ML20235M543

Forwards Comments on Operator & Senior Operator Exams Administered on 860912,for Consideration.Related Info Encl

ML20235M543
Person / Time
Site:	Summer
Issue date:	09/19/1986
From:	Nauman D SOUTH CAROLINA ELECTRIC & GAS CO.
To:	Arildsen J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
Shared Package
ML20235M447	List: ML20235M445 ML20235M518 ML20235M543
References
NUDOCS 8710060224
Download: ML20235M543 (19)

Text

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ENCLOSURE 3 goumyn.

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4 September 19, 1986 .

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Mr. Jesse Arildsen License Examiner U.S. Nuclear Regulatory Commission

,. Region.II, Suite 2900 101 Marietta Street, N.W.

Atlanta, Georgia 30323-

Subject:

Virgil C. Summer Nuclear Station Docket No. 50/395 License No. NPF Operator License Examinations

Dear Mr. Arildsen:

Enclosed are South Carolina Electric & Gas Company comments on the NRC Operator'and Senior Operator examinations administered at the Virgil C.

Summer Nuclear Station on September 12, 1986. Your consideration is appreciated.

rytrtly[urs.

N a a W

RCR/ DAN:jez Enclosures c: J. G. Connelly w/o attachment O. S. Bradham w/o attachment M. B. Williams w/o attachment K. W. Woodward w/o attachment

.A. M. Paglia w/o attachment NPCF File 8710060224 861113 5 PDR ADOCK 050

i. .

Question 5.01 b.3: ,

Match the items in column "A" to their respective core locations ..

in column "B". Assume full power.

"A" "B"

3. Maximum actual a. bottom heat flux b. between bottom and middle

c. middle

d. between middle and top

e. top d

Answer key response: c i

l

REFERENCE:

WJE 237, HT&FF Manual, pp. 227-229 VCS, Hot Channel Factors p.4, Figure 2 4

Suggested correct response: either b or c Reason: When operating with a negative 6I as we most frequently do (Cycle I was negative the entire duration), maximum actual heat flux will be between the bottom and middle. _

See Enclosure I, The Nuclear Design of the Virgil C.

Summer Power Plant Unit 1.

Question 5.13: (1.0)

Why are Rod Insertion Limits established?

Answer key response:

Rod Insertion Limits are established to insure that the reactor core can be shut down at any time under any condition or to insure the minimum shutdown margin is maintained.

REFERENCE:

V. C. Summer Reactor Theory p.I5.50, TS p.3/4 1-3 001/000 KS.04 (4.3/4.7)

Suggest correct response:

The above answer and/or the following:

o Minimize the reactivity consequences of an ejected rod (rod misalignment).

o Maintain uniform power distribution limits.

Reason: Technical Specification 3/4.1.3 Bases contain tht -

reasons for rod insertion limits. See Enclosure

~,

QUESTION 6.06. (1.0)

LIST 4 means by which crystallization of the Boron Injection

. Tank (BIT) is prevented during normal plant operation.

ANSWER 6.06 (1.0) '

(1) Electrical heaters on BIT L (2) Trace-heat all piping, valves, and pumps-(3) Continuous recirculation flow (4 8 0.25 each)

(4) Sparger inlet of BIT I

REFERENCE:

SU: AB-10, p 22-23, 4/10 K4.02 (3.2/3.8)

Suggestion action: Delete Reason: MRP #20482 removed the BIT. Training materials have not incorporated these plant changes. See Enclosure III.

I 1

1 1

i I

QUESTION 6.14 (2.0)

State the specific back-up source of emergency feedwater for

. each of the- emergency feedwater pumps, and the condition under which the back-up supply suction line isolation valves automatically open.

ANSWER 6.14 (2.0) turbine-driven pump ------ service water loops A or B motor-driven pump A ------ service water loop A motor-driven pump B ------ service water loop B The back-up supply suction line isolation valves automatically open on a low suction pressure condition in the combined emergency feedwater pur- header from the condensate storage tank. (The action occurt 2 seconds af ter 2 of 4 pressure switches decrease to 10.4 psig.)

REFERENCE:

SU: IB-3 p 10 Suggested correct response:

turbine-driven pump ------ service water [ loops A or B]

motor-driven pump A ------ service water [ loop A]

motor-driven pump B ------ service water [ loop B]

The back-up supply suction line isolation valves automatically open on a low suction pressure condition in the combined emergency feedwater pump header from the condensate storage tank. (The action occurs 2 seconds after 2 of 4 pressure switches decrease to 10.4 psig.)

Reason: The specific backup source of emergency feedwater for each emergency feedwater pump is service water. To designate loops is assumed and superfluous.

w _ _-_

QUESTION 7.11 (2.50)

A Safety Injection (SI) Signal has just occurred and it appears to have been spurious. Prior to resetting the SI and recovering ESF components:

a. List three containment building parameters that must be normal. (1.50)

ANSWER -7.11 (2 50)

a. Reactor building temp. </=120 F Reactor building sump levels indicate "0" Reactor building radiation levels normal (any 3 0 0.50 ea)

Reactor building pressure indicates 0-1.5 psig

REFERENCE:

E0P 1.0 p.12 Suggested correct response:

Reactor building temp. [</ = 120 F]

Reactor building sump levels [0]

Reactor building radiation levels normal Reactor building pressure [0-1.5 psig]

Reason: The question asked for the parameters, not parameters and associated values.

QUESTION 7.20 (1.0)

V. C. Summer has guidelines for radiation exposure during emergencies. In what situation can a person receive a radiation exposure of 100 REM, whole body?

ANSWER For life saving operations (0.50) such as search and rescue for KNOWN missing persons. (0.50)

REFERENCE:

VCS, Radiation Protection Fundamentals, p 5 5-14 Suggested correct answer: For life saving operations.

Reason: Emergency Plan Procedure, EPP-020, Emergency Personnel Exposure Control, addresses radiation exposure guidelines during emergencies. No reference is made to " search and rescue for known missing persons." See Enclosure IV. Although EPP-020 allows for planned exposures of up to 75 Rem for lifesaving operations, the 100 Rem exposure addressed in the question is covered by the note associated with step 3 2.4 that disallows the 75 Rem for " spontaneous reactions."

{

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J e, QUESTION f

7.22 (1.0)'

When is control rod cooling air required to be operating? l ANSWER 7 22 (1.0)

When the reactor is at hot shutdown temp. (> 200 F) or at power.

REFERENCE:

VCS, 10-5, Rod Control p 15 Suggested correct answer:

Accept also as correct response " anytime the control rods are energized (reactor trip breakers closed) and/or prior to exceeding 350 F RCS temperature."

Reason: The system operating procedure for Reactor Building Ventilation, SOP-ll4, states the above suggested answer as a precaution for control rod drive mechanism cooling. See Enclosure V.

QUESTION 7.23 (1.0)

Answer the following questions concerning the requirements associated with doing a reactor startup:

b. If the startup is delayed, when must the Estimated Critical Condition Calculation be reviewed? (0.50)

ANSWER

b. Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to criticality

REFERENCE:

VCS, 00P-3, pp 9 & 11 Suggested correct answer:

b. Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to criticality ~or within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> prior to

~~

criticality.

Reason: Reactor engineering procedure REP-109.001, Estimated Critical Conditions Calculation, notes recalculation /

reverification of ECC if criticality does not occur within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of criticality. See Enclosure VI. The GOP-3 pages 9 & 11, 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> requirements are Technical Specification 4.1.1.1.1.b requirements.

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, QUESTION. Ic14

.See SRO question 5.13 comments QUESTION 2.09 '

. 'See SRO. question 6.06 comments

-QUESTION 2.16

.See SRO question 6.14 comments i QUESTION 3.09 (1.50)

Excluding various logic test panel testing, list three of the four remaining faults or actions which will generate a Solid

. State Protection System general warning alarm.

ANSWER

1) Loss of either 48-V DC power supply-

2) Loss of either 15-V DC power supply

3) A protection train's bypass breaker is racked in (inserted) and closed.

4) Any solid state circuit card not properly inserted (3 of 4 8 0.50 each)

REFERENCE:

SU: IC-9 p 34, 35 Suggested correct answer:

Accept " ground return fuse open (blown)" as a correct response.

Reason: The Westinghouse Tech Manual for Solid State Protection System references a blown ground return fuse as a possible cause for a general warning. See Enclosure VII.

QUESTION 4.22 See SRO question 7.22 comments QUESTION 4.23 See SRO question 7 23 comments i

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s THE NUCLEAR DESIGN OF THE VIRGIL C. SUMbiER POWER PLANT UNIT 1 R. J. Fabean R. E. Radcliffe R. H. Pituiski March, 1980 APPROVED: bk.h. APPROVE A ^A M c/'n P. K. Doshi, Manager h.k.Iudwiczak,YrojectManager NE Product Group C NFD Fuel Projects Work Performed Under Shop Order CGWF-11001 l

eeun w co no uwA This document contains information proprietary to Westinghouse Electric

-WESTINGHOUSE-PROPRIETARYTATA-7*f/f'*/{.7"I",,f"("]

f Corporation; it is submitted in confidence and is to be used solely for

' the purpose for which it is furnished and returned upon request. This document and such information is not to be reproduced, transmitted, disclosed or used otherwise in whole or in part without authorization

, of Westinghouse Electric Corporation, Nuclear Fuel Division.

WESTINGHOUSE ELECTRIC CORPORATION Nuclear Energy Systems P. O. Box 355 Pittsburgh, Pennsylvania 15230 i (AC d$4 VR p,f

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Ifuel'loadin9 figure identifies the assemblywise average and peak power, the value location of. the pea,k power rod, the assemblywise burnup and BP fract1.-

,erating remaining, the regionwise power and burnup sharing, the critical boron-concentration, and cycle burnup.

flowthe rela tive 3.3 BURNABLE POISON DEPLET10N The depletion of Boron-10 in the burn'able poison rods as a function of 2 and 3) core depletion is shown in Figure 3.9. The burnable poison rods are used 3r than in gegions 2 and 3'only. Both the regionwise and core average depletion of Baron-10 is shown in Figure 3.9. The burnable poison worth including structural material and moderator displacement effects is 7.0%4p at BOL ttern and reduces to about 1.3%ao at E0L.

.2 and 2.5 AXIAL POWER DISTRIBUTIONS  !

3.4 .

Core average axial power distributions were calculated using the one-al boron =

dimensional APOLLO code.

.rnup, discrete The axial power distribution calculated at hot zero power conditions at the beginning-of-life is shown in Figure 3.10. Since the core contains all fresh fuel and the moderator temperature is constant axially, the power u

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n as 3.11 shows the HFP axial powdr distribution at BOL and EOL. Since the~ core t li fe-has atnegativectemperature coefficient:and the coolant temperature increases 1ths from the ' bottom.of. the: core to . the top of. the core. the axial' power;. dis-x tribution,is' shifted stoward - the core . inlet at- BOLT ^ As, the. core depletes .

the higher production in the- center of- the;. core;results. in a .non-uniform axial burnup: distribution and consequently the~ flattened power shape at Em  : E0L'as!shownlin Figure 3.11. The ' axial' offset changes fromL -9.8%iat' BOL

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DkCb6hMVt NEACTIVITY CONTROL' SYSTEMS BASES BORATION SYSTEMS (Continued)

MARGIN from expected operating conditions of 1.77% delta k/k after xenon decay and cooldown to 200*F. The maximum expected boration capability requirement occurs at E0L from full power equilibrium xenon conditions and requires 12475 ~

gallons of 7000 ppe borated water from the boric acid storage tanks or 64,040 gallons of 2000 ppa borated water from the refueiing water storage tank.

With the RCS temperature below 200'F, one injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single injection system becomes inoperable.

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable below 275'F provides assurance that a mass. addition pressure transient can be relieved by the operation o,f a . ~

single P.RV. O The boron capability required below 200*F is sufficient to provide a SHUTDOWN MARGIN of 2 percent delta k/k after xenon' decay and cooldown from 200'F to 140'F. This condition requires.either 2000' gallons of 7000 ppa borated water from the boric acid storage tanks or 9690 gallons of 2000 ppm borated water from the refueling water storage tank.

The cont [ained water volume limits include allowance for water not available because of discharge line lo. cation and other physical characteristics.

The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 8.5 and 11.0 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

The OPERABILITY of one boron injection system during REFUELING ensures that this system is available .for reactivity control while in MODE 6.

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3/4.li3" MDVABLE' CONTROL ASSEMBLIES k

The specifications of-this section ensure that'(1)' acceptable power distribution' limits.are aintained,L(2)'the m minimum SHUTDOWN MARGIN is, main-tained,;and.(3) limit the potential effects'of rod misalignment on associated accident analyses ~. OPERABILITY of tric control rod position indicators is~

required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits.

SUMMER - UNIT 1 B 3/4 1-3

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' MODIFICATION CHANGE NOTICE syggeig, 7[

PROBLEM DESCRIPTION g g ,,,, , gffgy e j gj,,,

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// INITIATOR DATE ASSOCIATE MANAGER DATE ORALLY REQUESTED BY ORALLY APPROVED BY DATE

_ _ _ _ TIME DOCUMENTS AFFECTED ~ ~"~

OTHER GROUPS AFFECTED FSAR: [ ] YES { I NO SIGNATURES SECTION SIGNATURES GROUPS (AS REQ'D.) GROUPS ESSENTIAL DRAWINGS: (AS REQ'D.)

[JC72. d'7/ rev fg NE/ELEC. SEC11tITY g Jod 7.5/ revf // NE/I&C TRAINING NE/STRUCT.- HEALTH PHYSICS CIVIL (SITE)

OTHER DRAWINGS AND NE/ MECH. MAINTENANCE DOCUMENTS: NE/ FIRE PRCTI. OPERATIONS p jej NUC. ANALYSIS TECH. SUPPORT Q,/ NUC. SYSTEMS COMPUTER SERVICES CORP. HEALTH PEYSICS AND ENV. PROGRAMS DISPOSITION: P u i,i,b M,*u /n ysi. i ,,., D,ps yt ;, / ,/,,, , , u r HAVE RESULTSjoF THE 10CFR50.59 EVALUATION ANALYSIS SHEET FOR THE BASE MRF CHANGEDY

{ } YES {V) NO IF YES, PROVIDE REVISED 10CFR50.59 EVALUATION ANALYSIS SHEET.

DOES THIS MCN AFFECT THE BASE MRF FINAL EVALUATION? [ ] YES [(,,f NO DID THE EVALUATION qESTIONS FOR THIS MCN RESULT IN "YES" ANSWERS? ( l YES (rINO I REVIEWSPERFORMEDINTHEDJSFOSITIONOFTHISMCN7(,)YESIFEITHERQU [ j NO (r N/A IF NO, PROVIDE EXPLANATION M iV jf(A/ j fefie/p //p snb;tw/ yy)/diing7w sy a sj'rnf A-m L /

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DOES THIS CHANGE INVOLVE: [ () UMENT REVISION 7 [ ] PHYSICAL PLANT CHANGE? [ ] BOTH ?

INTERFACE REVIEW REQUIRED? [ YES , { l NO VERIFICATION REQUIRED? [ ] YES [hi NO VERIFICATION COMPLETE? [d YES [ ] NO

]'YES ( NO RECORD IF NO, JUSTIFY FORM Afn PR0i'I,DF,Dchoiss.p,/inknrgL

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/ /V/9 kial' ba nr Awd hom i,fdd/ba se d A w ),, ga, VERIFIER DATE DIS SITION'APPR VAL ' '

MANAG ENT A OV I C/B;r ,r7 / d-d/*d?S ef f 4.+ s: .14d / 2 f//$ k O'GG1NEER/ LEAD ENGINEER DATE ASSOCIATE MANAGER DATE h hC. O N Vf. ~

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Revision 3

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'*h' SOUTH CAROLINA ELECTRIC AND GAS COMPANY 6

VIRGIL C. SUMMER NUCLEAN STATION

't NUCL$AROPERATIONS l' )

EMERGENCY PLAN PROCEDURE 4

EPP-020 EMERGENCY PERSONNEL EXPOSURE CONTROL REVISION 6 1 1

MAY 27, 1986 SAFETY RELATED G

'Date DISCI {LINESUPEfWISOR d2A$ d h APPROVAL AUTHORITY

& -3 0-&4 Date RECORD OF CHANGES CHANGE TYFE DATE DATE CHANGE TYPE DATE DATE NO. CHANGE APPROVED CANCELLED NO. CHANGE APPROVED CANCELLED-1 0

4

- _ - - _ - - - _ - _ - _ a

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. .. - EPP-020 REVISIOH 6 5/27/86 3.2.1 Normal Administrative exposure limits may be suspended or modified verbally by the I.terim Emergency Director / Emergency Director, Radiological Assessment Supervisor, or Shift Supervisor. Verbal exposure extensions should be followed-up with the appropriate documentation as defined by reference 2.4.

3 2.2 Pederal limits of 10 CFR 20 will not knowingly be exceeded, except as described in Section 3.2.4 of this procedure.

3.2 3 Maintain exposure ALARA.

3.2.4 Planned exposures'of up to 75 Rem to save. human life l or prevent the serious endangennent of human life, or up to 25 Rem to save or mitigate significant damage to vital equipment and/or facilities must be approved by the Interim Emergency Director / Emergency Director or Radiological Assessment Supervisor.

Persons performing the planned actions should be 30E&O employees who are volunteers broadly familiar with exposure consequences.

l NL'TE: 'Thiefdoes.not apply to " spontaneous' reactions" by on the scene personnel in a threatening situation.

3.2 5 Non-company employees such as medical support personnel should be limited to:

a. 3 Rem if there is an adequate number of personnel so that rotation may be accomplished.

b 5 Rem if the number of personnel is limited such that rotation can not be accomplished.

c. 25 Rem to save a life.

3.2.6 If an individual receives an annual dose equivalent in excess of twice the Annual Dose Equivalent Limit, the case should be referred to a medical physician for review (as per ICRP Report 26).

hk C fo$ ur-R ;f1 PAGE 2 0F 5 i

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. .. 30P-114

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REVI SION 11 1/03/86 I. PURPOSE This procedure outlines the steps involved in operating the It also includes l

Raactor Buildt.ng Ventilation System.

Off-Normal Operations.

II. PREC AUTIONS

1. For PI-8254, RB NR PRESS to accurately indicate pressure, the following valves must be open:

a. SVX-6054-HR, RB NR PRESS CNTMT ISOL.

LOOP A (IRB) .

b. SVX-6050A-HR, POST ACCID H2 2.

The concrete around the vessel supports and nozzles must be limited to less than 150*F.

The Control Rod Drive Mechanism cooling f ans must lessmaintain g

3 the temperature around the Control Rod Mechanisms at n than 170*F. j e

4.

The Digital Rod Position Indication Data Cabinet Cooling '

System must operate continuously to limit the space }

temperature in the cabinet area to less than 95 P. i l

5 The Control" Rod Drive Mechanism Cooling' Unit must be operating when one or more full length rods are energized and/or. prior to exceeding 350*F' Reactor Cooling System

' t empe rature. -

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h OENERAL REVISION PAGE 1 of 37 LL Eve osuve  !

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• REP-109.001 ATTACHMENT I I

PAGE 4 0F 4 )

REVISION 0  !

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Summary of Estimated Critical Conditions I

E.1 Bank Position (0.1.1 or D.3.2150 steps)

E.1.1 Minimum: Bank at Steps E.1.2 Maximum: Bank at Steps E.2 Boron Concentration (C.2.5 or D.1.1) opm NOTE:,, : Recalculate: or reverify. the ECC. if criticality

<willJnot:occuriwithin;2LhoursLof;the: estimated

~ '

'timelof~ criticality. a E.3 Time (A.1 1 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />)

E.3 1 Earliest: Date Time _

E.3.2 Latest: Date Time E.4 The Bank Position in E.1 is above the Rod Insertion Limits (Figure II-16) and. below Rod Withdrawl Limits (Figure II-19)

INITIALS Calculated By Checked By Shift. Supervisor

9. Actual Criticality Data F.1 Bank Position: Bank at Steps P.2 Boron Concentration opm Tave 'F Time P.3 Reason if Bank Position was not within the desired band for Criticality:

Determined By:

[5htfoSurf 17~f~

&~ &sure l/IL , p. I l The remaining three input circuits, each having a separate ED p.ator, require constantly grounded inputs in order to prevent Q17 from turned on and causing a general alam. The inputs are grounded through

1. h lly closed in positions where a general

' crie3 of contacts t at are norma da'* is not required. The breaking of the ground path, from a Test switch I from its nomal position, bypass breaker auxiliary contact open (i.e.

asM I

set) or the Ground Return Fuse blown or removed will cause a general geaker alarm.

hhen these three inputs are grounded, the 30.1K input resistors which are tied to +48V will insure breakdown of any switch contact film and These

,.aus.e the cathodes of diodes CR100, CR102, and CR104 to be at ground.

jiodes will then be fo m ard biased, producing approximately 0.6V at the anodes of the LED diodes (CR101, CR103, CR105). The ED diodes will be back-biased and non-indicating because the anode voltages are below the required junction voltages of the base-emitter of Q17, CR99, and the ED. If the ground is broken to one of the inputs, its ED and CR99 will be fomard biased and Q17 will be turned on because of the current path from 48V through the 7.5K pull-up resistor. This will cause a general alarm.

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Rev.1, 4/77 fl1C C$bff t f.l

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CR80

+ 48v P0wtR SUP NO I 4 +aSv R33' 1 147xl ' RS6< '

V 40s ll

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+40V P0wtR SVP NO 2 37 ,

RS6 g - -. - -  % TO TNf8 TRAig RS4 < 8(NERAL 64 7K D 147M  ;#$2 l i REACTOR Trip Y pl 4 7K tiSV P0wtR SuP 40 Ih ,

4 C#92

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g h---*

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44 7E <

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+iSv P0*ER Su*'4No.2 7x <

@ R38l l a

l g mARaiNo Rg CIRCuig k-- h - - - en

+ 48v CARD INTERLOCK g CR9)

' g ROW NOI v g3y l 14 7 p g -- - -p TO Cd l CORTROL

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+48V CARD INTERLOCK @ R40 < g Row No. 2 14 7 g' ' y i___p ____ L $'h4

+4 By CARD tNTERLOCg C #9,5 g l 'h"Di

@- - -+48v R0W NO 5 Ret  ; RS7 s4 7N '

' le?K Q C R,94 CRSS bi47K g

+4Sv CARD INTERLOCK '

' CR98 R0W NO 4 R 4 7 .'

to fu ,

+48V L---]

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+40V CARD INTERLOCK 0"3 age, CRIO6]L g

' 7SN ROW NO.S negj

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CRice 6 ev l( j(CRtio

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i Qie

, 4,y RS* b <

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30IRl ' D7 SK GND FROW GNO RET URN FUSE . (SFU S)

CRioO a

CRIOe p y (g0 "tNPut ERROR iNMiBiT" i SWITCH (5509) CRiO61I +4ev

+40V RIO 3 m!  ! SM ADoitioNAL output GND FROM g2 MW  : FOR GENERAL

• LOGIC ANDA"*WVLTiptExtR SwifCM (5501)v Q4 '

LED WARNING INoiCatt0N TEST

• Switch (5508)

C8,99 , /

GNO f ROM GNO  ; RSt 0 21 A605 <

T RETURN f uSE 458 U 9)

">ERW155;vtS[, $mitCM tSS05) ,, 4 CRIOS l 332M ! ,

2 MEMORIES SWITCH (5506), ' l QUTPuf REL AY ' MODE LED SELECTOR SWITCH 15802). Q g

AND tvPA55 BRE AnER AUILlU ARY CONT ACT j I (OPEN wMEN SET) I Figure 2-33, General Warning Circuits i

2-88 acy, 1, af77

_ _ _ . . .. ..