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Exam Rept 50-424/OL-90-03 on 900723-30.Exam Results:Six Reactor Operators (RO) & 11 Senior RO Passed Exams
	ML20059C929
Person / Time
Site:	Vogtle
Issue date:	08/20/1990
From:	Ernstes M, Munro J; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	;
Shared Package
ML20059C914	List: ML20059C911; ML20059C929;
References
	50-424-OL-90-03, 50-424-OL-90-3, NUDOCS 9009050416
	Download: ML20059C929 (188)
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Meuq UNITED STATES '

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'g. ' NUCLEAR REGULATORY COMMISSION REGION 11 '

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!g .t .101 >'iARIETTA STRE ET, N.W.

4- ATl ANTA, GEORGIA 30323 p \,**..

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/ ENCLOSURE 1 ,

EXAMINATION REPORT - 50 .424/0L-90-03

-Facility Licensee: Georgia Power Company T

P. O. Box 1295 - ['

Birmingham,-AL 35201 t

I ry Facility Name: Vogtle Electric- Generating Plant -. ;

. Facility Docket Nos.: 50-424 and 50-425 '

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l Written and Operating Requalification Examinations were conducted at.the- .

. , Vogtle Electric Generating. Plant site near Waynesboro, Georgia, u l

1 Chief Examiner: N [. d 8[te'[b .

Michael E..Ernstes- Date Signed ,

,ikpprov'edBy: /

N ohn F. Musto, Chief f* '

Date-Signed j 0perator' Licensing Section 1 !

Division of Reactor Safety i

SUMMARY

Examinations were conducted during.the weeks of July 23, 1990, and July 30, -

-1990.

Written and operating examinations were administered to six Reactor Operators i and 111 Senior Reactor Operators. All six of the Reactor Operators passed the examination. Ten of the 11 Senior Reactor Operators passed the examination. '

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I 9009050416 900823 PDR ADOCK 05000424 V PNU 1

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REPORT DETAILS l I

1. Facility Employees Attending Exit Meeting:

. G. Bockhold, Jr., General Manager 'I

' T. Greene, Assistant General Manager

.K. Holmes, Plant. Training and Emergency Planning Manager R. Dorman, Operations Superintendent of Training 1

,i J. Swartzwelder, Manager, Operatiens i

-J._Ho) kins, Operations Department J. Roaerts,-Emergency: Preparedness Supervisor C. Stinespring, Manager, Plant Administration H. Handfinger, Manager,' Maintenance E. Dannemiller, Manager, Nuclear Security ,

F. Ealick, Engineering Supervisor J. Williams,? Supervisor, Plant Engineering R. LeGrand, Manager, HP/ Chemistry H. Beacher, Senior Plant- Engineer E. Kozinsky, 0perations Superintendent q

2. Examiners: l

M. Ernstas, NRC, Region II

-M. Morgan, NRC, Region II sM.-Stein,-Sonalysts ,

K. Parkinson, Sonaylsts

Chief Examiner

3. Exit Meeting:

At the conclusion of the site visit, the examiners met with representatives of the plant staff to discuss. the results of the ,

examinations.

The licensee did not' identify as proprietary any material provided to or reviewed by the examiners, l

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=C90-06-014 IC *

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. g PLANT VOGTLE TRAINING DEPT. >

MASTER KEY

f FOR

. EXAM::SR-905-90-05.A Total' Points: 24.00 i

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' ASSEMBLED IN MANUAL MODE.

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STATIC SIMULATOR - PART A R3 S C E N A R I O 8 080 (1 ~;Y - 2 3 s

SCENARIO TYPEt Normal operations with minor malfunctions PLhWT CONDITIONS:

50% steady state BOL, rods in automatic, following closure of turbine stop valve.

t QUESTIONS APPLICABLE TO THIS EXAMINATION SCENARIO I

A2301 A2302 A2303 ;

A2304 A2305 A2306 A2307 1 A2309 '

A2310 A2311 A231?

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VALIDATED REVIEWED:

TRAINING:

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STATIC SIMULhTOR - PART A 13 I t

i SINULATOR SETUP INSTRUCTIONS FOR SS-23 INITIALIBE Tot IC 8, 50% Pwr, BOL 1 i

INSERT MALFUNCTIONS: !

No. WAME ENT91 K57300.

101C CV failed closed (2, 101C) 40D HL NR TEMP RTD Failure (3, 400) 62A PT505 failed low [4, 62A,0)

INSERT OVERRIDES:

NAME ENTRY METHOD TDAFW Stm supply SG #2 (1, HS3019/Close)

RWST to SI pump iso valve (2, HS8806/Close)

SELECT THE FOLLOWING SWITCH POSITIONS:

NUMBER NAME POSITION

1. 1 LS-4 590 Pressurizer level CNTL selector 461/460

2. 1HS-8806A RWST to SI pump iso valve B ON

3. TS-412T Tave defeat selector DEFEAT 442

4. Set ERP to Mode 1

5. Select ERF display 2 to Trend 7, (Select CRT trend)

6. Rod r?ntrol Selector Switch to Manual

7. All Pzr/B/U Heaters in Auto START THE SCENARIO PERFORM THE FOLLOWING ACTIONS:

I Manually drive control rods in to stabilize RCS Tave at

, approximately 570 degrees F. Rods should finally end up near 130 steps on control bank D. Select 130 steps on the control bank D step demand counters.

v . > . . -

e MF HB, EDG!IAf98 BI.8 RCS stable at 568 to 569 degrees F and control rods at 130 steps on control bank D and DRPI indicating 132 steps, i

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1 STATIO SINULATOR - PART A 23 j l

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SINULATOR SETUP INSTRUCTIONS FOR 88-23 !

VERIFY TRE POLLOWINg CONDITIONS:

POWER: 40 gog POSITIOut 130 on D RORON: 1560 l PSR LYLt 35-45% PER PRESS: Approx. 2235 I TAV3 Approx. 568 degrees P.

)

RCP STATUS: Running 3CCS !TATUS: Standby i Sg PRESS: Approx. 1020 psig gg [VLt 50% NR !

OTHER CONDITIONS 8 None VERIFY THE FOLLOWING SWITCH POSITIONSt TDAFW Steam supply from SG 2 (HV-3019) CLOSED i RWST to SI pump isolation valve (HV-8806) CLOSED ,

PRZR Level Control Selector Switch in the 461/460 position i

Lockout switch for HV-8806 in the ON position Tave Defeat Selector Switch in the 442 position Rod Control in Manual i All PRZR heaters in Automatic SELECT THE FOLLOWING DISPLAYS t ERE DISPLhY 1: Top level digital ERF DISPLAY 2: Trend 7, (Select CRT trend) l PROTEUS: RCS diagram J

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SR-905-90-05.Al KEY j Page 1 EXAM KEY l NRC REQUAL EXAM - WEEK 1,PART 1 - RO j

==================================================================

1.01 Q: Which of the following bistables associated with' Loop 4 Tave is NOT required to be tripped by Tech Specs 3.3.1 and 3.3.2?

a. OT delta T trip l

b. Lo Lo Tave steam dump block ,

c. Lo Tave TW Isolation l

d. OP Delta T Trip A: B R: TECH SPECSQ, EB#: LO-SS-23000-01-12 Point Value 1.00

==================================================================

1.02 Q: Which of the following best describes the operation of the steam dumps if a turbine runback occurred?

a. Steam dumps would open and control RCS temperaturn at Tref.

b. Steam dump error signal would increase but steam _ dumps would remain closed.

c. Steam dumps would open and stay open.

d. Steam dumps would have permissive to open (armed) but no demand signal.

A: C i 1

R: LO-LP-21201-00, TSAR Logic 7.2.1-1,,

EB#: LO-SS-23000-01-03 Point Value: 1.00

==================================================================

KEY CONTINUED ON NL'YT PAGE

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f SR-905-90-050A; KEY

s. Page 2 EXAM .SY ,

NRC REQUAL EXAM - hEEK 1,PART 1 - RO i

i. 1.003 Q: Why is there a mismatch between turbine load sr. and turbine load? i

a. PT-505 failed low :

b. Control valve 3 is closed

c. A load increase is in progress

d. S/G pressure is too high A: B R: LO-LP-30303-00, GEK46488B, EB#1 LO-SS-23000-01-04 Point Value: 1.00 KEY CONTINUED ON NEXT PAGE i

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Page 3 ,

EXAM KEY NRC REQUAL EXAM - WEEX 1,PART 1 - RO ,

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! 1.04 Q: Concerning the presence of the Lo-Lo RIL alarm, which of the- t following is true?

a. Emergency boration should be started in accordance with Tech Specs.

l. b. The alarm should have cleared when Tave was defeated, but no emergency boration is required. ,

c. The alarm can be cleared by placing the Delta T defeat switch to loop 4, and no emergency boration is required. j

d. The alarm can NOT be cleared, but no emergency boration l is required.

A: C :

R: LO-LP-16101-00, FSAR Logic 7.2.1-1,,

EB#: LO-SS-23000-01 06 Point Value: 1.00

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. Page 4 EXAN KEY NRC REQUAL EXAM - WEEK 1,PART 1 - RO I

1.05 Q: Which of the following correctly describes ~the loop 4 Narrow Range Temperature instrument failure? 1

a. Tcold, low l t 1 o b. Thot, low .

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c. Teold, high

d. Thot, high ,

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L R: LO-LP-16101-00, LO-LP-16101', NO:'

EB#: LO-SS-23000-01-07 Point Value: 1.00 1

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1.06 Q: From the existing plant conditions calculate the dilution required to raise power to 100%, with all rods cut, disregarding Xenon changes. (Assume DBW = 8.5 pcm/ ppm). Which one of the following ;

is correct?

a. 690 gal (+/- 50 gal)

b. 1120 gal (+/- 50 gal)

c. 2620 gal (+/- 50 gal)

d. 4040 gal (+/- 50 gal) l A: C ,

R: LO-LF-33440-00, Plant Technical Dat, EE#: LO-SS-23000-c.=09 Point Value: 1.00

.==.====.....===......... ========.============. ===========.================.

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, EXAM KEY I'

NRC REQUAL EXAM - WEEK 1,PART 1 - RO

==================================================================

lo07 Q: Which of the following is'true regarding the operability of the TDAFW pumpY

a. The pump is operable now, but would be inoperable if HV-3009 were shut.

b. If HV-3009 were shut, the plant would have to be in Mode 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c. The pump operability would not be effected by shutting HV-3009.

d. HV-3009 must be stroxe tested at least once per 31 days per Tech Specs regardless of its effects on pump operability.

A: C R: LO-LP-20101-00, Tech Spec 3.7.1.2',

EB#: LO-SS-23000-01-11 Point Value: 1.00 .

============================================================================== !

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Page 6 EXAM KEY NRC REQUAL EXAM'- WEEK 1,PART 1 - RO

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1.08 Q: 1HV-8806, RWST to SIP isolation, has been out of position for the last hour, which of the following would apply if the valve remains out of postition?

a. The plant must be placed in cold shutdown within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

b. The plant must be placed in hot shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c. The plant must be placed in hot shutdown within 83 hours9.606481e-4 days 0.0231 hours 1.372354e-4 weeks 3.15815e-5 months .

d. The plant must be placed in hot standby within 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

A: B R: LO-LP-13201-00, Tech Spec 3.5.2',

EB#: LO-SS-23000-01-02 Point Value: 1.00

==================================================================

1.09 Q: What automatic action would NOT occur if LT-461 (Pzr level transmitter) failed low?

a. LCV-459 would close

b. LCV-460 would close

c. Pzr backup heaters would turn off

d. All orifice isolation valves will clone A B i

R: LO-LP-16301-00, FSAR Logic 7.2.1-1,,

EB#: LO-SS-23000-01-01 Point Value: 1.00

==================================================================

KEY CONTINUED ON NEXT PAGE

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Page 7 EXAM KEY !

NRC REQUAL EXAM --WEEK 1,PART 1 - RO

.....................=.. .....................................................

1.10 Q: If rods were placed in automatic they would stept

a. In due to failed NR temperature instrument. I

b. In due to failed Impulse pressure instrument.

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c. Out due to failed NR temperature instrument.

d. Out due to failed Impulse pressure instrument.

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A: B R: LOGICS, EB#: LO-SS-23000-01-13 Point Value: 1.00

...=======.......-===...=====. .==..=====.====..==.==.====. ..==..==.=. .u.. .

1.11 Q: NOTE: THIS QUESTION POSES A SITUATION THAT IS NOT RELATED TO THE CURRENT PLANT / CONTROL BOARD CONDITIONS!!!!!!

s If BTRS were in service for dilution of the RCS, ALB07-E4 alarming would result in:

a. No effect to BTRS operation while in the DILUTE MODE.

b. Flow diverting around the CVCS demineralizers, while maintaining flow through the BTRS demineralizers.

c. Flow diverting arcund the BTRS demineralizers, while maintaining flow through the CVCS demineralizers.

d. Flow diverting around both the BTRS and CVCS demineralizers.

A: D ,

l R: 17007-1,R1,PG.,

EB#: LO-SS-23000-01-14 Point Value: 1.00

============.=====================================================

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, SR-905-90-05.A; KEY

<t Page 8 l EXAM KEY ;

NRC REQUAL EXAM - WEEK 1,PART 1 - RO '

= = n = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = t...: = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = r

.1.12 Q: NOTE:. THIS QUESTION POSES A SITUATION T!!AT IS NOT RELATED TO THE- !

CURRENT PLANT / CONTROL BOARD CONDITIONS!!!!!! ,

Wh!ch of the following correctly describes the effects of a loss ;

of .NB01 on the 1A DG? !

a. .The DG would be inoperable due to loss of Fuel Oil i Transfer Pump.

b. The DG would be considered operable and would remain in standby.

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c. The DG.would be inoperable due to the loss of lube oil circulating (keep-warm) pump. '

d. The DG would not auto start due to loss of Train A control '

power.

A: C

.R: TECH SPECS, EB#: LO-SS-23000-01-15 Point Value: 1.00

===========================================================================w== '

END OF SECTION KEY l

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1 STATIC SINULATOR - PART A 1

i SCENARIO

SUMMARY

- SS-54 SCENARIO TYPEt Emergency ,

l PLANT CONDITIONS Turbine trip with no S/D's stuck open PORV w/ loss of 23B06 QUESTIONS APPLICABLE TO THIS EKANINATION SCENARICt Ab401 A5412 AS422 - Same as A1812 AS402 AS413 AS423 - Same as A4705 AS403 AS414 A5424 - Same as A4712 A5404 AS415 AS405 AS416 AS406 'A5417 AS407 AS418 AS408 AS419 A5409 AS420 A5410 AS421 A5411 VALIDATED:

REVIEWED:

TRAININet

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STATIC SINULATOR - PART A SINULATOR SETUF INSTRUCTIONS FOR 88-54 INITIALISE Tot IC-13, 30% Pwr, MOL IWSBRT MALFUNCTIONS No. NAME ENTRY METHOD 59a Pzr Press Xmitter PT-455 (2, 59a, 0)

Fail Low 62c Pzr Relief Valve PV-456A (3, 620, 100, 0003)

Fail Open 62b Turb Imp Press PT-506 (4, 62b, 100)

Fail High 105 Loss of Main Turb Lube 011 (5, 105, 100) 135h Loss of 480V SWGR 1BB06 (6, 135h,, 0003)

INSERT OVERRIDESI l NAME ENTRY METHOD None SELECT THE FOLLOWING SWITCM POSITIONS:

i NUMBER NAME POSITION '

PS 455G PRZR Rec Sel SW P-456 i, HS 40041 Rod Bank Selector Switch Auto i

l START THE SCENARIO l

PERFORM THE FOLLOWING ACTIONS l Throttle AFW flow to 200 GPM to each SG. l l

l FREEEE THE SINULATOR AT: RCP Trip Criteria Met (1375 psig) l l

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STATIC SINULATOR - PART & :.

l SINULATOR SETUP INSTRUCTIONS FOR 88-54 VERIFY TER FOLLOWING CONDITIONSt

-7 POWER: Approx. 5 X 10 ROD POSITIOWs Tripped ,

DORON: Approx. 1020 ppa PER PRESS: < 1375 psig PSR LYLt > 20% !

TAV3 Approx. 540 degrees F ,

RCP STATUS: 4 Running ECCS STATUS: Injecting !

Ag RRasgt Approx. 940 33 LIL: 15-20% NR OTHER CONDITIONS N/A VERIFY THE FOLLOWING SWITCE POSITIONS:

Rod Control in Automatic PRZR Recorder Selector Switch in the 456 position SELECT THE FOLLOWING DISPLAYSt ERF DISPLAY lt Top level ERF DISPLAY 2: Heat Sink PROTEUS: RCS diagram i

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SR-905-90-05.At KEY Page 8 EXAM KEY r NRC REQUAL EXAM - WEEK 1,PART 2 - P.O !

! mumammmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmme !

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2.01 Q: Select the correct location where the leaking coolant is being collected.

U a. The RCDT

b. The Containment Sump .

I I. c. The PRT l

! d. The Containment Sump and the PRT l At C f R: SIM INDICATIONS, EB#: LO-SS-54000-01-02 Point Value 1.00 mummmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmma l'

2.02 Q: PZR level indication is increasing during this transient due to which of the following factors. 4

a. RCS pressure is lowest in the PZR, therefore the inventory of the RCS is flowing into it.

b. The inventory of the PZR is saturated and the RCS is I subcooled.

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c. BIT flow rate is greater than the leakage rate.

d. Rx Vessel Head Voiding is forcing water into the PZR.

L A: C R:-SIM INDICATIONS, L EB#: LO-SS-54000-01-04 Point Value 1.00

= = = = == = = = = ==== == = = ===== ====== =a m m mmmmmmmmmm m mmmmmmmmmum mm m m mm= mumm um m m mm m mm m

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Page 9 EXAM KEY :

NRL REQUAL EXAM - WEEK 1,PART 2 - RO J ====================e===w===================================================== ,

p 2.03 Q =What system / components actuated to replace the turbine generator as a steam release path immediately after the turbine generator tripped?

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a. Steam dumps on the load reject. controller r

b. Steam dumps on the plant trip controller i

c. S/G ARVs 7

d. S/G Safety Valves i A: C f

R: LOGICS, EBf: LO-SS-54000-01-05 Point Values 1.00

==================================================================

2o04 Q: Which of the following correctly explains how the plant responded to the preceding transient.

Following the turbine generator trip the:

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a. Reactor tripped immediately because of the P-9 !

permissive. ,

b. RCS heated up rapidly causing PZR level to increase to the high PZR level trip setpoint. !

c. RCS heated up rapidly causing a rapid insurge into the PZR which resulted in a Rx trip on rate compensated PZR high pressure,

d. SG shrink resulted in a reactor trip on low-low S/G levels.

A: D

'R: SIM INDICATIONS, EB#: LO-SS-54000-01-06 Point Value: 1.00

==================================================================

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I EXAM KEY Page 10 !

.NRC REQUAL EXAM - WEEK 1,PART 2 - RO I ,

2.05 Q: Which of the following describes the plant response that caused the automatic SI. ;

a. The turbine trip caused the ARVs to open. Steam line ,

pressure decreased resulting in a SI/SLI. ,

b. The reactor tripped. Steam dumpc opened on the plant trip controller. S/G ARVs also being open caused steam line pressure to decrease resulting in a SI/SLI. i

c. The turbine trip caused RCS pressure to increase and a PORV to open. The PORV failed to close resulting in a low PZR pressure SI.

d. The turbine trip resulted in higher than normal pressure' !

when the reactor tripped. RCS pressure lowered .

resulting in a rate compensated low PZR pressure SI. !

i A: C f R: SIM INDIACTIONS, ;

EB# LO-SS-54000-01-08 Point Value: 1.00 f

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i Page 11 EXAM KEY NRC REQUAL EXAM - WEEK 1,PART 2 - RO.

3.06 Q: Which of the effects described below did the failure of PZR pressure detector PT-455 have on this transient?

a. The failure was a benefit to the plant during the initial plant transient, because it called for PV-455A to be open.

Therefore the RCS pressure spike was lower than expected.

b. The failure was a detriment to the plant during the initial plant transient because it prevented PV-455A from opening.

c. The failure was~a detriment to the plant during the initial plant % snsient because PV-455A and the spray valves failed to ope, during the transient.

d. The fullure was a benefit to the plant during the initial plant transient because PV-455A and the spray valves opened during the transient.

A: C R: LOGICS, SIM INDICATIONS, ;

EB#: LO-SS-54000-01-15 Point Value: 1.00 !

2.07 Q: Immediately after the turbine trip, but before the reactor trip, control rods started to move. Which of the statements below correctly describes the control rod response.

a. Rods moved in at 8 steps per minute due to Pimp /NI '

input deviation.

b. Rods moved out at 8 steps per minute due to Tref /Tave deviation.

c. Rods moved in at 72 steps per minute due to Pimp /NI and Tref /Tave deviation.

d. Rods moved out at 72 steps per minute due to Pimp /NI and [

Tref /Tave deviation.

l A: C R: LOGICS, EB#: LO-SS-54000-01-17 Point Value: 1.00

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SR-905-90-05,At KEY l Page 12 ;

EXAM KEY NRC REQUAL EXAM - WEEK 1,PART 2 - RO !

saammmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmme==mumansme=============

l 2.08 Q The containment spray system is: j

a. In operation with chemical injection occurring.

b. NOT-in operation, but thould be.

c. In operation, but chemical injection is NOT occurring.

d. NOT in operation and is NOT currently needed.

A D R: SIM INDICATIONS, 19000-C,

'EB#: LO-SS-54000-01-16 Point Value: 1.00 maammmmmmmmmmmmmmmmmmmmmmmmmmmmme=============================================

h 2.09 Q: Select the correct system response to the failure of impulse .

pressure detector PT-506.

a. No control rod movement would have occurred as a result of the failure,

b. Steam dumps would have armed on the load reject controller ,

but would not have opened.

c. . Steam dumps would not have armed but would be set to open with maximum demand if an arming signal had been generated.

d. Control rods would not have moved because C-11 would have been blocking their movement.

At A R: LOGICS, EB#: LO-SS-54000-01-18 Point Value: 1.00

==================================================================

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SR-905-90-05.At KEY l 'f Page 13 EXAM KEY-NRC REQUAL EXAM - WEEK 1,PART 2 - RO 2.10. Q: Based on present plant conditions, the minimum required AFW flow ist

a. 570 gpm to each intact SG.

b. 570 gpm to all intact SGs.

c. 50 gpm to each intact SG.

d.- O gpm to all intact SGs.

At D R: 19000-C, EBf: LO-SS-54000-01-22 Point Value: 1.00 2.11 Q: The current Control Room HVAC system alignment is:

a. Incorrect because CRI should only be actuated on a high radiation sensed on the air intake line. l

b. Correct because CRI is automatically actuated on an SI.

c. Incorrect because both trains of filter units have automatically started on a SI.

d. Correct because CRI is automatically actuated when smoke is detected in the air intake line.

A: B R: LOGICS, EB#: ID-SS-54000-01-23 Point Value: 1.00 1

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, SR-905-90-05.At KEY

.= Page 14 5

EXAM KEY NRC REQUAL EX7.M - WEEK 1,PART 2 - RO

=================================,================================

2.12 Q . NOTE: THIS QUESTION POSES A SITUATION TRAT IS NOT RELATED TO THE CURRENT PLANT / CONTROL BOARD CONDITIONS!!!!!!

Which of the following is correct concerning the failures that required the quickest operator response prior to the Rx trip in terms of Tech Spec action requirements?

a. The PT-455 failure requires the least response time,

b. The PT-506 failure requires the least response time.

c. The PT-506 failure would require less time if power.were below P-13.

d. The PT-455 and the PT-506 both require the same response time.

At D R: TECH SP"Ca, LOLP39208-03, EB#: LO-SS-54000-01-21 Point Value 1.00

==================================================================

END OF SECTION KEY END OF TEST KEY l

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The following is an alpha-numeric list of figures and references which

, should accompany this test when it is administered.

Reference Question Static Sim 23 Intro. 2 Static Sim 23 Intro 3 Static Sim 23 Intro 4 i

Static Sim 23 Intro 5 Static Sim 23 Intro 6 Static Sim 23 Intro 7 Static Sim 23 Intro 8 Static Sim 23 Intro 9 9

END OF FIGURE AND REFERENCE LIST

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, EXAM PULLED FOR CCR ' F1LIN3 CUPVt DATE: l F i PLANT VOGTLE TRAINING DEPT. ]

QUESTION USE LIST for EXAMt SR-905-90-05.A Total Points: 24.00 FILE NotC90-06-614 Page 1 l i

Assembled by Chuck Stuhaan on 07/23/90 in MANUAL mode.

IIEM EB NUMBER' REVISION DESCRIPTION 1.01 LO-SS-23000-01-12 4 06/28/90 Tech spec useage 1.02 LO-SS-23000-01-03 3 06/20/90 Steam Dump Operation' 1.03 LO-SS-23000-01-04 4 06/20/90 'Turbina Control System' 1.04 LO-SS-23000-01-06 4 07/12/90 Emergency Boration Requirements' .

1.05 LO-SS-23000-01-07 5 06/28/90 Failure diagnosis *

~1.06 LO-SS-23000-01-09 8 07/12/90 Dilution calculation

1.07 LO-SS-23000-01-11 6 06/20/90 Tech Spec Application' ,

1.08 LO-SS-23000-01-02 4 06/20/90 Tech Spec Application

1.09 LO-SS-23000-01-01 5 06/20/90 Pzr level interlocks' 1.10 LO-SS-23000-01-13 00 . / / Rod Control response to instrument -

1.11 LO-SS-23000-01-14 1 06/20/90 BTRS divert on high temperature 1.12 LO-SS-23000-01-15 3 07/12/90 Loss of 1NB01 effects on DG operabi -

============================================================================== 5 END OF SECTION

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' EXAM PULLED FOR CCR FILING SUPV DATE:

PLANT VOGTLE TRAINING DEPT.

QUESTION USE LIST for EXAM: SR-905-90-05.A' Total Points: 24.00 FILE NOIC90-06-014 Page 1 Assembled by Chuck Stuhaan on 07/23/90 in MANUAL mode.

ITEM EB NUMBER REVISION DESCRIPTION 2.01 LO-SS-54000-01-02 6 06/20/90 leak diagnosis l 2.02 LO-SS-54000-01-04 5 06/20/90 event diagnosis 3.03 LO-SS-54000-01-05 5 06/28/90 post trip diagnosis 2.04 LO-SS-54000-01-06 5 06/20/90 turbine trip / reactor trip cause and 2.05 LO-SS-54000-01-08 3 06/20/90 causes of si 2.06 LO-SS-54000-01-15 4 06/20/90 effects 455 had on transient 2.07 LO-SS-54000-01-17 3 06/20/90 control rod reponse to pt-506

'2.08 LO-SS-54000-01-16 2 06/28/90 CS operation?

'2.'9 0 LO-SS-54000-01-18 4 06/20/90 rod response to pt-506 failure 2.10 LO-SS-54000-01-22 00 . / / AFW Throttling Limitations 2.11 LO-SS-54000-03-23 2 06/29/90 Control Room HVAC alignment 2.12 LO-SS-54000-01-21 9 06/29/90 tech spec useage

=============================:rs===================================

END 01 SECTION END OF QUESTION LIST

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['I n't Week l fact A (SRO) l k C90-06-016 l

l i

PLANT VOGTLE TRAINING DEPT. l

' MASTER KEY '

FOR EXAM: SR-905-90-05.AS Total Points: 24.00 l

ASSEMBLED IN MANUAL MODE.

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STATIC SINULATOR - PART A 23 i P

SCENARIO

SUMMARY

- 23 i l

I SCENARIO TYPE: Normal Operations with minor malfunctions PLANT CONDITIONS 50% steady state BOL, rods in automatic, following closure-of turbine stop valve.

QUESTIONS APPLICABLE To.THIS EXANINATION SCENARIO: l A2301 A2302 A2303 , [

A2304 A2305 A2306 -

A2307 -

l A2309 A2310

! A2311 i A2312 I

l VALIDATED:

REVIEWED:

TRAINING i

l'

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STATIC SINULkTOR - PART A 23 SINULATOR SETUP INSTRUCTIONS FOR SS-23 INITIALIBE TO: IC 8, 50% Pwr, BOL INSERT MALFUNCTIONS NO. MANE ENTRY METHOD 101C CV failed closed (2, 101C) 40D HL NR TEMP RTD Failure (3, 40D) 62A PT505 failed low (4, 62A,0)

INSERT OVERRIDES:

NAME ENTRY METHOD TDAFW Stm supply SG #2 (1, HS3019/Close)

RWST to SI pump iso valve (2, HS8806/Close) 1 SELECT THE FOLLOWING SWITCH POSITIONS:

NUMBER NAME POSITION

1. 1LS-459D Pressurizer level CNTL selector 461/460

2. 1HS-8806A RWST to SI pump iso V:lve B ON

3. TS-412T Tave defeat selector DEFEAT 442

4. Set ERF to Mode 1

5. Select ERF display 2 to Trend 7, (Select CRT trend) ;

6. Rod Control Selector Switch to Manual n

7. All Pzr/B/U Heaters in Auto START THE SCENARIO PERFORN THE F_OLLOWING ACTIONS:

Manually drive control rods in to stabilize RCS Tave at approximately 570 degrees F. Rods should finally end up near 130 steps on Contrcl bank D. Select 130 steps on the control bank D step demand counters.

E ,

a... %- 6 4 ;!

EF

. l M. . .

...r

=

raassa tal;sruoLaton-a;s _

RCS stable at'568 to 569 degrees F.and' control: rods at 130' steps on. control bank D and DRPI indicating 132csteps.

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't STATIC SINULATOR - PART L 23 >

4 SINULATOR SETUP INSTRUCTIONS FOR 88-23 l .

l

- VERIFY THE FOLLOWING CONDITIONS POWER: 40' ROD POSITION: 130 on D BORON: 1560 PER LVLt 35-45% PER PRESS: Approx. 2235 TAVE: Approx. 568 degrees F. .l FCP STATUtt Running.BCCS STATUS: Standby -

gg PRESS Approx. 1020 psig gg LYLt 50% NR l-l, OTHER CONDITIONS:

None VERIFY THE FOLLOWING SWITCH POSITIONS:

L 1

TDAFW Steam sup' ply from SG 2 (HV-3019) CLOSED I

! RWST to SI pump isolation valve (HV-8806) CLOSED PRZR Level control Selector Switch in the 461/460-position L

Lockout switch for HV-8806 in the'ON position j Tave Defeat Selector-Switch in the 442 position l

l Rod control in Manual i All PRZR heaters in Automatic 1

1 SELECT THE FOLLOWING DISPLAYSt BBZ DISPLAY 1: Top level digital ERF DISPLAY 2: Trend 7, (Select CRT trend) l PROTEUS: RCS diagram 1

4-. , - - - . _ _ --___ _ __ ___ _ _ __ _ _

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SR-905-90-05.AS; KEY

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Page 1-EXAM KEY NRC REQUAL EXAM - WEEK 1,PART 1 - SRO

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======,================= ...................--======....-====================== . 1 1.01- Q:.Of the-following bistables associated with Loop 4 tl ave, which bistable is most limiting with respect to tripping time by Tech-Specs. !

a. OT Delta T trip

b. Lo Lo Tave steam' dump block j

c. Lo Tave FW isolation .j

d. OP Delta T turbine runback {

.A: .C 4

R:: LO-LP-39207-00, AOP 18001-C, Table, EB#t LO-SS-23000-01-05 Point Value: 1.00

============================u=====================================

1.02i Q: Which of-the following best describes the operation of the steam dumps-if a turbine runback occurred?

o

a. Steam dumps would open and control RCS temperature at Tref. ;

b. Steam dump error signal would increase but steam-dumps would remain closed. :(

c. Steam dumps would open and stay open. l

d. Steam dumps would have permissive to open (armed) but no demand signal.

A: C R: LO-LP-21201-00, FSAR Logic 7.2.1-1,,

EB#: LO-SS-23000-01-03 Point Value: 1.00

.==============================================================================

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EXAM. KEY' ,

NRC REQUAL EXAM - WEEK 1,PART 1'- SRO !

-...--============..... u=============.....== ............--- .................-

r 1s03 Q:-Why.is there-a. mismatch between turbine load sct and turbine load?' ,

a. PT-505 failed low !

b. Control valve-3-is closed. ,

.'- w ,

c. A load--increase is'in progress !

t d.'S/G pressure =is too high A:'B [

t

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R: 'LO-LP-30303-00, GEK46488B, EB#: LO-SS-23000-01-04 Point Value: 1.00 -i

==================================================================

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~ SR-905-00-05.ASF KEY!

Pe. gen 3 EXAM KEY- :

NRC REQUAL EXAM - WEEK'1,PART 1 - SRO-1

..........e========== ..................---==========........-===....======== , u I

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l 1. 04 ' Q LConcerning the presence'of the Lo-Lo RIL alarm, whichcof'the -

following11s true?=

a. Emergency boration should be started in accordance with Tech Specs.

b.'The alarm should have cleared when Tave was defeated, but_

no emergency boration is required.

c. Th'e alarm can'be cleared by' placing the' Delta T defeat.

switch to loop _4, and no emergency boration is required,

d. The alarm can NOT be cleared, but no emergency boration ,

is required.

A: C -

t s

R: LO-LP-16101-00,. FSAR Logic 7.2.1-1,,

EB#: LO-SS-23000-01-06. . Point Value: 1.00

===============...,==...==========...........................==...............

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l,u - SR-905-90-05.AS0 KEY-i Page 4_ 1 EXAM KEY- '

NRC REQUAL' EXAM -LWEEK 1,PART 1 SRO' ]

==============================================================================> 0 2

1405- Q: WhichJof the following-correctly describes the11oop 4 Narrow Range- ..!

TemperatureLinstrument failure? '! -

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u

'a. .Tcold,-low l

.i

b. Thot,;1ow:

.f

c. Teold, high j d.. Thot, hich 't

?

A: d.

i I

{

R: LO-LP-16101-00, LO-LP-16101", NO:~ .

I EB# : - LO-SS-23000-01 . Point Value: 1.00- ;

==============================================================================. p 1.06 Q: From the existing plant conditions calculate the dilution requ.' red '

to raise power to 100%, with.all rods out, disregarding Xenon-changes. (Assume DBW = 8.5 pcm/ ppm). Which one of the following a is correct?

a. 690 galD(+/ .50 gal)

b. 1120 gal-(+/- 50 gal)- j

c. 2620 gal (+/ .50 gal)

d. 4040 gal'(+/- 50 gal)

A:-C

~

R: LO-LP-33440-00, Plant Technical Dat, EB#: LO-SS-23000-01-09 Point Value: 1.00 l

============================================================================== ;

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-Page 5 j i EXAM KEY.

NRC REQUAL' EXAM - WEEK 1,PART 1" SRO_

======n==============================================n========================-' ;

1. 0 75 Q: Which of thel following- is true regarding tha t operability of the l TDAFW: pump?

a. _ The punp' is operable now, but would be inoperable if HV-3009 were shut. ;

b. If HV-3009 were shut, the plant would have to be in Mode 3-witniw G-hours.

+

c. The pump operability would not be effected by shutting.

-HV-3C09. '

d. HV-3J09 must be stroke tested =at least once per 31 dayr,'per Tech Specs regardless of its effects on pump-operability. i A: C

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R: LO-LP-20101-00, Tech Spec 3.7.1.2',

EB#: LO-SS-23000-01 Point Value: 1.00

- ==============================================================================

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s EXAM KEY

>i j NRC REQUAL EXAM - WEEK 1,PART 1 :SRO o

..............................=.....................................===========

1.08- Q: 1HV-8806, RWST to-sip isolation, has been out of position for the i last hour,.which of the following would apply if the valve remains' j out of postition? y j

.a. The plant must be placed in cold shutdown within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months ,

, _b. The plant must be placed in' hot shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .c 'l '

u

c. The plant must be placed in hot shutdown.within 83 hours9.606481e-4 days 0.0231 hours 1.372354e-4 weeks 3.15815e-5 months ., .

d. The plant must be placed in hot standby within 5fhours.-

.i A: B 'I -

-i R: LO-LP-13201-00, Tech Spec 3.5.2~, ,

EB#: LO-SS-23000-01-02 Point-Value:'1.00

===========............. ...=====...========= .. .....======================== ,

s 1.09 Q: What automatic action would NOT occur if LT-461 (Pzr ~ 1r. vel' transmitter) failed low?

a. LCV-459 would close b.ELCV-460 would close

c. Pzr backup heaters would turn off

d. All orifice isolation valves will close I

A: B l

1

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R:'LO-LP-16301-00, FSAR Logic 7.2.1-1,,

EB#: LO-SS-23000-01-01 Point Value: 1.00

..====================================================================

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Page 7

EXAM-KEY. <

NRC REQUAL EXAM.- WEEK 1,PART 1 - SRO ,.

.============================================================================== . !-

1 1".10. Q .If rods were placed in automatic they;would: step: ,

-a. In due to' failed NR temperature;1nstrument.

b .- In due to failed Impulse pressure instrument..

c. Out due to failed NR temperature instrument.

d. 'Out due to failed Impulse pressure instrument.

A: B- -

R: LOGICS, EB#: LO-SS-23000-01-13 Point Value: 1.00

============== ====================================================

1.11 Q: NOTE: THIS QUESTION POSES A SITUATION THAT IS NOT RELATED TO THE CURRENT. PLANT /CONTPOL BOARD CONDITIONS!!!!!!

If BTRS were in service for dilution of the RCS, ALB07-E4 alarming l would result in:

a. No effect to BTRS operation while in the DILUTE MODE.

b. Flow diverting around the CVCS demineralizers, while maintaining flow through the BTRS demineralizers, c.- Flow diverting around'the BTRS demineralizers, while maintaining flow through the CVCS demineralizers.

d. Flow diverting around both the BTRS and CVCS demineralizers.

A: D R: 17007-1,R1,PG.,

EB#: LO-SS-23000-01-14 Point Value: 1.00

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SR-90S-90-05.ASt-KEY Page.8 '

.' EXAM KEY ~

F NRC.REQUAL EXAM - WEEK:1,PART 1 - SRO

==================================================================

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i;1d 12 - Q:1 NOTE:. THIS QUESTION POSES'A SITUATION ~THAT ISiNOT RELATED.TO1THE. .,

CURRENT PLANT / CONTROL BOARD CONDITIONS!!!!!!

'Which of the following correctly describes / the effects of'a loss-E a

of INB01:on the 1A DG?

i

a. The DG would be inoperable dueLto loss of Fuel 011'

. Transfer-Pump.

b. The DG Would be considered operable'and'would remain.in standby. :

c. The DG would be inoperable due to the loss of lube' oil I circulating (keep-warm) pump.

d. The DG would not auto start due to loss of Train A control power.

(

A: C'

, R: -TECH SPECS, EB#: LO-SS-23000-01-15 Point Value: 1.00

==================================================================

END OF SECTION KEY'

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1 8TATIC 81NULATOR'- PART A-II SCENARIO

SUMMARY

.- SS-54 1

8CENARIO!TYPEt. Emergency.

j PLANT CONDITIONS:

-Turbine trip with no S/D's stuck open PORV w/ loss of 1BB06' i

QUESTIONS APPLICABLE TO TEIS BEANINATION SCENARIO A5401 A5412 A5422 - Same as A1812 i, AS402 AS413 A5423 - Same as A4705 AS403 AS414 'A5424-- Same as A4712 A5404 AS415 AS405 A5416 AS406 A5417 i AS407 A5418

-A5408 A5419 A5409 A5420 1 1-L A5410 'A5421 l' AS411 L !

VALIDATED:

REVIEWED TRAININel ,_, -

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d' t ' STATIC SIMULATOR..PART A . ;>

f SIMULATOR SETUP INSTRUCTIONS FOR 88-54 )

l ,

INITIALISE Mt IC-13, 30% Pwr, MOL r

'l INSERT MALFUNCTION 8:

NO.- NAME ENTRY METHOD =l L

59a Pzr Press Xmitter PT-455 (2, 59a, 0)- .i-Fail Low 62c Pzr Relief Valve PV-456A (3, 62C, 100, 0003)

Fail Open- t l 62b Turb Imp' Press PT-506 ( 4 ', 62b,.100)

Fall High

'105 Loss of Main.Turb Lube Oil (5, 105, 100).

135h Loss of 480V SWGR 1BB06 (6, 135h,,-0003) ,

IN8ERT-OVERRIDES I. NAME ENTRY METJOD l

't l' .i l~

None i

SELECT THE FOLLOWING SWITCH POSITIONS _;

NUMBER NAME POSITION PS 455G PRZR Rec Sel SW P-456 HS 40041 Rod Bank Selector Switch Auto I

' START THE SCENARIO l

PERFORM THE FOLLOWING ACTIONS Throttle AFW flow to 200 GPM to each SG.

FREEEE THE SIMULATOR AT: RCP Trip Criteria Met (1375 psig) l l

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STATIC SINULATOR - PART A- r s

t SINULATOR SETUP INSTRUCTIONS FOR 88 .

VERIFY THEJFOLLOWING CONDITIONS $

POWER: Approx. 5 X 10 ROD POSITION: Tripped BORON . Approx. 1020 ppa PSR'LVL > 20% PER-PRESS <-1375 psig TAVEs Approx. 540 degrees F

.l

'RCP STATUS: 4 Running.. ECCS STATUS: Injecting Eg PRESS:' Approx. 940 89 LY.Lt 15-20% NR j

, OTHER CONDITIONS: N/A i VERIFY THE FOLLOWING SWITCH POSITIONS:

i Rod Control in Automatic PRZR Recorder Selector Switch in the 456 position l

i SELECT THE FOLLOWING DISPLAYS BRF DISPLAY 1 Top level ,

I ERF DISPLAY l2: Heat Sink PROTEUS: RCS diagram i

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~ .,-~, 7 SR-905-90-05.AST-KEY _. l Page'8' ,

EXAM KEY !

NRC REQUAL EXAM-'- WEEK 1,PART 2- .SRO 4 J2.01. Q:-Select the correct location-whereithe leaking coolant is_being ,

collected.

~

a. The RCDT-

b. The Containment Sump-

, c. The PRT ,

.d.'The Containment Sump and the PRT A: C i R: SIM INDICATIONS, .

EB#: LO-SS-54000-01-02 Point Value: 1.00

.========= .======....=..====...====.............-===..........===

'2.02 Q:.PZR level' indication is increasing during this transient due to-which of the following factors.

a. RCS pressure is lowest in the PZR, therefore the inventory

.of the.RCS is flowing into.it,

b. The inventory of the PZR is-saturated-and the RCS-is subcooled.

c. BIT flow rate is greater than the~ leakage rate.

d. Rx Vessel Head Voiding is forcing water into the PZR. -i J

A: C

, R: SIM INDICATIONS,

l. 'EB#: LO-SS-54000-01-04 Point Value: 1.00 L ================================================u=============================.

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EXAM. KEY l NRC REQUAL EXAM - WEEK 1,PART 2 - SRO.

======================= -=== .....----====--======================

_a.-

2.03 Q:.What system / components actuated to replace the' turbine generator as austeam. release path immediately after the. turbine generator i, '

tripped?.

-a. Steam dumps.on the load' reject controller

b. Steam. dumps on the plant trip controller

c. S/G ARVs

d. S/G Safety Valves j A: C 1

' R: LOGICS, EB#: LO-SS-54000-01-05 Point Value: 1.00'

======================================wmAsewa=================================-

't r 2 '. 0 4 . Q: Which of the following correctly explains how the plant responded 1

to the preceding transient, ,

following the turbine generator trip the:

a. Reactor tripped immediately because of the P-9 permissive.- ' 1 b..RCS heated up rapidly causing-PZR level to increase to the high PZR level trip setpoint. j 1

c. RCS heated up rapidly causing a rapid insurge into !

the PZR which resulted.in a Rx. trip on rate compensated.

PZR high pressure,

d. SG shrink resulted in'a reactor trip on low-low S/G l levels.

A: D R: SIM. INDICATIONS', ,

EB#: LO-SS-54000-01-06 Point Value: 1.00

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_NRC REQUAL EXAM WEEK 1,PART 2 - SRO

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. 2 .' 0 5 ~Q: Which of-the following describes'the plant response that caused the automatic SI.

a.c The turbine trip caused the ARVs to open.: Steam line pressure decreased resulting in a SI/SLI.

b. The reactor tripped. . Steam dumps. opened on-the -

plant-trip. controller. S/G ARVs also being.open caused steam line pressure to decrease resulting.in.a SI/SLI. ,

c. Tns. turbine trip caused RCS pressure to increase and' a 1

.PORV to open.. The PORV failed to;close resulting in a low PZR pressure SI.- 'l

d. The turbine trip resulted in higher than normal' pressure when the reactor tripped. RCS pressure lowered resulting in-a. rate compensated low PZR pressure SI.

A: C I R: SIM INDIACTIONS, !

EB#: LO-SS-54000-01-08 Point Value: 1.00 l: - =============================================.================================

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3 SR-905L90-05.ASf' KEY 1 Page 11 EXAM KEY

-NRC REQUAL EXAM - WEEK:1',PART 2 - SRO

, ==============================================================================-

2.06. Q: LWhich of'the effects described below did the failure of-PZR-o pressure detector PT-455 have on=this transient?

a. The failure was a benefit'to the plant'during'the initial-F.

plant transient, because it called for PV-455A to be'open.

Therefore the RCS-pressure spike'was lower than expected,

b. The failure was a detriment to the plant during the initial plant transient because it prevented PV-455A from opening,

c. The failure was a detriment to the plant during the(initial plant transient because PV-455A and the spray valves failed to open during the transient,

d. The failure was a benefit to the plant during1the initial. !

plant transient because PV-455A~and the spray valves opened. I during the transient.

I

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R: LOGICS, SIM. INDICATIONS, !'

EB#:- LO-SS-54000-01-15 Point Value: 1.00

-==============================================================================-

12.07 Q:'The RO recommends that the RCPs be tripped. Which statement. --!

below properly suppor's, or refutes, his recomendation.

e

a. RCPs should not b9 secured, because by doing so you would ,

[- remove the safest aeans of depressurizing the RCS. i l

L

b. RCPs should be secured because of primary pressure and ECCS operation.

c. RCPs-should not be secured-because of_the potential for :

loss heat removal capabilities from the core. !

d. RCPs should be secured because they are contributing to the loss of coolant.

L <

A: B R: '19000-C, EB#: LO-SS-54000-01-10 Point Value: 1.00

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NRC REQUAL EXAM - WEEK 1,PART 2 - SRO

.=== ........................====== ........--.....--..... -----...-==========

.2.08 Q: The containment spray system is: 1

a. In. operation ~with chemical injection occurring. ;

b.'NOT in_ operation, but should be.-

c. In operation, but chemical injection is NOT occurring. ;

d. NOT.in operation and is NOT currently:needed.

A: D R: SIM INDICATIONS, 19000-C, EB#: LO-SS-54000-01-16 Point Value: 1.00 i '

==================================================================

2.09 Q: Select - the _ correct system response to the failure of impulse' pressu. detector PT-506. 3

a. No control' rod movement would have occurred as a result of the failure,

b. Steam dumps would have armed on the load reject controller but would not have opened.- l

c. Steam' dumps would not-have armed but would be set to open with= maximum demand if an arming signal had been generated.

d. Control rods-would not have moved because C-11 would have

'been blocking their movement.

A: 'A R: LOGICS,

[ .

EB#: LO-SS-54000-01-18 Point Value: 1.00

================.=================================================

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.NRC REQUAL EXAM - WEEK 1,PART 2 - SRO. ;

................................................................==.........===

I 2.10 Q: Based on.present plant-conditions, the minimum required AFW flow ;

i is:

t

a. 570'gpm to each intact'SG. }

b. 570~gpm.to all. intact SGs.

c. 50 gpm'to each intact ~SG.

d.: 0 gpm to all intact SGs.

A: D R: 19000-C, -l EB#: LO-SS-54000-01-22 Point Value:El.00 '

==.======.===.=.-=. .=.............................=========.=...=.=.....==...

-2.11 Q:'The current Control' Room HVAC system alignment is:

a. Incorrect because CRI should only be actuated on a high-radiation sensed on the air intake line,

b. Correct because CRI is automatically actuated on an SI.

, c. ' Incorrect'because both trains of filter units have '

(. automatically started on a SI. .

d. Correct because CRI is automatically actuated when smoke'is ;

l' detected in the air intake line.

{

l A: B R: LOGICS, EB#: LO-SS-54000-01-23 Point Value: 1.00 4 l ================.=========.====================.============================== ,

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NRC REQUAL-EXAM - WEEK 1,PART 2 - SRO-

.....on.........n........................................................====.

I 2412 Q: 1 NOTE: THIS QUESTION POSES.- A SITUk 'ON THAT IS NOT RELATED TO THE CURRENT PLANT / CONTROL-BOARD CONDITIO. I!!!!

Which of _ the following is: correct concerning the failures that required the-quickest operator response prior to the Rx trip inL *

--terms of Tech Spec. action requirements?

a. ' The PT-455 - f ailure requires the least- response time, b.'The PT-506 failure requires the least response time.-

c. The-PT-506 failure would require less time if power were ~

below P-13. i a

d. The PT-455 and the PT-506 both require the same response time.

A: D R: TECH SPECS, LOLP39208-03,

.EB#: LO-SS-54000-01-21 Point Value: 1.00.

======================================....========================

END OF SECTION KEY f

1 I

END OF TEST KEY l l

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ag7:- , ,

h The';following is an alpha-numeric list of figures.and_ references which- :

' should' accompany.'this test when-it is administered. '

' Reference' -Question l

' Static . Sim 2 3 : Intro 1- is Static-Sim 23 Intro- 2

' Static Sim-23 Intro 3. i

~ Static Sim 23: Intro 4' l' Static Sim 23-Intro' 5 ,

' Static Sim~23-Intro :6 j I

Static,Sim'23 Intro 7 Static'Sim 23 Intro 8

' Static Sim 23' Intro 9i l i

i

+

i

.I

['

i i

i END OF FIGURE AND REFERENCE LIST

___-_----_-----,_------------7 (L

,,. EXAM' PULLED-FOR CCR FILING SUPV DATE8-PLANT VOGTLE TRAINING-DEPT.

QUESTION USE LIST for EXAM:' SR-905-90-05. AS D

' Total Points: 24.00 FILE NO:C90-06-016 Page'1 j Asso.1 bled by Chuck Stuhaan on 07/23/90 in MANUAL mode.

ITEM EB NUMBER REVISION DESCRIPTION

.{

1.01 LO-SS-23000-01-05 4 06/20/90 Tech Spec Application

L l

1.02 LO-SS-23000-01-03 3 06/20/90 Steam Dump Operation ~ 1 1

1.03 -LO-SS-23000-01-04 4 06/20/90 ' Turbine Contr'ol System' i 1.04: -LO-SS-23000-01-06 4 07/12/90 Emergency'Boration Requirements' I 1.05. LO-SS-23000-01-07 5 06/28/90 Failure diagnosis ~ !

1.06 LO-SS-23000-01-09 8 07/12/90 Dilution calculation ~

l 1.07 LO-SS-23000-01-11 6 06/20/90 Tech Spec Application ~ !

1.08 LO-SS-23000-01-02 4 06/20/90 Tech Spec Application ~

1.09 LO-SS-23000-01-01 5 06/20/90 Pzr level interlocks' 1.10 LO-SS-23000-01-13 00 / / Rod Control response to instrument j l

A.11 LO-SS-23000-01-14 1 06/20/90 BTRS divert on high temperature

.1.12 LO-SS-23000-01-15 3 07/12/90 Loss of 1NB01 effects on DG operabi ,

===================================================n========================== '

END OF SECTION i

I

--,-,,-----isi---i---isimii-i i ii

7y 3 "li ' .. 7 '

As ,

a EXAM = PULLED FOR CCRcFILING SUPV1 DATE:

4 1

PLANT VOGTLE TRAINING DEPT.

QUESTION USE LIST for EXAM: SR-905-90-05.AS.

C , .i TotalLPoints:124.00 FILE NO:C90-06-016 Page 1

-Assembled by Chuck Stuhaan on 07/23/90 in' MANUAL' mode. ,

ITEM EB NUMBER REVISION' DESCRIPTION '

'I

-2.01- LO-SS-54000-01-02 6 06/20/90 leak diagnosis !

x , '

2.02 -LO-SS-54000-01-04 5- 06/20/90 event diagnosis

'2 . 0 " ' LO-SS-54000-01 ~5 06/28/90 . post trip diagnosis 12.04' LO-SS-54000-01-06 5 06/20/90 turbine-trip / reactor trip'cause1and 2.05 LO-SS-54000-01-08 3 06/20/90 causes of si 2.06 LO-SS-54000-01 4 06/20/90 effects 455 had on transient- ' '

2.07 .LO-SS-54000-01-10 4 06/28/90 rcp trip criteria-I: 2.08 LO-SS-54000-01-16 2 06/28/90 CS operation? t R2.09 LO-SS-54000-01-18 4 06/20/90 rod response to pt-506. failure 1

2.10 LO-SS-54000-01-22 00 / / AFW Throttling Limitations-2.11. LO-SS-54000-01-23 2 06/29/90 Control Room HVAC alignment.

2 '.12 : LO-SS-54000-01-21 9 06/29/90 tech spec useage

============================================================================== l END-OF SECTION 4

(

f 1 L END OF QUESTION LIST l-p

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" f. , W4ek l- f'.cf8T(R0)L

_ vy-.

- til C90-05-002

. PLANT VOGTLE TRAINING _ DitPT.

M.A:S T E_R K E Y. ,f; FOR EXAM:=SR-905-90-05;B 1

-Total Points: 24.00 j l

ASSEMBLED IN MANUAL MODE.

i

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._-- _-_ _m_m__ _ _ . _ - _ _ _ _ _ _ .m.m_ .-. ___m.-_-_..m_ _. _ _-_ . . . -

yli , 'SR-905-90-05.B;! KEY' Page 1 EXAM KEY

=====================================================================w.

1. - Q: Which one of the following combinations of-improperly positioned-valves'would result in inadequate. train "A" low head! injection-

' flow to:the RCS following a large break LOCA? Assume all other y

= valves and controllers are aligned to their normal standby ;

condition.

i

a. FV-610, RHR~ pump A miniflow, OPEN and.

HV-606, RHR~ pump A-discharge,. CLOSED ~

j

-l l

b. -1205-U4-226, RHR test recirc to RWST,-OPEN and -

i 1205-U4-021, RHR.HX A outlet to'CVCS letdown,.OPEN

c. 1205-U4-226, RHR test recirc to RWST, OPEN and !

1205-04-027, RHR test.recirc to RWST, OPEN

-i

d. HV-8811A, RHR pump'A suction from CNMT sump, OPEN and FV-610, RHR pump A miniflow, OPEN

' e-A: c.

R: 1X4DB122, 19000-C,R9,PG. 8, EB#: LO-OR-13301-02 Point Value: 1.00-

,=========================================================..===================

9

2. Q: Which of-the following provide water for the Seismic Category I Dry Standpipe System? i i

a. River Water Makeup System

b. Fire Water Storage Tanks L

c. Makeup Well Water System

d. Nuclear Service Cooling Water System i

A: d.

1 R: 1X4DB133-1, FSAR 9.5.1, EB#: LO-OR-43101-13-01 Point Value: 1.00

==========================n=======================================

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$4 1 SR-905-90-05.B; KEY

. :Page;2:

EXAM KEY-

============================================================================== q

, 3.- Q:"Which:one'of the.following is a continuous action step?

a.: Check- SG levels narrow range level in at-least one SG --

GREATER:THAN-5%

~

b.. Control feed flow to maintain'S/G narrow range level between !

5%-and 50%

t

c. Transfer condenser steam dump to STEAM PRESSURE MODE

d. Determine if natural circulation ecoldown is required. .

l 1

A: b. 'l R: WOG BACKGROUND DOC, 19001-C,R8,PG. 6, ,

EB#: LO-OR-37002-04-01 Point Value: 1. 0 0 -. !

=============,================================================================.

l

/4. Q: A LOCA has. occurred. While the operators are performing 19010-C, Loss of Reactor or Secondary-Coolant, a red path _ develops on the containment CSFST. The control room operators enter'19251-C, Response'to High Containment Pressure. They complete all actions !

13f the FRP and return to 19010-C.- When they return to 19010-C, !

they observe that the containment CSF has not been restored. The containment.CSFST continues to show a red-path.- !

1 With these conditions, the operators should

a. Continue with the actions of 19010-C. 19251-C does not have to' I be implemented again.

b. -Implement 19251-C again, and continue the actions until the red path is cleared.

c. Return to the last step of 19251-C, and hold until the red l path is restored. ;

d. Stay on the step in effect in 19010-C until Reactor Engineering determines if 19251-C should be performed again.

A: a.

R: 19251-C,R3,PG. 1, G2F4 000-069-006, EB#: LO-OR-37002-08-03 Point Value: 1.00

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..e SR-905-90-05.Bt KEY-Page 3 EXAM KEY

==============================================================================-

5. . . Q : A-reactor trip occurs-from full power. The control room operators verify;that the reactor and the turbine are'both. tripped'and that the AC emergency buses; are both energized.: An operator checks to-see if-SI is requiredfand: notes that RCS pressure-is;1845 psig and 'l

. steadily decreasing. No SI annunciator or BPLP light is' lit.

For this reactor trip event, SI.has-a.- Not occurred but is required. The operator should manually-initiate'SI.

b.- .Not occurred and'is.not required. The operator should transfer

~

to 19001-C to stabilize the primary and. secondary. plants at- l no-load conditions.

c. . Occurred and11s required. The operator should continue with. -l the immediate-actions of 19000-C because SI is already in.

progress. -t d.- Occurred but is not required. The operator should immediately -j terminate SI. ;

A: q.

R: 19000-C,R9,PG.3, GREB 000-007-003,; I EB#: LO-OR-37011-06-04 Point Value: 1.00: ,

============================================================================== '

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.t SR-905-90-05.B;~ KEY:

Page'4, !

EXAM KEY 4

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6. Q: A total l'oss' of all AC power has occurred. Steam is being released locally _via the S/G ARVs in an effort to. reduce S/O' i pressure to 265 psig. A low steamline pressure SI' signal has been received.. When steamline pressure reaches'315 psig and-RCS cold leg. temperatures are between 320 degrees F and 330 degrees F, the licensed operator monitoring the Critical Safety Functions informs, you that the. source range startup rate is reading +0.2 dpm with no. ,

t

' indication as to why. Your-action should be-to:-

a. Begin emergency boration

.b. Secure dumping st9am and heat up to add negative.

l reactivity

c. Continue to lower SG pressure f

d. Try.to start one RCP A: b.

R: 19100-C,R4,PG. 12, .

EB#:'LO-OR-37031-09-04 Point Value:-1.00_

.====u==========....=========================....==.===a_==============n====

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SR-905-90-05.Bt KEY Page 5 ,

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7. Q: The reactor has failed to automatically trip u;>9h required and cannot be manually tripped. The turbine is tripped, the ArW pumps are running, power is still above 5%, and emergency boration is in progress._ Pressurizer pressure is at 2375 psig. Both PORV's are open, but.their associated block valvas are closed. A control room operator succeeds in opening one block valve and reduces PZR pressure Below 2135 psig Under these conditions, the main reason for reducing pressure is tot

a. Prevent the rapid overpressuritation transient expected with most ATWT events.

b. Minimize primary-to-secondary' lea'kage in case of the most 7.initing ATWT event, a SGTR, until other recovery actions can be taken.

c. Allow enough borated water to flow into the RCS to encare the addition of negative reactivity to the core.

d. Begin a slow, controlled cooldown and depressurization, thereby minimizing positive reactivity feedback via a negative MTC..

l A: C.

R: 19211-C,R3, PG. 3, GREB 000-029-003, EB8: LO-OR-37041-08-02 Point Value ?. 00

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SR-905-90-05.0; MEY 4 T Page 6 EXAM KEY-

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8.- Qt The plant is in an emergency condition, and the control room E operators are performing step 2 of 19231-C, Loss of Secondary Heat Sink. They verify that a secondary heat sink is required and attempt to restore AFW flow. They are unsuccessful, so they stop the RCP's.

The PRIMARY reason for stopping the RCP's at this point in the procedure is to

a. Reduce RCS pressure to ensure subsequent SI flow is adequate for ECCS requirements.

b. Establish natural circulation conditions that wil' tend to mitigate the transient.

c. Reduce the heat input from the RCP's, thereby delaying the need for feed and bleed and gaining time to establish a means of supplying FW to a S/G.

d. Prevent the hect added by the pumps from masking indications used to determine whether or not RCS feed and bleed will be required.

At C.

R: 19231-C,R11,PG. 3, GREB 000-054-003, EB#1 LO-OR-37051-04-01 Point Value: 1.00

=u================================================================

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. SR-905-90-05.B;' KEY I

, Page 7 EXAM KEY

==================================================================

9. Q: What is accomplished by performing the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months temperature soak of l 19241-C, Response to Imminent Pressurized Thermal Shock? l 1

a. Allows time for any bubble that may have formed in the vessel !

head area to collapse. i

b. Gives the operator ti4ne to depressurize the RCS to minimize brittle fracture concerns. i

c. The coak allows thermal gradients in the vessel wall to be j reduced, thus reducing corresponding stresses.

I

d. Gives the operator time to. terminate ECCS flow thereby ;

minimizing the threat of a repressurization accident. ]

i A: c. :

R: WOG BACI; GROUND DOC., 19241-C,R7,PG. 14, EBW: LO-OR-37071-06-02 Point Value: 1.00

=================,============================================================ ,

10. Q: A'SGTR '.tas ocurred on Unit one. The operating team is currently on step 14 of 19030-C (SGTR). In step 14, the operator correctly determined the required core exit T/C tempars',ure to be 506 degrees F. During the rapid RCS cooldown tise operating team observes that ruptured SG pressure has decreased to 900 psig. The team decides to continue the cooldown to 403 degrees F. Is this action appropriate? Explain. ,

a. No. The required core exit thermocouple temperature should be determined only once prior to commencing cooldown.

b. No. The crew should immediately transition to 19131-C.

c. Yes. It is required to ensure adequate subcooling exists after depressurizing the RCS.

d. Yes. It is required as per step 14c. of 19030-C.

A: a.-

r R: 19030-C,R8,PG. 12, EB#: LO-OR-37311-07-10 Point Value: 1.00

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. SR-905-90-050B; KEY Page 8 l EXAM KEY

.o...................,.........................................................

II. Q: While the plant is operating at 85%, due to feedwater heater maintenance, the POWER RANGE CHANNEL DEVIATION annunciator and th6 ROD AT BOTTOM annunciator both alarm. One rod is verified on the bottcm using DRPI. The QPTR is calculated and results in a QPTR of 1.03.

In rer-sw ' to this, the control room operators MUST perform which-of tre fol ting?

a alculate the QPTR each hour until it returns within its

.imits.

b. Reduce thermal power to less than 50% within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

c. Reduce the Power Range Neutron F1tx high trip setpoint to 91% within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

d. Immediately commence a power reduction and be in hot standby within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

A: a.

Rt T.S. 3/4.2.4, 18003-C,R6.PG. 4, GREB 015-000-006 EB# LO-OR-39206-03-06 Point Value: 1.00 l

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c., i b ., SR-905-90-05.B; KEY j Page 9 l

((

y EXAM KEY j .------------ ...............................---------------------------------- ,

I !

12. Q: The reacter was shutdown at 1600 hours0.0185 days 0.444 hours 0.00265 weeks 6.088e-4 months on Jaruary 3rd. While r operating at midloop three days later, a cose ate loss of R'.iR occurs at.1600 hours0.0185 days 0.444 hours 0.00265 weeks 6.088e-4 months and cannot be correcte;. Within a few minutes, core exit thermocouples begin to indicate >200 degrees F.

The Unit Shift Supervisor orders a charging pump started to remove the decay heat. The minimum flow rate required under these conditions is:

l

a. 300 gpm ,

b. '150 gpm

c. 120 gpm

d. 80 gpm i A: d. 1 Rt 18019-1,R6,PG. 15, ED#: LO-OR-12101-19-01' Point Value: 1.00 r

13. Q: The plant is operating at 100% RTP Ten hours ago, PZR level '

channel LC-460 failed off scale low. All actions that are required to allow continued plant operations have been completed. Now level channel LC-459 also fails off scale low.

The operators should

a. Select LC-461 as the controlling channel, and continue at-power operations.

i

b. Take manual control of charging flow, and continue 100% power operation.

c. Reduce power to belo'. P-7, where T.S. no longer applies for this condition. 3.0.3 is in effect.

d. Immediately trip the reactor.

A: c.

R: 18001-C,R7,PG. 8, GREB 000-028-003, T.S. 3.3.1 EL#: LO-OR-39207-03-07 Point Value: 1.00

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t SR-905-90-05..St KEY EXAM KEY p ..........................................................=...................

14. '

Q -Reactor power is 100%, Tavg is 588 degrees F, and pressurizer

, level is 60%. The pressurizer level control selector switch is in the "459-460" position with the PDP operating in AUTO. Indicated charging flow is 75 gpm. Channel I pressurizer level transnitter

, (LT-459) fails high. Prior to any operator actions, approximately what will charging flow indication be?

L a. O gpm

b. >0 and <75.gpm

c. 75 gpm

d. >75 gpm A: b.

R: LOGICS 7.2.1-1, 18 0 01-C , R7, PG . 7, 17011-1,R5,PG. 32 EB#: LO-OR-iG302-03-01 Point Value: 1.00

15. Q: The Unit i reactor is being refueled when an announcement is made to evacuate the containment building. Why are personnel directed to remain outside the airlock?

a. For accountability and radiological monitoring purposes.

b. To act as rescue team members, if needed.

c. To allow fission product gases time to decay from clothing.

d. This allows Security Department personnel time to set up a an evacuation route for contaminated personnel.

A: a.

R: LO-LP-25201 00, 18006-C, EB#: LO-OR-60306-01-01 Point Value 1.00

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, l SR-905-90-05 B) KEY Page 11 l EXAM KEY :

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16. Qt The plant is operating at power, when inverter 1AD1I11 fails causing 1AY2A to-be deenergized. Before 1AY2A can be reenergized a spurious SI occurs. Which of the following describes actions that will occur with 1AY2A deenergized.

a. INB01 will not strip, A D/G will not start, A CCP will not start, and A Train CIA will occur.

b. 1NB01 will not strip, A D/G will start, A CCP will start, and A Train CIA will occur. t

c. 1NB01 will strip, A D/G will not start, A CCP will not start, ,

and A Train CIA will occur.

d. 1NB01 will strip, A D/G will start, A CCP will not start, and h A CIA will occur.

A: d '.

R 1X3D-CE-H04Q, 1X3D-AA-G02C, '

EB#: LO-OR-60324-01-02 Point Value: 1.00

====================================v ============================

17. Q: Plant electricians would like to perform scheduled maintenance on inverter 1AD1I11. Which of the following prevents supplying bus ,

LAY 2A from inverter 1AD1I11 and 480 VAC MCC 1ABB simultaneously?

a. Mechanical. interlock.

b.- Electrical interlock.

c. Taking maintenance lockout switch to MAINT position on 1ABB.

I

d. Administrative guidelines.

A: a.

R: 13431-1,R4,PG. 2, LO-LP-60324-01, EB#: LO-OR-01103-03-01 Point Value: 1.00

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.Page 12 EXAM KEY

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18. Q: The B train of RHR is being aligned for plant cool 6own. Valves HV-8804A:and MV-8804B are currently inoperable due to problems with their motor operators. Both valves currently have power removed from them. How will-the condition of HV-8804A and HV-8804B affect the performance of the RHR alignment?

i

a. This will have no effect on RHR alignment as long as the 8804 valves were closed when power was removed.

b. This will prevent RHR alignment from the QMCB because HV 8701B will only open from the remote shutdown panel.

c. This will prevent RHR alignment from the QMCB because HV 8702B will only'open from the remote shutdown panel.

d. This will prevent RHR alignment totally until power is restored to the 8804 valves.

A: c.

R: 13011-1,R17,PG. 6, EB#: LO-OR-12101-08-03 Point Value: 1.00-

...u..........................................................................

19. Q: The Reactor Operator notes that No. 1 seal leakoff flow has risen above 6 gpm. Which of the following actions are required to be performed'with respect to the affected Reactor Coolant Pump?

a. Isolate seal leakoff within 5 minutes and trip the RCP when.

seal delta P decreases below 200 paid,

b. Isolate seal leakoff within 5 minutes and trip the RCP within !

30 minutes after the seal leakoff is isolated,

c. Trip the RCP if no. 1 seal delta P decreases below 200 psid or if no. 1 seal leakoff flow decreases below .2 gpm.

d. Isolate seal leakoff within 5 minutes and trip the RCP when no. 1 seal leakoff flow decreases below .2 gpm. I A: b. '

i R: 13003-1,R5,PG. 1-2, EB#: LO-OR-16401-04-01 Point Value: 1.00

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. ER-905-90-05.Bf KEY Page 13 EXAM KEY

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20. . Qt To reset the TDAFW pump trip and throttle valve (PV-15129) following a mechanical overspeed trip, which of the following actions must be performed?

a. Reset the mechanical overspeed trip linkage locally and manually open the trip an<l throttle valve using HS-15111.

b. Reset the mechanical overapeed trip linkage locally, open the steam supply valve, K't-5106, then manually close the trip and throttle valve using HS-15111.

c. Manually close the trip ant throttle valve using HS-15111 and then open the trip and thrcttle valve using HS-15111. 4

d. Manually open the trip and-throttle valve using HS-15111, reset the mechanical trip linkage locally, then close the trip and throttle valve using HS-15111.

A: a.

R: 13610-1,R9,PG. 6, 1X3D-BC-F02, 17016-1,R8,PG. 53 EB#: LO-OR-20101-10-01 Point Value: 1.00

==================================================================

21. Q: Containment pressure is 1.1 psig. Chemistry has issued a release permit to allow containment pressure to be reduced. Which of the following lineups best describes the flowpath to be used initially during the pressure reduction evolution.

a. HV2628B & HV2629B open; HV12592 AUT0; Mini Purge Exh Fan running

b. HV2628B & HV2629B open; HV12592 closed; Mini Purge Exh Fan running

c. HV2628B & HV2629B open; HV12592 AUTO; Mini Purge Exh Fan stopped !

I

d. HV2628B & HV2629B open; HV12592 closed; Mini Purge Exh Fan stopped A: d.

R: 13125-1,R12,PG. 10, .

EB#1 LO-OR-29110-03-01 Point Value: 1.00 l

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., SR-905-90-05.B; KEY.

Page 14 EXAM KEY

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22. Q: Which of the following events / conditions would cause the value of the estimated critical boron concentration to decrease?

a. Estimated startup time increases from 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months after a trip from 75% power, to 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months after the trip.

b. Desired critical rod height changes from 130 steps / Bank "D" to 140 steps / Bank "D".

c. The anticipated Tavg at startup is changed from 557 degrees.

F to 554 degrees F.- (Assume a negative MTC)

d. Reactor power history used in the estimated critical boron-calculation is corrected from 40% to 50% power.

A: d.

R PTDB TABl.4.1-T1,, 1.5.1-T3, 14940-1,R8,PG. 1 EB# LO-OR-33510-07-02 Point Values 1.00

..........-===...........-=.........................................n.........

23. Q: Which one of the following would require action to be taken within the next two hours to commence a unit power descent?

a. Four-loop operation with power at 90%, Tavg at 620 degrees F, "

and pressurizer pressure at 2235 psig.

b. 92% RWST level with boron concentration at 2430 ppm, and temperature oc 72 degrees F.

c. RCS activity at 1.2 microcurie per gram DOSE EQUIVALENT I-131 for 50 hours5.787037e-4 days 0.0139 hours 8.267196e-5 weeks 1.9025e-5 months at full power,

d. Discovering that only one NSCW pump was operable on NSCW train A 47 hours5.439815e-4 days 0.0131 hours 7.771164e-5 weeks 1.78835e-5 months ago.

At a.

R: T.S. 3.2.5, EBf: LO-OR-39206-03-02 Point Value: 1.00

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-. - SR-905-90-05.B; K0Y 1 Page 15-EXAM KEY j 2

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24. Q: Criticality in the Spent Fuel Storage Racks is prevented by:

I

a. Using a boron absorber in the storage racks. !

I

b. Decreasing the number of fuel assemblies stored in the racks. ]

l

c. Ensuring borated t*3ter'is used in the Spent Fuel Pool and ?

spacing of at least-(3; three inches exists, center-to-center, between assemblies.

I

d. Storing spent fuel in borated polyethylene wrapping. j I

A: a.

R: FSAR 9.1, V0GTLE TEXT CH. 18B, EBf: LO-OR-25102-04-01 Point Value: 1.00 L

f i

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i END OF TEST KEY l

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?

3:

1:

The following is an alpha-numeric list of figures and. references which should accompany this test when it is administered.

Reference Question L

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SUPV DATE:

PLANT VOGTLE TRAINING DEPT..

QUESTION USE LIST for EXAM: SR-905-90-05.B Total Points: 24.00 FILE No:C90-05-002 Page 1 Assembled by Chuck Stuhaan on 07/23/90 in MANUAL mode.

ITEM EB NUMBER REVISION DESCRIPTION

1. LO-OR-13301-02-02 11 06/29/90 Low head SI flowpath

2. LO-OR-43101-13-01 2 06/29/90 State the source of water to the Se

3. LO-OR-37002-04-01 10 06/29/90 EOP Usage - Continuous action steps

-4. LO-OR-37002-08-03 3 06/29/90 Crew responds to Red path, returns

.5. LO-OR-37011-06-04 6 06/29/90 Perform IOA's per 19000-C. SI if SI

6. LO-OR-37031-09-04 8 06/29/90 What to do with CSFST's while in 19 7 LO-OR-37041-08-02 4 06/29/90 Reason to reduce pressure on ATWT.

8. LO-OR-37051-04-01 5 06/29/90 Reason for stopping RCP's during a

9. LO-OR-37071-06-02 9 06/29/90 Function of the temperature soak us I

10. LO-OR-37311-07-10 8 '06/29/90 Performance of step 14 of SGTR proc j

11. LO-OR-39206-03-06 '4 06/29/90 OPERATOR RESPONSE TO QPTR OF 1.03 l i

12. LO-OR-12101-19-01 8 06/29/90 Minimum charging flow after Loss of

13. LO-OR-39207-03-07 5 06/29/90 What to do if two PZR level' channel

14. LO-OR-16302-03-01 9 06/29/90 Automatic PRZR Level Control (KA 3,

15. LO-OR-60306-01-01 12 06/29/90 Concerns during Containment evacuat

16. LO-OR-60324-01-02 9 07/12/90 Effect of loss of vital bus w/emerg

17. LO-OR-01103-03-01 13 06/29/90 120 VAC interlocks / Loss of 120V AC-

18. LO-OR-12101-08-03 9 06/29/90 RHR Operations while shifting to RC

19. LO-OR-16401-04-01 12 06/29/90 RCP ops w/ Seal Abnormality

20. LO-OR-20101-10-01 11 06/29/90 How to reset the TDAPW Pump T&TV '

21. LO-OR-29110-03-01 11 07/12/90 Describe how to vent containment wi

22. LO-OR-33510-07-02 9 06/29/90 ECC Boron Concentration Changes

=====================hf8T=89MTENWED=9N=MBMT=PAGE==================

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[,;. SR-905-90-05.B Page.2 QUESTION USE LIST

~L ITEM EB NUMBER REVISION- DESCRIPTION !

23. LO-OR-39206-03-02 12 06/29/90 Tavg T.S. interpretation using give

24. LO-OR-25102-04-01 3 06/29/90 !!ow criticality is prevented in the

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vi.4 Week ~l Part' 8 (SRD) ,

C90-05-003-i PLANT VOGTLE TRAINING DEPT. ,

t '

MASTER KEY FOR l EXAM: SR-905-90-05.BS Total Points: 24.00 ,

ASSEMBLED IN MANUAL MODE.

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L 4'L, SR-905-90-05.BS KEY Page 1 EXAM KEY

1. Q: Which one of the following combinations of improperly positioned valves would result in_ inadequate train "A" low head injection flow to the RCS following a large break LOCA? Assume all other valves and controllers are aligned to their normal standby condition.

a. FV-610, RHR pump A miniflow, OPEN and HV-606, RHR pump A discharge, CLOSED

b. 1205-U4-226, RHR test recirc to RWST, OPEN and 1205-U4-021, RHR HX A outlet to CVCS letdown, OPEN

c. 1205-U4-226, RHR test recirc to RWST, OPEN and 1205-U4-027, RHR test recirc to RWST, OPEN

d. HV-8811A, RHR pump A suction from CNMT sump, OPEN and FV-610, RHR pump A miniflow, OPEN A: c.

R: 1X4DB122, 19000-C,R9,PG. 8, EB#: LO-OR-13301-02-02 Point Value: 1.00 KEY CONTINUED ON NEXT PAGE

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EXAM KEY

2. Q: Both units are operating at 100% power. The operators are preparing to perform the weekly operability test of the Diesel ,

Fire Pump fl. Fire System pressure is at 98 psig. The first I attempt to start the Diesel Fire pump'in Auto did not work and a second attempt to start it also fails. The pump is declared inoperable. As a precautionary measure the control room operators verify the other components in the Fire Protection System are also i operable. Checks on the North and South Fire Water Storage Tanks ;

show level to be 26.5 feet.

What action (s) must be taken by the onshift operating crew?

a. Restore one storage tank to OPERABLE status within the next 24 l hours or be in hot standby in the next 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . !

l'

b. Restore both inoperable storage tanks to OPERABLE status (or establish a nominal backup system) in the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or ;

commence an orderly shutdown of the plant. 1

c. Restore Diesel Driven Fire Pump #1 to operable status within 1 hourt no further compensatory actions are required. $

d. Restore the system OPERABLE within the next 7 days or be in l hot standby in the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . ;

A: b.

R: 92035-C,R5,PG. 26, EB#: LO-OR-22101-06-01 Point Value: 1.00

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3. Q: Which one of the following is a continuous action step? j ,

a. Check SG levels narrow range level in at least one SG - l GREATER THAN 5% ;

b. Control feed flow to maintain S/G narrow range level between-L , 5% and 50% ,

c. Transfer condenser steam dump to STEAM PRESSURE MODE

d. Determine if natural circulation cooldown is required A: b.

R: WOG BACKGROUND DOC, 19001-C,R8,PG. 6, i EB# LO-OR-37002-04-01 Point Values 1.00 t

...........................................ao............r....................

4. . Q: A LOEA has occurrod. While the operators are performing 19010-C, ,

Loss of Reactor or Secondary Coolant, a red path develops on the' containment CSFST. The control room operators enter 19251-C, Response to High Containment Pressure. They complete all actions of the FRP and return to 19010-C. When they return to 19010-C, i they observe that the containment CSF has not been restored. The containment CSFST continues to show a red path.

With these conditions, the operators should

a. Continue with the actions of 19010-C. 19251-C does not have to be implemented again. ;

b. Implenunt 19251-C again, and continue the actions until the i red path is cleared.

t

c. Return to the last step of'19251-C, and hold until the red path is restored.

d. Stay on the step in effect in 19010-C until Reactor Engineering determines if 19251-C should be performed again.

At a.

R: 19251-C,R3,PG. 1, GREB 000-069-006, EB#: LO-OP-37002-08-03 Point Value: 1.00

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!. Page 4 EXAM KEY

5. Q: A reactor trip occurs from full power. The control room operators verify that the reactor and the turbine are both tripped and that the AC emergency buses are both energized. An operator checks to see if SI is required and notes that RCS pressure is 1845 psig and steadily decreasing. No SI annunciator or BPLP light is lit.

For this reactor trip event, SI has

a. Not occurred but is required. The operator should manually initiate SI.

b. Not occurred and ic not required. The operator should transfer to 19001-C to stabilize the primary and secondary plants at no-load conditions.

c. Occurred and is required. The operator should continue with the immediate actions of 19000-C because SI is already in progress.

d. Occurred but is not required. The operator should immediately terminate SI.

A: a.

R: 19000-C,R9,PG.3, GREB 000-007-003, EB#: LO-OR-37011-06-04 Point Values 1.00

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SR-905-90-05.BSt KEY Page 5 EXAM KEY

6. Q A total loss of all AC power has occurred. Steam is being released locally via the S/G ARVs in an effort to reduce S/G pressure to 265 psig. A low steamline pressure SI signal has been received. When steamline pressure reaches 315 psig and RCS cold leg temperatures are between 320 degrees F and 330 degrees F, the licensed operator monitoring the Critical Safety. Functions informs you that the source range startup rate is reading +0.2 dpm with no indication as to why. Your action should be tot

a. Begin emergency boration

b. Secure dumping steam and heat up to add negative-reactivity

c. Continue to lower SG pressure

d. Try to start one RCP At b.

R: 19100-C,R4,PG. 12, EB# LO-OR-37031-09-04 Point Value: 1.00 i

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7. Q: The reactor has failed to automatically trip when required and cannot be manually tripped. The turbine is tripped, the AFW pumps are running, power is still above 5%, and emergency boration is in progress. Pressurizer pressure is at 2375 psig. Both PORV's are open, but their associated block valves are closed. A control room operator succeeds in opening one block valve and reduces PZR pressure Below 2135 psig Under these conditions, the main reason for reducing pressure is to:

a. Prevent the rapid overpressurization transient expected with most ATWT events.

b. Minimize primary-to-secondary leakage in case of the most limiting ATWT event, a SGTR, until other recovery actions can be taken.

c. Allow enough borated water to flow into the RCS to ensure the addition of negative reactivity to the core.

d. Begin a slow, controlled cooldown and depressurization, thereby minimizing positive reactivity feedback via a negative MTC.

As c.

R 19211-C,R3, PG. 3, GREB 000-029-003, EB#: LO-OR-37041-08-02 Point Value: 1.00

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EXAM KEY i

8. Q: The plant is in an emergency condition, and the control room operators are performing step 2 of 19231-C, Loss of Secon.dary Heat Sink. They verify that a secondary heat sink is required and attempt to restore AFW flow. They are onsuccessful; so they stop :

the RCP's.

The PRIMARY reason for stopping the RCP's at this point in the l procedure is to l

a. Reduce RCS pressure to ensure subsequent SI flow is adequate for ECCS requirements.

b. Establish natural circulation conditions that will tend to !

mitigate the transient.

c. Reduce the heat input from the RCP's, thereby delaying the I need for feed and bleed and gaining time to establish a means of supplying FW to a S/G. ;

d. Prevent the heat added by the pumps from masking indications used to determine whether or not RCS feed and bleed will be required. .

As c.

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9. Q: What is accomplished by performing the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months temperature soak of 19241-C, Response to Imminent Pressurized Thermal Shock? t

a. Allows. time for any bubble that may have formed in the vessel head area to collapse. ;

b. Gives the operator time to depressurize ths RCS to minimize brittle fracture concerns.

c. The soak allows thermal gradients in the vessel wall to be reduced, thus reducing corresponding stresses.

d. Gives the operator time to terminate ECCS flow thereby minimizing the threat of a repressurization accident.

A c.

R: WOG BACKGROUND DOC., 19241-C,R7,pG. 14, EB#: LO-OR-37071-06-02 Point Value: 1.00

=====....m===... ===============================.....=...= .....==

10. Q: A SGTR has ocurred on Unit One. The operating team is currently on step 14 of 19030-C (SGTR). In step 14, the operator correctly i determined the required core exit T/C temperature to be 506 degrees F. During the rapid RCS cooldown the operating team observes that ruptured SG pressure has decreased to 900 psig. The team decides to continue the cooldown to 493 degrees F. Is this action appropriate? Explain. i

a. No. The required core exit thermocouple temperature should be determined only once prior to commencing cooldown.

I

b. No. The crew should immediately transition to 19131-C.

c. Yes, It is required to ensure adequate subcooling exists after depressurizing the RCS. ,

d. Yes. It is required as per step 14c. of 19030-C.

A: a.

R: 19030-C,R8,PG. 12, EB#: LO-OR-37311-07-10 Point Value: 1.00

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! EXAM KEY 1

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11. - Q: Unit 1 is in Mode 3 at 547 degrees F, when a sustained loss of instrument air occurs. Prior to the event, charging and letdown

.w ere in their normal alignment. Which of the following statements is TRUE concerning pressurizer level and VCT level?

a. Pressurizer level and VCT level are slowly increasing,

b. Pressurizer level and VCT level are slowly decreasing.

c. Pressurizer level is slowly increasing and VCT level is slowly decreasing.

d. Pressurizer level is slowly decreasing and VCT level is slowly increasing.

A: c.

R: 1X4DB116-1.R19, 18028-C,R7,PG. 9, EB#: LO-OR-09201-12-03 Point Value: 1.00

======================================================u===========

12. Q:-The reactor was shutdown at 1600 hours0.0185 days 0.444 hours 0.00265 weeks 6.088e-4 months on January 3rd. While operating at midloop three days later, a complete loss of RHR occurs at 1600 hours0.0185 days 0.444 hours 0.00265 weeks 6.088e-4 months and cannot be corrected. Within a few minutes, core exit thermocouples begin to indicate >200 degrees F.

The_ Unit Shift Supervisor orders a charging pump started to remove the decay heat. The minimum flow rate required under these ,

conditions is: !

a. 300 gpm

b. 150 gpm

c. 120 gpm

d. 80.gpm A: d.

R: 18019-1,R6,PG. 15, EB#: LO-OR-12101-19-01 Point Value: 1.00

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SR-905-90-05. cst KEY l Pace 10 EXAM KEY 1

BMWSEmmmmmmmmmmWBBMEMBBS3mmBENSBWassBBMWmmmmmmmmmmWMBRBsamBENESENBWWWEBemmWBMW 1

I

13. Q: The plant is operating at 100% RTP. Ten hours ago, PZR level

channel LC-460 failed off scale low. All actions that are required to allow continued plant. operations have been completed. Now level ,

channel LC-459 also fails off scale low. ,

The operators should H a. Select'LC-461 as the controlling channel, and continue ;

at-power operations.

b.- Take manual control of charging flow, and continue 100% power operation.

c. Reduce power to below P-7, where T.S. no longer applies for this condition. 3.0.3 is in effect.

d. Immediately trip the reactor.

A: c.

R: 18001-C,R7,PG. 8, GREB 000-028-003, T.S. 3.3.1 EB#: LO-OR-39207-03-07 Point Value: 1.00

14. Q: Reactor power is 100%, Tavg is 588 degrees F, and pressurizer level is 60%. The pressurizer level control selector switch is in the "459-460" position with the_PDP operating in AUTO. Indicated charging flow is 75 gpm. Channel I pressurizer level transmitter (LT-459) fails high. Prior to any operator actions, approximately what will charging flow indication be? ,

i. a. O gpm i p

E

b. . >0 and <75 gpm

c. 75 gpm L d. >75 gpm l-l l-L A: b.'

R: LOGICS 7.2.1-1, 18001-C,R7,PG. 7, 17011-1,RS,PG. 32 EBf: LO-OR-16302-03-01 Point Value: 1.00 a==mm==========m=m=====================m=====B================================

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SR-905-90-05.BS9 XEY Page 11 EXAM KEY l

15. Q: The Unit 1 reactor is being refueled when an announcement is made ,

to evacuate the containment building. Why are personnel directed !

to remain outdide the airlock?

a. For accountability and radiological monitoring purposes.

b., To act as rescue team members, if needed. '

c. To allow fission product gases time to decay from clothing.

d. This allows Security Department personnel time to set up a an evacuation route for contaminated personnel.

A: a.

R: LO-LP-25201-00, 18006-C, ,

t EB#: LO-OR-60306-01-01 Point Value: 1.00 '

..................rs.......................................................... !

.16. Q: The plant is operating at power, when inverter 1AD1I11 fails causing 1AY2A to be deenergized. Before 1AY2A can be reenergized a spurious SI occurs. Which of the following describes actions that will occur with 1AY2A deenergized.

i

a. 1NB01 will not strip, A D/G will not start, A CCP will not start, and A Train CIA will occur.

b. INB01 will-not strip, A D/G will start, A CCP will start, and A Train CIA will occur.

c. 1NB01 will strip, A D/G will not start, A CCP will not start, and A Train CIA will occur.

d. 1NB01 will strip, A D/G will start, A CCP will not start, and A CIA will occur.

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A: d.

R: 1X3D-CE-H04Q, 1X3D-AA-G02C, EB#: LO-OR-60324-01-02 Point Value: 1.00

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17. Q: Which of the following' currently states how the possibility of !

personnel overexposure from spent fuel is minimized while operating with spent fuel in the New Fuel Elevator. ;

a. Administrative controls'and guidelines (i.e. Fuel Handling Procedure restrictions) and the New Fuel Elevator is interlocked to prevent raising the basket with the weight of ,

a spent fuel assembly in it.

b. The Fuel Handling Machine is interlocked such that it cannot ,

be positioned over the New Fuel Elevator with a spent fuel- !

assembly in it,

c. The New Fuel Elevator is interlocked to prevent raising the basket with the Fuel Handling Machine positioned directly over it.

d. A radiation monitor is located to sense increasing radiation levels.in the New Fuel Elevator area and stop upward movement.

A: a ..

R: VOUTLE TEXT, CH.38, 93210-C,R3,PG. 1, i EB#: LO-OR-25101-06-01 Point Value: 1.00

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18. Q: The B train of RHR is being aligned for plant cooldown. Valves HV-8804A and HV-8804B are currently inoperable due to problems with their motor operators. Both valves currently have power-removed from them. How will the condition of HV-8804A and HV-8804B affect the performance of the RHR alignment?

a._ This will have no effect on RHR alignment as long as the 8804 4 valves were closed when power was removed.

b. This will prevent RHR alignment from the QMCB because HV 8701B-will only open from the remote shutdown panel,

c. This will prevent RHR alignment from the QMCB because HV 8702B will only open from the remote shutdown panel.

d. This will prevent RHR alignment totally until power is restored to the 8804 valves.

-A: c.

R: 13011-1,R17,PG. 6, EB#: LO-OR-12101-08-03 Point Value: 1.00

.............m=========== .........m............================ ......m..====

I, l19. Q: The Reactor Operator notes that No. 1 seal leakoff flow has risen above 6 gpm. Which of the following actions are required to be performed with respect to the affected Reactor Coolant pump?

a. Isolate seal leakoff within 5 minutes and trip the RCP when seal delta P decreases below 200 psid.

b. Isolate seal leakoff within 5 minutes and trip the RCP within 30 minutes after the seal leakoff is isolated.

c. Trip the RCP if no. 1 seal delta P decreases below 200 psid or if no. 1 seal leakoff flow decreases below .2 gpm. l

d. Isolate seal leakoff within 5 minutes and trip the RCP when

, no. I seal leakoff flow decreases below .2 gpm.

1 A: b.

R: 13003-1,RS,PG. 1-2, EB#: LO-OR-16401-04-01 Point Value: 1.00

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L20. Q: To reset the;TDAFW pump trip and-throttle valve-(PV-15129) following a mechanical overspeed trip, - which of ' ae following -

actions must be performed?

, I Resetithe~ mechanical'overspeed trip linkage locally and -

a. -

manually;open'the' trip and throttle valve using HS-15111. 1

b. : Reset the mechanical overspeed' trip 1inkage, locally, open

-the' steam supply-valve, HV-5106, then manually close the-trip; and throttle, valve using MS-15111.. .j c.- Manually;close the' trip and throttle valve using HS-15111 and then'open the trip and throttle valve using HS-15111.

d. Manually open the trip and throttle valve using HS-15111, -

reset the mechanical trip linkage locally, then close the trip and throttle valve using HS-15111.

A: a.- }

R:-13610-1,R9,PG. 6, 1X3D-BC-F02, 17016-1,R8,PG. 53 EB#: LO-OR-20101-10-01 Point Value: 1.00

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Pl.- Q: Containment pressure is 1.1 psig. Chemistry has issued a release permit to allow containment pressure to be reduced. Which of the=

following lineups best describes the flowpath to be used initially y during;the pressuro reduction: evolution. i a.: HV2628B & HV2629B open; HNN '97 AUTO; Mini Purge Exh Fan i running

b. HV2628B & HV2629B open; HV12592 closed; Mini Purge Exh Fan running-

c. HV2628B & HV2629B open; HV12592 AUTO; Mini Purge Exh Fan stopped

d. HV2628B & HV2629B open; HV12592 closed; Mini Purge Exh Fan stopped A: d.

R: 13125-1,R12,PG. 10, EB#: LO-OR-29110-03-01 Point Value: 1.00

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22. Q:Which of the:following events / conditions would cause the value h of-the estimated critical boron concentration:to decrease?

a .a. Estimated startup time increases from 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months after a trip }

from 75% power, to 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .after the. trip. j Desired critical rod height changes-.from 130 steps / Bank

~

b. "D" to 140 steps / Bank "D".

c. The anticipated Tavg at startup is1 changed from-557 degrees F to 554--degrees F. (Assume,a negative MTC). l

d. Reactor power history used in the estimated critical boron calculation is corrected from 40% to 50% power.

A:- d.

R: PrDB TAB 1.4.1-T1,,'1.5.1-T3, 14940-1,R8,PG. 1 .

EB#: LO-OR-33510-07-02 Point Value: .1.00 *

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23. Q: Which one of the following would require action to be taken=within !

the next two hours to commence a' unit power descent?

a. Four-loop operation with power at 90%, Tavg at 620 degrees F, and pressurizer pressure at 2235 psig,

b. 92% RWST level with boron concentration at;2430 ppm, and .

temperature at 72 degrees F. l RCS activity at 1.2 microcurie per. gram DOSE EQUIVALENT-I-131 c.

for 50 hours5.787037e-4 days 0.0139 hours 8.267196e-5 weeks 1.9025e-5 months at full power.

d. Discovering that only'one NSCW pump was operable on NSCW 4' train A 47 hours5.439815e-4 days 0.0131 hours 7.771164e-5 weeks 1.78835e-5 months ago.

A:-a.

R: T.S. 3.2.5, EB#: LO-OR-39206-03-02 Point Value: 1.00

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.24i Q:lThe' Auxiliary Hoist on the SIGMALrefueling. machine has failed l

' during. CORE: ALTERATIONS.' Several RCCA's need to be unlatched- prior: ,

to head removal, so the Outage Area Supervisor has written a ;

Temporary' Procedure' Change to allow the RCCA-disconnect tool, .

suspended from the Polar Crane, to complete the unlatching of the +

RCCA's.=As Fuel Handling Supervisor you should:

o

.a. Authorize use'of the Polar Crane since it-is_ allowed per? -

93500-C,-Manual Operation of Fuel' Handling Equipment. . ,

b.; Authorize use of the Polar. Crane.since.it meets Tech.

Spec. requirements of 3.9.7. ,

c.' Not authorize the use of the Polar Crane'since.'it-only1

~

meets the requirements of procedure 93260-C, Fuel Transfer .

System (FTS)-Operating Instructions.

d.- Not authorize the'use of the-Polar Crane since it is not .q allowed per Tech. Spec. 3.9.6.

-A: Ed..

R:LT.S. 3/4.9.6. 93300-C,R6,PG. 3, EB#:.LO-OR-39213-03-09 Point Value: 1.00 .

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SUPV: DATE:

%cEXAMPULLEDFORCCR. FILING: ,

I a . (

I "fD PLANT VOGTLE TRAINING DEPT. N QUESTION USE LIST for EXAM: SR-905-90-05.BS TotallPoints:-24'00

. FILE NO:C90-05-003 Page 1 ,

' Assembled by Chuck Stuhaan on 07/23/90 in> MANUAL. mode,

, i ITEM EB NUMBER REVISION DESCRIPTION

1. LO-OR-13301-02-02 11 06/29/90 Low head SI flowpath

2. LO-OR-22101-06-01 4 - 06/29/90 Fire Protection System' Operability-1 c3.-LO-OR-37002-04-01 10 06/29/90- EOP Usage - Continuous action stepa 0

4. LO-OR-37002-08-03 3 06/29/90 Crew responds to Red path _ iturns 5.:LO-OR-37011-06-04 6 06/29/90 Perform IOA's per 19000-C. SI if SI' I

6. LO-OR-37031-09-04 .8 06/29/90 What to do with CSFST's while~in-?l9- , .(

7. LO-OR-37041-08-02 4 06/29/90 Reason to reduce pressure on ATWT.' [

8l . LO-OR-37051-04-01' 5 06/29/90 Reason for stopping RCP's'during a

,9. LO-OR-37071-06-02 9 06/29/90 Function of the' temperature soak us

10. LO-OR-37311-07 8 06/29/90 Performance of step 14 of SGTR proc

11. LO-OR-09201-12-03 3 06/29/90 CVCS response to Loss of Instrument

12. LO-OR-12101-19-01 8 06/29/90 Minimum charging flow after Loss of

13. LO-OR-39207-03 5 06/29/90 What to do'if two PZR-level channel-

14. LO-OR-163u2-03-01 5L 06/29/90 Automatic PRZR Level Control (KA 3.

15. LO-OR-60306-01-01 12 06/29/90 Concerns during Containment evacuat

16. LO-OR-60324-01-02 9 07/12/90 Effect of loss of vital bus w/emerg

17. LO-OR-25101-06-01 17 06/29/90 Design features that min. exposure

18. LO-OR-12101-08-03 9 06/29/90 RHR Operations while shifting to RC

19. LO-OR-16401-04-01 12 06/29/90 RCP ops w/ Seal Abnormality ,

20. LO-OR-20101-10-01 11 06/29/90 How to reset the TDAFW Pump T&TV ;

21. LO-OR-29110-03-01 11 07/12/90 Describe how to vent containment wi

22. LO-OR-33510-07-02 9 06/29/90 ECC Boron Concentration Changes

-===========================b48T=60NTENWB9=0N=NENT=PAGE========================

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- SR-905-90-05.2S 'l

'. f Ai 4 p 4 , -Page 2

. QUESTION-USE.; LIST .;

f.. . ,

ITEM - EB' NUMBER' REVISION - DESCRIPTION = '

kj E23'9LO-OR-39206-03-02,

. 12 06/29/90 Tavg T.S. interpretation using.give -

. . . .. i

' 2 4-. - LO-OR-39 213-03-09 3 06/29/90 - Use of<the-RCCA disconnect tool vit ,

==============================================================================. ,

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t M-A'.S T E R K E;Y' .:)

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' EXAM:.SR-905-90-06.A -

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Total' Points: 124.00' M, p,

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,. 7 STATIC ^SINULATOR[-PART~A_-

SCENARIO:

SUMMARY

'FOR'58-44 SCENARIO TYPE Normal' PLANT CONDITIONS: Stuck rod recovery NI failure.-

QUESTION 81 APPLICABLE'TO TNIS EIAMINATION-SCENARIO

).-

A4401 A4419 - Same as'2601 A4403 A4420 - Same as 2603-A4404 A4421 - Same as 2605 ,

A4405- A4422 - Same as 2606 ;

A4406 A4408 A4409 A4411~.

A4412 A4413 A4414' ;

A4415-A4416 1 A4417- .i A4418 _

j VALIDATED REVIEWED TRAINING i

- - - - - - . . - . . . . . .... . . - is

__------------3------------

! s s

q.ll.i

, -STATIC *IMULATOR :PART A-L SIMULATOR SETUP INSTRUCTIONS FOR 88-44

'y INITIALIBE TQ:-IC14-INSERT H&LFUNCT.19.MA '

- M9_2.5 MM ENTRY ME9' HOD

70 L/D H/X' Leak-

1. 2,70,17.5 7B- PR' Failure 2. 3,7B,100,0001 USERT OVERR1 DES - ,

L M&M3 ENTRY RETEOD !

N/A

'I SELECT THE: FOLLOWING SWITCH POSITIONS ,

NUMBER M&MR POSITION

, HS8000A Pzr PORV 455A Block Valve Closed '!

TIC-130-LTDN HX Control-' Temp Pot -

6.46 l;. '

START THE SCENARIO PERFORM ~THE FOLLOWING ACTIOMB j s- !

,o Manually actuate CNMT spray (HS-40004 & 40005 to actuate) 1

, Drive:SBA in to 180 steps, then enter Malf. 27K. Withdraw SBA

=to ARO,.then remove-malfunction 27K. Open all disconnect' switches in SBA except M-2 and withdraw rod M-2 three (3)

C steps.

Set HV182 demand to full charging flow (o seal flow)

Set ~FIC 121 to 90% demand.

Manually reduce RCS press until less than 2185 psig and ALB12 -

D03 energizes.

EBRERE THE SIMULATOR &T RCS pressure less than 2185 psig and ALB12 - D03 energized.

l

_ . _ _ _ _ . . . - . . . . . -M

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STATIC SIMULATOR:- PART A SINU 1ATOR' SETUP' INSTRUCTIONS-FOR 88 ;

VERITY TER-FOLLOWING CONDITION 88 !

POWER::99 .BQQ POSITION: 217: BORON: 787 1 R&B LY1 .55- 21B PRES 8: <'2185 psig TAVE t ,- 590 RCP STATUS: '4-Run ECCSS STATUS -N/A

.19 PRESS:l'990 19 LY1 NOL OTHER CONDITIONS --

. Rod M-2. at: 183 steps, VERIFY THE FOLLOWING SWITCH POSITIONS:

Rod select switch.to.8BA.

- All 8BA rods in Hdisconnect" except M-2.

SELECT THELFOLLOWING DISPLAYS

-RBZ DISPLAY 134CNMT Rad Data (

RBE M RI la Trend.of PRT Level & Press PROTEUS: Any ( ,g p I' C " :

,.p,- . . . - , ,-

c, , , '

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W'!! i gj$y(ja ~y^ SR-905-90-06cA0 KEY'

'Jm n "'M *

'Page 1 -

EXAM KEY Q;b,. I NRC;REQUAL-EXAM,-WEEK 2,.PART l',-.RO -i segg==..=..............=.=........==..................................=...==.==== .

+

5

\

.' 1 '. 01 ~Q:..When-annunciator ALB12-D03,'"PRZR-PRESS LO PORV: BLOCK" clears what

~

automatic: action will occur?-

+

a. PORV block valve 8000A will.open.- "

s,

b. PORV block l valve 8000B will open.. '.

c. Both.PORV block valves, 8000A EjB,'will:open.-

d. Both PORV-block valves, 8000At &8B, will stay' shut, i

i , .i A: B '

b 2) ",

R: LO-LP-16303-00, ARP 17012-1, D03, '

H EB#: LO-SS-26000-01-01 Point-Value: 1.00 .

7 c ==========..===============.=============..=.. ....===.=.v..===.=.===========

L {

1.02 'Q:.Which action below-should be taken:to clear annunciator ALB08'-F06?- 1

a. Adjust HV-182 i o

i

b. Throttle closed FV-121

, c. Increase VCT pressure

d. Isolate seal return i

4 A: A a.

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i i-.

R: LO-LP-09001-00, ARP 17008-1, F06, ~

l ED#: LO-SS-26000-01-03 Point Value: 1.00

============================================================================== l KEY CONTINUED ON NEXT PAGE 1

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.SR-905-90-06.AI? KEY

,Page 2; , j

. . . EXAM' KEY ..

s ; NRCLREQUAL EXAM,. WEEK 2,.PART 1,HRO- i- !

.==============================================================================f j

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1.03'.Q:2Whatfis thel temperature.of the steam downstream of?the;f2lS/G:- ..

je ; leaking ARV?E 5 m

iu : . .

Y a. '212Ff(+/-:5F) ,

b. - 280F - (+/- 5F) sc

c. 300F.(+/- 5F), ,

, d.734 5F-f(+/- 5F) s u

O A:1C .i

- R: :LO-LP-34110-00,- ASME. Steam Tables, '

. EBf: .LO-SS-26000-01-05 Point Value::.1.00 ;

==============================================================================.,

El . 04 ; Q: When:;the' Reactor' Operator misadjusted TIC-130, which_of. thel -i

'following occurred? 1 a.1TV-130;went'No'thefullopenposition. i

'b.'TV-129: shifted!to the VCT position.

16 .

, c.EFlashing soon, began at letdown orifices. ,

o

. d .f Let'down . isolated r

a t

si - ,

, A: B

'I I

s R: LO-LP-09001-00, VEGP PLS-1X6AA04-30, '

.EB#: LO-SS-26000-01-06 Point Value: 1.00

====================================_____=========================

KEY CONTINUED ON NEXT PAGE

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, 1 SR-905-9 0-06' AT. . KEY u

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Page<3i .

4

~

EXAMEKEY' , . .

sl NRC REQUAL. EXAM;LWEEK 2,1PART_l',=RO a:

==================================================================

a j

i s 1105 ~ Q: '. Based:uponthecontrolboardiindicationrkidentifythelocatio$iof 'I th' -the:CVCS leak'from the following list: , J i

. a. ' Letdown-line, upstream oflthe regenerative' heat-exchangeri yq b.-

Regenerative heat exchanger tubeEleak. -

c. High-pressure, letdown relief valve PSV-8117' seat': leakage;

, d. Excess letdown line, upstream of valve'REACTORLCOOLANT TO' j

EXCESS LETDOWN HV-8154' a l

t l l

A: A- I a

R: P&ID1X4DB114,z EB#: LO-SS-44000-01-01 , Point Value: 1.00

==============================================================================.

gi .

?

1.06 Q: .Annunicator ALB-10-B6 Rod Control Urgent failure alarmed-j during recovery of rod M2. Which of the'following describes whyL this. alarm-was generated?

,< a. Lift coil disconnect switches'placed in the disconnect.

1 ; position /immediately resulted.in a regulation failure in L

being generated.

j

b. Stationary gripper' disconnect switches placed in the.

' disconnect position:resulted11nLa pulser failure when rod.

m movement was demanded.

a

- . l

c. Movable coil disconnect switches placed in.the disconnect .

I position resulted in a regulation' failure when rod movement-was demanded. U l

d. Lif coil disconnect switches.placed in the disconnect.

position resulted in a regulation failure when rod movement _

l was demanded. )

l A: D l

R: AOP 18003-ROD CONT., SYS DESC. ROD CONTRL, SYSTEM, REV. 1 EB#: LO-SS-44000-01-05 Point Value: 1.00

=======================n==========================================

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SR-905-90-06.Al KEY .

> R ' ,. l / .

Page14: y o

EXAO KEY ~ >

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NRC REQUAL EXAM,\$EEK 2,'PART,1,'RO=

{ .g =========================================o====================================;

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jjf W1.07 ~ Q:)A' method?to'determineithe leak rate is to perform a flowibalance

~

" ' d across the CVCS. Which of-the following most closely-describes:

LtheLleak rate as de* ermined-by this~ method.

l , , .J.

.a. 70'gpm (+/

2'gpm) ,

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b. 82;gpm:(+/ 2 gpm).

w a

c. 58 gpm (+/- 2'gpm) l W s d. 65 gpm (+/- 2 gpm)

f. : il)

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A: C1 -

R:L1X4DB114,115,

-Point Value:-1.00.

~

' EB# : -- LO-SS-4 4 000-01-14 -l

============================== ,=============================================== ,

. k'

l'.08;-Q: . Presume that letdown was isolated in attempts to. identify;the- i leaknWhich of the following statements isJoorrect.

=l

. :a.. Letdown pressure control valve'PV-131-will shut ?.o controli

. . pressure. -

b'. Letdown pressure control valve PV-131 will>open'to control pressure. .

c. TV-130-will open to control. temperature of the ACCW l through the letdown heat exchanger' l

d. TV-130 will open to. control' temperature of.the fluid in the letdown line.

A: A R:.1X4DB114,115, EB#: LO-SS-44000-01-12 Point Value: 1.00

.==============================================================================

KEY CONTINUED ON NEXT PAGE n

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.. SR-905-90-06.A; KEY Page 5 EXAM KEY NRC REQUAL EXAM, WEFK 2, PART 1, RO

==================================================================

1.09 Q: Which of the following statements regarding the misaligned rod is true?

a. It is permissable to leave the rod in its present location.

b. Per the PTDB this is the most reactive rod in the core.

c. If the plant were to trip, and this rod were to remain stuck, a boration of 104 ppm would have to be performed to compensate for the positive reactivity effects.

d. Insertion of the group to match stuck rod position would cause you to enter Tech Spec 3.0.3.

A: D R: T.S.3.1.3.5, EB#: LO-SS-44000-01-16 Point Value: 1.00

==================================================================

1.10 Q: With respect to the current RCS pressure and the N-42 problem:

a. Current RCS pressure has increased OT delta T setpoints,

b. Current RCS pressure has decreased OT delta T setpoints.

c. The N-42 problem has increased loop 2 OP delta T setpoint.

d. The N-42 problem has decreased loop 2 OP delta T setpoint.

A: B R: TECH SPECS, EB#: LO-SS-44000-01-17 Point Value: 1.00

=============================================================cn===

KEY CONTINUED ON NEXT PAGE

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o f 'E SR-905-90-06aA;4 KEY 7 q, y i >

jbr ,y o .l 'Page 6- 3 m" y* ,, EXAM KEY M 7%

iNRC REQUAL EXAM,-WEEKE2, PART 1,;RO.

W .===========================ma=================================================-

,4 y,

. 1 i 1'.*11 Q: As'a result of.the CNMT Spray' Actuation, whichlof the following-I y 'did NOT occur.

V .

9

a. HV-9017ALreceived an open signal.

T' >

'b. HV-9001A-received 1an-open signal.

t i

c. The pH of the RWST water increased as.it-passedithrough !

k the=C.S. pumps.= 1

,}j 9 d. HV-8994A received an open' signal A::A-yC

,, R: -CHAPTER 15, . . ,

EB#:-.LO-SS-44000-01-18 n

. Point Value: 11.00

============================================================================== l c te 1

. l

1.12 -Q: NOTE:- THIS. QUESTION POSES.A SITUATION NOT :RELATED;TO THE' CURRENT 4: SYSTEMS / CONTROL ~ BOARD INDICATIONS!!!!!!

If the BTRS were being used to borate the RCS:

{

o< a. Letdown flow:would bypassLthe.CVCS demineralizers while I flowing through the BTRSCdemineralizers. '

i

b. Letdown flow would'first go through the BTRS demineralizers'then the CVCS demineralizers. J 1

c. BTRS demineralizers' inlet temperature would be maintained j by TIC-381A which controls flow through the chiller heat exchanger.

d. BTRS return header' temperature would be maintained by. l l TIC-386 which controls flow through the BTRS chiller heat

! exchanger.

I' h

l l A: D R: P& IDS, L EB#: LO-SS-44000-01-23 Point Value: ~1.00

==================================================================

l END OF SECTION KEY l

l 1

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7, STATIC.SINULATORJ- PART A. -j

$ _ s r

T<

SCENARIO SUNNARY FOR SS-53' I s

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g . SCENARIO: TYPE , Emergency. .i

.. i

. PLANT CONDITIONSt: .RCP trip from 30% power. MI y j i

. f t

. QUESTIONS APPLICABLE TO THIS EXAMINATION SCENARIO: y o

, i a A5301 A5311 A5321 ,

A5302 .A5312 ,

iA5303 .A5313

, 4 l

A5304- A5314'

A5305 A5315 TA5306 -

A5316 A5307 'A5317

'3i- A5308 -A5318 1 A5309 A5319 /

A5310- A5320 l

l:

.': i i VALIDATED:

REVIEWED

~ TRAINING I 1'

L i

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1 1 1

1: - , _________.__.__________1

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..i g 4 8TATIC; 81MULATOR '- PART A li,

, SINULATOR SETUP INSTRUCTIONS FOR SS-53 l

l INITIALI5E Tot IC-13, 28%, Power, BOL IN8ERT KALFUNCTIONS:

NO. NAME ENTRY METHOD Y

i 14, RTB 'B' FAILS SHUT .2, 14 *

[L!

s1 62B, PT-506 FAILS hIGH 3, 62B:

I !

= INSERT OVERRIDES -[

NAME ENTRY METHOD-.

l NONE SELECT THE FOLLOWING SWITCH POSITIONSt l NUMBER NAME , POSITION l

HS-5208 A'MFP discharge valve P7L-OPEN START'THE SCENARIO PERFORM THE FOLLOWING ACTIONS 1

1. Stop #4 RCP and start all four RCP Oil Lift Pumps.

2. When FWI occurs after Rx trip, then stop other 3 RCP's. l

3. Trip the TDAFW TTV, then position TDAFWP'discharr valves t

to 90% open, l ~

4. Throttle AFW flow to each S/G to 200 GPM, after all S/G NR D levels are back on scale.

FREEEE THE SIMULATOR AT IR 0 5x10-7 Amps l

l

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,* f STATIC SINULATOR -'PART'A 8CENARIO SRTUP -~ATTATCEMENT-1. ,

i +

't i

'I Perform the following actions in the order specified, after thef l reactor trip. l q

t

.1. Isolate-SG #2 as follows: ,

t a.. Close Blowdown Isolation Valve,LHS-7603B. .,

J,; > b. Place MSIVs, HS3016A and HS3016B in Fast Close PTL. 'l

c. Close MSIV bypass valves, HS-13007A and HS-13007B.-

q 2.- Arm COPS by placing COPS controls, HS-8000G and HS-8000H, i

Lin ARM.

l- 3.- Isolate AFW flow to SG #2: "

a. Close TDAFW valve, HS-5125A L

b. <Close-MDAFW valve, HS-5132A ]

l l

L '4. Throttle AFW to-SGs 1, 3, & 4 to'200 gpm. H u

l S. RESET SI.

G

6. RESET CIA. I 1

1

7. Restore Instrument Air to CTMT.

8. Stop both RHR pumps.

9 .- Place steam Dump Mode Selector Switch in STEAM PRESSURE mode.

10. Raise Steam Dump controller, 1PIC-507, to 100%.

L 11. Open SG ARVs to 80 percent demand on SGs 1, 3, and 4.

., , u 4

i l ;- s';

~

-STATIC SIMULATOR >- PART.A b

-SIMULATdR SETUP INSTRUCTIONS FOR SS-53:

, I

- VERIPY THE,FOLLOWING CONDITION 81

~

, POWER: L 5E-7' ROD POSITIONt-TRIPPED ~ BORON 993 ;i

PSR LVL1 2.8% PSR PRE 88: ~2100 PSIG TAVE *545 NR RCP BTATUSt 4 OFF RCCB BTATUS STBY d SG' IRE 881 ~900 PSIG SG LYLt #4 GREATER TRAN 1, 2,.AND 3 OTHER: CONDITIONS l

N/A

]

VERIFY,THE FOLLOWING SWITCH POSITION 8: ;

i

-Rods in auto. 'l 88 LECT THE FOLLOWING DISPLAYSt i

ERF: DISPLAY la T.L.D. l ERP DISPLAY 22 T.L.D.

PROTEUS Any !

t

1 ; 70' . < .!9: , w- '

i

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*

3[\;

i

u. 7 SR L905-90-06.A;iKEY- 'i r

Page 6 j

-EXAM KEY _ . . .j NRC REQUAL EXAM,WEEI; 2, PART 2, RO'

===============================================3============================== ,

-)*'2.01 'Q:-Which.of the:following completelyfdescribes the signal /that wouldi

have generated theiturbine trip?

i a.~125'VDC trip signal to mechanical trip solenoid from

,r1 P-4 train A.

b. 24:VDC trip signal _to electrical trip solenoid fromi P-4-train B. ,

c. 125 VDC.end 24 VDC trip signals as;a' result _of P train A.

~

d. 125 VDC; and 24 VDC trip signals.as' a result of P-4 tr%in B. i A: C R: - LOGICS , .

EB# LO-SS-53000-01-01 Point Value: 1.00 3 1

+

{

2.02: Q: What signal initiated the reactor. trip? l

.o

a. Single loop-low flow; _s 1

.b; Turbine trip / reactor trip l c._S/Gil, 2, and 3 LO-LO levels -

d. S/G # 4 LO-LO level A: D i

R: SIMULATOR IND., .

EB#: LO-SS-53000-01 Point Value: 1.00

==============================================================================- r e e I

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' SR-905-90-06.At KEY!

, Page =.- 7 q 1

EXAM KEY 1

, NRC REQUAL EXAM, WEEK 2, PART 2, RO: I

==============================================================================:

+r -

O

{2 03 Q: S/G=4 level:is: higher =than~the other 3 S/Gs because: !

.a. Feed' flow increased to S/G'4 after,the RCP tripped but- N steam flow decreased. -Following the' shrink transient ;

e .-level-would-be expected to be higher, .

d

b. AFW flow was. inadequately throttled to S/G 4-compared tc N

'the othercS/Gs after the trip resulting;in the level

[7 deviation.. ',

a-

>c. A!SGTR ' occurred before the reactor tripped ~and is~ 0 ay t( . continuing to raise level in S/G 4.

d. The Main Feed Reg Valve for S/G 4.is~still open causing its j 1 <

1evel to'to riss at a faster rate.

' j! :A: A-r .R: SIM INDIACTIONS, / N EB#: LO-SS-53000-01-06 Poi.nt Value: 1.00

==================================2.=================.2============

3 2.04 Q: When reactor trip breaker.B failed to open following the reactor trip, what effect did thisthave on steam dump operation?-

< \ , a. Steam dumps armed, but did not open. >

-i

b. Steam dumps did not arm.

c. Steam dumpc armed and opened on the plant trip 4

controller.

[ '

d. Steam dumps armed and opened on the load reject ,

controller.

A: .D R: LOGICS,

- EB#: LO-SS-53000-01 Point Value: 1.00

.========================n=====================================================

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Page 8

EXAM-KEY NRC-REQUALl EXAM, WEEK 12, PART 2 .RO! 1 f========u==============================================,=======================- q 0

' 2.051 Q: Select. the. correct response concerning- the ' PER PORVs during , i this transient.

1

, a. No_PORVs opened, but-at-least one should have opened. 4

b. Atileastlone PORV opened immediately after the reactor.

2 I

trip,

, c. At lebst'one PORV opened immediately after the, turbine-3 l trip. !

g d.-No PORVs opened,'and none should have opened.

t A: D-

~

-R:'SIM INDICATIONS,

'EB#:cLO-SS-53000-01-17 Point Value: 1.00

==============================================================================.

'2'e06 Q: If the Train B'SR BLOCK / RESET'handswitch.(HS-40031) were placed!ini the RESETLposition:

a.-.

N-311would. energize.-

b. N-32 would energize ~. >

l

c. N-31:and N-32 would energize. '

d. . N-31 and N-32 would energize when both P-6 bistables i cleared.

A: C s

R: LOGICS, EB#: LO-SS-53000-01-25 Point Value: 1.00

==================================================================

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, l SR-905-90-06.A; KEY'

'*P

-Page 9l- >

EXAM-KEY 2 la NRC REQUAL EXAM, WEEK 2, PART 2',-RO

)

lf f.

1 u ; 2 L O7- Q: Given the current conditions, which of.the following is required

, 'to reset the feedLwater isolation signal?- j q a. Take HS-40049(FWI RESET) to ' RESET'.

~

o$ .b. .

Take HS-40050(FWI RESET) to.' RESET'.

i

c. Cycle 1 reactor trip breakers,-then take HS-40049 to ' RESET'. H

!iE d. Cycle reactor trip breakers,-then take HS-40050Ltc ' RESET'.- I h'

s A': A it R:' LOGICS, EB# -LO-SS-53000-01-20 Point Value:'1.00.

========================================================================= ====-

i 2.08 Q: If the running MFP were tripped right now, which of the following would.be true?

a. Both MFP discharge valves would shut, b'. All TDAFW pump discharge throttle valves would open fully.

c. All TDAFW pump discharge throttle valves would remain in the.ir current position,

d. All MDAFW pump discharge throttle valves would fully ,

open.

t A: C R: CAPTER 20, EB#: LO-SS-53000-01-21 Point Value: 1.00

==================================================================

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a

..< SR-905-90-06.Al KEY-PageL10 EXAM-KEY-NRC REQUAL EXAM, WEEK 2, PART 2, RO

==================================================================

b

[2'.09 'Q:' NOTE: THISfQUESTION POSES A SITUATION THAT IS NOT RELATED TO THE CURRENT PLANT / CONTROL BOARD CONDITIONS!.! ! ! ! ! !

-If.:RE-0002, Containment Area radiation monitor, were to fail high:;

a. CVI would actuate, and the containment atmosphere: hydrogen monitor would isolate.

1

b. CVI would NOT actuate, but the containment evacuation 1 alarm would~ sound.

c.- CVI would actuate, and.any open containment' purge. valves ,. c would close,

d. -CVI would NOT actuate,.but-ALPn5-A3, RMS channel. Failure, would alarm. '

.: 1

'A:-C

=!

R: P& IDS, EB#:. LO-SS-53000-01-22 Point Value: .1.00

=======================2==========================================

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I I,

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SR-905-90-06.A7' KEY Page 11 EXAM KEY- -

NRC REQUAL EXAM 2', PART 2, RO a ======================================,W=EEK =====================================

.i .

'2.10 -Q:. NOTE - THIS QUESTION POSES A. SITUATION NOT RELATED,TO_THE'PRESENT PLANT / CONTROL BOARD. CONDITIONS!!!!!!: !

A plant transient has occured thatiresulted'in a Reactor Trio and Safety Injection.- The Train B ReactorLTrip= Breaker did NOT. cen,-

and RCS pressure stabilized at 1500 psig.- Five minutes;later.both b SI RESET handswitches are placed in the RESET position'and results in:

a. No change to-the SI ACTUATED or AUTO SI BLOCK 3D lights because SI-cannot be resetcwithout BOTH reactor trip. 6 breakers opening.

b. SI ACTUATED (RWST) light' going out and the AUTO SI BLOCKED light turning on. .j

c. SI ACTUATED and SI ACTUATED (RWST)' lights going out.

d. The AUTO SI' BLOCKED light turning on and the RWST SI TEST lights staying on.

A: D R:cLOGICS, EB#: LO-SS-53000-01-23 Point Value: 1.00

.==============================-======================================n========

2.11- Q: NOTE - THIS QUESTION POSES A SITUATION NOT RELATED TO THE PRESENT PLANT / CONTROL' BOARD CONDITIONS!!!!!!!!!

DG1B is being used to' supply electrical power to bus 1BA03. A ;

loss of bus 1BB07 would result in:

a. Losslof one DFO Transfer Pump for each DG.

b. Loss of both DG1B air start compressors.

L i

c. Loss of Jacket Water Cooling for DG1B. l

d. Loss of both DFO Transfer Pumps for DG1B.

l A: D R: ONELINE DWGS, ED#: LO-SS-53000-01-24 Point Value: 1. .a

==================================================================

KEY CONTINUED ON NEXT PAGE

~4.;

'SR-905-90-06.A;, KEY H Page 12 EXAM KEY NRC REQUAL EXAM, WEEK 2, PART 2, RO i

==========.............==.m......= .................................m..mm .m..

=2.12. Q:/ NOTE: - THIS QUESTION ~ POSES A SITUATION' TRAT IS : NOT RELATED TO -

THE CURRENT ~ PLANT / CONTROL BOARD CONDITIONS ! ! ! ! ! !-

4 Which of the following explains'pe.ver operation limitations after;

4 RCP was secured, if the plant nad been' stabilized and no trips-occurred? . Assume that it would NO'r be necessary to: secure -

RCPs 1, 2 ', and 3.

a. Continued plant operations may be maintained without-any.

limitations.

b. Plant operations may be continued, but power must be reduced to less than 15% until the cause of the trip has been determined. '

c. Power may be maintained at the stabilized power level-but' f l for no longer than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ,

d. The plant must be in Mode 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months unless the pump is_made operable and restarted.

'l A: D !

R: TECH Spics, .

EB#: ID-SS-53000-01 _ ,

Point'Value: -1.00 l

===============================================================mmmmmm mmmmmm==.

END OF SECTION KEY ;

END OF TEST KEY

A l l' l ..-

K, "I The following is an. alpha-numeric list of figures and references which should accompsny this test when it is administered.

Reference' Question Static Sim 26 Intro -* . 01 Static Sim 26 Intro 1.02 Static Sim 26 Intro 1.03 Static Sim 26 Intro 1.04 o

i e

l FND OF FIGURE AND REFERENCE LIST

l i

, EXAM PULLED FOR CCR FILING SUPV: DATE ,

PLANT VOGTLE TRAINING DEPT. !

QUESTION USE LIST for EXAM: SR-905-90-06.A Total Points: 24.00 FIII NO:C90-06-019 Page 1 Assembled by Chuck Stuhaan on 07/30/90 in MANUAL mode.

ITEM EB NUMBER REVISION DESCRIPTION 1.?l. LO-SS-26000-01-01 7 07/12/90 PORV block valve interlocks' 1.02 LO-SS-26000-01-03 5 06/20/90 Problem Recognition * ;

1.03 LO-SS-26000-01-05 7 07/12/90 Isenthalpic Throttling Process 1.04 LO-SS-26000-01-06 4 06/20/90 CVCS interlocks and response

1.05 LO-SS-44000-01-01 10 0^/20/90 letdown line leak identification 1.06 LO-SS-44000-01-05 9 07/12/90 Urgent failure causes 1.07 LO-SS-44000-01-14 5 07/12/90 leak rate determination .

1.08 LO-SS-44000-01-12 6 06/28/90 isolate 1/d [

1.09 LO-SS-44000-01-16 8 07/30/90 ril of the s/d banks 1.10 LO-SS-44000-01-17 3 .06/28/90 OTdelta T response to res pressure 1.11 LO-SS-44000-01-18 2 06/28/90 CNTMT GPRAY ACTUATIONS 1.12 LO-SS-44000-v1-23 00 / / BTRS flowpath during dilution

===========================================================....................

END OF SECTION i

t L

1:

I: "' " '

. ; ~. . ,

? r vi

EXAM PULLED FOR CCR FILING SUPV: DATE _

i PLANT VOGTLE TRAINING DEPT.

f QUESTION USE LIST for EXAM: SR-905-90-06.A Total Points: 24.00 FILE NO:C90-06-019 Page 1 Assembled by Chuck Stuhaan on 07/30/90 in MANUAL node. l ITEM EB NUMBER REVISION DESCRIPTION L1 2.01 .LO-SS-53000-01-01 6 06/28/90 causes of turbine trip i 2.02 LO-SS-53000-01-03 3 06/20/90 reactor trip diagnoses !

2.03' LO-SS-53000-01-06 5 06/28/90 integrated diagnostics 2.04 LO-SS-53000-01-16 2 06/20/90 steam dumps ops 2.05' LO-SS-53000-01-17 3 06/20/90 porv operation indications i

2,06 LO-SS-53000-01-25 1 07/12/90 SRNI Reset operation between P-6 an 2.07' LO-SS-53000-01-20 2. 06/28/90 FWI reset requirements 2.08 LO-SS-53000-01-21 3 06/28/90 System response to MFP trip.

2.09 LO-SS-53000-01-22 3 06/29/90 Hi Rad effects on CVI 2.10 LO-SS-53000-01-23 3 06/29/90 SI reset indications with reactor t i

2.11, LO-SS-53000-01-24 3 06/29/90 LOSS OF 1BB07 EFFECT ON EDG OPERATI 2.12 LO-SS-53000-01-19 8 06/29/90 rcp teen specs

============================================================ a====

END OF SECTION P

I~

END OF QUESTION LIST 4

't c' . - . , . - . .

1

t l- h 1

1, Weer 2 Port n (.CR OT l

- J*e ! C90-06-018 l 1

1 i

l l- 1 i

PLANT VOGTLE TRAINING DEPT. l

\= !

l- ,

l-3 MASTER KEY !

FOR

l. EXAM: SR-905-90-06.A3 l Total Points: 24.00 l 1

l 1

l l l l l

j ASSEMBLED IN MANUAL MODE. q I

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-,,~.-e-- - -

..k . -l

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STATIC SINULATOR - PART A !

SCENARIO

SUMMARY

POR 88-44 SCENARIO TYPE: Normal -

PLANT CONDITIONS: Stuck rod recovery N1 failure.

QUESTIONS APPLICASLE TO THIS EXAMINATION SCENARI0t A4401 A4419 - Same as 2601 A4403 A4420 - Same as 2603 A4404 A4421 - Same as 2605 A4405 A4422 - Same as 2606 A4406 '

A4408 A4409 A4411 A4412 A4413 A4414 A4415 A4416 A4417 A4418 VALIDATED:

REVIEWED: ;

TRAININGt _

l-

(

o STATIC SIMULATOR - PART A SIMULATOR SETUP INSTRUCTIONS FOR 88-44 INITIALISE TQ: IC14 INSERT MALFUNCTIONS:

MQA NAME ENTRY METHOD 70 L/D H/X Leak 1. 2,70,17.5 7B PR Failure 2. 3 , 7 B ,100, 0001 INSERT OVERRID.B.f t NAME ENTRY METHOD N/A 8 ELECT TER FOLLOWING RELTCE 29JI ITIONS:

NUMBER NAME POSITION HS8000A Pzr PORV 455A Block Valve Closed TIC-130 LTDN HX Control Temp Pot 6.46 START THE SCENARIO PERFORM THE FOLLOWING ACTIONS Manually actuate CNMT spray (HS-40004 & 40005 to actuate) .

1 Drive SBA in to 180 steps, then enter Malf. 27K. Withdraw SBA to ARO, then remove malfunction 27K. Open all disconnect switches in SBA except M-2 and withdraw rod M-2 three (3) steps.

Set HV182 demand to full charging flow (0 seal flow)

Set FIC 121 to 90% demand.

Manually reduce RCS press until less than 2185 psig and ALB12 -

D03 energizes.

FREEEE TER SIMULRTOR &T RCS pressure less than 2185 psig and ALB12 - D03 energized.

h\

e STATIC SINULATOR - PART A SINULATOR SETUP INSTRUCTIONS FOR 88-44 VERIFY ZEE FOLLOWING CONDITIONS.:

POWER: 99 RQD 293]IIgM 217 BORON: 787 HR kYL 55 15313H3 < 2185 psig TRYE: 590 RCP STATUS: 4 Run ECCSS STATUS: N/A E9 PRESS: 990 39 LY1 NOL OTHER CONDITIONS:

Rod M-2 at 183 steps.

VERIFY ZEE E9LLOWING SWITCH POSITIONS:

Rod select switch to SBA.

- All SBA rods in " disconnect" except N-2.

SELECT THE FOLLOWIMg,DISPLAYR ERE DISPLAY la CNMT Rad Data ERE DISPLAY 1: Trend of PRT Level & Press PROTEUS: Any l

L

l.*

p

.. . .s f SR-905-90-06.AS: KEY

[(

Page 1

EXAM KEY

! NRC REQUAL EXAM, WEEK 2, PART 1, SRO

===============================================================mmummmmewmanwas L 1 01 Q: When annunciator ALB12-D03, "PRZR PRESS LO PORV BLOCK" clears what automatic action will occur?

a. PORV block valve 8000A will open.

b. PORV block valve 8000B will open..

c. Both PORV block valves, 8000A & B, will open.

d. Both PORV block valves, 8000A & B, will stay shut.

A: B R: LO-LP-16303-00, ARP 17012-1, D03,

EB#1 LO-SS-26000-01-01 Point Value: 1.00

=========================================================n========

1.02 Q: A QPTR was calculated and determined to be 1.02. Assuming ,

that no other actions have been taken and all failures began l simultaneously, select the technical specification whose time 3imit would be exceeded first,

a. 3.1.3.1, Movable Control Assemblies Group Height I l

b. 3.2.4, Quadrant Power Tilt Ratio j

c. 3.3.1, Reactor Trip System Instrumenthtion

d. 3.4.6.2, Operational Leakage A: a._ 3.1.3.1, Movable Control Assemblies Group Height R TECH SPEC. 3.1.3.1, 3.2.1, 3.2.4,3.3.1, EB#: LO-SS-44000-01-08 Point Value: 1.00

==================================================================

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[.--

i

,_. SR-905-00-06.ASt-KEY Page 2 t EXAM KEY ;

NRC REQUAL EXAM, WEEK 2, PART 1, SRO ;

1.03 Q: What is the temperature of the steam downstream of the #3 C/G j leaking ARV? t

a. 212F-(+/- Sr') ,

b. 280F (4/- SF) {

c. 300F (+/- SF)

d. 345F (+/- SF) !

i i

At C

l R: LO-LP-34110-00, ASME Steam Tables, .

EB#: LO-SS-26000-01-05 Point Value: 1.00 l 1.04 Q: When the Reactor Operator misadjusted TIC-130, which of the following' occurred?

a. TV-130 went to the full open position. ;

b. TV-129 shifted to the VCT position.

c. Flaching soon began at letdown orifices.

l

d. Letdown isolated i

i A: B ,

i i

R: LO-LP-09001-00, VEGP PLS-1X6AA04-30, i EBf: LO-SS-26000-01-06 Point Value: 1.00

==..........................n.........................=====.....==

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^ !

7; SR-905-90-06.ASt KEY l Page 3 i EXAM KEY NRC REQUAL EXAM, WEEK 2, PART 1, SRO maammmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmummmmmmmmmmmmmmmmmmmama l l

l 1.05 Q: Based upon the control board indications, identify the location of i the CVCS leak from the following list' i

\

a. Letdown line, upstream of the. regenerative heat exchanger I I

b. Regenerative heat exchanger tube leak

c. High-pressure letdown relief valve PSV-8117 seat leakage

d. Excess letdown line, upstream of valve REACTOR COOLANT TO l EXCESS LETDOWN HV-8154 l

l l' A: A l R P&ID1X4DB114, EB#: LO-SS-44000-01-01 Point Value: 1.00 maammmmmmmmmmmmmmmmmmmmmmmaremama .mmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmme=

1.06 Q: Annunicator ALB-10-B6 Rod Control Urgent failure alarmed during recovery of rod M2. Which of the following describes why this alarm was generated?

a. Lift coil disconnect switches placed in the disconnect position immediately resulted in a regulation failure being generated.

, b. Stationary gripper disconnect switches placed in the '

' disconnect position resulted in a pulser failure when rod movement was demanded.

c. Movab?e coil disconnect switches placed in the disconnect position resulted in a regulation failure when rod movement was demanded,

d. Lift coil disconnect switches placed in the disconnect-position resulted in a regulation failure when rod movement was demanded.

A: D R: AOP 18003-ROD CONT., SYS DESC. ROD CONTRL, SYSTEM, REV. 1 EBN: LO-SS-44000-01-05 Point Value: 1.00

==ma=ammmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmma KEY CONTINUED ON NEXT PAGE

r; i

, SR-905-90-06.AST KEY Page 4 .

EXAM KEY ;

NRC REQUAL EXAM, WEEK 2, PART 1,~SRO '

============================================================================== 3 1.07 Q: A method to determine the leak rate is to perform a flow balance .

L across the CVCS. Which of the following most closely describes '

the leak rate as determined by this method.

a l

a.'70 gpm (+/- 2 gpm) l i

l b. 82.gpm (+/- 2 gpm) !

l c. 58 gpm (+/- 2 gpm)

]

d. 65 gpm (+/- 2 gpm) l L >

A: C'

-l R: 1X4DB114,115, l EB#: LO-SS-44000-01-14 Point Value: 1.00 l

1 :

1 1.08 Q: Presume that letdown was isolated in attempts to identify the leak. Which,of the following statements is correct.

a. Letdown pressure control valve PV-131 will shut to control I pressure.

l

b. Letdown pressure control valve PV-131 will open to control {

. pressure. '

-)

c. TV-130-Will open to control temperature of the ACCW

-l through the letdown heat exchanger. '

~1

d. TV-130 will open to control temperature of the fluid in j the letdown line, l 3

A: A l R: 1X4DB114,115, EBf: LO-SS-44000-01-12 Point Value: 1.00 I

==================================================================

1 l- )

1 l

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m

', SR-905-90-06.ASt KEY !

Page 5 ,

EXAM KEY ,

NRC REQUAL EXAM,-WEEK 2, PART 1, SRO :

============= ....-===============================================

1.09 Q: Which of the following statements regarding the misaligned rod ,

is true? !

a. It is permissable to leave the rod in its present location.

b. Per the PTDB this is the most reactive rod in the core.

,. c. If the plant were to trip, and this. rod were to remain stuck, a boration of 104 ppm would have to be performed to

compensate for the positive reactivity effects.

d. Insertion of the group to match stuck rod position would cause you to enter Tech Spec 3.0.3.

A: D R: T.S.3.1.3.5, EB#: LO-SS-44000-01-16 Point Value: 1.00

==================================================================

1.10 Q: With respect to the current RCS pressure and the N-42 problem:

a. Current RCS pressure has increased OT delta T setpoints.

-b. Current RCS pressure has decreased OT deltu T setpoints. ,

c. The N-42 problem has increased loop 2 OP delta T setpoint.

d. The N-42 problem has decreased loop 2 OP delta T setpoint.

A: B P: TECH SPECS, EB#: LO-SS-44000-01-17 Point Value: 1.00

============================================================r..====

KEY CONTINUED ON NEXT PAGE I

l

' SR-905-90-06.AST KEY Page 6 ;

EXAM KEY ;

NRC REQUAL EXAM, WEEK 2, PART 1, SRO ;

1.11 Q:-As a result of the CNMT Spray Actue' ion, which of the following did NOT occur. ,

a. KV-9017A received an open signal. ,

b. HV-9001A received an open signal.

i

c. The pH of the RWST water increased as it passed through the C.S. pumps.

d. HV-8994A received an open signal f A: A R: CHAPTER 15, EB#: LO-SS-44000-01-18 Point Value: 1.00

==========================_=======================================

1.12 Q: NOTE: THIS QUESTION POSES A SITUATION NOT RELATED TO THE CURRENT SYSTEMS / CONTROL BOARD INDICATIONS!!!!!!

If the BTRS were being used to borate the RCS:

a. Letdown' flow wo'. tid bypass the CVCS domineralizers while l flowing through the BTRS demineralizers,

b. Letdown flow would first go through the BTRS i demineralizers then the CVCS uomineralizers, :

c. BTRS demineralizers inlet temperature would be maintained 4 by TIC-381A which controls flow through the chiller heat exchanger,

d. BTRS return header temperature would be maintained by TIC-386 which controls flow through the BTRS chiller heat exchanger.

t A: D-R: P& IDS, EB#: LO-SS-44000-01-23 Point Value: 1.00

=========================================n========================

END OF SECTION KEY

[

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

(

STATIC SINULATOR - PART A I

SCENARIO

SUMMARY

FOR'SS-53 4

-i SCENARIO TYPE Emergency PLANT CONDITIONS: RCP trip frem 30% power.

i f

_ QUESTIONS APPLICABLE TO TNIS EEANINATION SCENARI0t l

A5301 7.5311 A5321 ,

A5302 A5312 i

-A5303 A5313 A5304 iS314 '

A5305 A5315 A5306 A5316 A5307 A5317 A5308 A5318 A5309 A5319 A5310 A5320 VALIDATED REVIEWED:

TRAININGt P

v* e __. m

_ - . - . , , , , - . , ,,,,----,y-- - y-

STATIC SINULATOR - PART A SINULATOR SETUP INSTRUCTIONS FOR SS-53 INITIALISE Tot IC-13, 28%, Power, BOL INSBRT MALFUNCTIONS NO. NAME' ENTRY METHOD 14, RTB 'B' FAILS SHUT 2, 14 62B, PT-506 FAILS HIGH 3, 62B INSERT OVERRIDEst NAME ENTRY METHOD NONE i

SELECT THE POLLONING ,8 WITCH POSITIONS:

NUMBER NAME. POSITION i

HS-5208 A MFP discharge valve PTL-OPEN START THE SCENARIO PERFORN THE FOLLOWING ACTIONS:

1. Stop #4 RCP and ntart all four RCP Oil Lift Pumps.

2. When FWI occurs after Rx trip, then stop other 3 RCP's. l

3. Trip the TDAFW TTV, then position TDAFWP discharge valves to 90% open.

4. Throttle AFW flow to each S/G to 200 GPM, after all S/G NR levels are back on scale.

FREEEE THE SINULATOR AT: IR 9 5x10-7 Amps j

l 49 .

1

\ '

i STATIC SINULATOR - PART A

)

SCENARIO SETUP - ATTATCKMENT 1 ;

Perform the following actions in the order specified, after the ;

reactor trip.

1. Isolate SG #2 as follows:

a. Close Blowdown Isolation Valve, HS-7603B.

b. Place MSIVs, HS3016A and HS3016B in Fast close PTL.

c. Close MSIV bypass valves, HS-13007A and HS-13007B.

2. Arm COPS by placing COPS controls, HS-8000G and HS-8000H, in ARM.

3. Isolate AFW flow to SG #2: .

a. Close TDAFW valve, HS-5125A

b. Close MDAFW valve, HS-5132A

4. Throttle AFW to SGs 1, 3, & 4 to 200 gpm.

5. RESET SI.

6. RESET CIA.

7. Restore Instrument Air to CTMT.

8. Stop both RHR pumps.

9. Place steam Dump Mode Selector Switch in STEAM PRESSURE mode.

10. Raise Steam Dump Controller, 1PIC-507, to 100%.

11. Open SG ARVs to 80 percent demand on SGs 1, 3, t -. id 4 .

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

STATIC SINULATOR - PART A SINULATOR SETUP INSTRUCTIONS FOR SS-53 VERIFY THE POLLOWING _ CONDITIONS:

POWER 5E-7 RCD POSITION: TRIPPED BORON: 993 PER LYL 18% PSR PRESS: *2100 PSIG TAVE:*545 NR RCP STATUS 1 4 0FF ECCS STATDS STRY SG PRESS ~900 PSIG SG LYLt #4 GREATER TRAN 1, 2, AND 3 OTHER CONDITIONS:

N/A vERIPY TMs 70LLowzNG SuzTCn POSITIONS Rods in auto. '

(

SELECT THE FOLLOWING DISPLhYS:

ERF DISPLAY 1: T.L.D.

BRF DISPLAY 2: T.L.D.

PROTEUS,8 Any t

I i

i

'SR-905-90-06.AS; KEY Page 6 EXAM KEY NRC REQUAL EXAM, WEEK 2, PART 2, SRO ;

.......... .c ................................................................

1 2401- Q: Which of the following completely describes the signal that would !

have generated the turbine trip?

l i

a. 125 VDC trip signal to mechanical trip solenoid from 'l P-4 train A. l

b. 24 VDC trip signal to' electrical trip solenoid from.

P-4 train B.

i

c. 125 VDC and 24 VDC trip signals as a result of P-4 i train A. J l

d. 125 VDC and 24 VDC trip signals as a result of P-4 train B.

At C 3

R: LOGICS, EB#: LO-SS-53000-01-01 Point Value: 1.00 2.02 Q: What signal initiated the reactor trip? ,

a. Single loop low flow b.-Turbine trip / reactor trip j c.-S/G 1, 2, and 3 LO-LO levels

d. S/G # 4 LO-LO level A: D R: SIMULATOR 1ND., l EB#1 LO-SS-53000-01-03 Point Value 1.00

==............................................................................

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I b*.

~ SR-905-90-06.AS7 KEY I Page 7 i EXAM KEY !

NRC REQUAL EXAM, WEEK 2, PART 2, SRO I i

2.03 Q: Which of the following explains the SGWLC system response l following the trip of $4 RCP. l

a. MFP speed increased, causing the MFRVs for S/Gs 1, 2, and 3 to close while causing the MFRV.to S/G 4 to open. j

b. MFP speed remained constant, the MFRV for S/G 4 would j c' have opened and the MTRVs for S/Gs 1, 2, and 3 would not '

move. I 1

c. MFRV for S/G 4 would open, this would cause the discharge l pressure to drop at the MFP, this would cause the MFP to j speed up. j i

d. MFRV for S/G 4 would open, this would cause the discharge l pressure to drop at the MFP, this would cause the MFP to j slow down. '

I A C R: LOGICS, i EB#: LO-SS-53000-01-02 Point Value 1.00 1

.................................ru........................................... .

2.04 Q: When reactor trip breaker B failed to open following the reactor trip, what effect did this have on steam dump operation?

a. Steam dumps armed, but did not open.

b. Steam dumps did not arm. +

c. Steam dumps armed and opened on the plant trip controller,

d. Steam dumps armed and opened on the load reject controller.

A: D

- 1 L R: LOGICS,

.EB#: LO-SS-53000-01-16 Point Value: 1.00

....=......................==................................... ...====

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, SR-905-90-06.AS; KEY !

Page 8 t EXAM KEY {

NRC REQUAL EXAM, WEEK 2, PART 2, SRO

==============================================================================- l 2.05 Q: Select the correct response concerning'the PZR PORVs during !

this transient. !

a. No PORVs opened, but at least one should have opened.

b. At least one PORV opened immediately after the reactor trip. ;

c. At least one PORV opened immediately after the turbine trip. ;

d. No PORVs opened, and none should have opened.

t A: D R: SIM INDICATIONS, ,

EBf: LO-SS-53000-01-17 Point Value: 1.00

================================.......m...............r..............=.....== ,

i 2.06 Q: If the Train B SR BLOCK / RESET handswitch (HS-40031) were placed in the RESET positions

a. N-31wouldknergize.

b. N-32 would energize.

c. N-31 and N-32 would energize.

d. N-31 and N-32 would energize when both P-6 bistables cleared.

A: C R: LOGICS,

- EB#: LO-SS-53000-01-25 Point Value: 1.00

==============================================.-======= .-========

l l i

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t' a-

. SR-905-90-06.ASt KEY EXAM KEY NRC REQUAL EXAM, WEEK 2, PART 2, SRO 2.07 Q: Given the current conditions, which of the following is required to reset the feed water isolation signal?

a. Take HS-40049(FWI RESET) to ' RESET'.

b. Take HS-40050(FWI RESET) to ' RESET' .

c. Cycle reactor trip breakers, then take HS-40049 to ' RESET'.

d. Cycle reactor trip breakero, then take HS-40050 to ' RESET'.

A: A R: LOGICS, EB#: LO-SS-53000-01-20 Point Value 1.00

...........................................u..................................

2.08 Q: If the running MFP were tripped right now, which of the followP.sg would be true?

a. Both MFP discharge valves would shut.

b. All TDAFW pump di' true throttle valves would open fully,

c. All TDAFW pump discha, throttle valves would remain in !

their current position. -

d. All MDAFW pump discharge throttle valves would fully open.

1 A: C R: CAPTER 20, 1 EB#: LO-SS-53000-01-21 Point Value: 1.00

...== ....................................................................==,=

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Page 10 l EXAM KEY .!

NRC REQUAL EXAM,ECEK 2, PART 2, SRO 1 mummmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmammmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmam ;

1 2.09' Q: NOTE: THIS QUESTION POSES A SITUATION THAT IS NOT RELATED TO THE j CURRENT PLANT / CONTROL BOARD CONDITIONS!!! !!! l If RE-0002, Containment Area radiation monitor, were to fail hight

a. CVI would actuate, and the containment atmosphere hydrogen monitor would isolate.

b. CV.I would NOT actuate, but the containment evacuation al trm would sound.

. C.'I would actuate, and any open containment purge valves would close.

d. CVI would NOT actuate, but ALB05-A3, RMS channel Failure, '

would ale mi. ,

A'C R: P& IDS, EB#: LO-SS-53000-01-22 Point Value: 1.00 nummmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmme=============

I l.

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l

', SR-905-90-06.AS; KEY Page 11 ,

EXAM KEY '

NRC REQUAL EXAM, WEEK 2, PART 2, SRO

==================================================================

2.10 Q: NOTE - THIS QUESTION POSES A SITUATION NOT RELATED TO THE PRESENT PLANT / CONTROL BOARD CONDITIONS!!!!!!

A plant transient has occured that resulted in a Reactor Trip and Safety Injection. The Train B Reactor' Trip Breaker did NOT open, and RCS pressure stabilized at 1500 psig. Five minutes later both SI RESET handswitches are placed in the RESET position and results !

in:

a. No change to the SI ACTUATED or AUTO SI BLOCKED lights because SI cannot be reset without BOTH reactor trip breakers opening.

b. SI ACTUATED (RWST) light going out and the AUTO SI BLOCKED light turning on. j

c. SI ACTUATED and SI ACTUATED (RWST) lights going out.

d. The AUTO SI BLOCKED light turning on and the RWST SI TEST '

lights staying on.

1 A: D R: LOGICS, . I EB#: LO-SS-53000-01-23 Point Value: 1.00

==================================================================

2.11 Q: NOTE - THIS QUESTION POSES A SITUATION NOT RELATED TO THE PRESENT PLANT / CONTROL BOARD CONDITIONS!!!!!!!!! '

DG1B is being used to supply electrical power to bus 1BA03. A loss of bus 1BB07 would result in:

a. Loss of one DFO Transfer Pump for each DG.

b. Loss of both DG1B air start compressors.

c. Loss of Jacket Water Cooling for DG1B.

d. Loss of both DFO Transfer Pumps for DG1B.

A: D R: ONELINE DWGS, EB#: LO-SS-53000-01-24 Point Value: 1.00

==================================================================

KEY CONTINUED ON NEXT PAGE

it -

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=

) '

SR-905-90-06.AS; KEY Page 12 .

EXAM KEY '

l NRC REQUAL EXAM, WEEK 2,-PART 2, SRO ;

'2012 Q: NOTE:- 'THIS QUESTION POSES A SITUATION THAT IS NOT RELATED TO THE CURRENT PLANT / CONTROL BOARD CONDITIONS!!!!!!

Which of the following explains power operation limitations after

4 RCP was secured, if the plant had been stabilized and no j trips occurred? Assume that it would NOT be necessary to secure !

RCPs 1, 2, and 3. I i

a. Continued plant operations may be maintained without any limitations.

1

b. Plant operations may be continued, but power must be reduced to less than-15% until the cause of the trip has i been determined.

c. Power may be maintained at the stabilized power level but for no longer than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

d. The plant must be in Mode 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months unless the pump is made operable and restarted.

A: D R:-TECH SPECS, .

l EB# -LO-SS-53000-01-19 Point Value: 1.00

...===......===....=============........==............................======.. -

END OF SECTION KEY i

END OF TEST KEY

~

n L.

\;

The following is an alpha-numeric list of figures and references which should accompany,this test when it is administered.

Reference Question L-Static Sim 26 Intro 1.01 Static Sim 26 Intro 1.03-Static Sin 26 Intro 1.04-

+

1 END OF FIGURE AND REFERENCE LIST

7, i

EXkMPULLEDFORCCRFILING SUPV: DATE:

l i

PLANT VOGTLE TRAINING DEPT.

QUESTION USE LIST for EXAM: SR-905-90-06.AS !

Total Points: 24.00 FILE NO:C90-06-018 Page 1 :

Assembled by Chuck Stuhaan on 07/30/90 in MANUAL mode. I ITEM EB NUMBER REVISION DESCRIPTION 1.01' LO-SS-26000-01-01 7 07/12/90 PORV block valve interlocks' 1.02 LO-SS-44000-01-08 6 06/21/90 rod control tech specs 1.03 LO-SS-26000-01-05 7 07/12/90 Isentnalpic Throttling Process 1.04 LO-SS-26000-01-06 4 06/20/90 CVCS interlocks and response

f t

1.05 LO-SS-44000-01-01 10 06/20/90 letdown line leak identification 1.06 LO-SS-44000-01-05 9 07/12/90 Urgent failure causes 1.07 LO-SS-44000-01-14 5 07/12/90 leak rate determination ,

1.08 LO-SS-44000-01-12 6 06/28/90 isolate 1/d 1.09 LO-SS-44000-01-16 8 07/30/90 ril of the s/d banks 1.10 LO-SS-44000-01-17 3 ,06/28/90 OTdelta T response to res pressure 1.11 LO-SS-44000-01-18 2 06/28/90 CNTMT SPRAY ACTUATIONS 1.12 LO-SS-44000-01-23 00 / / BTRS flowpath during dilution

=================================================================n

END OF SECTION l

l 1

i l

I 1

, ~ ,EXAN PULLED FOR CCR FILING SUPV: DATE:

i PLANT VOGTLE TRAINING DEPT.

QUESTION USE LIST for EXAM: SR-905-90-06.AS Total Points: 24.00 FILE NO:C90-06-018 Page 1

]

Assembled by Chuck Stuhaan on 07/30/90 in MANUAL mode. f l

IIEd' Eb NUMBER REVISION DESCRIPTION ;

1 2.01 LO-SS-53000-02-01 6 06/28/90 causes of turbine trip 2.02 LO-SS-53000-01-03 3 06/20/90 reactor trip diagnoses 2.03 LO-SS-53000-01-02 4 06/28/90 sgwic response to rcp trip i 2.04 LO-SS-53000-01-16 2 06/20/90 steam dumps ops 2'05

. LO-SS-53000-01-17 3 06/20/90 porv operation indications 1

2.06 LO-SS-53000-01-25 1 07/12/90 SRNI Reset operation between P-6 an l i

2.07 LO-SS-53000-01-20 2 06/28/90 FWI reset requirements 2.08 LO-SS-53000-01-21 3 06/28/90 System response to MFP trip.

2.09 LO-SS-53000-01-22 3 06/29/90 Hi Rad effects on CVI 2.10 LO-SS-53000-01-23 3 ,06/29/90 SI reset indications with reactor t 2.11 LO-SS-53000-01-24 3 06/29/90 LOSS OF.1BB07 EFFECT ON EDG OPERATI 2.12 LO-SS-53000-01-19 8 06/29/90 rcp tech specs

==============================================================,===

END OF SECTION l

.l 1

1 l

1 l END OF QUESTION LIST l

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' ' Page 1

, EXAM KEY T 7==============================================================================

1 -- , .

5 1. '

.Q: Whichlof the following is the reason to make the transfer to hot U

leg recirculation at.the 11 hour1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months point as;ooposed to waiting-until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or. longer?-

, a.- To begin removal of_ boric acid off the-fuel cladding before allowable peak centerline temperatures are exceeded.

b. To begin reducing the boron concentration in the core ~ prior to exceeding theLsolubility limit for boron.

r c. To reduce cooldown stresses in the downcomer region of the

' reactor vessel thus minimizing brittle fracture potential.

V.

d.. To flush _the plated out boric acid back into~the core which wil'1 ensure _the RCS pH remains high to' limit corrosion.

k, A: a.

-R: "oG BACKGROUND DOC.,

EB#: Lv-OR-36101-02-01 Point Value: 1.00

j. .-==================================================n===========================

2. -Q: Which of the following describes the most probable progression J

through the EOP's to terminate a spurious train "A" SI.

a. EOP 19000-C through Step 27, EOP 19011-C to completion.

b. EOP 19000-C through Step 25,-EOP 19010-C to Step 7, EOP 19011-C to completion

c. EOP 19000-C to Step 27, EOP 19235-C to completion, EOP 19011-C to completion,

d. EOP 19000-0 to Step 25, EOP 19235-C to completion, EOP 19010-C to Step 7, EOP 19011-C to completiori A: a.

-= R: 19000-C,R9,PG. 13, 19200-C F-0.3, 19011-C EBN: LO-OR-37002-03-02 Poir.'c Value: 1.00

================================r=================================

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$ 45 SR-905-90-06.B; KEY -

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==================a===============================================

3. .Q: Unit 1~was operating at 100% poweriwhen a major steam line break

-occurred resulting in,an uncontrolled depressurization of all S/G's.-The reactor tripped and SI was initiated. Due to'the plant ,

cooldown to 194 degrees F over a.47 minute period, a' transition- ,

wac made to to'the appropriate FRP after monitoring CSFST's. All.

S/G's are currently. reading approximately 22% WR (wide range) level with AFW available.

~Under these conditions what is the recommended feedrate.to the S/G's if the steam line break cannot be isolated?

a. Under 10 gpm per S/G

b. 50. ppm to each-S/G

c. E70 gp a total AFW flow

d. 1260 gpm total AFW flow i

i A: b.

R: 19241-C,R7,PG. 3, EB#: LO-OR-37002-09-03 Poirt Value: 1.00

==================================================================

(4 9

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SR-905-90-06$B; KEY.

i ,

Page 3 EXAM KEY ,

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t 4' . . - 'Q: A-natural-. circulation cooldown is..in progress in accordance with 19002-C, Natural Circulation Cooldown. The RCS'is_at 510' degrees-F y and 1900 psig. All CRDM cooling fans have tripped and cannot be restarted.

Without the CRDM fans in operation, which of the following is the greatest concern?

4

a. Damage.may occur to the CRDM coils because of overheating.

b. NDT requirements are more likely to be exceeded for the; reactor head flange welds.

c. Damage may occur to the excore nuclear instrumentation because- '

of overheating.

d. The formation of a steam bubble in the reactor. vessel head-region.

A: d.

R:'19002-C,R6,PG. 7,-GREB 002-010-001, EB#:.LO-OR-37012-05-05 Point Value: 1.00

===================================================================== ========= <

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"R SR-905-90-06.B;= KEY-

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==================================================================

5.- Q: 'During a loss of'all onsite and offsite power,_the reactor' ,

operhtor failsLto isolate RCP seal injection and leakoff.-This i goes unnoticed-for some time. What adverse affect will this have-on the RCS? ;

a. An inadvertant dilution say occur-viaLthe seal injection !

' lines.

b. _ Unnecessary damage to the RCP thermal barrier HX will take place. prior to restoring AC power to the plant.

c .- .Will have an increased loss of RCS-inventory through the RCP seals.

d. RCS will be open to the Containment atmosphere allowing I i

non-condensible gases to enter the system.

i A: c.

R: WOG BACKGROUND DOC, 19100-C,R4,PG. 7, EB#: LO-OR-37031-01-01 Point Value: 1.00 L. .

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.ma=mmmmmmmmmmmmmm========= man ===ammmmmmmmmmmmmmmm=mmmm=mm=mpemm===m===========:

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6.. Q:)All FW/ flow has been lost,.and the control room operatorsLarei responding to a loss:of secondary heat sink. They are unable to.

restore. feed flow,-so the crew establishes RCS feed and bleed with one PORV open.

At'this point the operators should:-

a. Terminate ~ attempts to establish a S/G heat sink because one PORV-will allow sufficient bleed and SI flow for cooling.

b. Keep trying to open the failed PORV'and reduce SI flow-_as necessary to prevent rapid overpressurization of the RCS.4

-c.- Establish alternate bleed paths and cooling methods because one PORV may not allow the RCS adequate SI flow,

d. Terminate RCS feed and bleed because one PORV ogen will cause RCS pressure to increase, SI flow and PZR level will: decrease.

A: c.

R: 19231-C,Rll,PG. 10, GREB 000-054-004, . ,

EB#:- LO-OR-37051-08-02 Point Value:'l.00

=====

ra==========================================================

7. Q: .The control room operators are responding to a LOCA. An operator monitoring CSFST's observes-that all core exit thermocouple readings are greater than 1200 degrees F. The control room operators enter the correct FRP.

Which one of the-following methods-used in the FRP is theimost effective in restoring the CSF associated with these symptoms?

a. Rapidly depresLarize the secondary to depressurize the RCS.

b. Reduce RCS pressure by opening all available RCS vent paths to L-1 containment.

L c. Start all RCP's.

i l d. Establish high-head safety injection flow.

l>,

A: d.

l R: 19200-C,R6,PG. 5, GREB 000-074 'o4, EB#: LO-OR-37061-02-05 Point Value: 1.00

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< EXAM [ KEY M

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1 8.'. Q: The; plant is operating at power,-nnd the following conditions-e:

exist:

Reactor power - 58% and' trending;up,

RCS, pressure -2210 psig and decreasing slowly

Tavg - 569 degrees F andl slowly decreasing

Turbine power - 595 Mw steady - no= change

~

~

S/G. levels - 53%' slight increasing trend noticed ,

Steam pressure - 970 psig and slowly decreasing ~'

Containment pressure - 1 psig and slowly increasing

Makeup to_ condenser hotwell in progress Based on-the indications listed above, the most likely event in progress is which of the following:

a. Continuous rod withdrawal accid

b. RCS LOCA q

c. Steamline break inside containment -

l i

d. Steamline break outside containment q

)

A: c.

.j R: 19000-C,R9,PG. 11, GREB 000-040-003, l EB#: LO-OR-37121-05-01 ' Point Value: 1.00 )

==================================================================u=========== ,

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. EXAM' KEY

n==================en=========================================================.

~

H9 .- .Q: In'19030E(SGTR) if the ruptured SG's MSIV's and MSIV Bypasses can.

not be 1solated,-all remaining MSIV's and' bypasses ~are closed'and the secondary system-is isolated. Why is this, action taken?

a.

~

To allow usoto operate the turbine driven AFW pump from a S/G that is isolated from the ruptured S/G. '

b. To allow us to align the condenser steam dumps to the ruptured' l*

S/G only. ~

c. To allow Chemistry to. draw accurate' samples from the ruptured S/G that wi'l not be affected by the intact S/G flows.

1

~

d. To isolate the ruptured'S/G from.the intact S/G's as well as minimize potential radiological releases.

i i

'A: d; <

R: 19030-C,R8,PG. 4, EB#: LO-OR-37311-07-05 Point Value: 1.00

.=a============================================================================

10. Q: Spent fuel pool ~ cooling was swapped from B train to A train. When this was done CCW was not valved into the proper heat exchanger, u What would most likely bring this to the attention of-the control c room operator first if the improper valve lineup is not spotted j locally?

a. Spent Fuel Pit Hi level alarm.

]

b. Spent Fuel Pit High Temp alarm,

c. Radiation levels on Fuel Handling Building Area monitors increase to the warning alarm setpoint.

d. Spent Fuel Pit Lo level alarm A: b.

R: VOGTLE TEXT, 17005-1,R10,PG. 13,

'EB#: LO-OR-25102-03-02 Point Value: 1.00

============================================================================== ,

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f SR-905-90-066B; KEY g4 Page.8' l' EXAM KEY

==================================================================

-11'. :Q: A Safety [ Injection occurred 15 minutes ago.- '

l All pressurizer safeties and PORV's are closed and all air-operated valves?have' cycled to their failed positions.

i' The most' probable cause of PR'.' level still increasing is:.

d

a. Seal return relief lifting a

b. Lctdown line relief lifting c.- RHR discharge relief lifting

d. CCP suction relief lifting.

i

A: a.

R: 1X4DB114,fl8004-C,R6,PG. 18, j

.EB#: LO-OR-09202-03-01 Point Value: l'.00 l

==================================================================

l

13 . Q:: Which one of the'following conditions would result in the core becomingLuncovered for the longest period of time if a total loss.

of RHR occurred 120 hours0.00139 days 0.0333 hours 1.984127e-4 weeks 4.566e-5 months after shutdown? (Assume no operator action taken)

I

a. ~ Refueling pool filled to Tech Spec level with fuel movement j in progress in the containment building. i

b. RCS atJmidloop with all SG primary.manways removed. No nozzle

-dams are installed.

.c. RCS at midloop with all SG primary manways removed. Hot leg i nozzle dams are installed and there have been no vent paths 1 established.

d. RCS at midloop with all SG primary manways removed. Hot leg nozzle dams are. installed and the pressurizer manway has been removed.

A: c.

R: VOGTLE TEXT, 18019-C,R6,PG. 2, EB#:-LO-OP-121D1-15-02 Point Value: 1.00

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SR-905-90-06 BT KEY 0

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-adw===amammmmmmm=====ma=======================================================

-Q : A-low spent f'uel' pool' level. alarm is received.

13. The operator sent to investigate reports that the level is decreasing and the .

~

transfer canal is filling due to leakage past the spent fuel ~ pool gate. -If the service air. system is not available to inflate the igate seals,.what action (s) should be taken by the operators? ;

a.. Makeup to the Spent Fuel Pool from-the Potable Water Storage Tank.

'i

b. Useithe Nitrogen System to operate =the gate ram to stop the 3 Spent Fuel Pool _ Gate leakage.

c. Establish Feed and Bleed to the Spent Fuel Pool.

d. Makeup.to the Spent Fuel Pool from the RWST and use bottled-

. nitrogen to inflate the gate seals.

A:' d.

R: 18030-C,R4,PG. 2-3, 13713-C, EB#: LO-OR-25102-05-01 Point Value: - 1. 0 0 -

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14. -Q: The-plant is~ operating-at 80% power in a stable condition. All systems are operating normally, and the Rod Control System is in

~

AUTO. Without warning,.the rods continuously step-out at the maximum-rate. Shortly there-after, Tave begins to increase above

-Tref. Tref remains at the normal value for 80% turbine load. PZR' pressure and water level also begin to increase.

The proper ~immediate action is which'of the following?

a. Trip the reactor

b. Place the BANK SELECTOR switch in the MAN position, and adjust turbine load in' STANDBY to match Tavg to Tref. !

c. Place the BANK SELECTOR switch in the MAN position, and withdraw the control rods in manual as required to match Tavg.

to Tref

d. Place'the BANK SELECTOR switch in the MAN position, and insert'

the control rods manually as required to match Tavg to Tref.-

5 A: d. i R: 18003-C,RS,PG. 9, GREB 000-001-002, EB#: LO-OR-60303-12-01 Point Value: .1.00

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' 15. ~ Q: Initial Conditions:

1) Unit-1 at 100% RTP ,

2 )' Train "A" equipment running '

3). Normal lineup l

Initiating event:

1)- .The 1Bh03 normal supply breaker opens

-(

Assume that 1B Emergency - D/G has auto started, 1B diesel breaker did not automatically c1'se in and the control room operator was. I unable to manually-close the' breaker.- Which of the following actions should be taken? (Select one) a.. Depress both " Emergency Stop" pushbuttons on the. Control'. Room-

Panel

b. Local'ly close the emergency breaker

c. Depress the " Manual Stop" pushbutton on the Control Room Panel-

d. Manually initiate a Train "B" safety-injection LA : a.

R: 18031-1,R6,PG. 2, EB#: LO-OR-60323-01-02 Point Value: -1.00 ,

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. ~ pig 3-12

, EXAM KEY

==================================================================

16. .Q: The unit is at 100% power and is being readied for a ramp to 80%

power. The' Unit Shift Supervisor directs the RO to insert rods

=from 220.to.200 steps'which is equivalent to approximately 75 pcm. {

The amount of reactivity resulting from power defect is approximately 375 pcm. Ignore any changes in Xenon concentration and-assume a boron worth of 10 pcm/ ppm.. Boron concentration is i 450 ppm initially.. How many gallons.of demin water or~7000-ppm-boric acid must be added to complete this power change?

a. 282 gall'ons of boric acid >

b. -423 gallons of boric acid

c. 4232 gallons of demin, water

d. 6468.5Lgallons of domin, water l

A: a.

How calculation was done.

+375 pcm - '75 pcm = Delta + 300 pcm

--> bcrate 30 ppm Using PTDB Tab 2.0 to borate 30 ppm from 450 ppm requires 282 gals of boric acid.

R: 13009-1,R6,PG. 5, PTDB TAB 2.0, d

.EB#: LO-OR-09401-07-01 Point Value: 1.00

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17. Q: The plant:is in-hot standby with preparations 1almost complet'edifor taking the reactor critical. Transformer'1NB01X catches fire and. ..<

is' destroyed. '"

1The~ reactor startup.wil1$be delayed due to: I the Train'A' emergency diesel generator being-declared-3 a.,

' inoperable. .;

b. a less of essential 120 VAC bus 1NY1N.. ;;

.c. -a loss of adequate pressurizer pressure control'. <1'

.. i a.- the reactor trip breakers must be opened, due to a' loss'of.

digital rod position.

A: a. ,

ut

.R: LO-LP-3 9 212-03, 13145-1,R20,PG. 6, 17035-1,R5,PG. 8 L EB#: LO-CR-11104-08-01 Point Value: .1'. 00: - -

==============================================================================ft 18.. Q:DIf.a loss of A train power (AA02) were to occur,.which ofithe sfollowing-RHR loop suction valves would be affected?

a. HV-8701A'

'?

b. >HV1 8701B Tc, HV-8702A 1

d. 'HV-8702B A:=a..

R: 13011-1,R17,PG. 6, EB#: LO-OR-12101-08-04 Point-Value: 1.00 i

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19 . - Q: During solid pla'nt, operations, HV-0128 closes-due to a control 1 system feilure. The ensuing RCS pressure transient would best be

-described by:

a ._ Pressure will decrease and will continue to do so, trending _

towards; atmospheric pressure.

~

b. Pressure will decrease until the p'ressurizer heaters energize and stabilize pressure.- '

c. Pressure will increase until the:PZR: spray valves and/or PZR '

PORV's open.

d.

Pressure will increase until the RHR suction reliefs'and/or PZR PORVs open. !

'l A: d. o i

R: TS 3.4.9.3, 13011-1,R16,PG. 2, EB# : LO-OR-16501-03-02 Point-Value: 1.00 '

=================,============================================================-

4

20. Q: During a reactor startup, the RO pulls control rods and changes the' source range level from 100 cps to 250 cps. If the RO adds an i equivalent amount of. reactivity with a subsequent rod pull, which '

of'the followi'ng statements is TRUE?

a. Source range count rate will increase to 500 cps. ,

b. The reactor will be critical.

s

c. Source range SUR will double. '

d.. Not enough information available.

A: b.

R: VOGTLE TEXT, W NEP 211, CH 7, 12003-C,R14,PG. 1 EB#: LO-OR-33310-11-01 ~ Point Value: 1.00

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EXAM KEY

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$21. Q: -' Reactor power: is at ' 58% . The maximum allowable value for the-nuclear enthalpy rise hot chann61 factor for this powerclevel:is:- -

i y a.. 1.55: (+/- .02) '!

b. J.75 (+/- .02)- '

f

c. 1.82=(+/ .02) N

o> d. 1.93-(+/- .02) , i I

1 yl , , J ., A b.-

m,, ;

R:cVOGTLE TEXT, T.S.:3.2.3, -;

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EB# : - LO-OR-34 510-06 Point Value: 1;00; l1:

22. Q: 1The' plant is operating at-100% power. The TDAFW pump is being m" 3; started up to perform surveillance testing. While the. pump is.

,f '

operating, a'ste'am break develops on the steam supply line in the-TDAFW pump house. Personnel evacuate the building. The control-

-room operators isolate the break by shutting both HV-3009 and

HV-3019.

m- '

Under these conditions, the plant:

T

a. Must be shut'down to hot standby within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and '

to hot shutdown within the following16 hours. !

i

b. Can continue to operate as long as the remaining AFW' pumps are

'T verif.ied to be operable at least once every 31 days.

A

c. Must be tripped immediately, and 19000-C must be implemented. <

d '. - Can continue to operate'for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , by which time the ;

break must be repaired and the pump returned to operable status.

.,,\"

f ,

A: d.

R: T.S. 3.7.1.2, GREB 000-040-008, EB#: LO-OP-39211-03-03 Point Value: 1.00

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,==================================================,==================

4

23. Q: During-~ response to.a loss of all AC' power, the operator is t directed to-reset SI to: ,

t

.a. l Allow the sequencer to load non-essential loads.when l power.is restored to-a-1E bus.-

I

b. Prevent automatic loading of ECCS equipment-when power- '

is restored to'a 1E bus.

c. Allow automatic ~ restoration-of-ACCW flow to the RCP's Lwhen power is restored to the 1E bus. '

d. Prevent,any spurious ESFAS actuation signals from'being generated-when power is restored to a 1E bus.

A: b.

R: WOG BACKGROUND DOC., 19100-C,R4,PG. 6, 19100-C,R4,PG. 14 EB#: LO-OR-37031-09-01 Point Value: 1.00

>==============================================================================

-24. Q:. Reactor shutdown is logged in the Unit Control Log when:

a. Control rod insertion is commenced.

b. Control rods are fully inserted.

c.- Reactor trip breakers are opened.

d. Shutdown bank insertion is commenced. ,

A: b.

R: 12005-C,R8,PG. 5, T.S. PG. 1-9 MODES,

'EB # : LO-OR-39202-02-02 Point Value: 1.00 L

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l END OF TEST KEY

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F The following'is'an alpha-numeric list of figures and references whichi

- should accompany this test when-it'is administered.--

1 L

Reference Question i

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1 1

N.

EXAM PULLED FORLCCR FILING SUPVt. DATEt pV PLANT VOGTLE TRAINING ~ DEPT.

QUESTION USE LIST for EXAM::SR-905-90-06.B-Total' Points: 24.00 FILE NO:C90-05-004 Page 1 <

Assembled by Chuck.Stuhaan on 07/30/90 in MANUAL mode.

ITEM EB NUMBER , REVISION DESCRIPTION LO-OR-36101-02 .

' 1. 5 06/29/90 Boron plateout on fuel rods

'2. LO-OR-37002-03-02 9 06/29/90 EOP Usage - Terminating SI

3. LO-OR-37002-09-03 4 06/29/90 Use of the FRPs. Determine S/G feed

4. LO-OR-37012-05 3 06/29/90 Concerns about NC cooldown without

.5. LO-OR-37031-01-01 11 06/29/90 Adverse affects of not isolating th ;

6. LO-OR-37051-08-02 4 06/29/90 Reason for opening-2 PORVs vs. 1 PO

7. LO-OR-37061-02-05 5 06/29/90 Select the most effective-method to

~

8. LO-OR-37121-05-01 5 07/12/90 Given conditions, what accident is

9. LO-OR-37311-07-05 9 06/29/90 Isolation of Ruptured SG from the o '

10. LO-OR-25102-03-02 12 06/29/90 Loss of spent fuel pool cooling - i

11. LO-OR-09202-03-01 10 06/29/90 CVCS /PRT Interface / Leakage paths -

12. - LO-OR-12101-15-02 8 06/29/90 Loss of RHR with the RCS at Mid-loo

13. LO-OR-25102-05-01 11 06/29/90 Spent fuel pool gate seal operation

14. LO-OR-60303-12-01 8 06/29/90 IOA'S FOR UNCONTROLLED ROD WITHDRAW

15. LO-OR-60323-01-02 7 06/29/90 Actions on loss of Class 1E - BA03 16..LO-OR-09401-07-01 14 06/29/90 Boration calculation

17. LO-OR-11104-08-01 10 06/29/90 DG operability following the loss o

18. LO-OR-12101-08-04 8 06/29/90 RHR loop suction valve independant

19. LO-OR-16501-03-02 8 06/29/90 COMS response to solid plant condit

20. LO-OR-33310-11-01 9 06/29/90 Subcritical multiplicatior. theory

21. LO-OR-34510-06-02 13 07/30/90 Calculate FN Delta H gi'en reactor

22. LO-OR-39211-03-03 3 06/29/90 T.S. problem with TDAFW pump inop a

= = = = = r = = = = = = = = = = = = = = = = = = = = = bf 8 T = C O N TI N U E D = 0 N = N E X T = PAG E = = = = = = = = = = = = = = = = = = = = = = = = ,

.4},

Q?-

i.

J SR-905-90-06.B W ' '

Page~2 j ,

-QUESTION'USE LIST IllM- EB-NUMBER' REVISION- DESCRIPTION-d . .

-23 7LO-OR-37031-09-01 .

.11- 06/29/90. What is the bases for SI reset 11n 1:

24. LO-OR-39202-02-02 2 . 06/29/90 When' Mode 3 logged,when-

================================================================performingi

==

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END OF QUESTION LIST t

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' 9 C90-05ko05 1

's

.)

-PLANT VOGTLE TRAINING DEPT.

MASTER- KEY

'FOR a.

EXAM: SR-905-90-06.BS- -;

-- i 3 -.

i Total Points: 24.00 i

s. !

-1 ASSEMBLED IN MANUAL MODE.

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77 7 ,- 1

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.SR-905-90-06.BS;-KEY inq",

Page 1-EXAM KEY

-==============================================================================

l

1. Q: Which'of.the following is the reason to make the transfer to hot

. leg're' circulation at the-11 hour point as opposed.to waiting until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .or longer?

a. To begin removal.of boric acid off the fuel cladding before n.

allowable peak centerline temperatures are exceeded.. j 1

b. To'begin reducing the boron concentration in the core prior to' l exceeding the solubility limit for boron.

c.- To reduce cooldown stresses.in the downcomer region of the

~

reactor vessel thus minimizing. brittle fracture potential. !

d. To flush the plated out boric acid back into the core which '

vill ensure the RCS pH remains high to limit corrosion. e l

A: a.

R: WOG BACKGROUND-DOC.,

EB3: LO-OR-36101-02-01 Point Value: 1.00 i

======================:,===========================================

t

2. Q:~Which of the following describes the most probable progression through: the EOP's to terminate a spurious train "A" SI.

a; EOP 19000-C through Step 27, EOP 19011-C to completion, j b.- EOP-19000-C through Step 25, EOP 19010-C to Step 7, EOP 19011-C to completion

, c. . EOP 19000-C to Step :27, EOP 19235-C to completion, EOP 19011-C to completion,

d. EOP 19000-C to Step 25, EOP 19235-C to completion, EOP 19010-C to Step 7, EOP 19011-C to completion A::6 i

R: '19000-C,R9,PG. 13, 19200-C F-0.3, 19011-C EB#: LO-OR-37002-03-02 Point Value: 1.00

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SR-905-90-06.BS;= KEY !

, -Page<2- 1 EXAM KEY 4

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'i 3 '. . Q: Unit 1 was= operating at.100% power when a major steam line break

occurred resultingtin.an uncontrolled depressurization_of all.

S/G's. The reactor tripped-and SI was initiated. Due to the plant cooldown to 194 degrees F over a 47 minute period,..a. transition was made.to to the appropriate FRP after monitoring CSFST's.EAll-S/G's are currently reading approximately 22% WR (wide range) 1 level'with AFW available.

Under these conditions what-is the recommended feedrate to-the S/G's ifethe steam line break cannot be isolated? .

a '. Under 10 gpm.per S/G-b._ 50 gpm to each S/G t

c. 570 gpm total AFW flow i

d. 1260 gpm total AFW flow A:-b. !

'R: _19241-C,R7,PG. 3, EB#: LO-OR-37002-09-03 Point Value: 1.00 ,

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+

4. Q: A natural circulhtion cooldown is:in progress'in accordance with 19002-C, Natural' Circulation Cooldown. The:RCS is at 510 degrees F and 1900 psig. All CRDM' cooling fans have. tripped and'cannot be-restarted. -l Without'the CRDM fans in operation, which of the following is the greatest concern?

a. Damage may occur to the CRDM coils because of overheaf.h.ng.

Eb. NDT requirements are more likely to be exceeded for the !

reactor head flange welds. ,

c. Damage may occur '.o-the excore nuclear instrumentation because of overheating.

-d. The formation of a steam bubble in the reactor vessel head- -

region.

A: d. 4 Rt 19002-C,R6,PG. 7, GREB 002-010-001, EB#: LO-OR-37012-05-05 Point Value: '1.00

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. J,(t'f EXAM KEY:

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E 5 '. . Q:fDuring'alloss of all onsite and offsite power,_the reactor operator fails;to_ isolate RCP_ seal injection and leakoff.LThis

~

goes. unnoticed for'some time. What adverse affect will thio have

, on the RCS?

a. An inadvertant dilution may occur via the seal injection lines.

=

b. Unnecessary damage tx) the RCP thermal barrier HX will.take.

place' prior.to restoring AC power to the plant.

c. Will have an increased loss'of RCS inventory through the RCP seals.

1. RCS will be open-to the Containment atmosphere allowing non-condensible gases'to enter the system.

=

A:'c.

R:LWOG BACKGROUND DOC, 19100-C,R4,PG. 7, EB#: IO-OR-37031-01-01 -Point Value: 1.00'

==================================================================

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==============================================================================t 6 -- Q:5All FW flow has b'een lost,'and:the control room operators?are

~

. responding.to'a-lossLof secondary' heat sink. They:are unable to restore feed flow, so the crew establishes RCSl feed and bleed'with one PORV open, e

At this point the operators should:

a. Terminate attempts to establish a S/G heat sink because'one

.PORV will allow sufficient' bleed and SI flow for cooling.-

b. Keep trying to open the failed PORV and reduce SI flow as necessary-to prevent retpid overpressurization of the.RCS.

c. Establish alternate bleed paths and cooling methods because one PORV may not allow the RCS adequate SI flow.

d.: Terminate RCS feed and bleed because one PORV open will cause RCS pressure to increase, SI flow and PZR level will decrease.

A: c.-

R: 19231-C,R11,PG. 10, GREB 000-054-004, EB#: LO-OR-37051-00-02 Point Value: 1.00 ;

===========================================================c================== ,

7.. 'Q: The control room operators are responding to a LOCA. An operator 1' monitoring CSFST's. observes that all core exit thermocouple readings are-greater than 1200 degrees F. The control room operators enter the correct FRP.

Which one of the following methods used in the FRP is the most effective in restoring the CSF associated with these symptoms?

a._ Rapidly depre'ssurize the secondary to depressurize the RCS.

b. Reduce RCS pressure by opening all available RCS vent paths to containment.

c. Start all RCP's.

d. Establish high-head safety injection flow.

A: d.

l R: 19 2 0 0-C , R 6,'PG . 5, GREB 000-074-004, EB#: LO-OR-37061-02-05 Point Value: 1.00

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.8. -Q: The plant is operating at power, and the followingEconditions I

, exist: ,

1 J

Reactor power - 58% .and trending up

- .RCSipressure - 2210 psig and decreasing slowly;

.Tavg - 569 degrees F and slowly-decreasing Turbine-power - 595 Mw steady .no change' S/G levels.-'53% slight increasing trend noticed ~,

~

LSteam pressure - 970 psig and slowly decreasing-Containment pressure - 1 psig and-slowly increasing:

\

r;

Makeup to. condenser hotwell in progress 3 Based on the indications listed above, the most likely event in progress is-which of-the following:

a. Continuous rod withdrawal accident

'b. RCS LOCA.

c. Steamline break inside containment d .- Steamline break outside containment .

.A: c.

R: 19000-C,R9,PG. . 11, GREB 000-040-003,. ,

EB#: LO-OR-37121-05-01 l Point Value: .1.00

==================================================================

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. _ ;4 S .

P;go 7-

.- , EXAM KEY' i

s maammm%Amatammmmmmmmmmmmmmmmm=REMB3mmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmm=m=m===m=

9 '. Q: In;19030 (SGTRhiftherupturedSG's-MSIV's'andMSIVBypassescan )

not be, isolated, all remaining MSIV's and bypasses are closed and j the~ secondary _ system istisolated.;.Why is this action ~taken?_ l

a. To allow us to operate the turbine driven AFW pump from a S/G >

that-is isolated from the ruptured S/G.

b. To allow us to align the condenser steam dumps to_the ruptured: i S/G only. .

c. To allow Chemistry to draw accurate-samples from the ruptured-S/G that will not be affected by the intact S/G-flows.

1

d. To isolate the ruptured S/G from the intact S/G's as well as minimize potential radiological releases, 4

A: d. l R: 19030-C,R8,PG. 4,.

EB#: LO-OR-37311-07-05 Point Value: 1.00

============================================mm====================

l

10. ,Q: You.have entered 19100-C, ECA 0.0, Loss of All AC Power. The TDAFW pump will not start'and all S/G levels are 10% (NR) narrow range. Reactor power is <5% on all power range channels. Core exit TC's are 732 degrees F, subcooling is 28 degrees F and RVLIS is not functional.

l At this point you should:

H l

a. Remain in 19100-C, ECA-0.0, Loss of All AC Power.

l

b. Proceed to the Remote Shutdown Panels and implement 18038-1.

-c.- Exit-to 19221-C, FR-C.1, Response to Inadequate Co..' Cooling.

d. Exit'to 19231-C, H.1, Response to Loss of Secondary Heat' Sink.

l B A: a.

L l R: 19200-C, 19100-C,R4,PG. 2, i

EB#: LO-OR-37031-08-02 Point Value: 1.00

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SR-905-90-06.BSP KEY:

I ,; , Page. 8 EXAM KEY

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'11. -Q:sA Safety Injecti'on occurred 15 minutes ago.

All pressurizer safeties and PORV's are-c.osed and all air A operated valves have cycled to their failed positions.

The most probable cause of PRT level.still increasing ist i

a. Seal return relief lifting i

b. Letdown line relief lifting !

c. RHR discharge relief lifting i

d. .CCP suction-relief lifting. !

A: a.

l R:: 1X4DB114, 18004-C,R6,PG. 18, EB#: LO-OR-09202-03-01 Point Value: 1.00 '

.==============================================================================

12 . - Q: Which one of the following conditions would result in the core becoming r aovered for the longest period-of time if a total-loss of RHR octc.. red 120 hours0.00139 days 0.0333 hours 1.984127e-4 weeks 4.566e-5 months after shutdown? (Assume no operator action taken) ;

{

a. Refueling pool fille to Tech Spec level with fuel movement !

in progress in the containment building.

b. RCS at midloop with all SG primary manways removed. No nozzle dams are installed.

c. RCS at midloop with all SG primary manways removed. Hot leg nozzle dams are installed and there have been no vent paths established, j

d. RCS at midloop with all SG primary manways removed. Hot leg 3 nozzle dams are installed and the pressurizer manway has been removed.

A: c.

R: VOGTLE TEXT, 18019-C,R6,PG. 2, EB#: LO-OR-12101-15-02 Point Value: 1.00

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13. Q: A low' spent fuel pool level alarm is received. The operator sent to investigate reports that the level is decreasing and the transfer canal is f4lling due to leakage past the spent' fuel pool gate. If the service air s gate seals, what action (s)should ystembeissaken not available to inflate the by the operators?

a Makeup to the Spent Fuel Pool from the Potable Water Storage 6 Tank.

b. Use the Nitrogen System to operate the gate ram to stop the Spent Fuel Pool Gate leakage.

c. Establish Feed and Bleed to the Spent Fuel Pool.

d. Makeup to the Spent Fuel Pool from the RWST and use bottled nitrogen to inflate the gate seals.

c A d,'

R: 18030-C,R4,PG. 2-3, 13713-C, EB#: LO-OR-25102-05-01 Point Value: 1.00

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EXAM KEY j i

14. Q: The plant is operating at 80% power in a stable condition. All systems are opera *ing norrally, and the Rod Control System is in AUTO. Without warning, the rods continuously step out at the maximum rate. Shortly there-after, Tave begins to increase above ,

Tref. Tref remains at the normal value for 80% turbine load. PZR pressure and water level also begin to increase.

The proper immediate action'is which of tlie following?

a. Trip the reactor '

b. Place the BANK SELECTOR switch in the MAN position, and adjust turbine load in STANDBY to match Tavg to Tref.

c. Place the BANK SELECTOR switch in the MAN position, and withdraw the control rods in manual as required to match Tavg to Tref

d. Place the BANK SELECTOR switch in the MAN position, and insert the controi rods manually as required to match Tavg to Tref.

At d.

R: 18003-C,RS,PG. 9, GREB 000-001-002, '

EB#t LO-OR-60303-12-01 Point Value: 1.00

............. m.........................................re............... 9....

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EXAM KEY ammmmmedammmmmmmmmmmmmmarmaammmmmmmmmmmmmmemammmmmmmmmmmmabsmaammmaussammanama 3 t

15. Q Initial Conditions:

1) Unit 1 at 100% RTP

2) Train "A" equipment running

3) Normal lineup ;

Initiating event:

1) The 1BA03 normal supply breaker opens Asnume that 1B Emergency D/G has auto started, 1B diesel breaker did not automatically close in and the control room operator was unable to manually close the breaker. Which of the following actions should be taken? (select one)- ,

a. Depress both " Emergency Stop" pushbuttons on the Control Room -

Panel

b. Locally close the emergency breaker

c. Depress the " Manual Stop" pushbutton on the Control Room Panel

d. Manually initiate a Train "B" safety injection P

A: a. ,

R: 18031-1,R6,PG. 2, i EBN: LO-OR-60323-01-02 Point Value: 1.00

. mmmmm=ammmmmmmmmmmmmmmmmmmma.mmmmmmmmmmmmemummmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmma a

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1, SR-903-90-06.BS; KEY Page 12 EXAM KEY

16. Q _The unit _is at 100% power end is being readied for a ramp to 80% '

power.. The Unit Shift Supervisor directs the RO to insert rods '

from 220 to 200 steps which is equivalent to approximately 75 pcm.

The amount of reactivity resulting from power defect is approximately 375 pcm. Ignore any changes in Xenon cor. centration and assume a boron worth of 10 pcm/ ppm. Boron concentration is -

450 ppm initially. How many gallons of demin water or 7000 ppm t boric acid must be added to complete this power change? '

a. 282 gallons of boric acid

b. 423 gallons of boric acid

c. 4232 gallons of demin water *

d. 6460.5 gallons of demin. water ;

A: a.

How calculation was done. *

+375 pcm - 75 pcm = Delta + 300 pcm -> borate 30 ppm Using PTDB Tab 2.0 to borate 30 ppm from 450 ppm requires 282 gals of boric acid, t

R: 13009-1,R6,PG. 5, PTDB TAB 2.0, EB#: LO-OR-09401-07-01 Point Value: 1.00

..........................=....,..........................................e=== ,

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17. Q: The plant is in' hot standby with preparations almost completed for i taking the reactor critical. Transformer 1NB01X catches fire and is destroyed.

The reactor startup will be delayed due tot '

a. the Train A emergency diesel generator being declared inoperable,

b. a loss et essential 120 VAC bus 1NY1N.

c. a lost, of adequate pressurizer prescure control. .

d. the reactor trip breakers must be opened, due to a loss of digital rod position.

A: a.

R: LO-LP-39212-03, 13145-1,R20,PG. 6, 17035-1,RS,PG. 8 EB#: LO-OR-13104-08-01 Point Value: 1.00

.......................................a..................................=...

18. Q: If a loss of A train power (AA02) were to occur, which of the '

following RHR loop suction valves would be affected?

a. HV-8701A

b. HV-8701B

c. HV-8702A

d. HV-8702B '

A: a.

R: 13011-1,R17,PG. 6, EB#: LO-OR-12101-08-04 Point Value: 1.00

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'. i SR-905-90-06.BSr KEY Page 14 EXAM KEY

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19. Q: Unit 1 was operating at 100% power when it experienced a SGTR in {

1 S/G. The unit was tripped and a manual SI was initiated. While '

performing the steps in 13030-C to isolate the ruptured-S/G a Main I Steam Safety Valve (PSV-3001) failed fully open on loop 1 and j cannot be reseated. All safety systems functioned as expected. !

The Shift Superintendant assumed the duties of Emergency Director !

and classified the event as a. General Emergency.

While performing his required notifications to the state and local authorities he also makes recommendations to protect the 3

public from the anticipated release of radiation.

The Shift Superintendent / Emergency Director should :

a. Recommend precautionary evacuation of all people within a i 2-mile radius from the plant and evacuation of. people, expected to be located in the Plume Exposure Pathway for ;

a distance of 5-miles from the plant.

b. Recommend that non-essential plant personnel be evacuated.

With anticipated traffic problems associated with the departure of people from the plant, local residents should seek shelter from the plume.

c. Recommend precautionary evacuation of all non-essential i plant personnel and general population for a 5-mile radius around the plant. Also evacuate people in the Plume Exposure Pathway for up to 10 miles from the plant.

d.- Recommend all local residents take shelter until state and local authorities can respond to the emergency.

A: a. s R: 91305-C,R8,PG. 4-7, EB#: LO-OR-40101-36-01 Point Va lue: 1.00

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i SR-905-90-06.BS; KEY "

Page 15 ,

EXAM X;t '

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20. Q: During a reactor startup, the RO pulls control rods and changes '

the source range level from 100 cps to 250 cps. If the RO adds an-equivalent amount of reactivity with a subsequent rod pull, which i of the following statements is TRUE?

1

a. Source range count rate will increase to 500 cps.

b. The reactor will be critical.

c. Source range SUR will double.

d. Not enough information available.

i A: b.

R VOGTLE TEXT, W NEP 211, CH 7, 12003-C,R14,PG. 1 EB#: LO-OR-33310-11-01 Point Value: 1.00

==..................................n.............................,...==...m. 7

21. Q: Reactor power is at 58%. The maximum allowable value for the nucleat enthalpf rise hot channel factor #-r this power level is: ,

a. 1.55 (+/ .02) b.- 1.75 (+/ .02) c.- 1.82 (+/ . 02)

d. 1.93 (+/ .02)

A: b. ;

R VOGTLE TEXT, T.S. 3.2.3, EB#: LO-OR-34510-06-02 Point Value: 1.00 ,

....===...................=........=........................ .=...........===.

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t 7-

SR-905-90-06.BS; KEY Page 16 EXAM KEY

22. Q: The plant is operating at 100% power. The TDAFW pump is being started up to perform surveillance testing. While the pump is operating, a steam break develops on the steam supply line in the TDAFW pump house. Personnel evacuate the buildin?. The control room operators isolate the break by shutting both NV-3009 and HV-3019.

Under these conditions, the plant:

a. Must be shut down to hot standby within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to hot shutdown within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

b. Can continue to operate as long as the remaining AFW pumps are verified to be operable at least once every 31 dhys.

c. Must be tripped immediately, and 19000-C must be implemented.

d. Can continue to operate for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , by which time the break must be repaired and the pump returned to operable status.

At d.

R: T.S. 3.7.1.2, GREB 000-040-008, EB#: LO-OR-39211-03-03 Point Value: 1.00

................a.............................................................

O L

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e V~,

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SR-905-90-06.BS; KEY

., Page 17 EXAM KEY

==..........................................r -...............................

23. Q: Unit 1 is in Mode 6 with refueling in progress.-The Radwaste operator, while on his rounds in the Fuel Handling Building, notes a Spent Fuel Pool Level of 217.5 feet. He notifies the Control Room.

. The correct operator response would be to:

a. Suspend refueling operations immediately and initiate filling the Spent-Fuel Pool to a minimum Tech. Spec. level of 218.5 feet.

b. Commence draining the Spe*'t Fuel Pool since it is near-the overflow level of 219 feet,

c. Stop all activities involving movement of fuel assemblies or crane operations in the Spent Fuel Pool area.

d. Note that no actions are required. Level in the Spent Fuel Pool is still greater than the required 23 feet above the irradiated assemblies.

A: d.

R: T.S. 3/4.9.11, 13719-1,k13,PG. 36, EB4: LO-OR-39213-13-02 Point Valuet 1.00 t

9 l-KEY CONTINUED ON NEXT PAGE 9

4 SR-905-90-06.BS; KEY Page 18

...................... 2......................................................

24. Q: Degassing the RCS during a power descent / shutdown when planning to go to Mode 5 or 6 is important in order tot

a. Remove air and non-condensibles from the pressurizer to insure maxiram pressure response control when using the pressurizer sprays during the depressurization of the RCS.

b. Reduce hydrogen concentration in the RCS in preparation for opening the system to minimize subsequent hydrogen release and airborne potential,

c. Reduce the amount of gas needing to be processed by the Waste Gas System hydrogen recombiners.

d. R0 duce potential for hydrogen bubble formation in the S/G U-tubes (high point of the system).

A: b.

R: 12005-C,R8,19 3-4, '

EBW: LO-OR-61204-06-01 ,

Point Value: 1.00 4

END OF TEST KEY

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The following is an alpha-numeric list of figures and references which should accompany this test when it is administered.

Reference Question r

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END OF FIGURE AND WErikFNCE LIST

, EXAM PULLED FOR CCR FILING SUPV __ DATE: f PLANT VOGTLE TRAINING DEPT. '

QUESTION USE LIST for EXAM: SR-905-90-06.BS Total Points: 24.00 FILE NO:C90-05-005 Page 1 Assembled by Chuck Stuhaan on 07/30/90 in MANUAL mode.

ITEM EB NUMBER REVISION DESCRIPTION

1. LO-OR-36101-02-01 5 06/29/90 Boron plateout on fuel rods

2. LO-OR-37002-03-02 9 06/29/90 EOP Usage - Terminating SI

3. LO-OR-37002-09-03 4 06/29/90 Use of the FRPs. Determine S/G feed

4. LO-OR-37012-05-05 3 06/29/90 Concerns about NC cooldown without

5. LO-OR-37031-01-01 11 06/29/90 Adverse affects of not isolating th G. LO-OR-37051-08-02 4 06/29/90 Reason for opening 2 PORVs vs. 1 PO

7. LO-OR-3 7 061-02-05 5 06/29/90 Select the most effective method to

8. LO-OR-37121-05-01 5 07/12/90 Given conditions, what accident is

9. LO-OR-37311-07-05 9 06/29/90 Isolation of Ruptured SG from the o

10. LO-OR-37031-08-02 9 06/29/90 Should you transition from 19100 if

11. LO-OR-09202-03-01 10 06/29/90 CVCS /PRT Interface / Leakage paths

12. LO-OR-121:1-15-02 C 06/29/90 Loss of RHR with the RCS at Mid-loo

13. LO-OR-25102-05-01 11 06/29/90 Spent fuel pool gate seal operation J4. 10-OR-60303-12-01 0 06/29/90 IOA'S FOR UNCONTROELED ROD WITHDRAW A5 LO-OR-60323-01-02 .7 06/29/90 Actions on loss ,c Class 1E - BA03

36. LO-OR-09401-07-01 14 06/29/90 Boration calculation

17. LO- OR- 1110 4 01 10 06/29/90 DG operability following the loss o

18. LO-OR-12101-08-04 8 06/29/90 RMR loop suction valve independant

19. LO-OR-40101-36-01 5 07/12/90 EPIP usage for PAG's - SRO ONLY

20. LO-OR-33310-11-01 9 06/29/90 Subcritical multiplication theory

21. LO-OR-34510-06-02 13 07/30/90 Calculate FN Delta H given reactor

22. LO-OR-39211-03-03 3 06/29/90 T.S. problem with TDAFW pump inop a

s=====================ht8T=CONTINUND=0N=NEXT=PAGE===================

% i 14 SR-905-90-06.BS b Page 2 QUESTION USE LIST IIEM EB NUMBER REVISION DESCRIPTION

'23. LO-OR-39213-13-02 3 06/29/90 Determine adequt;e level in the sFP

- 24. LO-OR-61204-06-01 3 06/29/90 Reason for degassing when taking th

============================== .====.s=======================================u t, -

e END OF QUESTION LIST l

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1 ENCLOSURE 3 i RE0VAllFICATION PROGRAM EVALVATION REPORT Facility Generated Refer _qngs Material The reference material supplied by the licensee was revieweo and determined to be adequate to support the examination. The licensee supplied a sampling plan describing the requalification cycle and the selection process used for the topics to be included in the examinations. Proposed written, walk-through, and dynamic simulator examinations derived from this sample plan, were reviewed by the NRC exam team.

The validation times for questions on the static simulator exam and open reference exam were revised during the prep week to more accurately reflect the amount of time which a competent operator would require to correctly answer the question. This resulted in adding more test items to each exam.

Some of the JPMs were revised in order to better define the critical steps.

It is important to do this prior to the exam administration, in one instance a st?p was changed from critical to non-critical after the exam had been administered. Although the change was a valid one, it resulted in a change to the pass / fail grade for one operator. There were alto modifications made to steps which fit the definition of a critical step, but had not been designated as such.

Many of the initiating cues were changed to delete information that would not be available to the operators under actual conditions. This included -

information such as what procedure or step number should be used or cues that a key would be needed for a certain valve.

l There were several JPMs which would be better evaleated on the dynamic simulator portion of the exam. These JPMs entailed responding to an imminent instrument failure. Most of these type JPMs were deleted from the exam during the prep week.

The NRC selected some JPMs from outside the sample plan and also wrote three JPMs to be included in the exam. The success ratio on these JPMs was relatively low. One of the NRC developed JPMs, 60316-001-01, directed the operator to step 7c of 18020-C and determine " equal to or greater than 9000 gpm" flow rate on FIT-1720A. This meter is calibrated from 0 - 100 percent I with no means to determine where 9000 gpm would be on this meter, it was decided to cue the operator that CCW flow was greater than 9000 rather than penalize him for what is a problem with the procedure. The facility has l initiated a change to the procedure to correct this problem.

There were two questions asked with each JPM. Weaknesses assoc V ed with the JPM questions included:

- . c_

2 i

i Many of the JPM questions were of a yes/no type or required only two or three words to answer. These questions need to be revised to include the use of higher cognitive skills.

The small number of questions associated with the JPMs resulted, o l occasion, in verbatim repeat back of the previously released answers and tended not to discriminate. The number of questions associated with each JPM needs to be expanded to preclude memorization of answers vice I understanding of t1e concepts, i l

Several tasks in the dynamic simulator scenarios were reclassified as I critical. The majority of thes.e were procedure transitions within the E0P i network. It was also necessary to increase the number of Individual Simulator Critical Tasks (ISCTs) in order to ensure each operator would be evaluated on j more than one.

JPM Performance There was a discernible difference in the performance of JPMs on- Unit 2. The facility had scheduled all JPMs to be performed on Unit 1,however, the NRC requested JPMs to be conducted on both units. The operators tended to be less at ease in Unit 2 as videnced by a more labored search associated with r locating equipment and components. This concern had been 3reviously I identified during observation of training and was one of tie reasons for j requiring a plant differences exam prior to amending operator licenses to include both units. The facility is advised to train and evaluate JPMs on both units. !

Common JPMs were not used as one of the program evaluation criteria for this ;

exam. However, the training department needs to note areas of poor .

performance as feedback for their program. The following JPMs were evaluated I as unsatisfactory for two or more operators:

12101-002-01 Place RHR in service 37111-001-01 Establish condensate flow to SGs on loss of heat sink 60315-001-01 Establish RCS bleed path following a loss of RHR 60316-001-01 Verify CCW heat exchanger cooling capacity 60328-001-03 Locally energize switchgear following local diesel start 60328-001-10 Locally control seal inj. flow following CR evacuation It was noted that the facility had scheduled several JPMs associated with i diesel generators and the Losr of all AC event which had occurred earlier in the year. The results of the i JPMs showed that the training department has 4 incorporated identified problems into their requal program and trained on them effectively, j

Evaluation of Facility Evaluators _

l No facility evaluators were found to be unsatisfactory, i

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Recualification Proaram Evaluation Based on the examination results, the Vogtlo Requalification Program meets the criteria established in rS-601.C.3.b and has been determined to be satisfactory. The unsatisfactory Individual Evaluation is subject to the requirements set forth in ES-601.E.1. The facility is permitted to administer the reexamination for returning the individual to licensed duties. However, an NRC administered examination will be required for license renewal.

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ENCLOSURE 4 SIMULATOR FACILITY REPORT Facility Licensee: Georgia Power Company facility Docket Nos.: 50-424 and 50-425 Operating Tests Administered On: July 26 and August 2, 1990 This form is used only to repo-t observations. These observations do not constitute audit or inspection findings ar,d are not, without further verification and review, indicative of noncompliance with 10 CFR 55.45(b).

These observations do not affect NRC certification or approval of the simulation facility other than to provide _ information which may be used in future evaluations. No licensee action is required in response '.o these observations.

During the conduct of the simulator portion of the operating test, the following items were observed:

1125 Dalqriotion Accumulator The accumulator pressure increased at a rate slower pressure than what the Operations representative expected in the plant.

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