Low (2E41-N002, 2E41-N003) 2(b)lc))
b)'c 1, 2, 3 2(a) d.
Suppression Chamber Water Level-High (2E41-N015A,B) 1,2,3 1(
1,2,3 e.
Logic Power Monitor (2E41-K1) f.
Reactor Vessel Water Level-High (2B21-N017 B,D) 2 1,2,3 4.
AUTOMATIC DEPRESSURIZATION SYSTEM _
R a.
Drywell Pressure High (2E11-N011A,B,C,D) 2 1,2,3 b.
Reactor Vessel Water Level - Low Low Low Y
(2821-NO31 A,B,C D) 2 1,2,3 C3 c.
ADSTimer(2B21-K5A,B) 1 1,2,3 d.
Reactor Vessel Nater Level-Low (Permissive)(2B21-N042A,B) 1 1,2,3 e.
Core Spray Pump Discharge Pressure - High (Pemissive) 4 (2E21-N008A,B; 2E21-N009A,B) 2 1,2,3 4
f.
RHR (LPCI MODE) Pump Discharge Pressure - High (Permissive)
(2E11-N016A,B.C.D; 2E11-N020A,B,C D) 2/loog,)
1, 2, 3 g.
Control Power Monitor (2B21-K1A,B) 1/ bus 1,2,3 (a) Alara oMy. When inoperable, verify power availability to the bus at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or declare the system inoperable.
(b) Provides signal to HPCI pump suction valves only.
(c) When either channel of the automatic transfer logic is inoperable, align HPCI pump suction to the suppression pool.
4 l
HPCI and ADS are not required to be OPERABLE with reactor steam dome pressure < 150 psig.
.g
- s
,[
TABLE 3.3.3-2 EMERCENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS P
g ALLOWABLE o-TRIP FUNCTION TRIP SETPOINT VALUE l
1.
COE SPRAY SYSTEM a.
Reactor Vessel Water Level - Low Low Low 2 -146.5 inches
- 2-146.5 inches *-
i b.
Drywell Pressure - High 12 psig 12 psig c.
Reactor Steam Dome Pressure - Low 1500 psig 1500 psig d.
Logic Power Monitor NA NA 2.
LOW PRESSURE COOLANT INICTION MODE OF RIR SYSTEM a.
Drywell Pressure - High 12 psig 12 psig w
b.
Reactor Vessel Water Level - Low Low Low 2.-146.5 inches
- 1-146.5 inches
- S c.
Reactor Vessel Shroud Level - High ~
__-203.5 inches
- 2-203.5 inches
- l y
d.
Reactor Steam Dome Pressure-Low 1500 psig 1500 psig u
e.'
Reactor Steam Dome Pressure-Low 1335 psig
$_335 psig f.
RHR Pug Start - Time Delay Relay
- 1) Pugs A, B and D 10 + 1 seconds 10 + 1 seconds
- 2) Pump C 0.5 + 0.5r seconds 0.5 + 0.5 seconds
~
~
g.
Logic Power Monitor NA NA
~
~
'See Bases Figure B 3/4 3-1.
f S
l
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INSTRI.NENTATION BASES MONITORING INSTRUMENTATION (Continued)
FIRE DETECTION INSTRt.NENTATION (Continued)
In the event that a portion of the fire detection instrumentation is inoperable, increasing the frequency of fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY.
3/4.3.7 TURBIfE OVERSPEED PROTECTION SYSTEM This specification is 'provided to ensure that the turbine overspeed prctection system instrumentation and the turbine speed control valves are OPERABLE and will protect the turbine from excessive overspeed.
Protection from turbine excessive overspeed is required since excessive overspeed of the turbine could generate potentially damaging missiles which could impact and damage safety-related components, equipment or structures.
3/4.3.8 DL"lRADED STATION VOLTAGE PROTECTION INSTRtJENTATION The undervoltate relays shall autorutically -initite the disconnection of offsite power sources whenever the voltage setpoint and time delay limits have been exceeded. This action shall provide voltage protection for the emergency power systems by preventing sustained degraded voltage conditions due to the offsite power source and interaction between the offsite and onsite emergency power systems.
The undervoltage relays have a time delay characteristic that
,i provides protection against both a loss of voltage and degraded voltage condition and thus minimizes the effect,of short duration disturbances without exceeding the maximum time delay, including margin, that is assumed in the FSAR accioent analyses.
HATCH-UNIT 2 B 3/4 3-5 Amendment No. 27
-ew
NOTE: SCAL 2 II INCM S ABOVE VESSEL 2ERC WATER 12WL NOENCIATURE EXCHT A80kT 800 --
NO.
WSSEL 2130 READING 1mRM (INCHES)
(8) 575
+50 YARWAY (7) 559
+42 CE /MAC 750 --
(4) 549
+32 m /MAC (3) 529.5
+12.5 3ARTON IIIk (2) 479 38 YARWAY 723.06 N "GE (1) 370.5 146.5 YARWAY g
yon __
(0) 313.5
-203.5 YARWAY D
~~
- MAIN S1EAM
-- 640 ggyg 600 --
YARWAY YARWAY m/IMC BAN
" 577 0- -
+60 - =
',75 (8)
+60- -
+60-r 58
- 559 (7)
(8)
HFCI & (7)-42 Hi A1 ARM 550 ss549 (4)
- $29.5 (3)
TRIIS II
~
TOR BOTTOM OF STEAM 517 INSTRLMNT 0-0 0
0== = SCRAM YAPWAY DRYER SKIRT ZELO 500-- -
TEED---4 83.5
. 479 (2)
- -38 (2)
WATER CORE N
INITIATE RCIC
- 405 SFRAY
/
HPCI, TRIP RECIRC.
IUMF8' CIDSE MSIV'S 45 5 =
400- -
-146.5 (1)
- 370.5 (1)
- 367
- 15 0- - INITIATE BHR, C.S.,
350~- 352.54 START WEL AG
~
CONTRIBUIE TO A.D.S.
2/3 CORE
-203.5 HEIGHT -
313.5 (0)
PERMISS 7VE 300 - -
ACTIVE TLTL 250 --
317 - -
200 -C IC.
g j
-176.56-DISCHARCE REC 11C.
N0ZZIE SUCTION - 161.5 N0ZZ1Z 150 - -
g 100 - -
50 --
BASES FICURE B 3/4 3 1 REACTOR VESSEL WATER IIVEL HATCH - UNIT 2 B 3/4 3-6 Amendment No. 30
NOTE: SCAM IN INCHES AROVE VESSEL ZEFC gATER 2 VEL NCENCIATURE E XCHT ABOVE 90.
VESSEL EERO READINC INSTRLM N1 800 -.
(INCHES)
(8) 575
+5" YARWAY (7) 559
+42 CE/nat 750 - -
(4) 549
+32 CE /MAC (3) 529.5 412.5 3ARTop 723.06--h8M (2) 479 38 YARWAY r (1) 370.5 146.5 TARWAY (0) 313.5 203.5 YARWAY 700
~
.- 640 - MAIN STEAM LIE YARWAy YARWAy 2/NAC BARTON 600 - -
$77
+60- -
+60 - -
" 575 (8)
+60.-
+60 = 5 8
- 559 (7)
(8)
HFCI 6 (73-42 M1 A1 ARM 550 -sn549 (4)
RCIC (4) -3210 ALARM
- 529.5 (3) 7Rtys (3) - 12.5 0 ---
REACTOR BOTTOM OT STEAM 337 _ INSTRtM NT 0-O e
SCRAM YARWAY DRYER SKIRT ZERO Soo-- -
17 - -
TEED 183.5
. 479 (2)
- -38 (2)
WATER CORE gygygg7g gcgg
- 465 syRAy
/
EPCI. TRIP RECIRC.
FtwS, CloSE HS1v S 3e_ _
400- -
146.5 (1)
- }70.5 (1)
J67
- 150- - INITIATE RHR, C.S.,
352.56 START DIESEL AND 350 -=
CONTR18L7E TO A.D.S.
2/3 CORE 203.5 HElchy 313.5 (0)
FERMISSIVE 300 - -
ACTIVE FttL 250- -
= 03.5 % C.
g
- 317 - -
200 j
-176.56-DISCHARCE RECIRC.
N0122 Stu!ON - 161.5 N0Z22 150 100 --
50 --
BASES TICL'RE l 3/4 3-1 REACTOR VESSEL WATER MVlt HATCH - UNIT 2 B 3/4 3-6 Acendment No. 30
- - - - -