Revision as of 22:58, 31 January 2020

Text

Emergency Procedures for Facility,Including General Abnormal Procedures,General Abnormal Index & Operating Abnormal Index
	ML19326E034
Person / Time
Site:	LaSalle
Issue date:	05/22/1979
From:	; COMMONWEALTH EDISON CO.
To:	;
Shared Package
ML19326E033	List: ML19326E032; ML19326E034;
References
	PROC-790522, NUDOCS 8007250455
	Download: ML19326E034 (82)
	v • d • e

(~ ,9 (b) 9 U ff y 5/22/79 LGA INDEX O ,9 '

PAGE

'O,)

PROC. NO. TITLE $ REV. REV DATE DISKETT

.GA 000-01 ci (Ol%8_.01 LOSS OF COOLANT (FAST LEAK) 00 .3-/-74 o1 i.3,l'f76 L -

.GA 000-02 LOSS OF COOLANT (SLOW LEAK) -3 / 77- 01

.GA 000-03 MAJOR STEAM LEAKS (OUTSIDE THE OC 3/79 01 ORYWELL)

.GA 000-04 HIGH REACTOR WATER LEVEL 00 3/79 01 OI //ED

.GA 000-05 LOW REACTOR WATER LEVEL 44- -W 01

.GA 000-06 REACTOR HIGH PRESSURE 00 3/79 01

- Of G IT M L-

_GA 000-07 INADVERTANT REACTIVITY A DD I T,I ON 4M)- +fW 01

&v os ia I9 % L.

.GA 000-08 LOSS OF RECIRCULATION F L O W ++E PUMP fL 4)C* W 01

_GA 000-09 LOSS OF RECIRCULATION FLOW-BOTH LOOPS 00 3/79 01 ol /s /'74dC-.

_GA 000-10 LOSS OF TURBINE GENERATOR LOAD ativ 01 GREATER THAN 25%

GA 000-11 LOSS OF TURBINE GENERATOR LOAD LESS 00 3/79 01 THAN 25%

GA 000-12 LOSS OF AUXILIARY ELECTRICAL POWER 00 ,3/79 01

.GA 000-13 AREA HIGH RADIATION 00 0/00 01

.GA 000-14 HIGH AIRBORNE ACTIVITY 00 ,

4/79 01 GA 000-15 RADIOACTIVE SURFACE CONTAMINATION 00 3/79 01

_GA 000-16 FUEL ELEMENT FAILURE 00 3/79 01

.GA 000-17 BOMB THREAT RESPONSE _GG. tf 01 OI & h9 SL.

GA cob-18 Transien+ U)ith Fedlure To sercem 0h 10 h 9 3 ._.

9 8007250US .

l - 7a'.750G0" f'U!Y-

_,.5 3

nj"i""

( ,,'...s T'qcec (cCUOCTuS M "

' ~ " ; ,7* . u 'w=p.y ri:f,d m

~t.n'f'.

,,b un

~

];gy a rc IL.

Cur [Sil.

i cs.p!; n'. OiM j.

.p"si.-I D ;j - " [q s p,,]; a ..- p 4 ,b. ,v':a ' ,i p g

g ,,gt

. . . ...r rey;510nd af oo (**

1 Revision 2 Juir 11. 1980 1

fi!0S iOI USB 50 Onst* o: tc.atn,. c,9jgajge F

1 LOSS OF CO3LANT (FAST LEAK) i

[ A. SYMPT 3MS i

l 1. Control 'toam Panel 1113 0631 (2H13-P501) Alarms:

, o

o. E -t R ORYdELL CH 3/D PRESSURE nl.

.; 3. RnR R E A L' T O R CH S/D LEVEL LO.

2.

j- c. RHR SYTEM 1 A TJATED.

j 3. RHR SYSTEM 2 ACTUATED.

e. HPCS SYSTEM ACTdATES.

I

f. LPCS SYSTEM A*.TJATEJ.

f

. LEAA 3ETECTIO'd 3RI CO*4T TEMP HI. .,

n. ORYdELL PRESSJRE HI. *

i. REACTJR *ATEQ LEVEL L LO.

n J. REACTJR WATER LEVEL 2 LO.

I

<. ' ADS LOGIC A ACTJATE]/3 ACTUATED /C ACTJATE0/0 ACTUATED. j a

1. REACTOR VESSEL LOW 4ATER LEVEL 3 CONFIRMEJ. j l 2. Control Room Panel 1413-P633 (2H13-P603) Alarms:

a. DIV 1 RX VESSEL HATER LEVEL 2 LO.

i a. DIV 2 RX VESSEL WATER LEVEL 2 LJ.

s

c. CH A DRIM CONT PRESS HI.

I c.' CH 8 PRI CONT PRESS HI.

e. CH A RX VESSEL LO LEVEL 3.

> t. CH 6 RX VESSEL LO LEVEL 3.

' 3. Co7 trol Roorr. P anel IPM13) (2PM13J) Alarms:

4 %

i 9 g e?- r

' *- - -- - s---er! -

3 y9 r-e ,yy g- -y-, - - .

9y- . - - g q p ~r p - ?'g-, y-=n=^y

?--ne--?-t*"- '*D'*-

d LGA-01 Revision 2 7 July 11. 1930 I -

2

a. ORY'aELL AIR HJMIDITY HI,

c. ORYwELL EQUIPMENT CRAIN SJMP LEVEL HI-HI.

c. ORYWELL E2UIPMENT ORAIN SJMP TEMPERATURE HI.

j a. FLOa TO ORYaELL FLODR ORAIN SUMP LEVEL MI.

.e. ORYdELL FLCJP DRAIN SJMP LEVEL HI-HI.

1

f. ORYWELL SUMP FILL-UP RATE.

g. ORYWELL SUPPRESSION POOL CHAMBER TEMPERATJRE HI.

4 Co7 trol Moom Panel IPMJ65 (2DM245) Containment Mimic Incicates Isolation.

B. AJTJMATIC ACTIONS

1. Level 3, Reactor. Vessel dater Level 12.5".

a. Reactor SCRAM.

o. TRIP Reactor Retirculation Pumps to 15 Hz.

c. ADS Parmissive.

a. ISOLATE Reactor vessel anc Containment Groups 4, 6. anc 7.

[ e. Automatic Programmea Vessel Level DOWNSHIFT to f

13".

a i 2. Level 2. Reactor Vessel dater Level - 50".

a. AUT0 START RCIC.

.o. AJTOSTART HPCS and HPCS Diesel Generator 18(23).

c. TRIP Reactor Recirculation Pumps to CFF.

3. ISOLATE Reactor Vessel and Containment Groups

.1, 2, 3, anc 5.

~3. Level 1, Rsactor Vessel dater Level - 129".

l

LOA-01 Revision 2 July 11. 1980 3

3. AUTOSTART RHR LPCI Mode.

o. AJTOSTART LPCS.

c. AUT0 START Diesel Senerators 0 anc 1A(2A).

a. ADS Permissive.

4 1 69 Containment dign Pressure.

a. Reactor SCRAM.

. o. ISCLATE Reactor Vessel and Containment Groups l 2, 4, 6, and 7

+ .. ISOLATE Reactor Vessel and Containment 3roup 9 if 57: Reactor Pressure also present.

C* 1EEEid[f.8_f.M[E1 3.E[12Isi

1. VERIFY if reached:

a. Level 3 Actions.

c. Level 2 Actions.

c. Level 1 Actions.

d. 1.o90 Containment Actions.

2. MAINTAIN Feedwater Flow if possiola.

3. MONI TOR:

3 Vessel Level sna Pressure.

O. Containment Pressare, Temoerature. Hydrogan Concentration. anc Raaiation.

4 As necessary,

a. N3TIFY and-EVACJATE Personnel.

i o. RESTRICT Access.

4

c. 'N3TIFY Shift Engineer or GSEP Station Director to classify tne event anc Initiate GSEP if i- required.

I

+a;nsseJe u6! sec suewu!esuca A;en!;d 6u!ceeoxa auana;d os Aeads lueuu!eauc3 31v111NI *2s!xe Aen used ssecAs a6exea7 auewu!esuo3 e +6 ! sd 92 scaeaxa a;nssa;d Jeowsu uc!sse;cdns ao 6 ! sd og spea:xe aanssa;g [tamA;c JI o.

=apew 6u! toc 3 tood uo!ssaJdCns au2 up pase; ado ac peau wcssAsonS bHM alaeJedc (1) euc A tuo av307 aua 6u!;no asc s! Janoc a2!s-jjo 3I NOI1rv

(S/v960d-2132-OW) S/vP40d-2131

- h CNIsC73 Ao Je6ueuax3 2 Esp kHb 9/r eu2 c2 mc Ld g/" 13d7 133EIO *panen2:e ao 2ocueo we2sAs Su!tco; t ce uc!sse;ddns au2 JI *casol3 s! (9/v29Cd2132) ;

s/V WGd-2131 e^ tea uo!2caful g/v 13d7 usur pa;aeo: '

su!et 9; e;oo au2 J! Ae;ds auswu!e2 cog ove El-Hb-dC7 42!* e.uec;oose u! Eu!too; tocd uc!sse;ddrs givIIINI g

+2uawu!e2uo3 pue e;co sus Jc Su! goop w;e2 Suoi JCJ uc!2eaeca;d u! +50-Pb-dC7 42!= ecuec;co:e u!

s;e6cecov3 2eas eHy eus os peonoc;2u! ec ogneus ; e s e.= ,

so!^;as asy *v307 aus Je23e sesnu!w OT A tatew!xc;cdv +9

C2

eeb aus WCJJ Janew Jo uC!s;aA!o pue uc!2e?!;nsse;ceG lleMAJG P!Ce; 2uebe;d C2 Jeo;o u! a3usJn 3C V3C7 au2 JC saanu!u OI u!u2!= pa;eA!2oe ec 2cu ptncus Ae;oS suewu!cquo; e.O lif E3

'vo!2e;2ue3eo .

ve6CJDdb Due e;nse;eCual *s;nsse;d queWu!e2Lo; E01INOW *uo!2r!2!ul ;c2!ucw v307 2 sed A g I t s ,a at

oeuteau!en e;e uo!2ez!;nssa;deo lessaA Jo/pue 6u!!oc3 a;o; ,

seu2 a;nsse c2 Bauew;oJ;ec s;33 bolinow c2 enu!2uo; +2

+;amoe ;osoest EC11NCW on anu!2uoD pue s;o23e2e0 WPS pue WBI lb35NI *T SNC113v N01vbsca 1h3005505

C

etowes oue AaA;ns c2 wau;/Deh AdI1CN 'C 9

0961 *11 A lnr 2 uo! stead [

10-v0' ,

)

e 0

9

+ se e . 4<.mm

-< q .,%- ,

. .. . . . . . . ~ .. . . - - - . ..

l j

LGA-01 l Revision 2 July 11, 1980 5

CAJTION If Containment Soray is not availaole, and a Sypass Leawage Patn exists, DEPRESSURIZE the Reactor of JPENING Safety-Relief valves.

7 CLJSE 40-lS21-FCoSA anc a (MO-2321-F065A anc 3) anc MO-lG33-F040 (MO-2333-F040) Feeceater anc Reactor ,

water Cleanup vessel Return Valves, if tnese systems are unavailaole or unnecessary. >

J r

CAJTION 90-1321-F0o5A and B (M3-2521-F365A ana B) anc MO-IG33-F040 (MD-2G33-F040) Feedwater anc Reactor weter i

Clea1uo Vessel RetJrn Valves s noe l

oe Closeo witnin 20 mins. of tne initiation of the .OCA if a total

, loss of Feedsater Flow occurs curing this tine.

l S. START the SSIV Lea < age Control Systen in accorcance '

eitn LOP-MS-02 if the-following conditions are met:

f i a. Twenty (20) minutes elapsed since start of LOCA.

i

o. Main Steam Line Pressure i s Del ow Reactor Pressure.

3

c. Main Steam Liae Pressure is less tnan Dry-ell Pressure.

9. START the Hydrogen D.ecomoiner System in accoroance with L3P-HG-31, LOD-HG-02, LOP-HG-03, or LOP-HG-09 orior to four (*) nours after start of LOCA anc prior to Prinary Containnent Hyorogen Concentration exceeding 3 3 percent oy volume. i l CAJT104 Syoassed steam resal ting f rom operation of tne l Hydr. ogen Recomoiner System could lead to nign !

suopressio,;chamoer pressures. A portion of toe RHR System flon snoulo oe oiverted to the Sa'.oression .

Onamoer spray headers to quencn the oypasses team if tais action will not uncover the core. I i

1 l

t l l

e I

a

9 b

LSA-01 Revision 2 Jul y 11. 1980 lfinah

10. SHED unncessary loacs from tne Diesel Generators.

11. CHEC< Laxe Screen House Traveling Screens for clockage. If all screens are olocxec, OPEN JE12-F330, Screen Bypass Valve.

12. MANUALLY S A Ci n' A S H Jidsel Generator Cooling natar Strainers if nign oifferential pressure occurs.

13. MONITOR Stack Gas Activity to cetermine GSEP catagory.

INITIATE GSEP.

E. D I S*;U S S I O N Tnis crocedure assumes the ruoture of a Reactor connectec systen and a resultant loss of Reactor Coolant. It is furtner assumed tnat tne leak occurs inside tne Dry-ell and may De eitner steam or eater. The operator must use all available information to icentify tne source ana locatio1 of tne l eak. The tencerature and oressure transients resulting from tnis accicent maintain claccing integrity. Tnus, activity r el ea s ed to tne Orywell is a result of intrinsic coola7t activity ano activity r el easeo as a co7 sequence of Reactor scram and depressurization.

Suosequent to tne LOCA, fission proouct cecay heat will result in a continuing energy dump to tne Suppression Pool. Jnless this energy is removed from tne Containment, it will result in unnacceptable Suporession Pool Temperature. After ten (10) ninutes an RHR Heat Excnanger must ce acti vatea wi tn cool ing water flow.

Due to the uninerted 3r ywell , special attantion must ce g i ven to . noni tor i ng Contai nme nt Hydrogen Concentration to arevent exceecing a four (4) percent volume j concentration.

' As a fi7al last resort metnoa of acaing water to tne ,

reactor, cacaoility exists to aco laxe water to the i Reactor via 8 RHR Injection Line.

.l

. e e s

. . _ . _ . . . _ . . . _ . _ - . _ _ .. -..m.. _ _ .. . _ __ . .

=b,

.~* ,

e. . r~

{ ~ i . * ~ - ~ U" C f T C II Y.

. . " - t. -.-, , , ,_ ,

s

.c

, . . .,- r .

- T c,* e , , , ,r, . .r:- , ' , ' . ' ' . .

.v - 'i ........*m' : .:.n t.. . . -) --- '

. 1 ,,g.

'. L3A-02 4

4, . , . - '." -e * ; ,' . ..

,. .'s Revision 2 tv..<-"

a

+*',U,f...'..C.----'*

Julv 11. 19d0

, ,: . ,,; tw t v. s ; -

.- 1 LOSS OF CO3LANT (SLOW LEAK)

A. SYMPT 3MS

1. If leax is witnin the o rimary Containment:

a. Control Roon Panel IPM13J (2PM13J) Alarms:

1) ORYWELL AIR HUMIDITY MI.

2 )- ORYWELL SUP 300L CHAMBER TEMP HI.

, 3) ORYWELL SUJPRESSIO.1 CHAMSER RADIDACT MI.

f 4) ORYhELL ECDT DRN SUMP TEMP HI.

5) ORYWELL LJOR JRN SU4P LEVEL HI-HI.

6) 3RYWELL EQUIP OR1 SUMP LEVEL HI-HI.

7) ORYWELL SUMP FILL-UP RATE HI-LO.

8) ORYWELL ECdIP JRN SUMP FILL-UP RATE HI-

! LO.

9) FLOW T3 JRYWELL FLOCR ORN SUMP LEVEL HI. ,

1 3 Control Room Panel IP913J (2P913J) Incications:

1) Orywell Equioment Drain Sump Lea < age Volume

, and Punp Discharge and Flow Recorder increasing.

i

2) Drywell Floor Orain Sumo Fill up Rate and Pung Discharge Flo Recorcer Increasing.

3) Drywell ;ontainment 1 arrow Range Pressure 4

Recorcer Increasing.

4) Drywell Air Particulate, Noule Gas and

{ Ioc i ne R ec o r .1. e r Increasing.

5) ~ 0rywell ~ooler Concensate Flow 1F1-RF001 4

(2FI-RF001) Increasing aoove 2 gam.

5

6) Drywell Equiament Drain Suno Temperature

! Incicator Increasing.

l.

l

LOA-02 Revision 2 July 11, 1960 l' 2

7) Orywell Relative humicity Indicator Increasing.

i :. Control Roon Panel 1H13-PdOL (2H13-P631) A l a t -n s :

1) LEAK OETECTIJN PRIMARY CJNT TEMP MI.

f d. Other Symptoms:

, 1) Hign Drydell Cooler Jifferential Cooling Aater Tenperature.

[ 2) High Drydell Cooler Differential Air

Temperature.

2. If leakage exists outsica tne Primary Contaiament in tne Reactor Builcing:

a. Control Roon Danel LPM 13J (2PM13J) Alarns:

1) 4X BLOS ECD JRAIN TK TEMP HI.

i

2) Rx BLDS EOP 3RN TANK FILL-UP RATE Hl.

3. Control Room Janel LPM 13J (2PM13J) Indications:

1) Reactor 3uilding Equipment Orain Tan <

, Level Recorder Increasing.

2) Reactor Builaing Equio.nent Drain Tan < j Tenperstare Indicator Increasing.

4

c. Control Roon Panel LH13P631 (2H13P601) Alarns: !

{ 1) REACTOR BL3G QA01ATION HI.

2) RHR Punp B/C CUBICLE TEMP HI.

3) -RHR PUMP A CJBICLE TEMP HI.

i 1

-4) - LPCS/RCIC PUMP CUBICLE TEMP HI. '

! 5) RCIC PIPE RTE EQUIP AREA TEMP HI.

6) LEAK DETECTION RX WATER CLNUP AM8 TEMP l HI.

7) LEAK DETECTIJN RX WATER CLNUP OIFF. TEMP HI.

'M 4*M"M; W*M4ruiAMb *t 64 WM6M@ 7 9- $->h bNu+rW

t/ - -=

4 A

LGA-02 4

Revision 2 July 11. 1990 3 ;

8) LEAK OETECTIJN RdC R3095 AM8 TEMP HI.

04 LEAK-DETECTION Rsc ROOMS DIFF TEMP HI.

10) LEAK DETECTIJN STM PIPE TUNNEL AMb TEMP HI.

11) LEAK DETECTIJN STM PIPE TUNNEL DIFF TE4P HI.

5

12) RHR EOJID AREA OIFF T E *iP OR AMB TEMP Hl. ,

l

13) LD PIPE TUNNEL MS LINE AREA TEMP HI.

c. Other Symptoms

4 4

1) Excess Flo. Oneck Valve position (PM10J).

1

2) Area Raciation Monitor Alarms.

2

3) Increasing Incication on the following Meters:

a) Stack Gas Radiation Increase.

o) Reactor Building Exnaust olenum Radiation. J c) R3CCW Radiation Monitor.

i d) RHR Cooling Water Effluent Raciation.

4 Tna. f oli o.ing sy npto ns inaicate an-Instrumen*, Line draa<:

a. Instrument readings among instruments monitoring the same paraneters differs or is erratic.

7

c. Annunciation of alarms associateo wi tn a oroxen Instrument Li1e.

c. Poss i bl e hal f scram associated with an RPS Instrument Li7e.

i- .

l

5. Standby Gas. Treatment Auto Starts.

1 4

- .~.. . . . - . , .....-4.J... . . . . , . . -

t. , - , , - - - , . . . . - - , - ,. . ,.

LGA-02 Revision ?

July 11, 1980 4

d. AJT3MATIC ACTIONS

1. oossiole Reactor Vessel ana Containment IS3LATIO95 from Leak Jetectio, or Process Raciation Instrtmentation.

a. Group 1 (Main Steam).

a. Group 4 (Reactor 313g. Ventilation, SsGTS INITIATION).

c. Group 5 (Reactor .v a t e r Cleanup).

c. Group 6 (Resiaual Heat Renoval).

a. Group 8 (Reactor Core Isolation Cooling).

C. IMMEDIATE OPERATOR ACTIONS

1. ATTEMP T to l ocate l eak. EVALUATE effects of isolating leak, and ISQLATE leak if crudent, j 2. As necessary:

a. NOTIFY and EV ACJATE Per sonnel .

o. RESTRICT access,

c. NOTIFY Snift Enginear or GSEP Station Director-to cl assi f y tne event anc Initiate GSEP if raquirea.

d. N3TIFY Rad / Chem to survey onc sample.

D. SJS EDUENT JPERAT3R ACTIONS

1. REVIEW Tecnnical Saecification Requirements.

a. If tne unisol ataole l eak is part of tne Reactor Pressure Boundary in the Reactor Building:

1) INITIATE a. SCRAM anc COMMENCE a Normal Reactor Coolco.n so tnat tne Reactor is in Col Snutco4n witnin six (6) nours.

2) STARTUP Stanaby Gas Treatment anc ISOLATE Reactor Suilcing Ventilation.

LGA-C2 Revision 2 Jul y 11, 1930 5

o. If tne uni sol atable lean is insice the Primary Containment anc greater tnan 5 gpm unicentifiec, ,

or 25 gpm total leakage averagea over.any twenty-four (24) nour perica, or i s ceter ni ned to De Reactor Pressure Souncary leakage:

1) CCMMENCE a Nor nal Reactor Snutacan anc

, CoolJo n so tnat tne Reactor is in Hot Snutdo.n witnin t.elve (12) nours, anc l Cola Snutco.n witnin ene next twenty-four i- (24) hours (T.S.).

2) START Suopression Pool Cooling in accordance with LJP-RM-13 if Suopression Pool i Tenperature aparoaches 100'd (T.S.).

2. RE/IEW cne follo.ing procecures:

a. LSA-01; Loss of Coolant (Fast Leak).

o. LGA-13; Hign Area Raciation.

c. LGA-14; Hign Airborne Activity.

3. LGA-15; Radiation Surface Contamination.

, 3. RESTORE any Out-of-Service Emergency Core Cooling Systems to service.

4 0045 ULT aith the Rad /Cnen Supervi sor and INCREASE

the frequency of routine station anc environmental

nonitoring if necessary. DBTAIN a sample of Reactor-Coolant for anal y si s.

5. MO'41TG4 the leak r ate :l osel y. If P r i ma r y C onta i n:nent Pressure aaproacnes tne Scram setpoint or if tne leak encangers personnel or equipment, MANJALLY SCRAM the Reactor and PERFJRM a controllea coolcown.

E. DISCUSSION Tnis orocacure is written to specify actions in tne event of a Small Leak in the Reactor or connectec systems wnich coes not result in immediate Primary Containment ano Reactor Vessel Isolation.

A slo. leak will De ooservec as a result of.nign i .temoerature, nunidity, or racioactivity in the, vicinity of the leak ratner than cnanges in plant process variaoles, t

. ...--..:., . ..... . ...-e. .

9 . __ - -.

.y 4 ,--6. g

LGA-02 Revision 2 JJIy 11, 1960 6 (final) sucn 3s steam flon, feed flaw, reactor pressure, or reactor level.

Since astomatic protective action coes not occar i nmec i a te l y , the cecision mast ce maae to continua ooeracian, olace the ol ant in a more reliaole moce or to perform a plant saatco-n and coolcoon. Variaoles tnat mJst ce consicere] inclace tne assu.nption tnat a leak cue to a small crack can aropagate into a Fast Leak, 35 well as tne raciological consequences of the escaae of Ene Coolant.

Due to tne various leak aetection systems associatec nith Reactor connectec systems. Isolation of tnese systens may o:CJr as an automatic action. Thus, the operator must cetermine the effect on plant operation of tne loss of tne isolatec systen.

Any lea < insice tne Prinary Containnent tnat cannot ce treatec as a Slow Lea 4 nust De treated operationally as a Fast Leak.

s

=

r 'e

'fo y T.f,"' POSE only.

TeVjgjgASBle r '?OU?E15 LSA-cs a

Cur & ~ ens'on 1 files fOr use inID 0 in "n'

.IIf8! i!I:

T OE er""I ""'Y L

' t'8 00' OT Idaint c ...

MAJOR STEAM LEAKS ( C J T .>

CtIV;'kyN THE aELL)

ORY0I n $9llltg A. S Y ad P T 3 M S

1. Co, trol Room Panal 1813-0631 (2H13-Po01) Alarms:

a. LD PIPE TUNNEL MS LINE AREA TEMP MI.

o. M A I '4 STEAM LINE FLO4 HI.

2. Control Room Panel 1413-7533 (2H13-Po02) Alarms:

a. M A I '4 STM IS3L VLV NOT F'JLL JPEN.

a. CHAN A MSIV NOT FJLL JPEN.

c. CriAN 3 MSIV NJT FJLL JPEN.

d. .v N STEAM LINE F L C,4 HI.

3. Jt,er Syrotoms:

3. Hign Reactor dater Level.

o. Increase in steam flow, feec ater flo anc reactor po-er wit, turoine load unchanged.

c. Area Temperstare anc Radiation '4cnitors reaaing high.

c. Noise and aater accumulation.

3. AJT3MATIC A;TroNS

1. ISJLATE Reactor Vessel and Containment.

a. Group 1.

2. Indirect Reactor 5; RAM.

C. IMMEDIATE OPERATOR ACTI3NS

1. VERIFY:

a. SCRAM, rods i r. , poner decaying.

.s__ - . - . . . . _ . _ . _ . . : .. .. .

LSA-03 Revision 1 July lle 1960 2

f 3. Groap'l ISOLATIJN.

2.- As necessary,

s. NOTIFY and EV ACJATE Der sonnel . '

, 3. RESTRICT access.

c. NOTIFY Shift Engineer or 3SEP Station Direc or

, to classify tne event and Initiate GSEP it raquired.

3. NOTIFY Rad / Cham to survey and Samole.

O. SJ8SEJUENT 3PERATJR ACTIONS

l. CARRY OUT procedura LGP 3-2 using tne Motor Jriven Reactor Feed Puma to maintain Reactor dater Level.

1

2. CONTINJE to control Reactor Pressure and Water L ev el .

j 3. VERIFY Reactor Mode Saatch is in SHUTDOWN.

4. MAINTAIN Condenser Vacuun if possiole.

5. Afcer tne Main Steam Lines nave depressurized, LJCATE and ISOLATE the leak i f poss iol e.

1

! 6. If tne leak is isolated and Condenser vacuum is greater'than 7" Hg:

a. RESET tne Isolation Signal.

o. OPEN the Main Steam Isolation Valves in accordance

.itn LCP-MS-01.

I

. PERFORM a Nornal Reactor Cooldown in accordance l witn LGP 2-1; Nornal Unit Snutdown.

7. If all of.the conditions in step O.6 cannot ce met, i

PERFJRM a Normal Reactor Cooldown in accordance witn LOP-RH-09; Steam Condensing Startup ana Operation.

l i f6. REVIEW tne f ol lowi ng procedures:

a. LGA-13; Hign Area Radiation. I d

. - . - ,. .m. , - , . , . . . . . - . - . - - , e --.---n-m. .,~7 ,.,,er-

A t

l L54-03

. Revision 1 July 11, 1980 t- 3 (finalj

o. LGA-14; Hign AirDorne Activity.

c. LGA-15; Racioactive Surface Contamination.

9. CHEC( Stack Release Rate.

- E. DIS;usSION 1

A Main Stean Line failure of sufficient severity will

' ini ti ate closure of tne Main Steam Isol ation Val ves.

j Since initial automatic protection is initiatec.

verification of automatic actio1 ensures tnat MSIV I sol at i on occurs, the-Reactar Scrams, Reactor Water Level

f. is maintained anc Reactor Pressure renains under control.

i Aftar monitoring 4eactor orocess parameters, action is

, oirected towarc ensuring adecaate snutaown cooling via

, tne Fee 0 water and Conce1 sate Systems or tne Steam Condensing anc Snutcown Cooling Modes of RHR.

i i

Since l ar ge a.moants of energy and radioactivity may nave .

Deen releasec, the extent of equipment damage snoul0 oe ascertainec as soon as oossiole.

I i

t

.}

4 4

4 i

i s

T

a 4

% %' = -g -- *

y-L.- -.

v y.-p .y, ,,w,y _

.m, c _ ,_ , ,,, .,m..% ,- a,.~. y .

lliese &:uments in ~.;c.ining purp4::s only.

They rre nat cr.r:tirJ!i T5c are not authorized ICT P!CM C,.1,ctiM (:i TcDint;,:'.l;6',U50. Current reVIS!ODS 2r8 L'!ilil.ii2 in C6r.;ra! Fi!3 of 8210!!!t0 LGA-o,

't

Revision 1 l'105 1 iOr USB in Oper. or ra~i*nt.

s

. activmes. auiy 11, tuc

1 f

HIGH REACT 3R WATER LEVEL A. SYMDTJVS

1. Control Room Panal 1H13-Pe03(2H13-De03) Alarms:

3. Fa CONTROL 4X WATER LEVEL 7 HI.

c. FW CONTROL RX aATER LEVEL 8 TRIP.

c. Possiole FW VALVE C3NTROL SIGNAL FAILURE.

d. Possiale TURBINE RFP CONTROL SIGNAL FAILURE.

e. Possible RX AJT3 SCRAM.

f. Possiale RX VESSEL PRESSURE HI.

g. TSV NOT FULL JPEN.

n. TURS ST3P VLV CLCSURE TRIP.

2. Control Room Panel 1H13-P603(2913-P603) Incications:

a. Reactor Water Level Instrumentation Level Increasing.

o. Turbi ne Dri ven -or Motor Driven Reactor Puno increasing flow or Feecwater Heacer increasing fIow ditn constant $ team F1ow.

3. Control Roon Panel IPM33J(2PM03J) Alarmi:

a. Rx FEED WATER PUMP TRIP.

o. ' Possiale CNDS.H]T.4 ELL LEVEL HI-LO.

4 Control Room Panel 1PMC2J(2PM02J) Alarms:

a. TURS MN STEAM BYPASS VALVE NOT CLOSED. -l l

o. Possi01e FW PJMPS TURB VIBRATION HI.

c. Possiole TURBINE TRIP VIBRATIJN HI.

l

LGA-04 Revision 1 Jul y 11. 1980 2

4

5. Oo7 trol Room Panel 1H13-P601(2H13-P601) A l a r .n s :

a. REACT 3R' VESSEL .4ATER LEVEL d HI.

c. Possiole RCIC TJRCINE TRIP.

6 Otner Incications:

, a. Increasing Level InJication on POST LOCA i- Recoraer.

D. Loss of Main Turoine Generator Output.

i B. AJT3MATIC A C T I O *J S i

I i

1. Level 8, Reactor Vessel dater Level 59.5a.

3. TRIP Main Taroine.

I o. TRIP RCIC Turoine.

c. TRIP Reactor Feec Pumos.

c. CLOSE HPCS Injection Valve.

2 .- Reactor Scram from Turoine Stop Valve Closure aoove Main Turoine First Stage Pressure corresponcin.; to t-30s Reactor Power.

j- 3. Re:irculation Pung Trio (RPT) from Turoine Stoo Valve Closure aoove Main Turoine First Stage Pressure corresponcing to 33% Reactor Power.

4 Dossiole Lackout of the TurDine 3 riven Reactor Feed j Punps or Feedwater Regalation valve (FRV). l l

C. IwME01 ATE CDERATOR ACTIJNS

1. If tne Automatic Feedwater Control System nas nalfunctionec or A or B Jut of Service Light is illuminated:

a. VERIFY Manual / Auto si gnal s are Dalanceo.

3. DLACE Manual / Auto Transfer Station in MANJAL.

d 1

L3A-04 Revision 1 July 11. 1930 3

c. RESET Locxout, if one exists.

a. Manuelly C O *4 T R O L 2eactor Level.

7

e. If TO4FP Lockout cannot De clearec, CONTRJL T3RFP with Motor Sceed Chonger end Jack.

2. If Level S is reacnea, VERIFY Actions. Also,

, - c,

a. If Reactor Poner is greater tnan 300:

1) SC R Ali, roos in, coner decaying.

, 2) RPT to 15 42.

3) Sypass anc Safety / Relief valves control

pressure.

3. As level decreases.

1) Place Feacaater Reg. M/A Station to wanual and CLJSE valve.

2) START "DRFD.

3) CONTROL Leve' using Feedwater Reg. V31ve.

4) START RCIC if necessary to maintain level aoove Level 4.

D. SJBSE205NT JPERATOR ACTIONS

1. If control of Re3ctor water Level has oeen re-

estaclisneo prior. to automatic protective action

a. STATI3N an additional operator in tne Control Room to monitor anc control Reactor Vessel Water Level. He snould De assigned no otner duties until the Feecwater System is returned to normal.

o. M3 NIT 3R the Reactor nater Level and Toroine Supervisory Instrumentation. N3TIFY the $nift Supervisor of any aonormal indications.

c. CETERMINE tne.saecific caase for tne nign Reactor Water Level ana INITIATE appropriate corrective action as requirec.

't - o , -- p- -- ,--

-,vr.-- -,

w

I. -

7 LGA-04 Revision 1 July 11, 1980 4

2. Suosequent Action if Reactor Water Level 8 was exceecea:

a. CONTINUE to control Reactor water Level anc Pressore.

1) OPERATE Syoass Valves as necessary if tne Main Consenser is availaole. If t n e . vain a Steam . Isol ati on Valves were closea cae to a low Reactor dater Level Isolation Signal, RESET the Isolation Si gnal and OPEN tne MSIV's in accorcance witn LOP-MS-01.

2) If tne Main Concenser i s not availaole.

STARTUP the Steam Concensing Moce of R44 in accor3ance with LJP-RH-09, if req; ired, to control Reactor Pressure.

t

c. VERIFY that tne Main Generator Trips, 006s 7-1- 10 ana 10-11 (0;Bs 2-3 ano 3-4) OPEN, anc tne

. Transfer of AJxiliary Power to the System i- Auxiliary Transformer (SAT).

i c.- CsECK tne Main TurDine and the Turcine Jriven Reactor Feec DutDs vioration during coastcoan.

. MINIMIZE Water Inouction as follows:

j' 1) OPEN the following Drain Valves in order to minimize tne effects of possiole Turbine water Incuttion:

a) Main Turoine Leec Orain Valves ~4 0 -

IB21-;V2 and 4 (40-2321-CV2 anc 4) and AJ-1B21-F42LA and C (AO-1621-F421A anc C).

a) "A" Reactor Feea Pump Turbine.

i (1) Belod Seat Orain valves 40-1321-F425A and F427A(90-2321-F425A and F427A).

?- (2) .First Stage Jrain Valve, MJ-

l. 1821-F432A(MJ-2321-F4323).

c) "3".Recctor Feed Pump Turbine.

f- <v e u +---gn -

ag,- e -,,,e -p- --- - ,

... - - ~ ,

LGA-O'4 Revision 1 July 11. 1960 N (?na.9 (1) Selos Seat Drain val ves M0-1321-F4258 and F427d (MO-2821-F4258 anc F4273).

, (2) First Stage Grain Valve. MJ-IS21-F432S(v.0-2821-F4326).

c. VERIFY that tne 'lJrning Gear Oil Pump Starts.

e. START Suppression Pool Cooling in accoraance witn LOP-RH-13, if Suopression Pool temperature approacnes 103 F (T.S.).

f. REFER to LG3 3-2. Reactor Scram Procedure.

VERIFY that Reactor Safety Limits were not exceedec curing tne transient prior to restart of tne Unit.

E. DIS *,USSION Hign Re ac tor Vessel 'atar Level during steady state co-er a

operation can occur as a direct r esul t of erroneous Feedwater Control Systen increasec cenand, or a sy'otom of otner transients, sucn as a Pressure Regulator failure calling for cecreased Reactor Pressure.

Action is directec tonards restoring the level to normal Dy ta< i ng contr ol of the Feeddater Control System.

However, if Level EIGiT is exceedea, action is directea i towarc ensuring automatic actions ta limit Reactor dater Level rise, anc piovent Turoine camage. In accitio7.

protective features to nitigate the-associated pressure transielt are verified. Fi nall y,' contr ol or Reactor Water Laval anc Pressure is estaclished.

_It is inportant to monitor Turcine nacninery to minimize water iqdaction.

t

,- m. . _ ,. . _ . , _ -,

These documents for training purposes only.

They are not centrolled. They are not atithor,ized for plant operation or mairibnme use. Current revisions are available in Central Fib or Satellite g , _,3 s

- v'S' " 2 files for use in Oper. or li:ain .. ac ivitles. ai, 11. 1,3c ,

lod REACTOR aATER LEVEL A. SYMPTJws

1. Control Room Panel 1813-P633 (2H13-DoO3) Alarms:

a. FW CONTROL Rx WATER LEVEL 4 L0.

3. Fa VALVE CONTROL SIGNAL FAILURE.

. TURdINE RFP CONTR3L SIGNAL FAILJRE.

d. Possinle RX VESSEL WTR LEVEL 3 LO/ TRIP.

e. Possiole RX AJT3 SCRAM.

f. Possiale RX VESSEL aTR LEVEL 2 LO.

2. Control Room Panel 1H13 3603 (2n13-Po03) Incications:

a. TurDine Driven Feec Pump Decreasing Flow.

3. Motor D' ven Feed Puma Decreasing Flow.

c. Feedwater Heater low Cecreasing.

c. LOCAOJT lignt(s) illuminated.

e. LEVEL A or LEVEL 3 JUT OF SERVICE Lignt

, illaminated. 1

f. ADnornal Genano or Jutput Signal on a Feedwater

r. y s t e n Controller.

g. Reactor Water Level Decreasing.

3. ' Control Room Panel IPM33J (2PM33J) Alarms:

l

a. RX FEEDdATER DUMP TRIP. 1 l

c. Fd PUMP OISCH PRESS L3.

c. Rx FW PUMP TURBINE HY) OIL PRESS LO.

O. Fn PUMPS SUCTION PRESS LO.

7 LGA-05 Revision 2 July 11. 1950 1

2

e. CN05 and 300 STER DUMP AUT3-TRIP.

i.

f. CNOS BSTR P*PS SUCTION HDR DRESS LO.

g. CN05 DUMPS DISCd ,DR PRESS LO.

n. HEATE4 ORAIN PU 4P AJT3 TRIP.

o 4. Co1 trol Room Panel IDM03J (2PM33J) Incications

a. Feeaw3ter Pumo(s) Flo Oecraasing.

l

o. Turoine Drive 7 Reactor Feea Puma Speed Jecreasing.

c. Low Pressure or Hign Pressure heater Stri7g Isolation.

c. Motor Driven Fead Dump LC A:nps Dec r eas i ng.

e. Heater Orain Pump Dacreasing Amos.

f. Feecwater Puma Suction Header Pressure Changes.

, g. Feedwater Pomo Discnerge Header Pressure Cnanges.

n. Feedwater Turoine Control Oil Pressure Decreasing, L

Control or Stop V31ves Closing.

i. Concensate/Condensa te Sooster Pump Suction or 7 Discharge Pressure Changes.

J. Heater Drain Tank Hign or low Level.

, S. Otner ~ Indications:

a. Possiole Recirc. Flow Control Valve Runoack.

{ o. -PossiDie: Decreasing Reactor Power, Recirc.

. Flo. or Jetpumo : low.

Feedwater Isolation Val ves Closed Incication.

c.

1 c. Auto Start of Auxiliary Oil Pump.

e. Possicle Decreasing Turnine Generator Outout anc Bypass valve Stean Flow.

i

?

[.

4 4

4 t LGA-05 Revision 2

July 11, 19dC

> 3

f. Decreasing Level indication on Control Room Panels lH13-P501 (2413-P601) 3nd lH13-PdO2 (2H13-PoO2).

g. Possiole RHR 2036A(3)' - FJo4A(3) JPEN alarm.

S. AJT3MATIC ACTICNS

1. Level 4. Reactor vessel 4acer Level 31.5" actions.

D. Low Level Alarm.

o. Recirc Flow Control Valve RuncaCK permissive.

f 2. Level 3, Reactor Vessel dater Level 12.5" actions.

a. Reactor Scram.

o. TRIP Reactor Recirculation Pumps to 15 4Z.

4

c. ADS Permissive.

d. ISCLATE Vessel an Containment Grouos 4 O, ano 7.

e. SETBACK ProgrammeJ Level Control to 18".

C. IMMEDIATE 00ERATOR ACTIONS

]

i

' 1. If L evel 3 is reacned. verify actions, 00 10T continue this procedure. and go to LGP 3-2.

2. If Level 4, is reachec. verify actions.

3. Oatermine cause and correct if possiole.

4 Restore level if possiole Dy:

a. Increasing Feecsater Flow.

o. Recucing Reactor Power to maten Steam Flow and Feeanater Flos.

D. SJSSEOUENT JPERATOR ACTIONS

1. If-tne Reactor has scr3mned, complete performance of LGP 3-2, Reactor Scran.

4 1

LOA-05

, Revision 2 July 11, 19e0

'+j!nal}

2. If control of Reactor dater Level nas oeen re-estaolisned prior to aJtomatic protective action:

a. STATI3N an additional operator in tne Control Room to monitor ano control Reactor water Level.

He Snould De assigned no otner auties until t7e Feecwater Systen is returnec to normal.

3. F3 NIT 3R all Reactor deter Level instrumentation.

NOTIFY the Snift Suoervisor of any aDnormel incications.

, c. OETERMINE tne Soecific Cause for the Low 4eactor Water Level ana INITIATE corrective action as required.

E. DISCUSSION A cecreasing Reactor dater Level can occur et any poner lavel anc in any .noce of the Feeawater Level Control Systen. ProcaDie causes are loss of Feeawater or failure of tne Feedaater Level Control System.

Aatomatic Reactor protection is ultimately provideo Dy a Reactor Scram. cue to Low Reactor Water Level 3.

However, innerent designs in the Feedwater System a7c Recirculation System nay orevent a scram especially it supolementea oy operator action. Tne Dasic features are:

1. Auto Start of the Motor Jriven Reactor Feea Puno, if Dotn Turbine Driven Reactor Feed Punps Trio.

2. RunDac< of tne Reci rcul ati on Flo. Control Valva to limit Reactor Stean flow to the capacity of one 4

Turbine Driven Reactor Feed Punp.

3. Lockout of tne Turoine Driven Reactor Feea Pumos l and Seeowater Regulation V al ve , if an aDnormal l control signal is detectea.

Action is directea toward maintaining sufficient Feec.ater flo . However, snould a Reactor Scram occur, tnen scram recovery is initiated wnile oaying oarticular attention

'to core coverage ana. cooling.

i

. a ;_ _ . . ; , . _ . - _ _ _ _ _ . . _ . , _ . _ . . _ . . _

l

ThES dCOU30Et2 [0' t~B!D!DE DUT?CS0':, On!y.r n Thevu..rg r j

m o...

-: w.t . T.;=.: u<.

... ...; + ..1..u..,. .

r . ,f n n..s..

. .. e. . . u a, L c, A -0 3 r.u. ,m. .#o .n- a. : ,. u 4,, , p*l: ., n. . e. nr! l ~.:. 0 vG_1n..! ['s - ,

u.2, "avssion i T C.u;..n ganc e u' . o ...:...

s

u. . qi.. , : ~ n . . . + .

v n u . July 11, 1980 lVil85. 1 filCS 107 USB in Hign 003i. Ci f.ir:iG!. 20t. .t.

REACTOR pggg3URE A. SYw T'JMS

1. . Control Room Panel 1413-Pc]3 (2H13-POO3) Alarms:

a. Rx VESSEL PRESSJRE MI.

2. Co1 trol Room Panel 1413-P633 (2H13-PbO3) Incications:

a. Reactor Narro, Range Pressure Recorcer Hign.

o. Reactor Wide Range Pressure Indicator Hign.

3. Control Room Panel 1H13-P632 (2H13-oe02) Alarms:

a. RR PUMP 1A TRIP - ATd5 INITIATED.

o. RR PUMP LS TRIP - ATWS INITIATED.

4 Control Room Panel IM13-P601 (2H13-P501) Alarms: '

a. SAFETY / RELIEF VALVE FULLY OPEN.

5. Control Room Panel lil3-P601 (2H13-P601) Indications:

a. Post Accident Monitor Recorder A Pressure Hign.

c. Post Accident Monitor Recorcer B Pressure nign.

3. AJT3MATIC ACTIONS

'l. 10430 teactor Pressure Scrams Reactor.

2. 10760 Reactor Pressure anc aoove, SAFETY / RELIEF VALVES ACTUATE at setpoints. Also, oossiole LJa LEVEL SETBAC< (LLS) if 2 or more valves actuatec.

3. 1120r! Reactor Pressure TRIPS Reactor Recirculation Dumps to GFF (ATas).

C. IMMEDIATE' OPERATOR ACTIJNS'

-hp 1.- 'If 13430 is reacnea, verify actions.

L

2. .If.1075t is reacned verify $/R valve operation as aparapriate and LLS if 2 or more valves open.

e

4 LOA-Os Revision 1 July 11, 1930 2 (fina0

3. If 1123n is reacned. verify RPT.

4. Control level with Feedwater, RCIC, and HP;S, as necessary.

D. S JB S E )UE:9T JPERATOR ACTION

1. Continue to CONT 40L Reactor Pressure, Level anc Suopression Pool Temperature.

. 2. CORRECT the cause of the Hign Reactor Pressure Concition.

3. anen recovery can ce comnenceo, REFER to Proceaure LGD l-1; Nornal Jnit Startup.

E. DISCUSSION Hign Reactor Pressure is al ways an indication of anotner initiating event. For examole, a Turoine Trio, Reactivity accition, Pressure Regulator nalfunction (douol e failure),

MSIV closure or Feed System failure will cause transients wnich i1 crease Reactor Pressure. During power operation an APRM Scram should alnays preceea a Reactor Mign Pressure Scram except for the unli<ely event of a Turoine Trip wi tn Sycass Val ve failure Delow 30 Power. Snoulo failure of an increase in Neutron Flux to scram tne Reactor occur, Reactor Hign Pressure will initiate the scram. In tne event 1tnat tnis 03ckuo orotection fails. Anticipateo Transient Witnout Scram (ATWS) will trip the Reactor Recirculation Pamps.

Action is cirectea towarc ensuring the scram occurs ano tnat teactor Pressure is relievec to prevent over pressurization. In acditio7. Reactor Water Level anc core _ cooling must ce maintainea.

1 es - -- ewww@t.-m

.~ IheSG documents for training purposes only.

They are not controlled. They are not authorized for plant operation or maintenance use. Current loa _o7 revisions are availabic in Central Fi!e or Satellite Rev'Si n 2 July 11, 1960 f iles for use in Oper. or laaint. activities. 1 INADVERTENT REACTIVITY ADDITION A. SYMATOMS

1. ;ontrol Room Panel li13-P603 (2H13-P603) Alarms:

a. SRM MI/INOP.

3. SAM SHORT PERIO3.

c. RdM HI/INOP.

c. IRM Hl.

e. LORM HI.

f. APRW HI.

g. Possi ble R03 JUT 3LJCA.

n. Possiole CRD 3 RIFT.

i. Possiole FW FLOd CONTROL SIGNAL GAILURE.

J. Possiale FLJX C3NTR3LLER OUTPUT SIGNAL A3 NORMAL.

4 Possiole MASTER CJNTLR CUTPUT SIGNAL ASNORMAL.

1. Poss i al e R X AJT3 SCRAM.

m. Other alarms signfying increasing power.

2. ;oatrol Room Panel 1H13-0603(2H13-P603) Indications:

a. Posi ti ve Reactor period.

o. Increasing Reactor Power.

c. A transient in Reactor Pressure.

3 .' Ot,er Symptoms:

a. Possiale Increasing Turoine Generator Outout or dyoass Val ve Flo .

3. Possiole Increase in Feednater Flo a .

i

, p LGA-07 Revision 2 Jul y 11. 1960 2

c. Pos s i ol e Increase in Recircalation Flow.

3.' AJTJMATIC ACTIONS

1. Reactor Scram if Limiting Safety System Settings are exceeceC.

2. If in Master Automatic Flux Control, Recirculation

, Flow cecreases unlass tne cause of tne reactivity accition was a Recirculation Control System Pailare.

'C. 19 MEDIATE OPEEATOR ACTIJNS

1. If LSSS is exceecea and Reactor Scrams. 00 NOT continue this procedure anc go to LGP 3-2.

2. If LSSS is exceecea anc Reactor DOES NOT Scram. 30 10T continue this orocecure anc go to LGA-IS (ATaS).

3. If LSSS is .NOT exceeceo, cetermine cause of reactisity adaition, and initiate corrective action to terminate caJse.

4. Maintain Reactor Vessel Pressure and Level.

G. SUBSEUUENT JPERATOR ACTIONS

1. CHEC< stack Gas Activity and Off-Gas Activity for any increase that nould inoicate ore damage or aonormal rel ease of raciaattivity.

2. 3BTAIN a sample of Reactor Coolant. If indications of fiss si product release are present, PERFOR4 LGA-lo, Fuel Element Failure.

3. CHEC( that Reactor Power ana Flos nave not exceecea Reactor Doerational or Thermal Limits for Single and Tao Pump Recirculation Flow. CHEC.< that Reactor Cool ant Syst'm Temaerature Coolcown Rate Limits were not exceecea (T.S.).

4 If any Limiting Concition for Jperation nas oeen exceedea, EVALUATE tne necessity of Normal Snutcoon in accoraance .ith LGP 2-1 in orcer to be in Hot Snutaonn witnin 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> anc Cold Snutdown eitnin the following 24 nours (T.S.).

7 I

LGA-07 Revision 2 July 11. 1960 3iDG80

5. If corrective acti n succeecec in correcting tne Re ac t i v i t y Accition. AJJJ57 Control Rods as necessary to correspond to tne new Power level.

6. RE-ESTABLISH tne aopropriate Reactor Power Laval in

.accorcance with LGP l-1 Normal Jnit Startup aqc LGS 3-1, Power Cnanges.

4

[a E. DISCUSSION Inadvertent Reactivity Accition is an indication or another initiating caJse. fne following are possible I causes: Continuous Wit 7dra.al of a Hign dorth Rod l (Multiple Failure), Rod Droo Accident, Imaroper.5:artap ;

i of an idle Recirculatiol Punc (Joerator error). ;

Recirculation Flow -Control Valve Failure in tne open 4

olrection, ,a xenon transient or the 37jection of cola water into the core fron sources such as RHR Shutcown Cooling (Jperator error) or loss of Feeawater Heaters.

Action is inmediately cirected towards terminating tne i cause of the Reactivity Accition if tne cause is Kno.n anc tima is availaole. It the Reactivity Accition is rapic, action is directed towara verifying automatic

, action occurs and controlling' Reactor Pressure and Level.

In ootn cases Operational, Thermal Hydraulic and Raciation Release Limits are evslaatec prior to continuing recovery.

i 1

2

+

f

+

kg

,..,r - --

.OWiBLC ,.,-m.-" " ,;c.

, w ..

.g , g,__:..

pgh]a,.gG,u eec: pens od'l . ,ag g . ae nc; m . .g c w e r,,

q, f .

g ,',. ;;, s . g . J . ." -; ". "l 1'-l,'.;,to e cir Sa!SIIIIS

~;' " :. % u . t a "

uA_o3 u- .,p*.

J.y/1510... -

,; g g([\ l{.l6$*- . Revision 2 jd.% iO[ 05$ III h0 July 11, 1980 LOSS CF RECIRCULATION FLJn - SINGLE PUvP A. SvvpT3*S

1. Control Aoom Panel 1913-9603 (2H13-Po03) A l a r t.s :

a. RJO CJT BLO;K.

o 3. APRM HI.

2. Control Room Panel 1413-P603 (2H13-P503) Incications:

a. Total Core Plou and Pressure Recorcer Decreasing.

3. Reactor Power Decreasing.

c. Steam Floa ano Feedsater Flow Decres ing.

d. Reactor Feedpamp cl aw Incicators Secreasing.

e. Reactor aater Level Increases tnen Decreases to Normal.

3. Co1 trol Room Panel 1H13-2o02 (2H13-P602) Alarms:

a. Rx RECIRC PJM3 1A(2A) or 1S(28) AUTO TRIP.

4 Co, trol Room Panel lil3-P602 (2H13-P602) Indications:

a. Reactor Recirculation Pump Flow Decreasing in one (1) Recirculation Loop.

a. Jet Pump Flow associated with the aoplicaole Recirculation Loop 3ecreasing.

c. Jet Pumo Flow assaciated nitn the other Recirculation Loop Increasing.

d. Reactor Recirculation Pump Differential Pressure A or 3 Decreasing.

e. Reactor Recirculation Pump Motor A or B Carrent Decreasing.

5. Jt7er Symptoms:

l l

M LGA-Oi Revision 2 July 11 1980 2

a. Decreasing Mein Generator Amps and aatts.

o. MSR Outlet Pressures Decreasing.

A JTJM ATIC ACTIONS B.

1. LA or IB (2A or 23) Reactor Recirculation Puno TRIPS.

C. _I ___

w M E _D _I

_ _A T_E __O_ P_E R A _T __O _R_A C T I J N _S

l. Verify M/A Station transferreo to Manual.

2. Place tripoed recirc pumo dkr 3 in Pull-to-Lock (PTL).

3. Decrease Reactor Power to less tnan (later) oy c decreasing recirc. flo..

4 Maintain Reactor Pressure and Level.

O. SJBSEQUENT 3pfRATOR ACTIONS

1. VERIFY tnat the Reactor Recirculation Pump Discnarge i

and Suction Bloc < Valves are OPEN in tne triopec Recirculation Loop.

2. VERIFY Seal Staging Flow Stop Valve 50-1833-7079A/S (53-2833-F079A/8) is CLOSED as inoicatec on 1H13-1 P632 (2H13-Po02).

, 3. DETERMINE the cause of tne Reactor Recirculation Punp Failure and I'J ITI AT E aopropr i ate corr ec t i ve action as requirec.

4 After the cause nas been found and correctec anc i - '

is VERIFIED tnat no Operational Limits have oeen j exceedea. RESTART the tripoed Reactor Recirculation Puno in accordance aitn LOP-RR-Oo.

5. If any Limiting Condition for Operation has oeen exceeded, EVALUATE'tne necessity of Normal Snutoown

'in accordance with LGP 2-1 in order to be in Hot Snatdown witnin twelve (12) nours and Colc Snutcown nithin the follo.ing t-enty-four (24) nours (T.S.).

6. If two (2) Looo Reactor Recirculation cannot be restorec, RESTRICT Reactor Power to less tnan or eaaal to 75% of-Rates-Thermal Power. See Tecnnical Specification 3/4.4.1.

I a19y T

-re--- - -

3 1

i i

LGA-CS Revision 2 July 11, 1980 Sfinal) 7

7. CHEC( Stac< Gas Activity anc 03TAIN a sample of f i Reactor Coolant.

i E. DISCUSSION A loss of one-(1) Reactor Recirculation Pumo with normal l coastsoen will not normal l y result in a Reactor scram

ouring operation. Snoulo a Reactor Scram occur auring

j. tnis tr ansi ent. tne pos s ioil i ty of instrument mal f uncti on 1 j or th? Jnlikely event of Reactor Recirculation Pump l i seizure may exist. Tnerefore, orocecure LGP 3-2, Reactor l Scram. shoulo De follo-aa if a Reactor Scran occurs.

i Action is cirected tonard ensuring that tne Reactor Recirculation PJmo trip procescs to completion (e.g. that i tne Reactor Recire. MS Set coes not energize the ounp in Slon Spaec) and that Core T,ermo-nydraulic Linits are not exceeded.

! Tne trio of one (1) Reactor Recirculation Pump to Slon

! Speed may oe inoicative of 3- failure of Dotn Reactor l Recirculation Pumas to transfer to Slow Speed. Tnis

situation woula occur f ollosing tne failure of a cavitation protection trip to transfer botn Reactor Recirculation l PJmos to Slo Saeed.

i Tne f ollowing are likely causes for a single Reactor

, Recirculation PJmo Trip:

1

, 1. Suction or Discnarge Block Valve less than 90% ooen.

2. Dunp Electrical Mal f Jnct i on.

3. Los Fraquency MG Set Electrical Malfunction.

1 f

a 1

l

, a -, -. -_ , . .. - . . - . . . _ . . . . _ , --- -. . .. ., .-. . - . -

79 " , 4gcijg16his IOr trainIO.E N c 1 m e e a aneu .! s e n nsta}geg U,; ;!.Snt Op'didK'[ ,0'.

, MO0;SateUlte [{,n t July 11, 1980 c ggygging, g.; ;;3 ions a!O BVEiiGIO kla. zines for use

, o 1

LOSS OF RECIRCULATION ELJW - BOTH LOOPS

a. SvuPTJMS

1. Control Room Panel 1913-Pc03 (2H13-P603) Alarms:

a. ROD Qui BLO;K.

, 3. APRM il.

2. Co7 trol Room Panel l' P603 (2H13-Po03) Incications:

a. Total Core Flow and Pressure Recorder Decreasing.

D. Reactar Power Decreasing.

c. Steam Flow and Feecaater Flow uecreasing.

d. Feed'aater Pumo Floa Incicators Decreasing.

e. Reactor Water Level Incre.' s tnen Decreases.

3. Control Aoom Panel 1H13-P602 (2H13-P602) Alarms:

1 i a. RX RECIRC PJMD LA(2A) AUTO TRIP.

j o. RX RECIRC PUMP 1B(23) AUTO TRIP.

4. Control Room Panel 1913-P602 (2HL3-P502) Indications:

a. Reactor Recirculation Pump Flow Decreasing in botn Recirculation Loops.

3. Jet Pumo F1ow Decreasing.

l l c. Reactor Recirculation Pump A and 6 Differential i

Pressure Decreasi1g.

d. Reactor Recirculation Pumo Motor A and 6 Current Decreasing.

5. 3t7er Symptoas:

a. Main Generator output cecreasing.

3. MSR Outlet Pressures Decreasing.

I

h i

+

! . LSA-09 Revision 1 July 11, 1960

2

5. AJTJMATIC ACTIONS

1. None.

C. IMMEDIATE OPERATOR ACTIJNS

l. PLACE tha mode s-iten in Snutdown (T.S.) ana carry 1

out LGP 3-2.

3. SJBSE00ENT JPERATJR ACTIONS

! 1. VERIFY tnat tne Reactor Recircalation Pump Discnarje i and Suction diocx Valves are open in A and E Recirculation Looos.

2. VERIFY Seal Staging Flow Stoo Valves 50-LS33-F377 j - - -

A snJ B (53-2833-F377 A C d) are closeo.

3. 3ETERMINd the caase of tne Reactor Recirculation Duno Failure, and INITIATE appropriate corrective action as required.

4 After the cause nas ceen found and corrected anc it 4

is VERIFIED that no Operational Limits or Core Tharnal Limits have cean exceeded. RESTART Dotn Reactor . Recirculation Pumps in accordance witn L3P-4R-06.

5. CHEC< Stack Gas Activity. OBTAIN a sample of teactor I ;oolant.

6. CHECS Throine for excessive vioration.

E. DIS;USSION Tne comolate loss of two (2) Reactor Rec i rcul ation pumps is a nighly unlikaly e v en't signified oy tne loss of four (4) powcr supplies. Snould ootn Reactor Recirculation Pamos tria, ENSJRE that Anticipated Transient Witnout Scram - ATWS (Reactor vessel Pressure 1135 psig or Lon

Reactor Water Level 2) was not the cause. SCRAM the Reactor in accorcance witn LGP 3-2, Reactor Scram, it ATWS was the caJse.

In addition, a reactor Recirc. Flow Control Valve (FCV)

Runoack cae to a loss of Peedwater Flow or failure of tne Reactor Recirc. Flow Control Valves (ACV) in a closec direction will result in a similar transient response.

LGA-09 Revision 1 f"lf(nit s [' 19 60 A tion iS Cirected to=ard ensJring th3t Core TherPJ-nycraulic Limits are not exceecea anc towarc ensuring a timely restoration of Reactor Racirculation Flow if possi31a.

I 1

These documents for training purposes only.

They are not contro!!cd. They are not authorized

, for plant operation or maintenance use. Current d$Ifon t 3'

revisions are avaliaW in Central File or Satellite f*' " *'

c

files forLOSS useOFinTURSINE Oper.GENERATOR or Maint.LCAD activitie$4 EATER THAN 25; -

w l A. SYMPTO*5 4

L. Sontrol Room Panel 1Pu02J (2PM02J) Alarms:

a. Possiole TURBINE TRIP THRUST 3 EARING FAILJR.E.

. c. Possiole TUR2INE TRIP MSR LEVEL HI.

1 3 c. Possiole TURBINE TRIP VISRATION HI.

d. Possiclo TURBINE TRIP LOSS OF STATOR COOLANT.

J

e. Possiole TURSINE TRIP EXHAUST H300 T E.M P HI.

f. Possinle TURSINE TRIP EHC HYD PRESS ~LO.

J

q. Possinle TURoINE TRIP EHC MANUAL.

, n. Possiale TU451NE TRIP EHC MASTER,

i. PossiDie TURSINE TRIP EHC COMPUTER TRIP.

m J. Possiale TURBINE TRIP VACUUM LO.

. k. Possiale TURBINE TRIP SPEED SIGNALS LOST.

a

1. Possiola TUREINE TRIP SHAFT pumas DIS PPEiS LO.

t

- m. Possiole TUREINE TRIP LOSS JF PS 100 A/S.

n. Possiole TURBINE TPIP OVE4 SPEED.

o. PossiDie TURb!NE TRIP LOSS OF 2*V DC.

p. Possiole TURBINE TRIP BACAUP OVERSPEE0

q. Possiale TURSINE TRIP NON-EHC.

_ 2. Control Room Panel IPM0lJ(2PM01J) Alarms:

3. Possiole GEN 1(2) P40T RELAY TRIP.

e

d .

J LOA-10 W Revision 1 Cecemoer 13, 1979 2

$h D. Possible GEN 1(2) TRIP Sys 1 Lx0 TRID.

c. Fassioie GE4 1 TRIP Sf3 2 mAC TtIP.

J d. Possiole Alarms Incicating Main Transrcrmer Trouble.

9 a

a 3. Otner Incications:

.. a. Cecreasing Turoine Generator Output.

J c. Extraction Dump valves Open.

a

c. MSR Outlet Pressure Decreasing.

J

c. Bypass Valves Open.

e. Safety-Relief valves Open.

a

t. Possiole Loss or Transfer or Auxiliary Power.

d q. Qeactor Scram.

, n. Recirculation PJmps Trip to sl oa Speed (RPT).

m

i. Possinle :4i.)n Reactar aater Level Incications.

3. AUTDwAT!C ACTIONS a

1. Turoine stop Valves (vsv's) Close.

j 2. Direct Reactor S c r a T. from Turbine Stop Valve Closure.

3. Recirculation oump Trio to slow speed (RPT) from Turbine stop VJlve Closure.

4. Bypass Valves Open to control Reactor Pressure.

, 5. Safety-Relief Valves Open as necessary.

o. '4 a i n Generator cirect or incirect trip.

~

C. IMMEDIATE CPE9ATOR ACT!]NS

, 1. VEstFY tnat in. -e3ctor S c r a r ". . all Control on

_ tully insert aio React;r Po.er cecresses w i tti normal neutron Jecay.

m - -

4 a

LGA-10

'(

a Reviseon 1 Decemoer 13. 1979 r 3 i t ,

y 2. V ER IF Y tnat the Recirculation Pumps Trip (RPT) to -

, 513= Soeec. '

1 I

,9 3. V ER IF Y tnat tne Bypass Valves anc Safety-Reliet j-* Valves Open as necessary to control aeactor Pressure. [

j 4. VERIFY tnat Main Turoine Speed is decreasing. If

,y, Main Turoine Speec is not aecreasing, CHECA tnat r i :ne Turoine Stop Valves (MSV's) nave closea.

4 d a. If MSV's are not closed, CONSIDER closing tne i "" MSIV's.

t t 5. VERIFY tnat the Main Generator Trips, OCSs 9-10 anc hj 10-11 (CCBS 2-3 and 3-4) OPEN, ano tne Transfer of Auxiliary Poner to tne System Auxiliary Transfarner (SAT .

" 6. CHEC4 Reactor water Level. MAINTAIN Reactor water Level aoove Level 4, using tne Reactor Feed System if possiale.

.d

7. START RCIC if necessary to maintain Reactor Water .

. Level above Level 4.

as D. SU 6S E 'JU E N T OPERATJR ACTIONS 3 1 CONTINUE to control Reactor Water Level anc oressure:

. .s

a. OPERATE Bypass valves as necessary if tne Main Concenser is availeole. If tne Main Steam I Lines are closed, VERIFY Isolation Signal is RESET. ana OPEN M51V's in accotaance oi tn L3P-MS-31.

o. If the Main Condenser-is not ava i l ao l e . STARTUP Steam Canaensing in accorcance witn LOP-Rw-09 if required to control Reactor Pressure.

~

2. VE41FY tnat the Tur ni ng Gear Oil Pump Starts.

3- 3. . VERIFY tnat the Turning Oear Engages. ;

4. ST ART ' Succres si on Pool Cooling in accorcance witn LOP 3H-13 it Suppressinn Pnci Tn perature approaches i 100 r(T.S.).

t 4

i l

l U , l l

I

8 LGA-10

,1 Revision 1 Decemoer 13, 1079 (fin 30 a 5. OETERMINE ene cause of tne Sain Turoine Genera cr Trio anc INITIATE aporo:riate correc-ive action as

. requirec.

A ed

6. REFER to LGP 3-2, teactor Scram. VERIFY t r. a t deactor

, Safety L i mi ts were not etceeced during tne transient j prior to restart of the unit.

.4 E. O!SCUSSION (j Loss of tne Main Turoine Generator at P.e a c to r P3-er Levels aoove 30% results in a direct Reactor Scram. Safety-

,, Relief Valves may operate to mitigate the Reactor Pressure

- Trt 7sient. Tne initial Aeactor Power Level directly

~J affects the namoer of Safety-Reliet valves requirac to operate. A Turoine Trio from nign Reactor Powr. r L ev e l s witn loss of 6yaass Valve cacacility requires all Satet(-

8 Relief Valves to operate.

Action is cirectec towarc verifying Reactor Scram. RPT anc proper 6ypass and Safetv-Relief Valve operation.

d Once tne integrity of t7e Feactar Coolant Pressare roundar y is verifiec, Turoine Generator Protective Actions ar e verified. Subsequentl /, R r_3 c t o r Level is controllec to j ensure acequate core cooling, and a timely r ec over y.

At niih Reactor Power Levels, tr ansi ent response is very r3pic. Tnus, the verification of Aatomatic Action is d

reouired prior to initiating nanual recovery.

4 m

0 t

i 1

1 i

e

. {c \i Y O '5 ( :

}hgte 60 p ,.i'c d. N N " Y .;., C UII a

They2.!'2US

$ c.; mdh I

. v.

. . . i s ;. ^.

c; N 10Tl! . li,,SSB ' h ". s W,;iyitiBSe LGA-11 Revis'on 0 d 56Yj,;irAS $0 '

6 D; )3,,L iti..

t Marcr 21, 1979 j'gyg 10I USD ggg (lG3 1

' LOSS OF TURBINE GENERATOR LOAO LESS THAN 25%

5 A. _S _Y _M _P _T O_.M._S L. Control Room Panel 1PM02J(2PM02J) Alarms:

i J a. Possible TURBINE TRIP THRUST BEARING FAILURE.

j

b. Possible TURBINE TRIP MSR LEVEL HI.

J

c. Possible TURBINE TRIP VIBRATION HI.

d. Possiole TURBINE TRIP LOSS OF STATOR COOLANT.

e. Possible TURBINE TRIP EXHAUST HOOD TEMP HI.

J f. Possible TURBINE TRIP EHC HYD PRESS LO.

. g. P os s i b'. e TURBINE TRIP EHC MANUAL.

h. Possible TURBINE TRIP EHC MASTER.

i. POssible TURBINE TRIP EHC COMPUTER TRIP.

J. Possible TURBINE TRIP VACUUM LO.

J k. Possible TURBINE TRIP SPEED SIGNALS LOST.

, , 1. Possible TURBINE TRIP SHAFT PUMPS DIS PRESS

.J LO.

m. Dossible TURBINE TRIP LOSS OF PS 100 A/B.

n. Possible TURBINE TRIP OVERSPEED.

, o. Possible TURBINE TRIP LOSS OF 24VDC.

p. Possible TURBINE TRIP BACKUP OVERSPEED.

. . q. Possible TURBINE TRIP NON-EHC.

2., Control Room Panel IPM01J(2PM01J) Alarms *

a. Possible GEN 1(2) PROT RELAY TRIP.

f

- - _ _.,y _.

J LGA-ll y Revision 0 March 21, 1979 2

j b. Possible GEN 1(2) TRIP SYS 1 LKD TRIP. .

. c. Possible GEN 1 TRIP SYS 2 LKO TRIP.

o E d. Possible Alarms Indicating Main Transformer Trouble.

i '

S 3. Otner indications:

a. Decreasing Turbine Generator Output.

W

b. Extraction Dump Valves Open.

c. MSR Outlet Pressure Decreasing.

d. Bypass Valves Open.

_ e. Poss ible Safety-Rel ie f Valves Open.

f. Possible Loss or Transfer of Auxiliary Power.

g. Reactor Power below 30%.

. h. Possible High Reactor Water Level Indications.

B. AUTOMATIC ACTIONS a 1. Turbine Stop Valves (MSV's) Close.

2. Bypass Valves Open to Control Reactor Pressure.

J

3. Main Generator direct or indirect trip.

C. IMMEDIATE OPERATOR ACTIONS

1. CHECK that Reactor Power is below 30% and NOT increasing.

2. If Reactor Power is increasing, MANUALLY SCRAM the Reactor. REFER to LGA-10, Loss of Turbine Generator Load Greater than 25%.

3. VERIFY that the Bypass Valves are controlling Reactor

! Pressure via the Pressure Regulator.

4. VERIFY that Main Turbine Speed is decreasing. If Main Turbine Speed is not decreasing, CHECK that

_ the Turbine Stop Valves (MSV's) have closed.

3 s

LGA-ll d Revision 0 Marcn 21, 1979 3

h a. If the MSV's are not closed, CONSIDER Scramming

the Reactor and Closing the Main Steam Isolation Valves (MSIV's). VERIFY that tne Safety-Relief j Valves open as necessary to control Reactor Pressure.

8 5. VERIFY that the Main Generator Trips, OCBs 9-10 and a 10-11 (CCBS 2-3 and 3-4) OPEN, and tne Transfer of Aux il i ar y Power to the System Auxiliary Transformer

? (SAT).

m 4

6. CHECK Reactor Water Level, MAINTAIN Reactor Water Level above Level 4, using the Reactor Feed System if possible.

D. SUBSEQUENT GPERATOR ACTIONS

= 1. If the Reactor Scrammed, REFER to LGA-10, Loss of Turbine Generator Load Greater than 25%.

, 2. If the Reactor has not Scrammed:

a. VERIFY that the Turning Gear Oil Pump Starts.

b. DECREASE Reactor Power to limit Bypass Val ve Steam Flow.

9 d

c. MAINTAIN Main Condenser Vacuum.

d. VERIFY that the Turning Gear Engages.

e. OETERMINE the cause of the Main Turbine Generator Trip and INITIATE appropriate corrective action

. as required.

f. CONTINUE Normal Unit Startup or Normal Unit Snutdown as appl i cabl e.

E. gijCUSSIQN

- Loss of the Main Turbine Generator at Reactor Power Levels below 30 percent should not result in a direct Reactor Scram, since this Reactor Power Level is within the cadacity of the Bypass Valves and Auxiliary Steam loads.

Shoul d the Bypass Valves fail to respond to accomodate Reactor Output, a Reactor Power or Reactor Pressure Scram wi.Il be initiated.

= - .

s LGA-ll

= Revision 0 March 21, 1979 ffinal)

Action is directed toward verifying Reactor Power Level and proper Bypass Valve operation. Once these criteria are established, Turbine Generator Protective Actions are verified. Subsequently, depending upon plant response, Reactor Power, Pressure and Level are controlled to prepare for a Unit Pestart or Controlled Shutdown.

A

.J w

4 k

J M

d d

n J

w

= .

These decmifents for training purposes only.

" They are not ecittrolied. They are not authorized

, for piant cptration er mainten:r.ca use. Current CMon a Maren 21, 19T9 revisions are avalinle in Caniral File or Satellite 1

- files for use in Oper. or Maint. activities.

J LOSS OF AUXILIARY ELECTRICAL POWER A. SYMPTOMS

1. Control Room Panel 1PM0lJ (2PM01J) Alarms:

3 a. TR 1E-IW (2E-2W) b-CKUP OIFF TRIP.

I

b. GEN 1(2) PROT RELAY TRIP.

1 J c. GEN 1(2) SYS 1 LKO TRIP.

d. GEN 1(2) SYS 2 LKO TRIP.

e. MAIN T-lW (T-2W) PROT RELAY TRIP.

f. MAIN T-1E (T-2E) PROT RELAY TRIP.

a

g. UNIT AUX T-141 (T-241) PROT REL TRIP.

, h. All 6.9 KV and 4.16 KV Feeder Breaker Trip alarms annunciate.

E i. Undervoltage alarms on Non-essential Buses annunciate.

SYS AUX T-142 (T-242) PROT RCL TRIP.

J J 2. C( ntrol Room Panel IPM01J (2PM01J) Indications:

a

a. Main Condenser Output Breakers, 5AT and UAT Supply Breaker to 6.9 KV and 4.16 KV Buses OPEN.

b. Voltage. Current and Power to all Buses Decreasing.

c. Diesel Generators O and 1A (2A) START. l

_ 3. C on,t r o l Room Panel 1H13-P603 (2H13-P603) Al ai ms :

a. ROD DRIVE CONTROL SYSTEM INOP.

b. CONTROL ROD ORIVE FEED PUMP AUTO TRIP.

i l

l l

1 4.

a s

LGA-12 d ,

Revision 0 Marcn 21, 1979 2

h

c. RPS MG TRDUBLE.

d. MAIN CONDENSER VACUUM LO.

e. MAIN STEAM ISOL VLV NOT FULLY OPEN.

, f. Possible RX AUTO SCRAM.

w

g. PRI CONT PRESS HI.

j 4. Control Room Panel 1H13-P603 (2H13-P603) Indications:

a. CRD Drive Water and Cooling Water Decreasing.

J b. CRD Pump Amp *; Decreasing.

c. Reactor Pressure Increases then Decreases.

A

d. Reactor Water Level increases then Decreases.

. e. Steam Flo.' Decreasing.

. f. Turbine Driven Reactor Feed Pump Steam Flow

" Decreasing.

g. Motor Driven Reactor Feed Pump Amps Decreasing.

" h. Rod In Lights Come On.

l

i. Decreasing Power.

a%

5. Control Room Panel 1H13-P601 (2H13-P601) Alarms:

l

c. 4 KV BUS 143/143 ~ '243/2 '-1) UNDERVOLTAGE.

I

b. DG 1B(2B) ENG.Nd RUN tNG. ;

c. Other Possible ECCS .r ' '

.6 tion Alarms.

6. Control Room Panel 1H13-P601 (2H13-P601) Indications:

a. Post Accident Monitor Reactor Pressure Increases 7

then Decreases.

~ '

b. Post Accident Monitor Reactor Water Level .

Increases then Decreases.

2 LGA-12 d Revision 0 Marcn 21, 19/9 3

~

t

~

".' c. Possible Start of Low Pressure ECCS Systems.

. d. Auto Start of RCIC and HPCS.

3 d

i 7. Other Indications:

3 a. Reactor Recirculation Pumps Trip.

4

b. Reactor Water Cleanup Pumps Trip.

1 s c. Possible Auto Start of Standby Gas Treatment. -

d. Trip of Plant Ventil ation and Plant C i rcul at i ng

, j Water Systems.

e. Turbine Generator Trip.

i J f. Loss of Feedwater, Condensate and Condensate Booster Systems.

. 9 Loss of Condensate Vacuum.

g h. Containment M:~ic Bus indicates Isolation.

i. Drywell Pressure Increase.

B. AUTOMATIC ACTIONS

*

1. Direct or Indirect Reactor SCRAM.

_1 2. AUTO START of Emergency Diesel Generators and Load Transfer to the associated essential division:

' a. Diesel Generator 0 - ESS Division 1.

b. Diesel Generato- 1A(2A) - ESS Division 2.

-I

c. Diesel Generator 1B(43) - ESS Division 3.

3. Reactor Recirculation Pumps TRIP.

4. Main Turbine, Unit Aux. Transformer (UAT) and System

, Aux. Transformer (SAT) TRIP.

3. Reactor Feedpumps TRIP.

1 1

a LGA-12 d Revision O Marcn 21, 1979 4

~

6" RCIC and HPCS AUTO START on Level 2.

i. AUTO START of Low Pressure Emdrgency Core Cooling 1 sequence and system realignment if Reactor Water Level decreases to Level 1.

j 8. AUTO START of Emergency Bearing Oil Pump as Turbine a Speed oecreases.

'a 9. All Reactor and Containment Isolation Valves CLOSE

J on Loss of Reactor Protection Buses.

] C. IMMEDIATE OPERATOR ACTIONS j

N 1. CHECK Reactor Water Level, Pressure and Containment Pressure response to ensure that a Loss of Cool ant Accident has not occurred.-

J

2. VERIFY Reactor SCRAMS, all Control Rods fully inserted and Reactor Power decreases with normal decay.

~

3. VERIFY AUTO ST ART of Di esel Generators 0, lA(2A),

1B(28) and that ESS Divisions are energized from d their respective Diesel Generators.

4. CHECK Reactor Water Level and Pressure:

J a. OPEN Safety-Relief Va'Ives if necessary to cecrease Reactor Pressure below 1076 psig. ,

l j b. START RCIC or HPCS if necessarysto maintain Reactor Water Level.

CAUTION Al ternate Saf ety-Rel i ef Valve Operation to promote uniform Suppression Pool Cooling.

5. VERIFY Reactor Recirculation Pumps TRIPPED. PLACE ;

A and 8. Reactor Recirculation Pump Breakers 3 in

, PULL to LOCK.

6.

r

-VERIFY that Standby Gas Treatment (SGTS) AUTO STARTS J

and that Reactor Building ventilation nas ISOLATED.

O. SUBSEQUENT OPERATOR ACTIONS 5

6

. - - . . , -- n, , . - - --, ,,,,--n

2 .

J LGA-12 y Revision 0 March 21, 1979 5

g 1. After Reactor Water Level is' restored and it is -

VERIFIED that a Loss Gf Cool ant Accident does not exist:

I w a. STOP HPCS, LPCS and RHR A/B/C Pumps.

.,j o. START Suppression Pool Cooling in accordance

, with LOP-RH-13 if Suppression Pool Temperature approaches 1000 F (T.S.) or if Drywell Pressure exceeds 4 psig. If high Containment Pressure j exists, START Containment Spray to decrease the pressure.

2. VERIFY proper operation of the Diesel Generators.

d

3. VERIFY the AUTO START of the Er.ergency Bearing Oil Pump and the Turning Gear Oil Pump.

a

4. VERIFY that the Main Turbine and Generator have TRIPPED and that Generator Output Breakers 9-10 and w

10-11 (2-3 and 3-4) OPEN.

5. VERIFY that the Turbine Driven Reactor Feed Pumps W have TRIPPED.

6. If possible RESTORE Auxiliary Power from SAT 142(242)

, as follows:

, a. RE-ENERGIZE the System Auxiliary Transformer with applicable steps of LOP-AP-Ol.

o. RESTORE Bus 141Y(241Y) Power from SAT 142(242) in accordance with LOP-AP-16.

c. RESTORE Bus 142Y(242Y) Power from SAT 142(242)

< in accordance with LOP-AP-17.

- d. RESTORE Bus 143(243) Power from SAT 142(242) in accordance with LOP-AP-18.

7. PLACE the following Control Switches in PULL to LOCK en order to minimize the possiblity of high currents when restoring power distribution:

1

- a. Circulating Water Pumps IA, 18 ano 1C (2A, 28 and 2C).

b. Heater Drain Pumps 1A, 18, 1C and 10 (2A, 2B, 2C and 20).

J a

LGA-12 y Revision 0 March 21, 1979 6

j c. Service Water Pumps lA, 1B (2A, 23), Service -

Water Jockey Pump OW502PA (OW502PB) and Service Water Pump "0" if Unit 2 Bus 241x was lost.

d

d. Condensate and Condensate Booster Pumps lA, 18, LC, 10 (24, 2B, 2C, 20).

e a e. Motor Driven Reactor Feed Pump LC (2C).

f. Electrode Boilers OA, OB (OC, 00).

i rd

g. Inerting Steam Electrode Boiler LA (2A).

J j h. Primary Containment Water Chiller lA, IB (2A, 28).

8. RESTORE Station Air as follows:

J

o. If the System Aux. Transformer is energized:

, 1) ENERGIZE Switchgear 151 and 152 (251 and 252) in accordance with LOP-AP-01.

"i

2) ENERGIZE Buses 131 A and B (231 A and B) and 132A and B (232A and B).

3) START one (1) Turbine Building Closed s Cooling Water Pump in accordance with LOP-WT-02.

, 4) START Unit 1(2) Station Air Compressor in accordance with LOP-SA-02.

b. If the System Aux. Transformer can not ce energized START the opposite Unit Station Air Compressor (or Common Air Compressor).

- 9. OBTAIN a sample from the Drywell in preparation for venting the containment (T.S.).

10. VENT the Primary Containment to clear the High Pressure Containment Isolation Signal.

1 1 ". RESET the following Emergency Core Cooling System Initiation Signals:

J w

LGA-12 mi Revision 0 March 21, 1979 7

h a. LPCS - RHR A.

b. RHR B-C.

I 2 c. AOS.

a

] d. HPCS. ,

12. PLACE Emergency Core Cooling System Pumps in NORMAL 9 unless the Pump is performing a Core or Cont'ainme1t j Cooling Function. STARTUP Steam Condensing mode of RHR in accordance with LOP-RH-09 if required to control Reactor Temperature and Pressure.

13. STARTUP RPS in accordance with LOP-RP-01.

14. RESET the Scram.

a

15. RESET Reactor and Containment Isolation Signals.

i 16. START Primary Containment Ventilation Fans in accordance with LOP-VP-05.

4 17. ENERGIZE BUS 142X(242X) and START Service Water in accordance with LOP-WS-01. Using Service Water Jockey Pump OWSO2PA (0W502PB) maximize flow to the 3

Reactor Building Closed Cooling Water and Fuel Pool a Cooling Systems.

. 18. ENERr,1ZF Buses 133 and 134X and Y (233 and 234X and a Y).

19. STARTUP Reactor Building Closed Cooling Water to supply the Containment in accordance with LOP-WR-02 using one RBCCW Pump.

20. STARTUP the Reactor Water Cleanup System in accordance

- with LOP-RT-02 in order to promote mixing in the Reactor and RESTORE reject capability.

21. STARTUP the Control Rod Drive Hydraulic System in accordance with LOP-RD-01.

2 2 '. ENERGIZE Bus 141X (241X).

23. STARTUP Pr i mar y Contai nment Coolinq in accordance with LOP-VP-02. OPERATE the Chiller Unit that will equalize loading on Bus 141Y anc 142Y. If absolutely necessary, SUPPLY Primary Containment Chill Water

w w

LGA-12 g Revision 0 March 21, 1979 8

j with Re:ctor Building Closed Cooling Water in -

accordance with LOP-vP-09.

j 24. START an additional Service Water Pump if required.

25. STOP venting of the Primary Containment.

?,

e 26. VERIFY that the Fuel Pool Cooling System is operating.

4 27. VERIFY AUTO ST ART of the Main Turbine and Feed Pump j Turbine Turning Gears.

28. TERMINATE Lake Blowdown and Radwaste Discharge.

J 29. If CFFSITE Power has not been restored, assess Unit Shutdown Power requirements.

a. ENSURE an available supply of Diesel Oil is available.

. b. ENERGIZE 480 Volt AC Buses as necessary maintaining Breaker and Diesel Load Limitations.

i d 30. CHECK Stack Gas Activity. OBTAIN a sample of Reactor Coolant.

31. COMPLETE LGP 3-2, Scram Recovery.

32. STARTUP Reactor Building Ventilation in accordance l

with LOP-VR-01 and SHUTDOWN Standby Gas Treatment

- in accordance with LOP-VG-02 after power is restored to the 6.9 KV - 480 Volt Distribution System.

33. If STARTUP electrical lineup is available and Limiting Condi: ion for Operation will preclude a normal startup (T.S.), REFER to LGD l-1; Normal Unit Startup or LGP l-3; Unit Hot Standby to Power Operation as applicable.

E. Ol}CU}ON Loss of the System Auxiliary Transformer with simultaneous

      .       loss of the Main Generator or a loss of all grid connections will cause a Loss of Auxiliary Electric Power.' Initial
         ,    automatic action'is initiated by any combination of the following events:     Turbine Generator Load Rejection, Loss of Condenser Vacuum and the Undervoltage trip of Reactor Protection-Motor Generator Sets.

1 LGA-12 d Revision 0 Marcn 21, 1979 9 (final) s' Action is directed toward ensuring that a Reactor Scram occurs, Auxiliary Power is transferred to tne Diesel Generators, and that Reactor Pressure and Water Level remain under control. In addition, ensuring continuous oil flow to the Main Turbine and Reactor Feedpumps is included to prevent equipment damage. d The Steam Condensing Mode of RHR will be used to control Reactor Pressure and Water Level due to the loss of the

   ,    Main Condenser.

J Subsequent action is directed toward restoring power via the 4160 Volt and 6900 Volt distribution systems to reset

 ;      the Isolation, depressurize the Drywell, reset the Scram, and minimize temperature stratification in the Reactor.

4 4 4 e d 9 M t

These de:nments fc- training perpe:cs only. T.hev are not Ccocolied.

                                                                      ' IE'. ;, ,~ " ' " W"..n
                                                                                                   " " ~* ' .MiB r- M 'dOMd G +*

f Or piEM sic' !'.'!. .. - c;;., r., 0 ..""B,h5I.O i .o. . j .a. ,..,.

                                    . ". .: a   o f v* E"<    t i M 4 ! a, ;"'   . . . ,. .,  *,,,
                                                                                                 ,,                                                                   t34 13                                          ;

{ U O $ i C,7 U S O k.". O h 0 I. 0 7 .. : ;.I @ WIIbS* Revision 0 July 11. 1980 1 AREA HIGH RA31ATION ; A. SYMPTOMS i

1. Control Room Panel 1913-3631 (2H13-Po01) Alarms:

f ] 'a. NEW FJEL ST] RAGE AREA RADIATION nl.

  .                                                      o.         REACTOR BLOG RA?IATION HI.
                                                           .        TURSINE SLDS RADIATION Hl.

c. AREA MONITORS RADIATION HI.

1 4

a. REACTOR BLDS TIP R03MS RA3IATION HI.

i

f. REFUELING FLO3R AREA RADIATION HI.

g. FJEL POOL VENT RADIATION HI.

2. Control toom Panel IN62-P630 (2N62-P500) Alarms:

4 a. CARdoN 3ED VAJLT RAJ HI. ., c. STATION VENT STACA RAD HI/HI-Hl.

c. BUILDING VENT EXHAUST RAD HI.

3. 3t,er Symptoms.

i a. Any Area Ra3 vonitor Reading aoove Normal.

o. Routine Survey indicates Hign Radiation.

I c. Racnaste Suil ding Hign Raci ation.

3. AJT7MATIC ACTIONS

1. None.

C. IuMEDIATE OPERATOR ACTIJNS

1. As necessary.

3. NOTIFY and EVACJATE Personnel.

e ,- - +n. ,. , <-mx-.- --,y..-~ - , , ,---n- . -- ,-y , - - , ,.-w,. -

LGA-13 j Revision O July 11, 19o0 . 2

c. RESTRICT access.

I

c. h3TIFY Shift Engineer or GSEP Station Director t o c l a s s i r' y tne event and Initiate GSEP it recuirec.

3. NOTIFY Rad /Cnem to survey and samole.

D. SJBSEQUENT OPERATOR A C T I O '4 5 O d

1. LOCATE the source of Radiation using tne folloning t

information: 3 CHECK Radiation Monitors for 30 normal reading.

c. CHECK Stack Gas Rel ease Rate.

c. CHECK Off-Gas Release Rate and Fl ow.

4

d. CHECK Area Tenparatares and Leak Jetection

, Systen T e.np e r a t u r a s .

e. CHECK Continuous Air ionitors.

,                         f.      CHECK tne Areas for visible System Leakage or Loss of Snielding.

t

g. C H " ". K Areas witn dortaDIe Radiation Monitors. '

2. STJP tne cause of the 4adiation. If tne cause is j a leak, ISOLATE the leak if possiole oy using nanual Stop Valves, or oy snutting cown tne affected system.

Consicer using temoorary sni elding i t necessary.

3. Reauest Raa/ Chem to SU4VEY ano ESTABLISH a Controllea Area. wnen a controlled area is estaolisnea.

personnel are no longer requirec to limit access. 4 ASSEMBLE PERSONNEL wra may nave Deen exposec anc read tneir dosimeters. If necessary. restrict tneir furtner exposure and nave film Daages develooed.

5. REFER to tne following procedures if tne cause of tne' Area Radiation is apolicable:

a i a. Fuel Element Failure; LCA-16. y y-- ew'- g m:-+-, cw- =sy.r ,y g - S se +---- 7'-ur- -e

LSA-13 Revision 0 July 11. 1900 3;ffnq

o. Cast Leak; LGA-31.

C. Slo. Leak; LGA-02.

d. Major Steam Lea <; LSA-C3.

e. Load Reduction for Off-Gas Emergency; LJA-GJ-01.

~

f. Coeration of Control Room HVAC Curing Raciation.

S noxe . or Cnlarine Detection; LJA-VC-31.

g. Doeration of Auxiliary Electric Equipment nvAC During Hign Radiation, Smoke or Cnlorine Detection; L0a-VE-01.

n. Filling the Goerating Reactor aall anc Drfsr Separator Pit; LCD-FC-09.

E. DISCUSSION Annormally ni gn area radi ation is an indication or nany possiale conditions. A l oc al situation may cause nign area ra3iation; system leaks and movement of contaminatea waste are examples. iign raciation in several areas is a symotom of -na j o r leaks. major equi ent failure, and scr eadi ng c onta ni nati on. A0 tion is takan to mininize p9rsonnel exposure, stoa tne source of tne radiation, stop tne spread of contamination anc clean up the contamination.

A These docurnents fo7 uaining purposes only* The1, are not controlierr -

                                            .    'IO 0ni BUth0iIZed                  L 3 A - v.

for plant og]c7,3Q'T9

                                                 *..'.Gk0n3nce uso."**'               Current res ;-i00,c are gygg. 33~",.? . le ;,i C "~t '-l rh                        July 11        lodo files for US* in 0      -                           rSata!Ilte              1 I H II4EJ            N)VITY A. SYMPTJMS

1. ;catinuous Air Monitor alarm.

2. Port 3 Die Air Monitor sdmale aDove limits for an antontrolIsc area.

3. selding. flame catting, grinding or neating of naterials <nown to ce contaminatea.

4. Hign Area Radiation w onitor alarms.

d. AJTJMaTIC ACTIONS

1. Refueling Flaor Ventilation E r.n a u s t or Reactor 3uiloing ventilation Exnaust System High Raciation will Auto-Start tne StJnCDy Cas Treatment Systam ano Isolate tne Reactor 3uil Jing Ventilatian System.

C. IMME0! ATE OPERATCR ACTIJNS

1. As necessary.

a. N3TIFY and EVACJATE Personnel,

o. RESTRICT access.

C. NJTIFY Shift Engineer or SSEP Station Diractor to classify tne, event and initiate GSEP if required.

. NOTIFY Rad /;nem to survey and sample.

Q. SJBSEJUENT JocRATOR ACTION 5 )

1. STJP the cause of the 4adiation. If tne cause i5 a leak. IS3 LATE ene lean if possible oy using Manual l Stop Valves or by SHJTTING DOWN the affected s y s tein . )

2. Request Rad /Cnem to SURVEY ana ESTABLISH a Control led Area. Wnen a controlleo area is estaclishec.

Dersonnel are no langer required to limit access.

3. If'the Airnorne Activity can be diluted or fi1tered sitnout spreacing, MAXIMIZE Ventilation fl ow to tne atrectea area.

LGA-1+ Revision 1 July 11, 1980 2 'ena& j 4. If tne Stack Gas R el ease Rate is approacning Limits (T.S.) or if Airoerne ;ontamination is spreaaing tnroagnout a DuilJing, C3NTAIN tne Contamination. ! CONSIDER stopping ventilation as follows:

a. Turoine Builcing in accordance nitn L39-VT-32. :
: 3. Off-Gas Builaing in accordance with L3P-V3-]2.; c. Radwaste Builaing in accordance witn LOP-va-03.

l . d. Macnine Snos Suilding in accordance witn LOD-VJ-02. I i e. Aux. Building in accorcance aitn LOP-VV-02. (

5. I f tne conta.ni nat i on is in the Reactor Builcing,
CONSIDER startinj the Grimary Containment Filter i s

Train or Standoy Gas Treatment System to ventilate l the Reactor 3uilding. VERIFY Reactor 3uiloing } i 4 Ve.ntilation System Isolation, if applicaole. ' 5

6. If the Control Room or AJx Electric Room are Airoorna, REFER to L30-VC-35; Startup and Shutdown of Control Room HVAC Emergency "a<eup Train.

j E. DIS;USSIO1 1 ;Aonormal Airborne Activity is a symptom of AirDorne Rel ease of Particulate, Iocine, or Noole Gas Activity.

!                          Action is taken to mini ni ze personnel exposut, e l i mi na te j                           the source ano spreaa of contamination ana effect clean-us.               Analysis of the sanole will nelp to determine the source.

Since Airoorne Contamination results in an internal aose.

                          .use of. respirators anc face masxs may De necessary to minimize internal dose.

e

                                                   - . -     w-          m ,m . p   - >--e , - w, m~           - ~      w-p=

                                                                                                        .p -               <

These docua',s,. , . 7, . .tsic. b;a"., 4.,.gan' e pUmeera' QUkV.c.,,, e - --- S, C , .:. . a,-"..m-* " orro I h 6), ,'. .$-,- g12' _ ,.j e e ,a 7,0 ( U i.:v. . W ' g t' v,.s.>.

                                                           ' ' , "d'y,,

] ( C 7 p ,i 3 I _; ,' ' ,: "9

                                                                   , ,   13   .
                                                                                           , ;. . . t '{;' , 3;, Q[ pcc, as , itt t ye,v A"'*           ".       v...'.     .

LGA-15 I4~'**;; E N ,u,i:

                             ,,,-r-
                                                                                   ',       . . s :. .: A :, -

p"t,7,1. r < .s v U !*- *

                                                                                                                                              Revision 1 illcs ici US E' IU N,, ','

July 11, 1930 , 1 RADIOACTIVE SJRFACE CONTAMINATION A. SYMPT 3M5

1. WoJtine survey or Smear,satDie indicates surface contamination.

2. Ra3icactive water spill or lea <.

3. Contaminated material inadvertentiy moved from a controlled area in violation of contamination control orocedures.

4 Imoroper handling of raciological orotective clotning.

5. .ielaing, flame catting, grinoing or neating of

                                             ~ ' contaminated materials.

3. AJTJMATIC ACTIONS

1. None.

C. IwHDIATE OPE.4ATOR ACTl]NS

1. As necessary.

a. NOTIFY and EVACJATE Personnel.

o. RESTRICT access.

c. NDT[FY Shift Engineer or.GSEP Station Director

. to classify tne event and initiate GSEP if

requirec.: d. NOTIFY Rac/ Chem to survey anc sample.; 3. SJ8SE.)UENT JPERATOR AQQy; 1. -Reauest Rao/ Chem to SURVEY and ESTABLISH a Controllea A r e'a . Wnen a controlled area is establinsed, personnel are no longer requireo to limit' access.

4 2.' OE;DNTAMINATE personnel uno nave oeen contaminatec.

3. DE"ONTAMINATE the Area.

LGA-15 R3visio7 1 July 11. 1980 2' Hag E. DIS ussfos Racr oacti ve Sar f ace Canta ni nati on could De present cue to small or l ar ge contani nat i an releases. Spilling a sa-.la sottle of contaminate- liquid. .elcing or grinaing o.. :o7tsminated systems anc seaxage from certai, systems are soma arobacle causes. Actio1 is cir3cta0 towarc s topping the leak, warneng personnal. isalating tne casse anc ninimizing tne saraac. Tne person fincing tna contani net i on shoulc cetermine tne source of tne conta ni nat i on and eliminate tne source of the contamination, if possiole, e.g. u;rignting a spillac samole cottle.

                         **;f.

Io C' -. ., ...

                                                   . . , f.,    /. ,   . . . a .- ; , : , 7. ; . . ,, , , -                    t
                                                               ..-.t..               b.4..6..-        .    ..y,...a c .u     g ;,f.
                          ,s..,._,                                                                                ,.
                                                 ...,...7-6           .'..               .
                                                                                                 ,.g.,
                                                                                                . . ,   m..
                                                                                                                        .,...,-.,.4

7. .

l '. ; , , . . ; . L .. ; , t m '. 7 .. ..

                                                                                           .            .. -    :     g. ".1. w ."

LGA-13

                                              .          . s.    .       . _ . - .. -. .                     .

R isi n 1

                         . ? ' , , .: . gs., j                    *
                                                                                              -- ;
                                                                                              ...-.....u..v.....
                                                                                                              , - : I i.... 1%.
                                                            ..L c..    ,

e.

                                                                                                                         .                              July 11, 1980 1

FUEL EL E ME NT FAILURE l

1 A. SYMPTJus

1. Control Room Panel 1No2-P60C(2N62-9600) Alarms; a.> OFF GAS SYS OJTLET ANJ ORAI1 ISJLATEG.

o. OFF GAS POST TREATMENT RA0 dl.

c. CARBON SED VAJLT 4A0 HI.

c. STATION VENT STAC< RA3 HI/HI-HI.

e. BUILDING VENT EXHAUST RA0 HI. _ - - -

I

f. OFF GAS SYSTEM DILL ISOLATE.

, g. TECH SPECS LIMIT WILL SE EXCEEDED.

!                                   2.          Jt7er Alarms:

a. ORYWELL SUP CHAMBER RADIOACT HI.

o. MAIN STEAM LINE RAD HI.

c. Any Area or Ventilation Exhaust Raciation Monitor Alarms.

3. Jtner Symptons:

a. A significant increase in Off-Gas or Stack Gas Activity not attricatacle to pl ant evol ut i ons .

o. An increase i7 long-lived activity followec ey a cecrease in tne ratio of short-livea activity to long-lived activity (Off-Gas grao sample).

c. An increase in the Iodine isotopic Content of tne Reactor water.

d. An increase in tne gross gamma activity of the Reactor Watar.

j e. Increase in Dose Rate from the following alant areas: e 6 . - , , . , - .

LSA-1s Kevision 1 July 11. 1990 2

1) Turoine Cloor.

2) Feecaater ieetar Area.

3) Stean Heacer ano "oisture Seoarator Area.

4) Concensate D am i ner al i z er Area.

. 5) teactor :eec Pumo Area.

6) Condensate Puma Area.

7) General Plant 3ackground.

8) Primary Containment.

9) Steam Tunnel.

13) Reactor Suilding.

d. AJTJMATIC ACTIONS

1. Possicle Main Stean sign Raciation (3x normal)

Actions.

a. Peactor SCRAM,

3. ISCLATE Vessel anc Containment Isol a ti on Grouos i anc 3

2. oossiDie Off Gas Post Treatment High Radiation Actions.

a. Off Gas Systen I sal at i ons.

1) Off Gas 3iscnarge Valve

2) 3ff Gas Condanser Drain Valves

3) Hola Ua Line C. IwM:OIATE OPERATOR ACTIJNS

1. If Main Stea.n Hign Raciation is reachec, verit'r actions, ca 10T :ontinue tnis orocedure, and refer to LGP 3-2. Reactor Scran.

2. If Main Steam High Radiation is NOT reacnec, verify 3ff Oas Charcoal Aosorcer Train is in service.

1 LGA-lo Revision 1 July 11, 1930 3

3. As necessary,

a. NOTIFY and.EVACJATE personnel.

o. RESTRICT access.

c. N3TIFY Shift Engineer or GSdP Station Director to classify ano initiate SSEP if requireo.

o

d. NOTIFY Rad /; nam to survey and samol e.

D. SJBSEUUENT 30:RATOR ACTIONS 3

1. 3BSERVE cne Continuous Air Monitors for trencs.

2. CHEC4 local areas for increasec Oose Rates during

, operation with cofactive fuel, sucn as tne Turoine i Area, tne Reactor 3uilding.~tne Reactor Water Cleanuo j De ni ne ral i zer Area, the Condensate Demineral i zer area, tne Feedwater de3ter Compartments, tne Ory. ell and Turcine Building Atmospnere Sampling Statians. and the general oackground around the Plant ventilation Stac<.

3. IN;REASE ene frequency of analysis for Reactor water Onamistry during ooeration witn defective fuel.

4 dnan shutting down and opening tne Reactor vessel with a known " l e'a k a r a present, soecial care should De taken to vent tne Reactor Vessel Head correctly

;                           to minimize tne release of fission gases during nead renoval. Any nandling of suspect fuel elements should be cone wita caution. Reactor Water should oe kept as cold as passiole during the entire 4

operation unless element sipping is to oe cone. Tne following items are nethods to minimize raciation ex>osure and snoald oe includea in tne normal snutcown for refueling sequence:

a. SHUTDOWN the Reactor in tne nor nal manner.; o. PLACE tne Primary Containment Vent anc Purge Systen in' service.; c. MAINTAIN Reactor Water Temperature at or aelow 105 F for at least 24 hours oefore renoving

tne Reactor vessel Head. i I

L3A-lo Revision 1

July 11. 1980 4

i CAJTIO'4 Jo not lower Reactor nater Temperature oelow 9]OF if tne Reactor Vessel Mead Bolts are unaer tension. 3 M3 NIT 3R the Reactor SJilaing atmospnere continuousl / dhen ooening tne Reactor vessel i to atnospnere.

e. PROCEED witn Jnoclting the Reactor Vessel Head if air sample activity renains less tnan tne l Maximum Per ni ssi cl e Concentration (MPC).; f. TAKE frequent air Samol es cur ing uncolting of tne Reactor Vessel Meaa.; j. REMOVE the iesctor Vessel Head usinj filter masks or indeaeacent air suopliea masxs for personnel protection as recommencac by the Raa/Cnen Deoartnent.; n. After sipoing, RE40vE tne susoect fuel anc PLACE it in tne saecial Dasket for transf3r to tne Fuel Storage Pool. On r ec o:nm en ca t i on of ;
tne Rad /Chen 3epartnent. tnose elements icentified as " gross lea <ers" shoulc ce placed in sealaa

,                          containers and stored uncer water until sniapac offsite.

j. E.- DISCUSSION A massive fuel element failure shoulc not normal l y occur unless a severe condition anich exceeds fuel strain l imits or enernal l i ni ts occar s. Autonatic action is initiatea oy the Reactor protection anc Reactor Containment Isolation

! Systens. Action is directed'toward ensuring the aporopriate automatic action and taking steps to prevent environmental release. Sucsecuent recovery includes assessing tne , camage, c omo l et i ,:t shJtcoon, and preparing a corrective

              . maintenance plan.

A minor fJel element failure will normall y De inciC3ted oy increasing coolant fission procuct activity anc oy ) increasing area ano stack rel ease raciation levels. Posi tiva acti 36 is required to nininize tne effects'of L tneLcondition. Consiceration snould ce given to recuting loa 3, or to snutting co.n anc replacing tne leaky fael. Routine coerations should acaitionally De conductec so I m- t-etl , u.-w-,- -q Y w g

LSA-lo Revision 1 JJ1y 11, 1480 5(final) as to m i n i ra i z e aersonnel exaosure due to changing pl ant r ac i a ti on levels. Tne attiun fol1owing a ;3'iPLETE CORE FAILURE IS NOT C 3v 8 E0 IN THIS PROCEDURE. sioce it is covereo in GSEo. O 1

                                                                                                                                              ;b
                                                                                          . ? ,, - e , L,. , n :. .- ',. O
                                                                                                                      ~
                                                                                                                                   .~ ~; * ~ .};
                                                                       ~
                                                 ,_,...-~~r i : " 't -                        *,                    ,

ce s .. r,, ... -

                                                                                                                           ,;                            ,

i.

                                                                                                                  ~
                                                                                                                               ,,           ,   ,,=.1      ,
                                                                ,                                               ..            i.    .

u' e d, = k

                            .              .                                                                                           , c.

L3A-17

                                                                                                        .      ,,,'v.                  "'*

Revision 2

                                                                                       ,i           '

July L1, 1930 1 SOMS THREAT RESPJNSE A. SYMPT 3MS

1. Sono tareat receivac.

S. AJTJMATIC ACTIONS

1. None.

C. IwEDIATE OPER.ATOR ACTIJNS

1. If poone cal l:

i

a. Do NOT nang up anone.

3. Attemot to transfer to Shift Engineer.

2. Jotain and compl et e narc copy of checklist ( A t ta en.nent A).

3. Notify Snift Engineer or GSEP Station Director to cl assi f y anc initiate 'SEP as necessary.

i D. SJBSE0UENT 3PERATOR ACTIONS

,                                          1.            NOTICY Security of any support required.

i t.

CAJTION JO NJT touch or attempt t inspect or .nove a cevice suspected of oeing a bomo. Clear area anc request Securi ty to ootain cemol ition exper t.

l

2. CONDJCT Sono Searca in accordance with Station Somo

_ Sear:n Protecures.

a. Each Departmen. saoulo oe notifiec of tne enreat and possiole locations of tne device.

o. Tne preferred action oy each Geoartment is to have personnel renain-in their current location and review that area for any recent cnanges tnat may indicate tne location of the device.

c. Personnel snoald m oe Ossemoled in general areas.

m 7 W - - -

                                                                                                                               .-r                  w        4   - - -   =>

_. . ~. LOA-17 4evssion 2 J u l_ y 11 1960 2...

              .d . Tne Station Director ill evaluate the tnreat anc cetermine if any evacuation is necesserf.

E. DISCUSSICN Once a Dono tnreat has oeen receiveo, tne Snift Engineer must cetide wnat actions are necessary to protect personnel anc equionent basec on nis evaluation of the validity or o tne tarcat. The more iaformation ne nas availaole to him tne cetter nis decisions and actions will be. For tnis reason transfer of a como threat pnone call to nim saould oe attemated so ne can talk directly witn the person na<ing tne threat. Tne first decision the Shift Engineer must maKe is anat level of validity to attach to tne call. Next ne must Cecide .h3t verification of tne call is necessary sJCn as visu31 s i gnt i e'g o f or se 3rCn for tne DomD device. Once the location .,as oeen Jetermined, ne must weign the oossiole consequencec of the device cetonating and ta<e actions to mitigate tncse consequences. In acJition he nust decide anat assistance is needec from outsice sources. For exanple, tne inspection ana removal of a suspected conD must only be done by experiencec, qualified indivicuals. l

                        . Y:.~"(

s 7sa..*,:.. -. .

. - f V- : . : :;;.i,_ .;2.

                                                                                                           .     .x. .::.; ?: .:w~.

c. j+&. f.,

                                                                                                                                                                                                          , .~. , .. .. .y 3..m, ;.n. y- ., . .

w-

                                 .                                                                                                                                                                            LSA-17 ATTACHMENT A                                                                                               Revision 2 July 12.,1980 i
                   .                                                            BOMS THREAT CALL CHECXCFF LIST                                                                                           3 (final) i C Nurcer of rings 1.2,3 i

1. Date , and Time...

Leng:n i

    -                          2.         Exact wording of the threat:

4 i i . 3.. Questiens ::: ask of caller: ? a. When is becc scing to explode? i -

b. Where is .it right now?

c. What dces it lock like?

  -  !                                    d.        What kind of bceb is it?

e. Did you place the bomb? '

f. Why? m

g. What is your name?

h. What is your address?

4

     ;                         4          Cal.l er's: Sex                        , Approx. Age                  , & Race                                                                             .
     .                                    Caller's Voice:

c. CALM.... LAUGHING. LISP.......... DISGUISED.. ANGRY.. CRYING.. RASPY......... ACCENT..... [ ]' EXCITED NORMAL.. OEEP.......... FAMILIAR.. O . SLOW.. DISTINCT RAGGED........ IF VOICE IS FAMI:.IAR, I V RAPID. SLURRED. CLEARING THROAT WHO DIO IT SCUNO LIXE? SOFT. NASAL..- OEEP BREATHING LCUD. STUTTER _ CRACXING VOICE

    ,.                                    Backercune Scunds:

STREET...... HCUSE.... CLEAR......... .

                                       . CROCKERY...                               MOTOR....                            STATIC........

t VO!CES...... CFFICE... LOCAL......... P.A. SYSTEM. FACTORY.. LONG OISTANCE. MUSIC....... ANIMAL... 300TH......... OTHERJ...... Threat Lancua;e: 4' WELL SPCXEN.. INC0HERENT................... !

                                       . FOUL.........                                TAPED........................

IRRATIONAL...' MESSAGE READ BY THREAT MAKER. - 1 S. Remarks:

6. -Notify imediately the Shift Engineer (Ext. 202 or 203) or curing regular daytime hours the Super.intencent (Ext. 212).

c O 7. Fill cut completely: Name Position Date Phone Numcer

           "                       .             * ' * * *            . m6 e m o'          e.-  e                 __ _ . ewm e ,    em m o m ,,, ,,                            y                                    . , _ , , ,

w -

                                                                ,    - . - - .      -    i.m          .
                                                                                                                        - . , - .                                     p          , - -                  - = - - . ,                v.,.         .      =   e   -

b' E-e. 1. ,,'" r* [\*.e s.n Vo'* e rr^ %j*

               ,, ,, .. f r e

u *-c. r.,, y.ta..u+,% -

                                                                                                                                  .. .3;;;  .- r ,_
           .y        ,
                                                                                                  .,,~r. 'y_                     ,
                                      ,           ,      *x=
                                                                                                         - '         w,
                                                                                                                                               *6
                                                                                                             ,.                      ,                    .s.      ,
             <l""[                                                 ,,          ,
                                                                                                                                    ',. m. d.1
                                                                                                              - ** .. "' .-'~..
                                                                                                                           . . v v ,e ,i L;A-13
                                   , , , - ,,                        ,,L.           .l. . -          4

Revision 1

                ; . .v'. . v '-             ' " ' "

JJ1y 11e 1C00 1 .

TRANSIENT WITH FAILURE TO SCRAri A. SYMOTJMS

1. The following true signal (s) aae to a Reactor transient are indicate] oy alarm or indication ano tne requirec full scran does not insert control rocs
.' as indicatea on tne full core cisplay, roa posi tion orintout on the conputar, or four roc di spl ay: a. Reactor Low Water t evel 3 (12.5").; o. High 3rywell. Pressure 1.69 osig.; c. Scram Di schar ge Vol ame High Water Level (Scram Alarm).; d. Pain S t a a.n Line Pign Radiation (lateru).; e. Turoine Stoo salve Closure.; f. Control Valve Closure (Turoine Generator Load Reject).; g. Hign Power Thernal Trip (flow Di3 sed AP4M).

i J

n. ' H i gn 'Jeut ron FlJx (120% in RUN. 15 in Startuo).

5

i. MSIV Cl osu r e (l ess enan 90% open in Run).

4

                                                  .J.            Nign Rea. tor Pressure aoove 1043 ,) s i g .

2. Reactor pressure ana/or neatron flux indication o

                                                  ~in:reases abruptly, and may go off-scale on recorcers and neters.                             This is tne key indication to recognizing an ATwS event.

3. Saf e ty-Rel i e f Valves may Lift.

4. Ocner. Indications:

a. . Increasing 3rywell Pressure anc Temperature.

o. Increasing Suopression Pool Temperature.

t- c. Possiold increase in containment radiation n 1evals. i

    ._   .        .                     _ _ _ _               ..a.___             - _ . . _ . _       _, , ,                                 . . . _ _ _ _ _ - , _   _ - .      _ - _ _ . _ _ . - . . . _ .

LGA-lS Revision 1 July 11, 1960 2

d. Possiola Hign Stacx anc Off-Gas release rdtes.

a. Recirculation Pamas trip sitn jet pump anc core flow cecreasi,g.

d. AJTJMATIC ACTIONS

1. IC7oc Reactor vessal Fressure anc acove actuates various safety relief valves, .ith possiola lo.

level setoac< ( L *. 5 ) .

2. 1120: Reactor Vessel Pressure initiates Reactor Recirculation pump TRIJ (ATWS).

3. Level 3, Reactor vessel aater Level 12.5" Actions:

a. Reactor SCRAM signal.

D. TRIP Recirc Pamps to 15 Hz.

c. ADS Permissiva.

3 ISCLATE vessel and containment grouas 4, 5, 7

a. SETSACK programmec level control to 13".

4. Level 2. Reactor Vessel aater Level -50" Actions:

a. AUT3 START RCIC.

a. A 'J T J 3 TART HP;S anc H?CS Jiesel Generstar.

c. T4IPS R. circ Pu nps OFC (ATWS).

c. ISCLATE Vessel anc Sontainment Groups 1 2, 3.

5.

5. Level 1, Reactor Vessel aater Level -129" Actions:

a. AUTOSTART RHR LPCI Moce.

a. AJTOSTART LPCS.

c. AJT3 START Dicsel Generator 3 and 1A(2A).

a. AJS Permissive. ,

m S t l' L5A-13 Revision 1 l Jul y 11, 1990 i 3

6. 1.59: ;ontainment Pressure Actions:

i

a. ISCLATE Vessel anc Containment Groups 2, 4, e.

7. Also Group 9 if 570 Reactor Pressure.

i j o. Reactor SCRAM signal. i 1 C. IuMEDIATE OPERATOR ACTIJN5 .I l.- Manual l j SCRAM Raactor. I a. ARM anc CEPRESS Manual pushouttons. .I

c. PLACE moce switch in Shutdonn.

c. If S C R A '4, all rocs in, and coder decaying, go to LGD 3-2, Reactor Scram, and ao NgT continue tnis.orocedure.

2.- dithin C3E minute of tne start of the event, if no 5 cran: i a. START 13TH SSLC PJmos.

o. VERIFY RWCU ISOLATION.

 ,                 3.        VERIFY actions if reacned.

i 1

a. 107sn Safety / Relief Actuations.
o. 11200 Recirc ?unp Trio. I: c. Level 3.

4

c. Level 2.

e. Level 1.

f. 1.690 Containnent.

4. MAINTAIN Reactor Pressare ana Level.

S '. ditnin'Tsy minutes of the event:

a. START.RHR Service Water A E B Loops (LOO-RH-

                                  '05).

, 3. Place corn-RHR A 6 3 in Suppression Pool ;ooling l Mode (L3P-RH-15). )

/:

'

i

          .       . . ~ .       .

m___.- ,-- . ._.. ..~

l + LGA-13 Revision 1 July 11, 1980 4

3. SJBSE2UENT JPCRATJR A ; T I C ':S

                       =1.      VERI:Y tne following inoications:

3. Reactor power is Jecreasing.

3. Feactor pressure i s cel ow 1076 asig.

c. Reactor level is 3oave Level 2 (-50"). CONTROL Reactor level .ita RCIC, IPOS, or Feedwater floa, if possiole.

d. Cor? flow cecreases to natural circulation level, consistant witn operating nap.

4

e. If containment oressure reaches or exceecs 30 l .psij, USE-wetwell saray to CONTROL containment pressure.

l t. Botn Stancey Liquic Control Pumps are running i anc Stancoy Ligaia Control Solution Tank level

,                                               is cecreasing.

J CAJTION j l. De-energizing RPS cusses will result in a loss of neutron monitoring instrumentation.

2. Tne folloning attempts to scram the Reactor

;                                               are to De performec concurrently if manpo-er
,                                               is sufficient.                        .,
    .                   2.      CHEC( Control RoJ Dositian Incication. .If tne control rods have not fully insertec, PROCEEJ to scran the Reactor in tne following order:                                                                 i

a. DE-ENERGIZE R35 Suochannel Logic Oy ooening

<                                               oreakers from control rocn panels lH13-Po09 anc 1413-P611 (2H13-P609.anc 2Hl3-P611).

o. . TRIP RPS Scran Logic dreakers CS-2A anc C3-23 at the RPS Distrioution Bus in tne Auxiliary Electric Room.

c. OPEN tne following RPS Po er Supply Brea<ars i for 2 minutes lacelly at the RPS Power Suoply l Buses anc MS Set control.canels anc tnen ;E; LOSE:

s _ _ - _ . , .. _ ,_ , _, , . . - , , . _ _ _ . , - ,,m. .

LGA-lS Revision 1 July 11. 1980 5

1) RPS MC Set "A" output creaxer.

2) .RPS MG Set "3" output areaxer.

3) CB-1, RPS Alternate Poaer Suoply Sreakar from MCC 1325-1.

d. INDIVIOUALLY SCRA1 Control Rods at Local j , hyoraulic Control Units (HCJ's) cy olacing Brancn Junction Module Switenes to tne Scr a n-Test oosition.

e. ISOLATE air from tne scram air system by closing IC11-5095 (2C22-F395); Scram Air Supply Valve.

f. MANUALLY INSERT Control Rods from Control R oo-n Panel 1H13-P633 (2H13-Po03) using t.ne Reactor 4

Manual Control Systam. f

3. CHEC< Stacx Gas Release Rate anc Off-Gas Release Rate. INITIATE 3SEP if necessary.

4. AFTER the Reactor is snutacan to tne level wnere the only source of power is aecay neat, PRJCEE3 to stabilize Plant Co7aition in HOT SHUTDOWN as fallows:

i

,                                           (AJTION Do not snutcown SBLC injaction once it has oeen started until tne SSLC Solution Tank is verifisc to se enpty.

t

a. SAMPLE Reactor Coolant frequently to VERICY Boron. concentration is aDove tne level determinea to, maintain tne plant shutcown (acove 750 pom

                             - itn all roas oat).

c. PERFORM eitner step 0 4.D.1 or 0.4.o.2 as follows:

1) MAINTAIN Reactor at 1000 to 1050 psig oy operating a Relief' Valve.anc removing neat

                                   'from tne containment using Suppressian Pool Cooling.

J

2) MAINTAIN Reactor at 1030 to 1050 psig as follows:

4 c -. _ , , , , , . _ , _ _ _ - . . , _ - .y, - . , , , , y---

p hwae6 e m+ e a + = 4

LGA-13 i Revisisn 1 July 11, 1960

,                                                                                                  6 a)    VERIFY that Goron concentration in tne Reactor will oe sufficient to maintsin :ne R e ic to r shutdoun after accounting for a normal startuo of tne Steam Condensing Moce of RHR.

, 0) STARTdP tne Steam Condensing Mode of i RwR in Sccorcance .itn . LGP-RH-07,

    ,                                                     Steam Condensint; Startuo anc Goeration.

i l S. anen tne Reactar is to oe snutcown to COLD SHUTD0aN. 4 PR3CiEU using the following considerations: I

a. Sufficient negative reactivity nas oeen inserted to tne Reactor to account for tne positive reactivity effects te nper ature and oilution.

l i E3 ! Carryover snoal a not s i gn i f i cantl y affect Reactor Boron concentration. I

c. If the plant is not contaninated and the Resctor is not isolated, a normal snutcown and coolcoun

 ;                                            in accorcance wita LGP 2-1 can ce performec.

c. CAUTION must ce taken that tne unoorated water in the AHR Snatoo n Cooling lines coes not temporarily cilute the coron in tne core to allow criticality as follows:

1

                                             -1)    ESTA3LISS an excess doron Concentr3 tion to accomodate for tne effect of RnR dilution

(30% excess is require] aoove tne 75d-1000 i PPM concentration).

i

2) STARTUP the Reactor Recirculation Pumps in SLOW spee3 in orcer to nomogenize vessel'

.                                                   Boron concentration. If tne Primary Containment is isolated, tne isolation.

signal mast ce reset in order to suppl y RSCCd to tne Recire. Pumus.

3) START RHR Shutcown Cooling floa to tne Reactor vessel gr aduall y oy tnrottling 3 PEN tne Snutcown Cooling Injection Valve
(The RSR'Pumo' Minimum Flow Valve must'oe 1

V

LGA-13 Revision 1 Jul / 11 1930 7 averridden in tne closed position to prevent the loss of coratea water).

c. Do not exceea a 130 F/hr cooldown rate (TS).

e. Wnen fl ooding ne Reactor vessel up to tne steam come, use a source af -oter Dorateo to tne same concentration as tne -ater in tne

 ,                       Reactor to arevent Reactor doron concentration ailution. The SELC Solution Tank can ce ased.

However. if using tne 53LC Pumps, Reactor floccing will reqaire one to two days. An alternate pumo can ce used.

f. Concentratian l evel s of Baron in the Reactor Vessel will oe 750-1000 pom. Tne minimum soluDility of Soron in water at 32 F is greater tnan 5C00 pam.

g. If a fuel element failure is suspected, refer to LGA-16, Fuel E l e me nt Failure.

E. DISCUSSION An AT45 (Anticipatec Transient without scran) is extremely unlikely 3ut will require pronpt operator action to mitigate the consequences. Goerator concerns are as follows: 1) VERIFY that Recirculation pumps trio, 2) Snutdow, of Reactor. 3) Li ni t Reactor peak pressure 4) Maintain the core coverec, 5) Limit the tenperature or tne Suppression Cnameer, ana 3) Long-tern cooloo.n. Tne caerator must attemat to scram the Reactor with tne most readily availaole neans. Jpon recognizing tnat the Reactor does not scran, tne operator snould INITIATE STANO3Y LIQJID CCMT.ROL LS R L C1 W I T H I'J Tn0 MINUTE 5 0F THE EVENT to vi ni.ni ze Reactor p3.er production, wnich oaulc heat-up tne containment. HACS or RCIC operation is necessary to maintain tne core covered if' feed flow is stooped,.and should be initiated if Level 2 (-50 inches) is approachec. Suppression Pool Cooling using two RH4 , Heat Exchangers must ce initiatea to ensure tnat Suppression Cnamoer temoerature limits are not exceeded. Subsequently, tne operator nust insert enough negative reactivity into tne Reactor so tnat an uncontrolled restart will not occur. Thus, a cooldown must not ae initiatea until Control Rocs are insertec or Scron concentration is determined sati s f actor y. The consecaences 9 -

                                                  .,               +                 --

LSA-13 Revisier 1

July 11, 19%C 8 (final) of-this accident to tne containment anc environnent must j be eValJated. SSEP snoJi a ce i ni ti ated if necessary.

i Assaming 100% Reactor power and a two minute tine celay to the start of Boron injection, Suporession Pool temoerature will ceak o at 177 F after 28 minutes oita tne

MSIV's closeo or 105 F after too minutes with Bypass Valve capacility. Containmant aressure will oeek at 3.5

  ,     psig -ith MSIV closure or .3 asig witn dypass Valve capaoility. .i o t snutcown snoul 3 oe acnievec witnin 13 minutes of 3oron injection. Ho.ever, once coron injection is started, it must ce run to complation: 00 NOT 580TC0dN SSLC JNTIL POSITIVE VERIFICATION THAT THE S3LC SOLUTION TANr. IS EMPTY.

It nust ce notec that FAILURE OF A MANUAL SCRAM aITHOJT , AN A3 NORMAL TRANSIENT REOJIRES THAT REACTOR RECIRCULATION PdMPS REMAIN OPCRATING TO EXPEDITE BORON MIXING. Tne release limits of 10 ;FR 103 apply to tne ATWS event. 1 l I 1 i J

Th650 dOCUnlentYiOT't?Efding purp0SeS ODIy. PAGE ] __l_E:__N_:_________T_hyJMe not controlled. They cre noj_authosed nev 0 ATE DISKETH ___ __

 ' '    ^^~           for plant operation er maintensne use. Current _

reNsidrE are z.valisi;ic ir: Cutral Fik or 3ateRfte */79 LOA AA-02 ggIggR {0g g jg Rpfg pjjtjgg, 00 I1/78 01 Of t/k5 0 ) h ,a L LOA AA-03 FAILURE (H( DIKE ee- 11 / 7 5- 01 LOA AA-04 ACTION TO BE,TAKEN IN THE EVENT OF AN C/ llll5/ 7C1X01

                                                                               -00'    --l-r79 OIL SP jT(                         015 RIVER OR ON THE     Ilt.d NDIP[fiR LOA     AA-05               At T O                              CONTROL ROOM of
                                                                              -Ge-s///Po6L H77T'         01 LOA     AA-06
                                                              /

ACTION TO BEd T-AtfEN /f{ gE EVENT OF AN OI IIdlSO.L

                                                                               -etr    M-/+8        01

~ y OIL SPILL NjM VCd&LIGG LAKE q- Ol iI?O5L- 1 LOA AP-01 LOSS OF SYSTEM AUXILIARY TRANSFORMER, -4119' 02 l SAT 142 (242), OURING POWER OPERATION LOA AP-02 FAILURE OF BUS 141Y (241Y) OR BUS 142Y Ol ll?OK 02

                                                                                                         \
                                                                                   /-R (242Y) TO TRANSFER TO UNIT AUXILIARY TRANSFORMER, UAT 141 (241) UPON LOSS OF POWER FROM SYSTEM AUXILIARY TRANSFORMER SAT 142 (242)

LOA AP-03 LOSS OF A 4 KV ESS BUS Oi ll2O A 00 - *-tT9 02 LOA A 04 LOSS O NON ESS BUS Ot IIfDSL. 02

     /][gDie ACTMLQ C/t %.f Ahfr"QQv,w GLL&L WG^vnm CQ
                                                                                      -477{w%X, c2/

LOA AR-01 ACTION ON Aff ARE A RADIATION MONITOR 00 10/77 01 ALARM ol //00SL-LOA CO-OL LOSS OF CONOENSATE PUMP -ett W 01 LOA C6-01 LOSS OF CYCLE 0 CONDENSATE GLAND WATER O/

                                                                             -00
                                                                                      //$W
                                                                                     -Hr7tt        01 l

LOA CP-01 OPERATION OF THE CONDENSATE POLISHING 00 3/79 01 SYSTEM FOLLOWING AN EXCESSIVE CONCENSER TUBE LEAK LOA CP-02 CONTINUED OPERATION WITH EXCESSIVE ot t/2O3-GO 3711 01 CONDENSATE POLISHEO DIFFERENTIAL PRESSURE LOA CW-01 O/ / / 7M_.

LOSS OF ALL CIRCULATING WATER PUMPS -00' 1-171'7 01 4/ ///796L LOA Ch-02 A1540RMAL CONDENSER n' A T E R B OX 10/77- 01 OIFFERENTIAL PRESSURE LUA CW-03 C/ /// 7E CIRCULATING WATER HIGH INLET 16fM' 01 T E M F'E R A T UR E LOA CY-01 HIGH CONOUCTIVITY IN CYCLEO CONOENSATE 00 11/77 01 STORAGE

6/12/79. LOA INDEX PAGE PROC. NO. TITLE REV. REV OATE DISKETT LOA DC-01 pi it / ~79&- 250 VOC SYSTEM FAILURE 40- 11 f i7 01 LOA OC-02 125 VOC SYSTEM FAILURE 00 11/77 01 LOA 0C-03 Of 48/24 VOC SYSTEM FAILURE I {SObL 4-&ft7 01 LOA EH-01 Of

                         -M AtFtfMC-T1QN TJF~ THE PRC55UR t t, uN TROL-                    I {{ "

S ETftt--HL.LO<3, MOg -@_tl ^>E U-PJ

                                                                               %          W            O2 LOA     EH-02                                                                 OI         i f0b' -

TURBINE CONTROL VALVE FAILURE LOA el-03 FC-01 myi Lpxf -3 f M 02 LOSS OF FUEL POOL COOLING o/

                                                                               -69
                                                                                         /805 r779       01 10 A     FC-02                                                                 Of LOSS OF NORMAL LEVEL CONTROL IN THE                  -Ge
                                                                                         //b3 FUEL POOL                                                      .5774         03 LOA     FW-01 LOSS OF FEEDWATER HEATERS                             00        5/79         03 LOA     GA-01 LOSS OF HYOROGEN COOLERS 0/
                                                                                                                                                                       //MN C7-      01 LOA     GA-02                                                                Oi        14r$
                                                                                         //      S-LOSS OF GEN H2 TEMPERATURE CONTROL                  -OG-       -3 /7V        03 LOA     GC-01 LOSS OF GENERATOR STATOR CO                                      ts/'/93L.

65 -Ol Seam aeo / Euc<poro WlltalbLINGud . U f/p gGO-O/ 3779- 03 LOA HD-01 LOSS OF PUMPED FORWARD FLOW HEATER oi OO-i/8xsc.

                                                                                        +;rTV DRAIN                                                                        01 LOA      HD-02 OPERATION WITH REDUCED PUMPED FORWARD Oi
                                                                            - GO-
                                                                                        //[ W
                                                                                       +rNr          01 HEATER DRAIN FLOW

.0A [0-0.3 HS-01

              'OO f LOSS OF MAIN HYDROGEN SEAL OIL PUMP                   00       11/77         01

.0A HY-01 LOSS OF GENERATOR HYDROGEN PURITY 00 5/79 00 .DA IA-01 L05S OF INSTRUMENT AIR 0I

                                                                           -0 0 -

llSW 02 .0A IN-01 LOSS OF NORMAL DRYWELL PNEUMATIC AIR OI

                                                                           -7J        -3fl746 Dff SUPPLY                                                          4M           02

.OA IN-02 LOSS OF 1000 DYRWELL PNEUMATIC ol ShW AIR SUPPLY -*J777 03 OA MC-01 CLEAN CONDENSATE STORAGE TANK HIGH 00 11/77 CONOUCTIVITY 01 ! DA Hb-01 or -3 /2MC ! PRIMARY SYSTEM LEAKS -GO- AfM 03 . OA NB-02 og 5/&DSL 'l' FAILURE PRpP ER LOF A RELIEF Y Qr VALVE TG SEAT ,, lnC< Cli/Crf -ee Sf?9 01 OA NB-03 OF Ci SQ $6iy' hellC t bVIv' C tiA Q D O D INADVERT/tJTo g RELIEF y,Ald$ 0F A SAFETY 4-/t 03 JA Nb-04 Of / d-[tO1 REACTOR COOLANT HIGH CONOUCTIVITY 00- 4f79- 03 h 0I

                                                                                     //bi')$-.

i. E

   . EH-03 Aso.O 'y M. Oznvl. DaLutQ) qp5dZf gy yy @a Gw. Ag' u9tuhA 1isen dw_jx-HD-03     Ru) .O Yb               Loss      o-F H.P. F e e g ( p f ,.                       y f r, MbOV k 0Y2O bX=S of fl L.P. N&- c% -in 1

gp-os Sw.O %o HW. Dr Tnt. Le.ve.l H1/ oct) . l l I l I l t

               ~
                          ,-. ~       -.     ,            , . , , , , , - - -   - - . . , - . .   . . _ . , - . . -

PAGE 3 LOA INDEX 6/12/79 REV. REV DATE DISKETTE PROC. NO. TITLE ------ ---

--/

C2 //h)34- 03 NC-05 RECOVERY FROM AN ECCS INITIATION UNDER MWT 'T779 LOA POST ACCIDENT CONDITIONS c3 ll20 Sf- 03 4MF AwQir NR-01 SRM AND IRM INSERT OR WITHDRAW FAILURE LOA O/

                                                                        -e&-
                                                                                       //D9 C03
                                                                                      -vrPP LOA    NR-02            LPRM FAILURE /LPRM HIGH FLUX OR LPRM DOWNSCALE of iMr lol%L03 LOA    NR-03            LOSS OF NEUTRON FLUX INDICATION                                        L Gl             /

Of 03 LOA OG-01 LOAD REOUCTION FOR OFF GAS EMERGENCY 00 4/79 03 LOA OG-02 OFF GAS HYOROGEN EXPLOSION O& 4791r 02 LOA OG-03 ACTIONS TO BE TAKEN IN THE EVENT OF A jkh-- FIRE IN THE OFF GAS CHARCOAL ADSORBER O) TRAINS OG-04 LOSS OF ONE OR BOTH HYDROGEN ANALYZERS 00 3/79 03 LOA GD96L03 FAILURE OF THE STEAM JET AIR EJECTOR 00 En4HF LOA OG-05 STEAM PRESSURE CONTROL W Y' (d itC ALdC blVrt bqS. ll$)P/?[.fs ~~ , ? lor, . Cs' ICA r00C- 00 3/79 [ PWDI LOSS OF PRIMARY AN0/OR SECONDARY 03 LOA PC-01 CONTAINMENT INTEGRITY 00 11/77 01 LOA PR-01 RELE ASE RATE SPIKES AFTER POWER CHANGE 00 10/77 01 LOA PR-02 RELEASE RATE EXPONENTIAL WITH POWER 00 11/77 01 LOA PR-03 HIGH RELEASE RATE C>l I l-

                                                                          -G&                          03 LOA    RD-01            STUCK CONTROL ROD 00             5/ 9         03 UNCOUPLED CONTROL ROD LOA   RD-02
                                                                          &>l               { bl-      03 HWT LOA   RD-03            MISPOSITIONED CONTROL ROD 00            5/79         03 LOA   RC-04            CONTROL ROD DRIVE SYSTEM FLOW CONTROL FAILURE 00            5/79         03 LOA   RD-05            CONTROL ROD DRIVE STABILIZER 00                9        02 RH-01             LOSS OF SHUT 00WN COOLING LOA                                                                     Oj             3/[6)dhL.

l 4}& 3/4v 02 LOA RH-02 LOSS OF SUPPRESSION POOL COOLING 00 3/79 02 LOA RH-03 LOSS OF RHR SERVICE WATER 00 10/77 01 LOA RI-01 RCIC FAILS TO START ON AUTOMATIC INITIATION SIGNAL 00 11/77 01 LOA -RI-02 RCIC FAILS TO PUMP WATER AFTER AUTOMATIC START l

6/12/79 LOA INDEX PAGE PROC. NO. TITLE REV. REV DATE DZSKET T Of / /&dE/- LOA RL-01 FAILURE OF REACTOR WATER LEVEL CONTROL $9- afft 03 SYSTEM IN AUTO OR SINGLE Ol LOA RL-02 FAILURE OF THE TORFP M/A XFR STATION to- }{r PTl$.)bl~ 03 LOA RR-01 - AUTOMATIC TRANSFER OF MEACTOR 00 4/79 03 RECIRCULATION HYDRAULIC POWER UNIT pf/,.(,)S[(JyrA, ) $GCAbCW [0&f. [leCe g p,-Cw] (OtsA. FROM LEAD TO BACKUP SYSTEM W~OI LD ~S S O+ h Q( og nu.P l/h l lbDO-- fh0A > LOA SA-01 LOSS OF SERVICE AIR 3779- 03 i I l ~7CI S!._. LOA SC-01 INADVERTANT INJECTION OF BARON INTO 00 -0700- 01 THE REACTOR COOLANT SYSTEM OURING COLD SHUTOOWN LOA SC-02 7/'7 ML_ INITIATION OF STANDBY LIQUIO CONTROL 00 <ntTF 02 LOA SH-01 o/ // /~195L-LOSS OF STATION HEAT RECOVERY -OP" dh4WP 03 LOA TG-01 TURBINE HOOD SPRAY REGULATOR VALVE 00 3/79 01 FAILURE LOA TG-02 ABNORMAL VIBRATION OR NOISE FROM 00 3/79 01 TURBINE GENERATOR LOA TG-03 LOSS OF SHAFT GROllNDING 00 3/79 01 LOA TO-01 of 6 /795C. CONTINUE 0 OPERATION WITH ABNORMAL HMr 1C/77 01 BEARING OR BEARING OIL TEMPERATURE LOA -TO-02 TURBINE LUBE DIL COOLER LEAK 00 4/78 02 LOA VC-01 OPERATION OF CONTROL ROOM HVAC DURING 00 4/79 03 HI RADIATION,JiMOKE OR CHLORINE DETECTION LOA VE-01 OPERATION OF AUXILIARY ELECTRIC 00 4/79 03 EOUIPMENT HVAC DURING HIGH RADIATION, SMOKE OR CHLORINE DETECTION LOA VP-02 tilbEL PRIMARY CONTAINMENT COOLING PRESSURE- 00 0/09- 02 RELIEF AFTER HIGH DRYWELL PRESSURE, NON-LOCA LOA WL-01 OPERATION OF RIVER SCREEN HOUSE WITH or 10/796L G} 3ff9 01 ONE TRAVELING SCREEN INOPERATIVE LOA WR-01 O/ /o/ ~79S_. LOSS OF REACTOR BUILDING CLOSED -se -4f79 01 COOLING WATER (RBCCW) LOA WS-01 O! lbSY 'N-LOSS OF SERVICE WATER -Os- -6f78- 01 LOA WT-01 LOSS OF TURBINE BUILDING CLOSE0 O/ HMr

                                                                                 ///M S -02 J/44-COOLING WATER (TBCCW)

QB -03 & c Z t o M 6 . d l ' d d # I ,' N' R.6 Y]f 3wan zo o4n, Z> c~ st s- . SS ~ON WSkLC . &m On zk. 010. G%W - g'g . WJN4 $mevx W.

LOA INDEX ' 6/12/79 PAGE PROC. NO. TITLE REV. REV DATE DI (ETTI LOA .WX-01 NO C AP IN DRUM ol /0/%3

                                                                                 -ett            +f?+        03 LOA      WX-02 Of            (OlY1dL NO FILL SELECTION                           6               -4f74-      03 LOA      22-01                      OPERATION DURING EARTHOUAKE CONDITIONS 0I             10l106L-0
                                                                                 -Ofr           4f74                  -

27--03 JKq ve &&gcW W 00 "h[1 ZZ O 7 Gkg OS2cJe go /o/ R

     +22-04                         Ac+;on +o be deben In Even&

oS gn (), ] 3PI II +o the 111 Bid ,o yg u- og Df. On +ne zu. h ve-r 8-2 2 -oc PI'"1t Oeo~ Wtin *e el inS *i IDIC- % AcBon To Be Taten in -lhe 5tleni o o be of pn O-) , & ll To 7he Cwling s-l

                                   -Lek e 22-08                  uiwdg dwL DomagG lW!'E                                                        (~i
                                'A p ujk /g (d.ul c1 % dirny 00/aabdj 4

I i 1

,}}

ML19326E034: Difference between revisions

Revision as of 22:58, 31 January 2020

Text

Navigation menu

ML19326E034: Difference between revisions

Revision as of 22:58, 31 January 2020

Text

Navigation menu

Search