ML032591121

Draft - RO & SRO Written

ML032591121
Person / Time
Site:	Susquehanna
Issue date:	08/01/2003
From:	Roush K Susquehanna
To:	Conte R NRC/RGN-I/DRS/OSB
Conte R
References
50-387/03-301, 50-388/03-301
Download: ML032591121 (100)

Text

SSES LOC I 9 NRC Exam

\\--

I Unit I startup was in progress with Reactor Power at 74% when RECIRC PUMP A DSCH HV-143-F031A open position logic failed causing a run back. The limit switch logic problem has been repaired and the runback logic is to be reset.

Which of the following actions must be completed to reset the runback logic?

I A.

lower GE 1A & 1 B SPEED control signal to slightly lower Recirc pump speed, reset limiter # P

, monitor for speed change.

I lower GEN 1A SPEED control signal to slightly lower Recirc pump speed, reset limiter

1. 2, monitor for speed change.

B.

C.

lower GEN 1A SPEED control signal to slightly lower Recirc pump speed, reset limiter

1. I & #2, monitor for speed change.

D.

lower GEN 1A SPEED control signal to slightly lower Recirc pump speed, reset limiter

1. I, monitor for speed change.

Question Data Answer: D lower GEN 1A SPEED control signal to slightly lower Recirc pump speed, reset limiter #I, monitor for speed change.

ExplanationNustification:

A.

B.

C.

D.

Correct answer Runback does not affect both pumps, only 'A'

1. 2 limiter is caused by Low Reactor Level or Circ water pump trip Valve position does not feed both limiters.

Sys#

System Category KA Statement 295001 Partial or Complete Loss of Forced Core Flow Circulation FORCED CORE FLOW CIRCULATION:

Ability to operate andlor monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF Recirculation flow control system WA#

295001.~~1.05 WA Importance 3.3 Exam Level RO References provided to Candidate None Technical

References:

ON-164-002 Question Source:

New Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 0611 9/03

SSES LOC I 9 NRC Exam 2

Unit 1 & 2 are operating at 100% power with a normal electrical lineup when a loss of Startup Bus 10 occurs.

Assuming no Opera d r actions, and all equipment functions as designed. What will be the

\\

status of the Unit 1 'A' ESS bus breakers following this event?

A.

Transformer 101 (0x201) to Bus 1A - (1A20101) Breaker Closed Transformer 201 (0x203) to Bus 1A - (lA20109) Breaker Open D/G A to Bus 1A - (1A20104) Breaker Closed B.

Transformer 101 (0x201) to Bus 1A - (1A20101) Breaker Open Transformer 201 (0x203) to Bus 1A - (1A20109) Breaker Open D/G A to Bus 1A - (1A20104) Breaker Open C.

Transformer 101 (0x201) to Bus 1A - (1A201.01) Breaker Open Transformer 201 (0x203) to Bus 1A - (1A20109) Breaker Closed D/G A to Bus 1A - (1A20104) Breaker Open D.

Transformer 101 (0x201) to Bus I A - (1A20101) Breaker Closed Transformer 201 (0x203) to Bus 1A - (1A20109) Breaker Open D/G A to Bus 1 A - (1 A201 04) Breaker Open L-Question Data Answer: C Transformer 101 (0x201) to Bus I A - (lA20101)

Breaker Open Transformer 201 (0x203) to Bus I A - (1A20109) Breaker Closed D/G A to Bus 1A - (1A20104) Breaker Open Explanation/Justification:

A.

B.

C.

Correct answer D.

No power available through 01 breaker and 04 breaker would not close if 01 breaker still closed on bus.

The DIG output breaker would be expected to be closed if both the 01 and 09 breakers open.

No power available through 01 breaker

-4th.tcr.cd WA#

295003.AK2.02 WA Importance 4.1 Exam Level RO

&3\\ *,b m*udj Sys#

System Category KA Statement 295003 Partial or Complete Loss Knowledge of the interrelations between PARTIAL OR Emergency generators of A.C. Power COMPLETE LOSS OF A.C. POWER and the following:

2 4 bur References provided to Candidate None Technical

References:

ON-003-001 Question Source:

New Level Of Difficulty: (1-5) frrcrld I,&*

dbfc Question Cognitive Level:

Comprehension 10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAFom.doc Printed on 06/19/03

3 Which of the following describes the effect of losing 125 VDC 1 D614 power to the Auto Depressurization System (ADS) system with a valid initiation signal present?

A.

No ADS valves will open B.

ADS logic " B will C.

ADS logic "A" will D.

ADS logic "BI will still initiate ADS, Only 3 ADS valves will open receive backup power from 125 VDC 1 D 6 4, All ADS valves will open still initiate ADS, All ADS valves will open 2

Question Data Answer: D ADS logic "B" will still initiate ADS, All ADS valves will open Explanation/Justification:

A.

6.

C.

D.

All ADS valves will open.

6 ADS valves will open 1 D614 does not have an alternate correct answer, Either division of ADS will provide actuation of all 6 ADS valves Sys#

System Category KA Statement 295004 Partial or Complete Loss Knowledge of the operational implications of the Electrical bus divisional of D.C. Power following concepts as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER separation

'-4 WA#

295004.~~1.03 WA Importance 2.9 Exam Level RO References provided to Candidate None Technical

References:

TM-OP-083 7

Question Source:

Modified Fitzpatrick, 1992 Level Of Difficulty: (1 -5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForrn.doc Printed on 0611 9/03

SSES LOC 19 NRC Exam

-Le-4 The following alarm is received for Division lo TURB STOP VLV CLOSURE TRIP AR-104-EO2 Which of the following&sts the items causing the alarm?

A.

Turbine tripped Stop Valves 1 and 3 or 2 and 4 95% open.

Loss of power to RPS Channel B.

B.

Turbine tripped Stop Valves 1 and 3 or 2 and 4 < 95% open.

Loss of power to RPS Channel A.

C.

Turbine reset and speed not selected.

Stop Valves or Control Valves 1 and 3 or 2 and 4 < 95% open.

Loss of power to RPS Channel B.

D. Turbine reset and speed not selected.

Stop Valves 1 and 3 or 2 and 4 < 98% open.

Loss of power to RPS Channel B.

-A-Question Data Answer: A Turbine tripped Stop Valves 1 and 3 or 2 and 4 Loss of power to RPS Channel B.

95% open.

ExplanationlJustification:

A.

Correct answer B.

C.

D.

Valve position is <95%

Loss of power from B RPS causes Div 2 alarm.

Control valves will not cause alarm.

Sys#

System Category KA Statement 295005 Main Turbine Generator Ability to determine andlor interpret the following as Turbine valve position KIA#

295005.~~2.03 WA Importance 3.1 Exam Level RO Trip they apply to MAIN TURBINE GENERATOR TRIP:

References provided to Candidate None Technical

References:

AR-104001104001Question Source:

New Level Of Difficulty: (1 -5) 2 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC I 9 NRC Exam

• q 5

During a normal plant shutdown, a reactor scram is manually inserteddhich of the following criteria is utilized to determine if EO-1 00-1 13, "LeveVPower Control" entry is also required?

A.

The status of the Average Power Range Monitor (APRM) "Downscale" lights.

B.

The value of reactor Source Range Monitor (SRM) period after rod movement and detector insertion is complete.

C.

The position and number of control rods inserted.

D.

The ability to monitor instrumentation for valid, current reactor power level.

Question Data Answer: C The position and number of control rods inserted.

ExpIanatiodJustifictio entry (>5% power) the reactor is "shutting down" not "shutdown" C.

D.

correct answer, S I rod >DO, enter EO-I 13, no EO-I02 entry required power not known is EO-I 02 entry Sys#

System Category KA Statement 295006 SCRAM Knowledge of the operational implications of the WA#

295006.~~1.02 WA Importance 3.4 Exam Level RO Shutdown margin following concepts as they apply to SCRAM:

Technical

References:

ON-loo-101 Level Of Difficulty: (1-5) 2 Memo 10 CFR Part 55 Content:

55.41 References provided to Candidate Question Source:

Modified Question Cognitive Level:

L-NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03 m

SSES LOC 19 NRC Exam 6

ON-100-009, PLANT SHUTDOWN FROM OUTSIDE CONTROL ROOM, directs opening HPCl TEST LINE IS0 VLV HV-155-FO11.

The basis for opening HV-155-FO11 is to:

A.

allow RClC System to be cycled between level control and pressure control.

B.

provide a flow path for HPCl in the full flow test mode if it is needed for reactor pressure control.

C.

ensure that keepfill protection is available for both HPCl and RCIC.

D.

permit use of HPCl if RClC is unable to provide sufficient makeup.

Question Data Answer: A allow RClC System to be cycled between level control and pressure control.

ExplanatiodJustification:

A.

correct answer, B.

C.

D.

Valve position does not affect HPCl Flow path.

Does not affect keep fill for HPCl HPCl runs in auto from -38 to +54.

Sys#

System Category KA Statement 295016 Control Room Knowledge of the reasons for the following Disabling control room controls Abandonment responses as they apply to CONTROL ROOM ABANDONMENT:

WA#

295016.~~3.03 WA Importance 3.5 Exam Level RO References provided to Candidate None Technical

References:

ON-100409 Question Source:

Modified Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam 7

Given the following conditions:

- Unit 2 is operating at 100% power

- The "2B" Reactor Protection System (RPS) Bus is on the alternate power supply

- The "2A" RPS MG Set has just tripped Unit 2 may operate in Mode 1 for a limited amount of time based upon:

A.

the length of time for restoration of Emergency Switchgear cooling.

B.

the availability of the Reactor Building Sump Pumps.

\\@$e us4 dl ESc C~U-CcIC'C

'w"g9 cp +cii

-.>lo*

C.

D.

the Containment instrument gas supply to the Inboard MSIVs.

the availability of the Reactor Recirculation Pumps.

+ lend Question Data Answer: D the availability of the Reactor Recirculation Pumps.

ExplanatiordJustification:

A.

B.

C.

not an immediate concern D.

cooling will be restored if 'A' equipment in service or no effect if 'B' equipment in service.

continued operation based on being able to complete leakage surveillance, not an immediate concem correct answer, loss of RPS causes containment isolation and loss of cooling water to the recirc pumps i,

Category KA Statement Exam Level RO Technical

References:

ON-158401 Level Of Difficulty: (1-5) 3 10 CFR Part 55 Content:

55.41 Question Source:

Modified Question Cognitive Level:

NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam

\\/.

8 With Unit 1 at full power, annunciator AR-124-801, INSTRUMENT AIR HEADER LO PRESSURE alarms.

What is this alarm actually monitoring and what is expected to happen if the condition continues to degrade?

A.

Scram air header pressure is less than 65 psig. If not corrected, Scram Discharge Volume Vent and Drain Valves will fail closed preventing a scram.

B.

Scram air header pressure is less than 75 psig. If not corrected, the scram valves will begin to open, scraming in rods.

C.

Instrument air header pressure is less than 80 psig. If not corrected, the scram valves will begin to open, drifting in rods.

D.

Instrument air header pressure is less than 80 psig. If not corrected, Scram Discharge Volume Vent and Drain Valves will fail closed preventing a scram.

Question Data Answer: C Instrument air header pressure is less than 80 psig. If not corrected, the scram valves will begin to open, drifting in rods.

ExplanationlJustification:

A.

B.

C.

D.

Alarm is measuring IA header pressure SDV vent & drain valves closing will not prevent a scram.

Alarm is measuring instrument air header pressure.

correct answer ON-I 18-001 discussion section :scram inlet and outlet valves will begin drifiting open causing random control rod insertion.

SDV vent & drain valves closing will not prevent a scram.

L-.

Sys#

System Category KA Statement

/

WA#

295019.2.4.46 WA Importance 3.5 Exam Level RO 2/

Technical

References:

ON-I 18-001 Level Of Difficulty: (1-5)

,d Fundamental 10 CFR Part 55 Content:

55.41 References provided to Candidate Question Source:

Modified Question Cognitive Level:

NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC I 9 NRC Exam Why is the operator directed to ensure Reactor water level greater than +45 inches when RHR Shutdown Cooling is lost?

A.

to ensure natural circulation occurs since forced circulation is lost.

B.

to ensure the feedwater spargers effectively mix the feedwater added to the Reactor Vessel.

C. to maintain Reactor Narrow range water level instrumentation on scale when the Reactor water temperature increases.

D.

to ensure the water from the annulus area of the Reactor flows into the core shroud area as level decreases inside the shroud.

Question Data Answer: A to ensure natural circulation occurs since forced circulation is lost.

ExplanationlJustification:

A.

B.

Any feedwater addition has no affect on why level is maintained >+45 - fi.L<&& & AtWS, 4& $#C.

C.

0.

correct answer, ON-149-001 discussion, >+45 raises water level above steam seperato to establish natural circulation Narrow and wide range upscale e200 psig in vessel with actual level >+45 with level >+45 there is no difference in water level inside or outside of shroud.

c-Sys#

System Category KA Statement 295021 Loss of Shutdown Cooling Knowledge of the reasons for the following Establishing alternate heat responses as they apply to LOSS OF SHUTDOWN COOLING:

removal flow paths KIA#

295021.~~3.05 KIA Importance 3.6 Exam Level RO References provided to Candidate None Technical

References:

ON-149-001 Question Source:

Modified Monticello 1,1999 Level Of Difficulty: (1-5) 2 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC I 9 NRC Exam

.Le-10 Given the following conditions:

- A Railroad Access Shaft Exhaust Duct Radiation High signal is received.

- Standby Gas Treatment responds as required.

- All Zone I differential pressures are -.27" WG.

Select the Standby Gas Treatment system response to Zone II differential pressure decreasing to -23" WG.

A.

Outside Cooling air dampers will modulate open.

6. Outside makeup air dampers will modulate closed.

C.

Standby Gas Treatment dampers will not adjust for Zone II delta P.

D.

Standby gas treatment fan inlet vanes will modulate open.

Question Data Answer: C Standby Gas Treatment dampers will not adjust for Zone II delta P.

ExplanationlJustification:

A.

B.

C.

D.

Outside air darnplers modulate based on gas stream temperature., not any Z-I, II or 111 delta pressure.

Outside air darnplers modulate based on suction pressure, not Z-I, I I or 111 delta pressure.

correct answer, relay circuit excludes the signal from a zone that does not have an isolation signal present Inlet vanes vary position to maintain flow rate Y

Sys#

System Category KA Statement 295023 Refueling Accidents Ability to operate and/or monitor the following as Standby gas treatmenUFRVS they apply to REFUELING ACCIDENTS:

UtA#

295023.~~1.07 WA importance 3.6 Exam Level RO References provided to Candidate None Technical

References:

TM-OP-070 /

Question Source:

New Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis J 10 CFR Part 55 Content:

55.41 i

NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC I 9 NRC Exam I 1 Given the following parameters:

- Drywell pressure

- Drywell temperature

- Suppression chamber pressure

- Suppression pool water temperature 3.5 psig and rising 145 degrees F and rising 4.6 psig and rising 87 degrees F and steady What event from below would cause the conditions listed above?

A.

A safety relief valve tail pipe has broken in the Suppression Chamber while the valve is open.

B.

A pipe break into the drywell has occurred with a suppression chamber to drywell vacuum breaker open.

C. A downcomer vacuum breaker has failed open during a recirculation leak to the drywell.

D.

A recirculation line partial break has occurred with all containment parameters responding as designed.

Question Data Answer: A A safety relief valve tail pipe has broken in the Suppression Chamber while the valve is open.

-1 ExplanationlJustification:

A.

B.

C.

D.

correct answer, energy into chamber but not into pool, vacuum breakers opening back to drywell when dlp high enough downcomer vacuum breakers are designed to be open for these conditions, equalize pressure across the drywell floor when drywell pressure less than chamber pressure only one vacuum breaker failing would not provide a vent path to Suppression pool atmosphere.

all parameters way too low, especially pool temperature Sys#

System Category KA Statement 295024 High Drywell Pressure Ability to determine andlor interpret the following as Suppression chamber J WA#

2 9 5 0 2 4. ~ ~ 2. ~ WA IrnportanceQ Exam Level Question Cognitive Level:

Analysis I O CFR Part 55 Content:

55.41 they apply to HIGH DRYWELL PRESSURE:

• /

pressure: PlantSpecific RO References provided to Candidate No Technical

References:

Question Source:

Modified Sus uehanna, 199 Level Of Difficulty: (1-5) 3 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam b'

12 The reason for maintaining suppression pool temperature below the Heat Capacity Temperature Limit EOP-I 00-1 03, PRIMARY CONTAINMENT CONTROL is:

ke-pc-

~

If A.

to assure primary containment vent valve opening capability following RPV depressurization to provide adequate subcooling in the pool to prevent chugging of the SRV downcomers.

to assure the containment design pressure will not be exceeded due to compression of the non-condensable gasses.

j f i o( Wf B.

-D C.

D.

to assure the containment design pressure will not following RPV depressurization Question Data Answer: A to assure primary containment vent valve opening capability following RPV depressurization ExplanationlJustification:

A.

correct answer B.

C.

D.

Sys#

System KA Statement WA#

295025.2.4.18 WA Importance 2.7 References provided to Candidate None Technical

References:

EO-000-103 suppression chamber pressure of 13 psig pertains to chugging of SRVs assumption that all non-condensables are in the suppression chamber, reason for shape of curve, not reason for curve the concern is vent valve operation not design pressure of the Containment Question Source:

Modified Clinton 1,2000 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam 13 Given the following conditions:

- Unit I is in an ATWS condition

- Reactor pressure is 950 psig

- Suppression Pool Based on the

-me, when would an emergency blowdown be REQUIRED after the rods are inserted?

A.

19.5feet

6.

22feet C.

17 feet D.

15.5 feet Question Data Answer: 0 15.5 feet ExplanationlJustification:

B.

C.

D.

A.

19.5 is RClC room equalization value.

minimum normal level of the suppression pool.

at 17 feet go to RPV control Correct answer, using HCTL curve.

Sys#

System Category KA Statement 295026 Suppression Pool High Ability to determine andlor interpret the following as Suppression pool level WA#

295026.EA2.02 WA l m

p o

f l

y Exam Level References provided to Candidate Emergency Operating Technical

References:

EO-000-103 Water Temperature they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE:

RO Procedures Question Source:

Modified Peach Bottom, 1996 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Memory 10 CFR Part 55 Content:

55.41 NRC 2003 Rev I H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 0611 9/03

SSES LOC 19 NRC Exam

'L 14 What would be the consequence of NOT limiting Drywell Spray flow for the first 30 seconds of operation?

NOT limiting Drywell Spray flow for the first 30 seconds of operation will:

A.

cause the Drywell Spray Initiation Limit (DWSIL) of 13 psig to be exceeded.

6. cause a pressure drop fast enough to exceed the Containment design differential pressure.

C.

cause rapid pressure drop with minimal Suppression Chamber vapor to support Sup Pool - Drywell vacuum breaker operation.

D.

cause excessive thermal and mechanical shock to the Drywell downcomers.

Question Data Answer: 6 cause a pressure drop fast enough to exceed the Containment design differential pressure.

ExpIanatiodJustification:

A.

B.

Correct answer.

C.

D.

not a consideration limiting flow allows initiation of sprays at any temperature (not pressure), or without concern in all regions of DWSIL curve.

Vapor in drywell is issue not suppression pool

.4-Sys#

System Category KA Statement 295028 High Drywell Temperature Ability to operate and/or monitor the following as Drywell spray: Mark-l&ll K/A#

295028.~~1.01 KIA Importance 3.8 Exam Level RO they apply to HIGH DRYWELL TEMPERATURE:

References provided to Candidate None Technical

References:

EO-000-I03 Question Source:

Modified OYSTER CREEK, 1991 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Fundamental I O CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExarnPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC I 9 NRC Exam 15 The plant is in a LOCA with ECCS systems injecting to the reactor.

Suppression Pool level has dropped to 20.5 feet.

Which of the following is a condition that exists due to Suppression Pool level?

y fit$. l d C d c d A.

Suppression Pool average temperature-ined.

B.

The SRV tailpipe T-Quenchers have been uncovered.

C.

D.

Containment Pressure -ed.

The HPCl Turbine Exhaust has been uncovered.

s &;4&

Question Data Answer: A Suppression Pool average temperature cannot be determined.

ExplanationlJustification:

A.

B.

C.

D.

Correct answer, must use lower SPOTMOS sensors.

Not applicable, SRV tailpipe ehausts uncover at 5 HPCl Turbine exhaust is a concern at -47 Contsainment pressure idication is available.

L-Sys#

System Category KA Statement 295030 Low Suppression Pool Ability to determine and/or interpret the following as Suppression pool temperature Water Level they apply to LOW SUPPRESSION POOL WATER LEVEL WA#

295030.~~2.02 KIA Importance 3.9 Exam Level RO References provided to Candidate None Technical

References:

EO-OOO-io3 Question Source:

Modified Grand Gulf 1,1998 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Comprehension 10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H :\\N RCExarn Prep\\Rich\\NRCAForrn.doc Printed on 06/19/03

SSES LOC 19 NRC Exam 16 Under certain ATWS conditions, the EOPs direct the operators to take action to deliberately lower RPV level in order to reduce reactor power.

Which of the following describes why reactor power decreases as RPV level is lowered?

Lowering RPV level A.

reduces the natural circulation flow which increases the rate of steam removal, resulting in an increase in Void fraction which adds negative reactivity.

B.

uncovers the feedwater spargers to reduce core inlet subcooling, resulting in an increase in Void fraction which adds negative reactivity.

C.

reduces the amount of carryover through the steam higher Void fraction, adding negative reactivity,..

parators, resulting in a I

in a Question Data Answer: 6 uncovers the feedwater spargers to reduce core inlet subcooling, resulting in an increase in Void fraction which adds negative reactivity.

v ExplanatiodJustification:

A.

B.

Correct answer C.

D.

steaming rate decreases due to increased core inlet temperature and reduced moderation steaming rate decreases as power decreases, total carryover decreases quality is a measure of steam to moisture, not applicable to core inlet parameters Sys#

System Category KA Statement 295031 Reactor Low Water Level Knowledge of the operational implications of the Natural circulation: Plant-following concepts as they apply to REACTOR LOW WATER LEVEL Specific KIA##

295031.~~1.02 KIA importance 3.8 Exam Level RO References provided to Candidate None Technical

References:

EO-100-113 Question Source:

Modified WPPSS 2,1996 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Comprehension 10 CFR Part 55 Content:

55.41 C.

NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam

.-+-.-

17 Following a LOCA on Unit I the following conditions exist:

- All control rods have inserted.

- RPV water level is -1 62 inches slowly decreasing

- There are NO injection sources

- RPV pressure is 815 psig and Which of the A.

The reactor D.

ATable5 Answer: B RPV water level remains between -161 and -205 inches.

ExplanationlJustification:

A.

B.

Rapid depressurization accelerates the rate of inventory loss and pressure drop, thereby shortening the time that steam cooling is maintained.

correct answer, Steam cooling occurs when water heated in the core boils turns to steam and rises in the bundles cooling the upper portions. This occurs below -161 inches and continues to -205 inches (MZIRWL). For this to occur there must be a steam flow path and zero injection. From the stated conditions the break is providing a flow path (pressure is NOT rising).

RPV pressure must be stable or decreasing. Raising reactor pressure will be detrimental to the steam cooling. If RPV pressure is rising, the assumptions of the MZIWL calculation are no longer valid and the core may not be adequately cooled.

If an injection source is line up and running (injecting), the assumptions of the MZIWL calculation are no longer valid and the core may not be adequately cooled. RPV depressurization is required to maximize injection flow rate and cool the core by submergence.

L-.

C.

D.

Sys#

System Category KA Statement 295031 WA#

295031.~~2.04 WA Importance ~l.s Exam Level RO Reactor Low Water Level Ability to determine andlor interpret the following as they apply to REACTOR LOW WATER LEVEL Adequate core cooling References provided to Candidate None Technical

References:

EO-000-102 Question Source:

M w d Susquehanna, 2001 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

tlb SSES LOC I 9 NRC Exam 18 A failure to completely scram occurred on Unit I with Reactor Power at 15%.

- No high pressure systems are available.

- Reactor Water level is at -100" and slowly dropping with several rods not fully inserted into the core.

Why is Core Spray NOT a system listed in Table 15 of EO-I 00-1 13, LEVEUPOWER CONTROL for use to control water level during a a failure to scram t ansient?

A.

B.

injects inside the shroud cannot be throttled, not easy C.

provides indirect core cooling

\\,'

D.

required time to line-up wpath.

p Question Data Answer: B injects inside the shroud Explanation/Justification:

A.

B.

Correct answer, C.

D.

cold water injects inside shroud, cause power changes.

not during an A W S situation cold water on fuel largest concern.

L-S y s #

System Category 295037 SCRAM Condition Present Knowledge of the interrelations between SCRAM and Reactor Power Above APRM Downscale or Unknown CONDITION PRESENT AND REACTOR POWER Exam Level Question Source:

Modified Level Of Difficulty: (1-5)

Question Cognitive Level:

10 CFR Part 55 Content:

NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC I 9 NRC Exam EO-1 00-1 05, "Radioactivity Release Control", directs isolation of all primary systems discharging into areas outside Primary Containment or Reactor Building except those systems required to support EOP/DSP actions.

These systems are specifically exempted from isolation because:

1. $W, these additional isolations would require an unnecessarily e classification.

they are required to support alternate reactor depressurization methods.

B.

CQ!*

cquJ( M C.

additional&tkkr elea B es from them are unlikely.

D.

isolation may ultimately result in a much larger uncontroied radiologiBal release.

Question Data Answer: D isolation may ultimately result in a much larger uncontrolled radiological release.

ExplanationlJustification:

A.

B.

nottrue C.

D.

correct answer not a consideration for these conditions alternate depress methods are part of EOPs

-u Sys #

System Category KA Statement 295038 High OffSite Release Rate Knowledge of the reasons for the following System isolations responses as they apply to HIGH OFF-SITE RELEASE RATE:

WA#

295038.~~3.02 WA Importance 3.9 Exam Level RO References provided to Candidate None Question Source:

Modified Susquehanna, 1999 Question Cognitive Level:

F u n d a m e n i 55.41 V

NRC 2003 Rev I H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam v

20 The T-20 Transformer developed an internal fault causing a fire and automatic actuation of the fire protection deluge system. The Fire Protection System functions as designed. For these pressure drops to:

conditions the Diesel Driven Fire pump will automatically A.

105 psig B.

95 psig C.

85 psig.

D.

125 psig I

Question Data Answer: C 85 psig.

ExplanationlJustification:

A.

6.

C.

correct answer, D.

The auto start of the Jockey Fire Pump The auto start of the Motor Driven Fire pump The auto shutdown of the Jockey Fire Pump Sys#

System Category KA Statement

-1_

/

WA#

6ooooo.2.1.31 WA Importance Exam Level RO Question Question Cognitive Source: Level:

New f12 Memory References provided to Candid e Technical

References:

TM-OP-013 Level Of Difficulty: (1-5) 2 10 CFR Part 55 Content:

55.41 NRC 2003 Rev I H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam 21 Unit 2 is starting up with power at 50%.

The third Reactor Feedpump has just been placed in service.

A Reactor Recirc system runback to 30% has occurred. The runback is due to the failure of the Feedwater Level Control System total feed flow signal.

What is the expected plant response with no operator actions?

A.

6.

C.

D.

Reactor Recirc Pump Trip signal High Water RPV level alarm Feedwater level control system Setpoint Setdown signal Low Reactor Water Level Alarm Question Data Answer: B High Water RPV level alarm ExplanationlJustification:

A.

B.

No trip signal to recirc pumps on high level.

Correct answer, Recirc runback caused by failure of feedwater flow signal to less than 20% flow, FWLC system sees a steam feed mismatch with feed flow low as compared to steam flow, W L C will increase feed flow to cancel SteamlFeed mismatch UNTIL level signal error over-rides steam feed mismatch thus, vessel level will increase.

Level will fail high not low.

Level will fail high not low.

C.

D.

Sys#

System Category KA Statement 295008 High Reactor Water Level Knowledge of the operational implications of the Feed flowkteam flow following concepts as they apply to HIGH REACTOR WATER LEVEL:

mismatch K/A#

295008.~~1.03 KIA Importance.2 Exam Level RO References provided to Candidate None Technical

References:

TM-OP-045 Question Source:

Modified Lasalle 1, 1996 Level Of Difficulty: (1-5) 4 Question Cognitive Level:

Analysis I O CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam c

22 Unit 1 is at 100% power in a normal lineup when annunciator RX WATER HI-LO LEVEL AR-101 -B17 actuates.

The following indications are observed at 1 C651:

- The Master Feed Water Level Controller LIC-C32-1 R600 is in AUTO and set at 34.

- I'A' Reactor Feed Pump speed controller SIC-C32-1 R601A demand signal is slowly lowering.

- 1 'B' Reactor Feed Pump speed controller SIC-C32-1 R601 B demand signal is slowly rising.

- 1 'C' Reactor Feed Pump speed controller SIC-C32-1 R601 C demand signal is slowly rising.

- The failed RFP A speed controller is placed in MANUAL.

Which of the following is the CORRECT, operator response for this situation?

A.

adjust feed flow to equal B & C Reactor Feedpumps, or, lower Motor Speed Changer, activate Hydraulic Jack, control speed to control flow.

B.

adjust feed flow to equal B & C Reactor Feedpumps, or, activate Hydraulic Jack, lower Motor Speed Changer, control speed to control flow.

L.

C.

adjust Master Feed Water Level Controller to maintain RPV water level -35", dedicate B & C Reactor Feedpumps, or, activate Hydraulic Jack, lower Question Data Answer: A adjust feed flow to equal B & C Reactor Feedpumps, or, lower Motor Speed Changer, activate Hydraulic Jack, control speed to control flow.

Explanation/Justification:

A.

correct answer, B.

C.

D.

Lower MSC first then set Hydraulic Jack.

'A' RFP controller has failed, attempting to control level with master controller will not work.

Lower MSC first then set Hydraulic Jack.

Sys#

System Category KA Statement 295009 WA#

295009.~~1.02 WA Importance 3.0 Exam Level RO Low Reactor Water Level Ability to operate andlor monitor the following as they apply to LOW REACTOR WATER LEVEL Reactor water level control References provided to Candidate None Technical

References:

ON-145401 Question Source:

Modified Duane Arnold 1,1999 Level Of Difficulty: (1 -5) 4 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC I 9 NRC Exam

'b-23 A Reactor scram resulted in water level dropping t 0 ~ 3 1 inches on Wide Range Level.

- Reactor level has since recovered to + I O inches on Narrow Range.

- Reactor Pressure is being maintained with the Turbine Bypass Valves.

- The maximum Reactor Pressure during the transient was 1080 psig.

- The operators have placed the 'A' RHR pump in Suppression Pool Cooling.

- Suppression Pool Water Temperature is rising.

Which of the following is the cause of rising Suppression Pool water temperature?

A.

RClC in operation, F048A HX A SHELL SIDE BYPS open, F047A RHR HX A SHELL SIDE INLET closed.

B.

RClC and HPCl in operation, F048A HX A SHELL SIDE BYPS open, F047A RHR HX A SHELL SIDE INLET closed C.

HPCl in operation, F048A HX A SHELL SIDE BYPS open, F047A RHR HX A SHELL SIDE INLET open D.

SRV actuation, F048A HX A SHELL SIDE BYPS open, F047A RHR HX A SHELL SIDE INLET closed L--

Question Data Answer: A RClC in operation, F048A HX A SHELL SIDE BYPS open, F047A RHR HX A SHELL SIDE INLET closed.

ExplanationlJustification:

A.

6.

C.

D.

correct answer, some cooling being provided with some flow forced through heat exchanger, heat input from RClC auto start at -30, level not low enough to auto start HPCI, pressure not high enough to open SRV.

HPCl did not receive a auto start signal.

Pressure did not get high enough to open SRV.

Pressure did not get high enough to open SRV.

Sys#

System Category KA Statement 295013 High Suppression Pool Ability to operate andlor monitor the following as Suppression pool cooling they apply to HIGH SUPPRESSION POOL TEMPERATURE:

Temperature RO WA#

295013.~~1.01 WA Importance 3.9 Exam Level Question Source:

Modified Grand Gulf 1,2000 Level Of Difficulty: (1-5) 3 References provided to Candidate None Technical

References:

OP-149-005 Question Cognitive Level:

Comprehension 10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAFom.doc Printed on 06/19/03

SSES LOC 19 NRC Exam

--.,/'

24 Given the following on Unit 1 :

- A reactowram signal has o c c m

- The full core display has several full-in lights NOT displayed

- The Four Rod Display is unavailable

- Computer rod position data is not available You are directed to verify the scram and you are aware of industry experiences where several light bulbs on the full core display have failed concurrently.

Which of the following indications would cause you to conclude that the scram may--

W s s f u l ?

4 -

A.

B.

All LPRM "downscale" lights are illu&ated

'f Reactor period stable at -80 seconds on C.

D.

All APRM "downscale" lights are not illuminated Div I and Div II RPS trip annunciators are lit on Question Data L-Answer: C All APRM "downscale" lights are not illuminated ExplanatiotdJustification:

A.

B.

C.

Correct answer D.

Expected downscale indication on scram Expected negative period on scram Expected indication on a scram f

Sys#

System Category 295015 Incomplete SCRAM Ability to determine and/or interpret the following as Control rod position they apply to INCOMPLETE SCRAM:

WA#

295015.~~2.02 WA Importance 4.1 Exam Level References provided to Candidate None Technical

References:

ON-100-101 Question Source:

Modified Hope Creek Unit 1,1998 Question Cognitive Level:

Comprehension 10 CFR Part 55 Content:

RO Level Of Difficulty: (1-5) 3 NRC 2003 Rev I H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam 25 The following statement exists in EO-I 00-1 05, RADIOACTIVITY RELEASE CONTROL:

I' IF TURBINE OR RW BUILDING HVAC NOT IN SERVICE RESTART APPLICABLE HVAC AS REQ'D" Which of the following is the basis for keeping the Turbine Building Ventilation System in operation while executing EO-I 00-1 05, RADIOACTIVITY RELEASE CONTROL?

Having Turbine Building Ventilation in service:

A.

maintains Turbine Building pressure above Reactor Building Pressure.

B.

prevents having an unmonitored ground release from the Turbine Building.

C. prevents a reactor scram due to high temperature in the MSL tunnel.

D.

ensures adequate dilution of the gases discharged through the Turbine Building Vent.

~

~

~~

~

~

Question Data Answer: B prevents having an unmonitored ground release from the Turbine Building.

Explanation/Justification:

A.

B.

C.

D.

Turbine Building pressure has no corelation to Reactor Building pressure correct answer, operation with no ventilation in service will lead to an unmonitored ground level release the EOP is not addressing high temperature in the steam tunnel.

the EOP is not addressing dilution of releases from the ventilation system

~ _ _ _ _ _ _ _

_ _ _ _ _ ~

~~

~

Sys#

System

~

category KA Statement 295017 High OffSite Release Rate Knowledge of the operational implications of the following concepts as they apply to HIGH OFFSITE RELEASE RATE:

KIA#

295017.~~1.02 KIA Importance 3.8 Exam Level RO Protection of the general public References provided to Candidate None Technical

References:

~0-000-io5 Question Source:

Modified Nine Mile Point 1,1998 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC I 9 NRC Exam 26 Suppression Pool Cleanup is in operation discharging to Rad Waste to lower the level of the Unit 1 Suppression Pool.

An inadvertant trip of the Division I logic for High Drywell Pressure is actuated and confirmed by alarms received in the Control Room.

What would be the trend of the Rad Waste Collection Tank level indicator as a result of the High Drywell Pressure?

A.

LRW Collection Tank Level trend is rising, Suppression Pool Level is lowering, Suppression Pool Cleanup pump tripped and SUPP POOL WTR FlLT PP SUCT IB IS0 HV-15766 and OB IS0 HV-I 5768 are open.

B.

LRW Collection Tank Level trend is level, Suppression Pool Level is constant, Suppression Pool Cleanup pump tripped and SUPP POOL WTR FlLT PP SUCT IB IS0 HV-15766 is closed.

C.

LRW Collection Tank Level trend is level, Suppression Pool Level is constant, Suppression Pool Cleanup pump running and SUPP POOL WTR FlLT PP SUCT OB IS0 HV-15768 is open.

D.

LRW Collection Tank Level trend is level, Suppression Pool Level is constant, Suppression Pool Cleanup pump Running and SUPP POOL WTR FlLT PP SUCT IB IS0 HV-15766 and OB IS0 HV-15768 are closed.

Question Data Answer: 8 LRW Collection Tank Level trend is level, Suppression Pool Level is constant, Suppression Pool Cleanup pump tripped and SUPP POOLWTR FlLT PP SUCT IB IS0 HV-15766 is closed.

ExplanationlJustification:

A.

6.

C.

D.

Suction valve would close stopping transfer to LRW correct answer, Pump trips and IB suction closes on Div I LOCA signal, stopping transfer to LRW.

Pump would trip and LRW tank level constant Pump trips, on Div I only IB valve closes.

Sys#

System Category KA Statement 295020 Inadvertent Containment Knowledge of the reasons for the following Suppression pool water level Isolation responses as they apply to INADVERTENT response CONTAINMENT ISOLATION:

References provided to Candidate None Technical

References:

ON-159-002 WA#

295020.~~3.06 WA Importance 3.3 Exam Level RO Question Source:

Modified Limerick I, 1995 Level Of Difficulty: (1-5) 4 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam

.~.--

27 Unit 2 is operating at 90% power.

- The 2B CRD pump in service and the in-service Flow Control Valve has failed closed.

Which of the following are the expected alarms for these conditions?

A.

AR-204-HO5 ROD DRIFT AR-207-COI CRD PUMP SUCTION FILTER HIGH DIFF PRESS B.

AR-202-F03 RECIRC PUMP A SEAL CLG WATER LO FLOW AR-202-FO6 RECIRC PUMP B SEAL CLG WATER LO FLOW C.

AR-203-HO5 CRD PANEL IC007 HI TEMP AR-207-AOI, CRD CHARGING WATER HI PRESS D.

AR-207-C02 CRD PUMPS DRIVE WATER FLTR HI DlFF PRESS AR-207-EO2 CRD PUMP B MOTOR OVERLOAD Question Data Answer: C AR-203-H05 CRD PANEL IC007 HI TEMP AR-207-A01. CRD CHARGING WATER HI PRESS ExplanatiodJustification:

A.

B.

C.

D.

Flow control valve could cause rod drift if failed open, suction DP due tio high flow or strainer clogged.

Seal cooling alarm is RBCCW flow.

Correct answer, Flow control valve or Drive Pressure Control valve closed will cause alarm and cause decreased cooling water flow.

Hi DP caused by high flow or filter clogged, CRD pump will have reduced load and not trip.

-.4 Sys#

System Category KA Sktemeat 295022 Loss of CRD Pumps WA#

295022.~~2.03 Exam Level RO Ability to determine andlor interpret the following as they apply to LOSS OF CRD PUMPS:

C 6

mechanism temp*

1 References provided to Technical

References:

AR-207-001 Question Source:

New Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Comprehension 10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

L-28 Unit 1 is operating at 75% power

-Valve stroke time testing is in progress on the "A" RHR Pump Suppression Pool Suction Valve HVI 51 F004A

-HV151 F004A is currently closed

-All other RHR system components are in their normal standby lineup

-A steam break causes drywell pressure to reach 2.0 psig Which of the following describes the response of the F004A valve and the "A" RHR pump?

A.

The HVI 51 F004A valve automatically opens and the "A" RHR pump will run after the HVI 51 F004A is fully open.

B.

The HV151 F004A valve must be opened and the "A" RHR pump must be manually started after HVI 51 F004A is fully open.

C. The HVI 51 F004A valve automatically opens but the "A" RHR pump must be started by the operator after HVI 51 F004A is fully open.

D.

The HVI 51 F004A valve must be opened and the,"A RHR pump will auto start after HV151 F004A is fully open.

L.

Question Data Answer: B The HV151 F004A valve must be opened and the " A RHR pump must be manually started after HV151 F004A is fully open.

ExplanationlJustification:

A.

B.

C.

No auto action associated with the 04 valve.

correct answer, valve must be manually opened and pump may be started manually. Pump will not auto start since there is no low level or low reactor pressure No auto action associated with the 04 valve.

suction head (interlock suction valve open): PlantSpecific Sys#

System 203000 RHWLPCI: Injection Mode Adequate pump net Positive (Plant Specific)

WA#

2o3000.~4.06 WA Importance 3.5 Exam Level RO References provided to Candidate None Technical

References:

TM-OP-049 Question Source:

Modified Hope Creek Unit 1,1998 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis /

10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 0611 9/03

SSES LOC 19 NRC Exam 29 The following conditions exist during refueling operations:

- The Reactor Mode Switch is in REFUEL

- Irradiated fuel is in the reactor

- The reactor vessel head has been removed

- The Fuel Pool gates are installed

- Reactor water level is 18" above the reactor vessel flange Which of the following defines the operability status required for shutdown cooling?

A.

One shutdown cooling subsystem is required to be operable and in service. The other subsystem is NOT required to be operable.

B.

Shutdown cooling is NOT required to be operable with RWCU in service.

C.

Two subsystems of shutdown cooling are required to be operable and one subsystem is required to be in service.

D.

Two subsystems of shutdown cooling are required to be operable and in service.

Answer: C Question Data Two subsystems of shutdown cooling are required to be operable and one subsystem is required to be ExplanationlJustification:

A.

6.

C.

correct D.

both shutdown cooling subsystems are required to be operable shutdown cooling is required to be operable one subsystem is required to be in operation NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC I 9 NRC Exam

-1 30 Unit 1 is in cold shutdown with the "A" RHR pump in the shutdown cooling mode of operation.

Reactor water level decreases to the LPCl system initiation setpoint.

What is the expected plant response assuming no operator action five minutes after the LPCl system initiation signal was generated ?

A.

B.

C.

D.

RHR loop suction and discharge valves HV-151-FO08 SHUTDOWN CLG SUCT OB ISO, HV-151-FO09 The "A" 81 "C" RHR are closed.

The "B" and "D" HV-151-FO15A RHR INJ OB IS0 J

open/closed).

into the reactor vessel.

RHR loop suction and discharge valves, HV-151-F008 SHUTDOWN CLG SUCT OB are closed.

The "A", "B, "c" and "D" RHR pumps are injecting into the reactor vessel.

ISO, HV-151-FOO9 SHUTDOWN CLG SUCT IB ISO, HV-151-FO15A RHR INJ OB IS0 RHR loop suction and discharge valves HV-151-F008 SHUTDOWN CLG SUCT OB remain open.

The "A" RHR pump remains in shutdown cooling mode of operation.

The "B", "C" and "D" RHR pumps are injecting into the reactor vessel.

ISO, HV-151-F009 SHUTDOWN CLG SUCT IB ISO, HV-151-FO15A RHR INJ OB IS0 RHR loop suction and discharge valves HV-151-FO08 SHUTDOWN CLG SUCT OB remain open.

The "A" RHR pump remains in shutdown cooling mode of operation.

The "B", "C" and "D" RHR pumps are not injecting into the reactor vessel.

ISO, HV-151-FO09 SHUTDOWN CLG SUCT IB ISO, HV-151-FO15A RHR INJ OB IS0 Question Data Answer: A RHR loop suction and discharge val SUCT IB ISO, HV-151-FO15A RHR INJ OB IS0 are closed.

The "A" & "C" RHR pump breakers are pumping (cycling openlclosed).

The " B and " D RHR pumps are run ExplanationlJustification:

A.

correct answer B.

C.

D.

no injection occurs without operator intervention F008, FOO9, and F015 auto close below +13" reactor wate F008, FOO9, and F015 auto close below +13" reactor water level. -

Sys#

System Category 205000 Shutdown Cooling System (RHR Shutdown Cooling Mode)

Ability to monitor automatic operations of the SHUTDOWN COOLING SYSTEMlMODE including:

KIA#

205000.~3.02 KIA Importance 3.2 Exam Level References provided to Candidate None Technical Question Source:

Modified Limerick 2,1995 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAFom.doc Printed on 0611 9/03

SSES LOC I 9 NRC Exam 31 During a LOCA, HPCl automatically initiated, then tripped. The operator notes the following indications:

- Turbine Stop Valve Closed

- HPCl Turbine RPM Zero

- HPCl TURB TRIPPED Alarm Sealed In.

- HPCl TURB TRIP SOLENOID ENERG Has NOT alarmed What caused the HPCl turbine to trip?

A.

loss of oil pressure.

6.

high exhaust pressure.

C.

high reactor water level.

D.

low steam supply pressure.

Question Data Answer: A loss of oil pressure.

ExplanationlJ ustification:

A.

B.

C.

D.

Correct answer, Turb Tripped alarm (AR-114-A01) is from valve position, trip solenoid alarm is direct turbine trip, all other distracters are direct turb trips.

Direct turb trip would cause trip solenoid alarm Direct turb trip would cause trip solenoid alarm Direct turb trip would cause trip solenoid alarm

..d Sys#

System Category KA Statement KIA#

2o600o.2.4.10 WA Importance 3.0 Exam Level RO References provided to Candidate None Technical

References:

AR-114-001 Question Source:

Modified Brunswick, 1995 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC I 9 NRC Exam 32 The Core Spray Loop "A" initiation logic channel has experienced a loss of power from 125 VDC Class 1 E Bus A (1 D614).

Which group of alarms from IC601 Alarm Panel (AR-109-001) and BIS Display for Core Spray (AR-153-001) would be the expected response to the loss of the ID614 power supply to the Core Spray Logic?

A.

LOOP A OL OR POWER LOSS (AR-153-AOI)

LOOP A OUT OF SERVICE (AR-I 53-B01)

LOOP A OL OR POWER LOSS (AR-153-AOI)

LOOP A RELAY LGC PWR LOSS (AR-153-AO3)

LOOP A RELAY LGC PWR LOSS (AR-153-A03)

B.

C.

CORE SPRAY LOOP A OUT OF SERVICE (AR-IO~-BOZ)

CORE SPRAY LOOP A OUT OF SERVICE (AR-I 09-802)

D.

LOOP A OUT OF SERVICE (AR-153-B01)

Question Data Answer: C LOOP A RELAY LGC PWR LOSS (AR-153-A03)

CORE SPRAY LOOP A OUT OF SERVICE (AR-109-602) v ExplanationlJustification:

A.

Loss of l2OVAC control power to any of following valves due to loss of associated power source or breaker racked Thermal overload on any of above valves with CORE SPRAY LOOP A MOV OL BYPS HS-E21-1S12A in TEST.

74 Relay failure on any of above valves.

Loop A out of service caused by the handswitch on the vertical panel not a loss of logic power.

LOOP A OL OR POWER LOSS (AOI)

Loss of 12OVAC control power to any of following valves due to loss of associated power source or breaker racked out.

Thermal overload on any of above valves with CORE SPRAY LOOP A MOV OL BYPS HS-E21-1S12A in TEST.

74 Relay failure on any of above valves.

Loop A out of service caused by the handswitch on the vertical panel not a loss of logic power.

B.

C.

correct answer, D.

Sys#

System Category KA Statement 209001 Low Pressure Core Spray Knowledge of electrical power supplies to the initiation logic WA#

209001.~2.03 Exam Level RO System following:

Technical

References:

AR-153-001, AR-109-001 Level Of Difficulty: (1-5) 2 Memo 10 CFR Part 55 Content:

55.41 References provided to Candidate Question Source:

New Question Cognitive Level:

NRC 2003 Rev I H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam 33 How does the Automatic Depressurization System (ADS) logic determine that the Core Spray System is available prior to depressurizing the reactor?

A.

Both Core Spray Pumps in one loop are operating at normal discharge pressure.

B.

At least one Core Spray Pump breaker is closed and the associated loop minimum flow valve is closed.

C.

Both Core Spray loop flow rates must be greater than 2000 gpm with the associated minimum flow valve closed.

D.

At least one of the two Core Spray Pumps in one loop is operating at normal discharge pressure.

Question Data Answer: A Both Core Spray Pumps in one loop are operating at normal discharge pressure.

ExplanatiodJustification:

A.

Correct answer B.

C.

D.

Breaker closure is not a permissive.

Flow is not a permissive Both pumps to be running with normal discharge pressure.

?---

KA Statement Sys#

System Category 209001 Low Pressure Core Spray System Knowledge of the effect that a loss or malfunction of the LOW PRESSURE CORE SPRAY SYSTEM will have on following:

ADS logic WA##

2o9001.~3.02 WA Importance 3.8 Exam Level RO References provided to Candidate None Technical

References:

TM-OP-83E Question Source:

Modified FITZPATRICK, 1993 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam c-34 An AlWS transient occurs and boron must be injected with the Standby Liquid Control System (SLC).

Which of the following describes the effect of a successful initiation of SLC on these indications:

1) Squib Valve Ready (Continuity) lights

2) Alarm on Panel 1 C601, SBLC SQUIB VALVES LOSS OF CKT CONTINUITY AR-107-A03

3) Pump discharge pressure A.

B.

C.

D.

I ) Illuminated

2) Annunciator in Alarm

3) 200 psig greater than reactor pressure

1) Extinguished

2) Annunciator in Alarm

3) 200 psig greater than reactor pressure

1) Illuminated

2) Annunciator not in Alarm.

3) Just above reactor pressure

1) Extinguished

2) Annunciator not in Alarm

3) Just above reactor pressure Question Data Answer: B

1) Extinguished

2) Annunciator in Alarm

3) 200 psig greater than reactor pressure ExplanationlJustification:

A.

Continuity lights go out.

B.

C.

D.

correct answer, Squib valve tires, causing loss of continuity, alarm annunciates indicating loss of continuity, pumps start with discharge pressure slightly greater than reactor pressure.

Continuity lights go out. Alarm is annunciated, discharge pressure 200 psig reactore pressure.

Alarm is annunciated, discharge pressure 200 psig above reactor pressure.

Sys#

System Category KA Statement 21 1000 Standby Liquid Control WA#

211000.~4.08 WA Importance &z Exam Level RO Ability to manually operate andlor monitor in the System initiation: Plant-Specific System control room:

References provided to Technical

References:

OP-I 53-001, TM-OP-053 Question Source:

Level Of Difficulty: (1-5) 3 Question Cognitive Level:

10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 0611 9/03

SSES LOC 19 NRC Exam

---/

35 Given the following conditions:

- Unit 2 has experienced a failure-to-scram (ATWS)

- The Standby Liquid Control (SLC) system was initiated and injected for 52 minutes. at normal flowrate before both SLC Pumps failed

- Reactor power is in the Source Range How does this failure affect the planned reactor cooldown and depressurization?

A.

Reactor Engineering must make the determination if current boron concentration will

/

allow a complete cooldown.

Cooldown can be accomplished if completed before Xenon decays out of the core.

Reactor boron concentration is sufficient to allow a complete cooldown with a maximum of 8 control rods not fully inserted.

Reactor boron concentration is sufficient to allow a complete cooldown under any pla conditions.

B.

C.

D.

Question Data Answer: D ExplanatiorVJustification:

A.

not required.

B.

C.

D.

Reactor boron concentration is sufficient to allow a complete cooldown under any plant conditions.

\\-

CSBW will account for Xe decay as well.

CSBW will handle any number of rods out correct answer, 2 pumps at TS minimum of 41.2 gpm for 52 minutes is 4284 gallons, greater than the CSBW of 4191 gallons Sys#

System Category KA Statement K/A#

211000.2.1.7 Exam Level RO References provided to Ca Question SOUrCe:

Modifi Level Of Difficulty: (1-5)

Question Cognitive Level:

Technical

References:

E0-000-IF

(,

10 CFR Part 55 Content:

c NRC 2003 Rev 1 H:\\NRCExarnPrep\\Rich\\NRCAForrn.doc Printed on 06/19/03

SSES LOC I 9 NRC Exam 36 During plant shutdown, which of the following Reactor Protection System automatic scrams is bypassed by taking the Mode Switch from RUN to STARTUP ?

A.

B.

C.

D.

Turbine Stop Valve Closure.

Scram Discharge Volume Level High.

Main Steam Isolation Valve closure Turbine Control Valve Fast Closure.

Question Data Answer: B Main Steam Isolation Valve closure ExplanatiodJustification:

A.

not bypassed in startup

6. correct answer C.

D.

bypassed by turbine 1st stage pressure bypassed by turbine 1st stage pressure Sys#

System Category KA Statement 212000 Reactor Protection Knowledge of the physical connections andlor cause-Main steam system System effect relationships between REACTOR PROTECTION

'V SYSTEM and the following:

WA# 212000.K1.14 WA Importance

~6 Exam Level RO nJ References provided to Candidate None Technical

References:

TM-OP-58 Question Source:

Modified Nine Mile Point 1,1996 Level Of Difficulty: (1-5) 9-,

Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

55.41 NRC 2003 Rev I H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC I 9 NRC Exam 37 With the plant is operating at 80% power, a loss of u

Which of the below are expected alarms?

A.

RPS CHANNEL B1/B2 AUTO SCRAM (AR-104-A01), NEUTRON MON CHAN B 1 13-BO2 )

SYSTEM TRIP (AR-104-A04), and CORE SPRAY LOOP B OUT OF SERVICE ( AR-B.

RPS CHANNEL AI/A2 AUTO SCRAM (AR-103-A01), NEUTRON MON CHAN A 1 13-BO2 )

SYSTEM TRIP (AR-103-A04), and CORE SPRAY LOOP B OUT OF SERVICE ( AR-C.

RPS CHANNEL AI/A2 AUTO SCRAM (AR-103-A01), NEUTRON MON CHAN A 109-BO2 )

SYSTEM TRIP (AR-103-A04), and CORE SPRAY LOOP A OUT OF SERVICE ( AR-D.

RPS CHANNEL B1/B2 AUTO SCRAM (AR-104-A01), NEUTRON MON CHAN B SYSTEM TRIP (AR-104-A04)

Question Data Answer: C RPS CHANNEL AI/A2 AUTO SCRAM (AR-103-A01), NEUTRON MON CHAN A SYSTEM TRIP (AR-103-AO4). and CORE SPRAY LOOP A OUT OF SERVICE ( AR-109-802 )

ExplanationlJustification:

A.

Wrong division B.

C.

correct answer, D. Wrong division d

Core spray valves F004 and 05 powered from 18217

/

Sys#

System Category KA Statement 212000 Reactor Protection Knowledge of electrical power supplies to the RPS motorgenerator sets WA#

212000.~2.01 KIA Importance 3.2 Exam Level RO System following:

Technical

References:

TM-OP-058 Level Of Difficulty: (1-5) 3 References provided to Candidate Question Source:

New Question Cognitive Level:

Compr ension F 4 10 CFR Part 55 Content:

55.41 NRC 2003 Rev I H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC I 9 NRC Exam 38 A reactor startup is in progress.

- Power is on Range 2 of the IRM's.

- The " A IRM increases to 124/125 of scale and the 'IF" IRM increases to 123/125 of scale.

A Division I half scram occurs.

Which of the following describes the unit RO expected actions?

A.

B.

C.

D.

Range up on "A" IRM and reset the Division I side half scram, continue the startup Range up on "F" IRM and continue the startup Place the reactor Mode Switch in Shutdown.

Range up on "A" and "F" IRM, reset the Division I side half scram continue the startup Question Data Answer: C Place the reactor Mode Switch in Shutdown.

ExplanationlJustification:

A.

6.

C.

correct D.

a full scram should have occurred for these conditions

'F' IRM already above trip setpoint a full scram should have occurred for these conditions L

Sys#

System Category KA Statement 215003 Intermediate Range Ability to monitor automatic operations of the RPS status Monitor (IRM) System INTERMEDIATE RANGE MONITOR (IRM) SYSTEM including:

WA#

215003.~3.03 WA Importance 3.7 Exam Level RO References provided to Candidate None Technical

References:

OP-AD-001 Question Source:

Modified M+;ii;ick 1,1997 Level Of Difficulty: (1 -5) 3 Question Cognitive Level:

10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam b

39 Which of the following describes how a gradual DECREASE in Argon fill gas pressure will affect the response of a Source Range Monitoring (SRM) detector?

/I A.

Gamma sensitivity would Decrease Neutron sensitivity would Decrease B.

Gamma sensitivity would NOT change Neutron sensitivity would NOT change C.

Gamma sensitivity would Increase Neutron sensitivity would NOT change D.

Gamma sensitivity would NOT change Neutron sensitivity would Decrease Question Data Answer: A Gamma sensitivity would Decrease Neutron sensitivity would Decrease ExplanationlJustification:

A.

Correct answer B.

C.

D.

Fission causes recoil within the detector chamber and ionize the Argon gas, decrease in argon gas will reduce sensitivity##Gamma causes direct ionization of argon, decrease in argon gas will reduce sensitivity Fission causes recoil within the detector chamber and ionize the Argon gas, decrease in argon gas will reduce sensitivity.

Gamma causes direct ionization of argon, decrease in argon gas will reduce sensitivity

-\\--

Sys#

System Category KA Statement 215004 Source Range Monitor Knowledge of the operational implications of the Detector operation (SRM) System following concepts as they apply to SOURCE RANGE MONITOR (SRM) SYSTEM:

KIA#

215004.~5.0i KIA Importance 2.6 Exam Level RO References provided to Candidate None Technical

References:

TM-OP-078 Question Source:

Modified Hope Creek Unit 1.1996 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExarnPrep\\Rich\\NRCAForrn.doc Printed on 06/19/03

SSES LOC 19 NRC Exam 40

- A Unit 2 startup is in progress with power at 20%

- Recirculation flow is 30%

- The "A" APRM F l

o w

m o

u t

v a

s recirculation f1o-w is raised As the plant startup continues, what will be the FIRST protective action to occur and the reason for that action?

~

-\\ i\\

A.

B.

A half scram will occur due to a flow unit A control rod block will occur due to a flow biased neutron flux upscale C.

A control rod block will occur due to a flow unit comparator trip. '

D.

A full scram will occur due to a flow biased neutron flux upscale Question Data Answer: C A control rod block will occur due to a flow unit comparator trip.

ExplanatiodJustification:

A.

B.

does not occur first C.

D.

not a half scram signal Correct, trip set at 10% difference between units not the first event to occur.

\\--

Sys#

System Category KA Statement 21 5005 Average Power Range Monitor/Local Power Range Monitor System Knowledge of the effect that a loss or malfunction of the following will have on the APRMILPRM:

Flow converterlcomparator network: PlantSpecific WA#

215005.~6.07 WA Importance Q Exam Level RO References provided to Candidate None Technical

References:

TM-OP-078 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

55.41 Question Source:

Modified Peach Bottom, 1998 Level Of Difficulty: (1-5) 3 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

v 41 Unit 1 at 100% power.

125V DC PANEL 1L610 SYSTEM TROUBLE AR-106-Al2 annunciates, indicating a loss of power to Division I I D614.

How will a loss of Division I I D614 impact auto initiation and/or steam leak detection isolation of RCIC?

/--

A.

Auto initiation will not function if needed. Steam leak detection will not function if needed.

B.

'Auto initiation will function if needed. Steam leak detection will function if needed by closing steam supply header inboard (HV-149-F007) and warmup isolation (HV-149-F088) valves.

C.

Auto initiation will not function if needed. Steam leak detection'will function if needed by closing steam supply header inboard (HV-149-F007) and warmup isolation (HV-149-F088) valves.

D.

Auto initiation will not function if needed. Steam leak detection will function if needed by closing steam supply header outboard isolation valve (HV-149-FOO8).

Question Data Answer: C Auto initiation will not function if needed. Steam leak detection will function if needed by closing steam supply header inboard (HV-149-F007) and warmup isolation (HV-149-F088) valves.

ExplanatiodJustifiction:

A.

B.

C.

Correct answer.

D.

Div II of steam leak detection isolation will function, closing inboard isolation valves.

Auto initiation will not function.

Div I1 of steam leak detection isolation will function, not Div I.

~~~~

~

Sys#

System Category KA Statement 217000 Reactor Core Isolation Knowledge of electrical power supplies to the RCIC initiation signals (logic)

K/A#

217000.~2.02 WA Importance 2.8 Exam Level RO Cooling System (RCIC) following:

References provided to Candidate None Technical

References:

TM-OP-050 Question Source:

New Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

55.41 NRC 2003 Rev I H:\\NRCExamPrep\\Rich\\NRCAFom.doc Printed on 06/19/03

SSES LOC 19 NRC Exam L'

42 The following plant conditions exist:

- A reactor transient is in progress.

- 'A' Core Spray pump in operation

- Drywell pressure

- Reactor water level The following alarm conditions are noted:

- 'B' loop of RHR is in operation.

_ / - - -

RX LO LEVEL SIGNAL A CONFIRMED AR-1 10-BO1 - Alarm window Not lit RX LO LEVEL SIGNAL B CONFIRMED AR-1 IO-BO3 - Alarm window Lit.

ADS LOGIC C & D TIMER INITIATED AR-1 IO-A03 & AR-1 IO-A04 alarms came in 40 seconds ago.

Which of the following states the response of t h m l a n t conditions with no operator action?

\\

A.

6 ADS valves will open in approximately 60 seconds from Div I & II logic.

B.

3 ADS valves will open in approximately 60 seconds from Div II logic.

C.

3 ADS valves will open in approximately 520 seconds.

D.

6 ADS valves will open in approximately 60 seconds from Div Ii logic.

Question Data Answer: D 6 ADS valves will open in approximately 60 seconds from Div I1 logic.

ExplanationlJustification:

A.

B.

C.

D.

Div I does not have +I 3 " confirmation signal nor Divl pp permissive Div II will initate all ADS valves.

Drywell pressure does not require to be bypassed with 7 minute timer.

Correct answer, Div I does not have +I3 " confirmation signh,

~~

Sys#

System

~

Category J

/ KA Statement 218000 Automatic Knowledge of the effect that a loss or mglfunction of Nuclear boiler instrument the following will have on the AUTOMATIC DEPRESSURIZATION SYSTEM:

Depressurization System system (level indication)

WA#

218o00.~6.03 WA Importance 3.8 Exam Level RO References provided to Candidate None Technical

References:

TM-OP-83 Question Source:

Modified Hope Creek Unit f, 1995 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC I 9 NRC Exam

'v 43 Which of the statements below is correct regarding the arming and de-pressing of the four (4)

NSSSS pushbuttons on panel 1C651.

A.

B.

C.

D.

"A" and "CY pushbuttons will cause a full MSIV isolation.

"A" and "BI pushbuttons will cause a full inboard isolation including all 8 MSIVs.

"B" and "c" pushbuttons will cause an inboard MSlV isolation.

"A" and "Do pushbuttons will cause an outboard MSIV isolation.

Question Data Answer: B

" A and " B pushbuttons will cause a full inboard isolation including all 8 MSIVs.

ExplanationNustification:

A.

cause no isolation B.

correct answer C.

D.

'B & 'C' button causes an inboard isolation, inboard and outboard MSlV isolation

'A' & 'D' causes a full inboard and outboard isolation Sys#

System Category KA Statement 223002 Primary Containment Knowledge of PClSlNSSSS design feature(s) andlor Manual initiation capability:

Isolation SystemlNuclear interlocks which provide for the following:

PlantSpecific Y

Steam Supply Shut-Off KIA#

223002.~4.03 KIA Importance 3.5 Exam Level RO References provided to Candidate Technical

References:

TM-OP-059 Question Source:

M o d i f i e d 0 USQUEHANNA, ne

~ - 1989 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Memory 10 CFR Part 55 Content:

55.41 NRC 2003 Rev I H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 0611 9/03

SSES LOC 19 NRC Exam 44 With the reactor at 100% power, which of the following would be an indication of an open Safety Relief Valve (SRV)?

A.

Total indicated steam flow increase.

B.

Indicated Feed flow less than Steam f l

o r

u" C.

SRV Tailpipe temperature stable at 500,deg F D.

Reactor thermal power increase.

Question Data Answer: D Reactor thermal power increase.

ExplanationlJustification:

A.

B.

C.

D.

Indicated total steam flow will decrease.

Feed flow will indicate greater than steam flow.

Tail pipe temperature will peak at approximately 320 degrees.

correct answer, power will increase slightly due to reduced feed heating due to loss of extraction steam heating.

Sys#

System Category KA Statement 239002 RelieflSafety Valves Ability to predict andlor monitor changes in Reactor power parameters associated with operating the RELIEFlSAFETY VALVES controls including:

WA#

239002.~1.06 WA Importance 3.7 Exam Level RO References provided to Candidate None Technical

References:

ON-183401 Question Source:

Modified Nine Mile Point 1,1996 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Comprehension 10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam L-45 Unit 1 is at 84% power with Feedwater Control in 'Average Level Control' and Three Element Control.

What would be the response if the Unit 1 "A" Narrow Range Level instrument,were equalized.(assume no operator action)

A.

All three feedpumps and the main turbine trip.

B.

RPV level will lower to approximately 23 inches and remain steady at the new lower level.

C.

Division I scram on low level, High level for Division II.

D.

RPV level will rise to approximately 47 inches and remain steady at the new higher level.

Question Data Answer: B RPV level will lower to approximately 23 inches and remain steady at the new lower level.

ExplanationlJustification:

A.

B.

D.

C.

signal only affected one instrument, need two out of three to trip turbines correct answer, zero delta p is a failure upscale.

RPV level will not lower to scram setpoint because of average level circuit the zero DP signal will simulate a high level signal and cause FW control lower level not contro L:

Sys#

System Category 3 -

J~ Statement Control System the REACTOR WATER LEVEL CONTROL SYSTEM; input 259002 Reactor Water Level Ability to (a) predict the impacts of the following on and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Loss of reactor water level W#

259002.~2.03 WA lmportanc Exam Level RO References p r o v i d g didate N\\=

Technical References :

TM-OP-045 Question Source:

Level Of Difficulty: (1 -5) 4 Question Cognitiv L el:

Analysis 10 CFR Part 55 Content:

55.41 c

NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam

,R 46 The following conditions exist:

- Unit 2 is operating at l m w e r.

- The Standby Gas Treatment System (SGTS) is in the standby lineup.

- A valid Unit 2 high drywell pressure initiation signal oqcurs.

Which of the following are the locations from which A.

U-2 drywell, Reactor Building Zone 2 and Reactor Building Zone 3.

B.

Reactor Building Zone 2 and Reactor Building Zone 3, Unit 2 HPCl Barometric Condenser.

C.

U-2 drywell and U-2 HPCl Barometric Condenser.

D.

Only the U-2 Reactor Building Zone 2 and Reactor Building Zone 3.

Question Data Answer: B Reactor Building Zone 2 and Reactor Building Zone 3, Unit 2 HPCl Barometric Condenser.

ExpIanatiodJustification:

A.

B.

correct answer C.

D.

U-2 Drywell does not align automatically U-2 Drywell does not align automatically HPCl Barametric Condenser exhausts to the SBGTS suction.

14 Sys#

System Category 261 000 Standby Gas Treatment Knowledge of the physical connections andlor caus System effect relationships between STANDBY GAS TREATMENT SYSTEM and the following:

K/A#

261000.K1.11 K/A Importance 3.2 Exam Level RO References provided to Candidate None Technical

References:

TM-OP-073 Question Source:

Modified Browns Ferry 1,2, & 3,1996 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

55.41 NRC 2003 Rev I H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC I 9 NRC Exam i---

47 Which of the following ESS 4 Kv bus feeder breaker trips will NOT auto transfer the bus to its alternate source?

A.

Incoming feeder overcurrent B.

ESS Transformer pressure relay.

D.

ESS Transformer differential current Question Data Answer: A Incoming feeder overcurrent ExplanationNustification:

A.

correct answer, B.

transformer lockout signal only C.

D.

transformer lockout signal only causes transfer to alternate source Sys#

System Category KA Statement 262001 AC. Electrical Distribution Knowledge of the effect that a loss or malfunction of the AC. ELECTRICAL DISTRIBUTION will have on Off-site power system i---

following:

WA#

262001.~3.05 KIA Importance g Exam Level RO References provided to Candidate None Technical

References:

TM-OP-004 Question Source:

Modified Peach Bottom, 1996 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam u

48 1E Instrument AC UPS Distribution w u p p l i e s power to 1Y12: and OY3Ol.

Which of the foll 1Y128 and OY3 the correct e ctrical configuration for 1 D130 to supply power to iately after (/"

the prefered source 18246 is lost?f 7

A.

1 B226, Alternate source stepped down to 12OVAC B.

Battery, 1 D133 inverted to 12OVAC C.

1 B226, Maintenance Backup, rectified to 250VDC, inverted to 12OVAC D.

lB226, Maintenance Backup, rectified to 120VDC, inverted to 12OVAC Question Data Answer: B Battery, 1D133 inverted to 12OVAC Explanation/Justification:

A.

B.

Battery is first alternate power supply.

Correct answer, battery, UPS being supplied from the dedicated battery can carry the distribution panels for approximately 20 minutes. When the external battery's output decreases to less than 210 VDC, the Static Transfer switch operates to supply alternate power to the distribution panels, and alternate supply is via 480-208/120 V step down transformers Maintenance backup is used for maintenance and is a manual transfer.

Maintenance backup is used for maintenance and is a manual transfer.

C.

D.

u Sys#

System Category KA Statement 262002 Uninterruptable Power Knowledge of UNINTERRUPTABLE POWER SUPPLY Transfer from preferred power Supply (AC./D.C.)

(AC.1D.C.) design feature@) and/or interlocks which provide for the following:

to alternate power supplies WA#

262002.~4.01 WA importance 3.1 Exam Level RO References provided to Candidate None Technical

References:

TM-OP-017 Question Source:

New Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam c

49 The Nuclear Plant Operator reports that both ground detection lights on a 250 V DC Battery Charger are dimmer than normal and that one light is brighter than the other.

What is the status of the 250 VDC bus?

A.

Unable to determine ground exists based on indications provided.

B.

A ground exists ONLY on the bus with the brighter light.

C.

A ground exists ONLY on the bus with the dimmer light.

D.

Grounds exist on both busses with the ground of the greater magnitude on the bus with the dimmer light.

Question Data Answer: D Grounds exist on both busses with the ground of the greater magnitude on the bus with the dimmer light.

ExplanationlJustification:

A.

B.

C.

D.

lights dim with grounds on both buses bright light occurs with single ground on the ungrounded bus dim light occurs with single ground Correct answer If one light burns bright and the other light is dim or out, then a single ground has occurred on the side with the dimly lit light. If grounds are present on both buses then the side with the dimly lit light has the grounds of greater magnitude.

Sys#

System Category KA Statement 263000 D.C. Electrical Distribution Ability to (a) predict the impacts of the following on Grounds the D.C. ELECTRICAL DISTRIBUTION; aprl (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

wA#

263000.A2.01 WA Importance 2.8 Exam Level RO References provided to Candidate None Technical

References:

TM-OP-002 Question Source:

New Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC I 9 NRC Exam LJ 50 During normal operation, the AC power supply to the 1 D663 Battery Charger is lost.

Which of the following describe the automatic actions for the loss of the AC power to this battery charger?

A.

The alternate battery charger immediately feeds the bus, 250V DC PANEL 1 L650 SYSTEM TROUBLE AR-105-AI 1 alarms on 1 C651, BATTERY CHARGER TROUBLE alarms on local alarm panel.

The battery charger is immediately supplied by alternate AC power, 250V DC PANEL 1 L650 SYSTEM TROUBLE AR-105-AI 1 alarms on 1 C651, BATTERY CHARGER TROUBLE alarms on local alarm panel.

B.

C.

The battery backfeeds the inverter and forward feeds the bus, 250V DC PANEL 1 L660 SYSTEM TROUBLE AR-105-B11 alarms on 1 C651, BATTERY CHARGER TROUBLE alarms on local alarm panel.

D.

The battery immediately feeds the 250 VDC bus, 250V DC PANEL 1 L660 SYSTEM TROUBLE AR-I 05-B11 alarms on 1 C651, BATTERY CHARGER TROUBLE alarms on 'IO=-.

/

M Question Data u

Answer: D The battery immediately feeds the 250 VDC bus, 250V DC PANEL 1 L660 SYSTEM TROUBLE AR-105-B11 alarms on 1 C651, BATERY CHARGER TROUBLE alarms on local alarm panel.

ExplanationlJustification:

A.

B.

C.

D.

No alternate battery charger on Div I I, only on Div I.

No alternate AC power supply for battery charger.

Battery feeds DC distribution bus directly.

correct answer, Control room alarm received from local alarm panel, local alrm panel alarm due to AC supply to battery charger lost.

_ _ _ _ _ _ _ _ ~

~

Sys#

System Category KA Statement 263000 D.C. Electrical Distribution Ability to monitor automatic operations of the D.C.

Meters, dials, recorders, ELECTRICAL DISTRIBUTION including:

K/A#

263000.~3.01 WA Importance 3.2 Exam Level RO alarms, and indicating lights References provided to Candidate None Technical

References:

AR-106-001 Question Source:

Modified Browns Ferry 2,2001 Level Of Difficulty: (1-5)

Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

\\--

NRC 2003 Rev I H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC I 9 NRC Exam 51 The A Diesel Generator (DG) is being run for the monthly operability check.

The PCOP has just closed the diesel output breaker to parallel the DG to the bus, when he notices the following:

- DG Frequency is 60 Hz

- DG Kilowatts is 5 Kw

- DG Kilovars is -4500 Kvar Attempting to lower Kvars on the DG, the operator takes the SPEED ADJUST switch to lower.

SELECT the DG response to this action:

A.

The DG will trip on @voltage.

B.

Frequency will decrease rapidly.

C.

D.

The DG will trip on reverse power.

The DG will slip a pole.

Question Data Answer: C ExplanationlJustification:

A.

trip on reverse power B.

C.

D.

The DG will trip on reverse power.

Frequency can't change with D/G tied to grid correct answer, D/G not loaded any decrease in fuel will reduce load further.

trip on reverse power not slip a pole Sys#

System Category KA Statement 264000 Emergency Generators Ability to predict andlor monitor changes in Maintaining minimum load on (DiesellJet) parameters associated with operating the emergency generator (to EMERGENCY GENERATORS (DIESEUJET) controls including:

prevent reverse power)

WA#

264000.~i.09 WP, Importance 3.0 Exam Level RO References provided to Candidate None Technical

References:

S0-024-001 Question Source:

New Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Comprehension 10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam c

52 Maintenance is requesting that both Containment Instrument Gas (CIG) compressors be removed from service due to a common mode failure potential. The plan requires that Instrument Air be are repaired.

What would be A.

Pressure

6. No repositioning.

Question Data Answer: C ExplanationNustification:

A.

No pressure for the 90# CIG header, Closure of Inboard MSIVs, loss of Drywell Cooling.

150# header storgae bottles can not feed the 90# header.

L-7 the effect that a loss or malfunction of ENT AIR SYSTEM) will have on the following:

wA#

300000.K3.01 WA Importance 2.7 Exam Level RO References provided to Candidate None Technical

References:

~M-oP-025 Question Source:

New Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExarnPrep\\Rich\\NRCAForrn.doc Printed on 06/19/03

SSES LOC I 9 NRC Exam L

53 Unit 1 is at 100% power when a pipe break occurs on the Unit 1 TBCCW idel& the in service

'LI What alarms and plant indications would be expected as a result of the pipe break?

A.

TBCCW HEAD TANK HI-LO LEVEL AR-123-GO6 TBCCW HX AREA FLOODED AR-123-HOl TBCCW Heat Exchanger Discharge Pressure indication (PI-I 4409) downscale Off gas precoolers Hi temperature B.

TBCCW PUMPS DISCHARGE HEADER LO PRESS GO3 Standby TBCCW pump running Instrument Air Compressors tripped TBCCW HEAT EXCHANGER HEADER LO PRESS AR-123-GO4 C.

TBCCW PUMPS DISCHARGE HEADER LO PRESS AR-123-GO3 TBCCW HEAD TANK HI-LO LEVEL AR-123-GO6 TBCCW HX AREA FLOODED (H01)

PASS Sample cooler Hi temperature D.

TBCCW HX AREA FLOODED AR-123-HOI TBCCW Heat Exchanger Discharge Pressure indication (PI-I 4409) downscale RFPT Lube Oil Cooler Hi temperature u

Standby TBCCW pump running Question Data Answer: 6 TBCCW PUMPS DISCHARGE HEADER LO PRESS GO3 TBCCW HEAT EXCHANGER HEADER LO PRESS AR-123-GO4 Standby TBCCW pump running Instrument Air Compressors tripped ExplanationlJustification:

A.

Offgas precooler RBCCW B.

Correct answer C.

Pass cooler RBCCW ',

D.

RFPT Lube Oil Cooler Service water 400000 Component Cooling Water System (CCWS) control room:

Ability to manually operate andlor monitor in th WA#

400000.A4.01 WA Importance References provided to Candidate None Technical

References:

Question Source:

New Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam 54 The reactor is at 100% power CRD PUMP 1 P I 32B 1s out of service) Annunciator CRD PUMP A TRIP alarms.

/

How is the movement of the control rods from the control room affected?

A.

Rods can be scrammed and withdrawn, but not individually inserted.

B.

Rods can be scrammed and inserted, but not individually withdrawn.

C.

Rods can only be scrammed.

D.

Scram Accumulator pressure assures that all functions work, although rod motion is slow, and CRD mechanism cooling is lost.

Question Data Answer: C Rods can only be scrammed.

ExplanationNustification:

A.

6.

C.

correct answer D.

No insert or withdraw available No insert or withdraw available accumulator function is for scram only KA Statement A.C. power

-i Sys#

System Category 201001 Control Rod Drive Hydraulic System Knowledge of the effect that a loss or malfunction of the following will have on the CONTROL ROD DRIVE HYDRAULIC System:

K/A#

~OIOOI.K~.O~

WA Importance 3.3 Exam Level RO 55.41 References provided to Candidate None Technical

References:

ON-155407 Question SOUt-Ce:

Modified Aw d, 1996 Question Cognitive Level:

Level Of Difficulty: (1-5) 10 CFR Part 55 Content:

NRC 2003 Rev I H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam 55 Select the method by which reactivity insertion rate is regulated for control rod withdrawals and insertions.

Reactivity insertion rate is controlled by:

A.

throttling the water flow entering and leaving the over-piston area of the drive mechanism.

automatically varying the position of the Control Rod Drive Hydraulic Pressure Control Valve.

6.

C.

automatically varying the operating position of the Control Rod Drive Hydraulic Flow Control Valve.

D.

throttling the water flow entering and leaving the under-piston area of the drive mechanism.

Question Data Answer: D throttling the water flow entering and leaving the under-piston area of the drive mechanism.

Explanation/Justification:

A.

B.

C.

D.

correct answer over-piston area flow is leaving is not thottled.

pressure control valve is not automatically positioned.

flow control valve controls total system flow not regulated flow to each mechanism

-+'

Sys#

System Category KA Statement 201003 Control Rod and Drive Mechanism Ability to predict andlor monitor changes in parameters associated with operating the CONTROL ROD AND DRIVE MECHANISM controls including:

CRD drive water flow WA#

2010~3.~1.03 WA Importance 2.9 Exam Level RO References provided to Candidate None Technical

References:

TM-OP-055 Question Source:

Modified River Bend 1,1997 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Comprehension 10 CFR Part 55 Content:

55.41 NRC 2003 Rev I H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC I 9 NRC Exam v

56 What is the design bases for allowing the Rod Sequence Control System to be automatically bypassed above 20% power?

A.

No combination of Operator errors could result in fuel damage due to a Control Rod Drop Accident.

B.

The power excursion caused by a Control Rod Drop Accident would be terminated by an APRM hi-flux SCRAM.

C. The RBM prevents any control rod from attaining the rod worth necessary to damage fuel on a Control Rod Drop Accident.

D.

Rod Worth Minimizer wiil continue to enforce the rod control sequence.

Question Data Answer: A No combination of Operator errors could result in fuel damage due to a Control Rod Drop Accident.

ExplanatiodJustification:

A.

B.

C.

D.

RWM also bypassed correct answer, voids in large enough quantity to minimize differential rod worths.

possibly but not the basis RBM minimizes local power increase by limiting amount of power change, does not affect rod worth Sys#

System Category KA Statement 201004 Rod Sequence Control System (Plant Specific)

Knowledge of the operational implications of the Prevention of clad damage if a following concepts as they apply to ROD SEQUENCE control rod drop accident CONTROL SYSTEM:

(CRDA) occurs: BWR4,5 WA#

201004.K5.01 WA Importance.s Exam Level RO References provided to Candidate None Technical

References:

TM-OP-056 Question Source:

New Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

55.41 NRC 2003 Rev I H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam b-57 With the Rod Worth Minimizer keylock in NORMAL, a loss of rod position signal frophe selected rod to the Rod Worth Minimizer will cause:

A.

B.

C.

only a withdraw block if power is less than the Low Power Set Point.

only a "SYSTEM ERROR display.

a withdraw and insert rod block if power is less than the Low Power Set Point.

D.

a withdraw and insert rod block at any power.

Question Data Answer: C a withdraw and insert rod block if power is less than the Low Power Set Point.

ExplanationlJustification:

k also insert block B.

withdraw and insert block C.

correct answer D.

blocks are bypassed above 20% power Sys#

System Category KA Statement 201006 Rod Worth Minimizer Ability to predict and/or monitor changes in Rod position: PSpec(Not-System (RWM) (Plant parameters associated with operating the ROD BWR6)

Specific)

WORTH MINIMIZER SYSTEM (RWM) controls including:

KIA#

~OIOO~.AI.OI WA Importance 3.2 Exam Level RO References provided to Candidate None Technical

References:

TM-OP-031 D Question Source:

Modified Susquehanna 1 81 2,1996 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

55.41 NRC 2003 Rev I H:\\NRCExarnPrep\\Rich\\NRCAForm.doc Printed on 0611 9/03

SSES LOC 19 NRC Exam 58 A Design Basis Accident has occurred on Unit 1 with a lockout on the 1A201 ESS bus.

Which of the following describes the

_J) cogre& ;&pa&

?

&mpact and why?

r e A.

RHR Outboard Injection Vlv HV-151 -F015A and Reactor Recirc Pump "A" Discharge Vlv HV-143-FO31A will not operate, power is lost from 1821 9.

9. RHR Outboard Injection Vlv HV-151 -F015A and Reactor Recirc Pump "A" Discharge Vlv HV-143-FO31A will operate with power from 16219.

C.

RHR Outboard Injection Vlv HV-151-FO15A and Reactor Recirc Pump "A" Discharge Vlv HV-143-FO31A will operate with alternate power from 16229.

D.

RHR Outboard Injection Vlv HV-151-FO15A and Reactor Recirc Pump "A" Discharge Vlv HV-143-FO31A will not operate power is lost from 16229.

Question Data Answer: B RHR Outboard Injection Vlv HV-151-FOI5A and Reactor Recirc Pump " A Discharge Vlv HV-143-FO31A will operate with power from 18219.

ExplanationlJustification:

A.

B.

ATS will supply power from 1 B230.

Correct answer, ESS Load Center 18210 is normal power supply via Swing Bus MG Set for 1 8219, loss of power from ESS bus I N 0 1 to 18210 will cause ATS to seek power from 18230.

The 18219 alternate source is ESS Load Center 16230 1 8229 is power supply for Div 2 valves.

18229 is power supply for Div 2 valves.

'd C.

D.

202001 Recirculation System Knowledge of electrical power supplies to the Recirculation system valves Question Cognitive Level:

Comprehension J W M 202001.~2.03 WA Importance 2.7 References provided to Candidate None Question Source:

New following:

Exam Level RO Technical

References:

TM-OP-004 Level Of Difficulty: (1-5) 3 10 CFR Part 55 Content:

55.41 NRC 2003 Rev I H:\\NRCExarnPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam L

59 While making a tour of the Unit 1 Lower Relay Room, you notice an alarm light on a RBM channel on top of panel 1 C608. The light is labeled, REF DNSCL.

What RBM function is associated with this alarm?

A.

Automatically bypasses RBM B.

Initiates Rod Block C.

Bypasses Rod Insert Blocks D.

Indication only for the reference APRM at 30% power Question Data Answer: A Automatically bypasses RBM ExplanationlJustification:

A.

correct B.

Not a RBM rod block C.

RBM does not block rod insertion D.

provides control function as well as indication.

Sys#

System Category KA Statement 215002 Rod Block Monitor KIA#

215002.~3.03 WA Importance 3.1 Exam Level RO Ability to monitor automatic operations of the ROD Alarm and indicating lights:

System BLOCK MONITOR SYSTEM including:

BWR-3,4,5 References provided to Candidate None Technical

References:

TM-OP-078 Question Source:

Modified Dresden 2 & 3,1996 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Memory 10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAFom.doc Printed on 06/19/03

SSES LOC 19 NRC Exam

=-./

60 Unit 1 is operating at 1-

r.

-Wide Range level indicates +!XLkhes.

- RX WATER HI LEVEL AR-101-AI7 alarm on I C651

- Green light above the LEVEL LOGIC RESET A HS-C32-1 S04A switch is lit on I C651.

- Green light above the LEVEL LOGIC RESET B HS-C32-1 S04B switch is lit on I C651.

- Green light above the LEVEL LOGIC RESET C HS-C32-1 S04C switch is NOT lit on 1 C651.

Which of the following is the expected response for the above indications A.

No Reactor Feedpumps tripped, no HPCl or RClC trip alarms.

B.

A, B & C Reactor Feedpumps tripped, no HPCl or RClC trip alarms.

C.

A & B Reactor Feedpumps tripped, C Reactor Feed Pump feeding, trip alarms annunciated for HPCl or RClC D.

No Reactor Feedpumps tripped, trip alarms annunciated for HPCl or RClC Question Data Answer: B A,

B & C Reactor Feedpumps tripped, no HPCl or RClC trip alarms.

\\

-/-

Explanation/Justification:

A.

B.

C.

D.

Trip logic for RFPs is actuated.

correct answer, two out of three logic for trip of feedpumps, HPCI-RCIC hi level trip come from different level switches thus possible not tripped.

No other alarms as mentioned in stem, thus no HPCllRClC alarms RFPs tripped, no alarms annunciated for HPCllRClC Sys#

System Category KA Statement 216000 Nuclear Boiler Knowledge of NUCLEAR BOILER Protection against filling the INSTRUMENTATION design feature@) andlor interlocks which provide for the following:

Instrumentation main steam lines from the feed system WA#

216000.~4.09 WA Importance Exam Level RO References provided to Candidate None Technical

References:

~ ~ - 0 P - 0 8 0 Question Source:

New Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

55.41

\\.--

NRC 2003 Rev I H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam 61 Suppression Pool cooling is in service on Unit 2 with the 2A RHR pump in service when a electrical fault causes -01 to All other buses remain energi

\\

r--

Which of the following describes how Suppression Pool cooling will be affected by loss?

A.

The 2D RHR Pump will be deenergized; Loops A & B are available for Suppression Pool Cooling.

B.

The 2A and 2C RHR Pumps will be deneergized; Loop B is available for Suppression Pool Cooling.

C. The 2A RHR Pump will be Cooling.

The 28 and 2D RHR Pool Cooling.

Loop B is available for Suppression P ~ o l D.

Loop A is available for Suppression Answer: C The 2A RHR Pump will be deenergized; Loop B k available for Suppression Pool Cooling.

Question Data

-./

ExpIanatiodJustification:

A.

only '2A' pump not available.

B.

1, only '2A' pump not available.

C.

correct answer, '2A' pump powered from 2A ESS bus.

D.

only '2A' pump not available.

219000 RHWLPCI:

Ability to predict andlor monitor changes in ToruslSuppression Pool Cooling Mode parameters associated with operating the RHRILPCI:

TORUSlSUPPRESSlON POOL COOLING MODE controls including:

WA#

~I~OOO.AI.O~

WA Importance 3.5 Exam Level RO References provided to Candidate None Technical

References:

TM-OP-049,0N-204-210 Question Source:

Modified Quad Cities 1,1996 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

C d e n s i o n 10 CFR Part 55 Content:

55.41 F

NRC 2003 Rev I H:\\NRCExamPrep\\Rich\\NRCAFom.doc Printed on 06/19/03

SSES LOC 19 NRC Exam L

62 Following an Auxiliary Bus Load Shed with all of the emergency busses energized, which of the following components will require operator actions to provide a source of coolin A.

Diesel Generators B.

Condensate pumps C.

HPCl room coolers D.

RHR pump room coolers Question Data Answer: B Condensate pumps ExplanationlJustification:

A.

B.

C.

D.

No additional operator actions required.

Correct answer, TBCCW available but will need ESW lined up to substitute for Service water.

No additional operator actions required.

No additional operator actions required.

L Sys#

System Category KA Statement 256000 Reactor Condensate Loss of equipment component System cooling water systems w/4#

256000.A2.12 RO References provided to Candidate None Technical

References:

ON-IO~-OOI Question Source:

Modified Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExarnPrep\\Rich\\NRCAForrn.doc Printed on 06/19/03

SSES LOC I 9 NRC Exam

+

63 Unit 1 & 2 are operating at 100% power when the following annunciator is received in the Control Room; REFUEL FLOOR WALL EXHAUST HI-HI RADIATION (AR-lOI-AO5)

Which of the following is the expected ventilation response?

A.

Isolation of Reactor Building Zone Ill Ventilation Automatic start of both SGTS fans RB Zone Ill Filtered Exh Fans 1V217A(B) and 2V217A(B) start.

B.

Isolation of Reactor Building Zone Ill Ventilation Automatic start of both SGTS fans Automatic start of Reactor Building Zone Ill Recirculation System C.

SGTS Train A or B start.

Emergency Outside Air Supply Fan A(B) starts.

RB Recirc System to SBGT HD07543A(B) closes.

D.

SGTS Train A or B start.

RB Zone Ill Is0 Dampers HD27502A(B), HD27514A(B) HD27564A(B), HD17502A(B),

HDI 7514A(B) and HD17564A(B) closes.

RB Zone I I I Filtered Exh Fans 1 V217A(B) and 2V217A(B) start.

-L-Question Data Answer: B Isolation of Reactor Building Zone 111 Ventilation Automatic start of both SGTS fans Automatic start of Reactor Building Zone 111 Recirculation System ExplanatiodJustification:

A.

B.

Correct answer C.

D.

RB Zone 111 Filtered Exh Fans trip not start.

RB Recirc System to SBGT HD07543A(B) open instead of closing.

RB Zone 111 Filtered Exh Fans trip not start.

KA Statement Sys#

System Category 272000 Radiation Monitoring Ability to monitor automatic operations of the Ventilation system isolation WAff 272000.~3.06 WA Importance 3.4 Exam Level RO System RADIATION MONITORING SYSTEM including:

indications References provided to Candidate None Technical

References:

TM-OP-079 Question SOUrCe:

Modified Big Rock Point 1,1995 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

55.41 NRC 2003 Rev I H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam has occurred with UnLl at 100% p L-Reactor Building Ventilation m a

l procedure has been entered, room cooling initiated and maintenance support requested.

Which area is expected to have the most rapid temperature increase:

A.

ESS Switchgear rooms B.

HPCl Room C.

Main Steam Pipe Tunnel D.

RWCU Pump Room Question Data Answer: D RWCU Pump Room ExpIanatiodJustification:

A.

B.

C.

D.

Correct answer ESS Switchgear rooms has its own cooling system which remains in service with a loss of Zone I HVAC.

HPCl Room has it's own cooling system which remains in service with a loss of Zone I HVAC.

Main Steam Pipe Tunnel has it's own cooling system which remains in service with a loss of Zone I HVAC.

Sys#

System Category KA Statement 288000 Plant Ventilation Systems Knowledge of the effect that a loss or malfunction of the PLANT VENTILATION SYSTEMS will have on following:

Reactor buildjng temperature:

PlantSpecific WA#

288000.~3.02 WA Importance 2.9 Exam Level RO References provided to Candidate None Technical

References:

ON-134402 Question Source:

New Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam 65 Unit 1 had a loss of Drywell Cooling while operating at 100% power due to a spurious high Drywell pressure signal.

The high Drywell pressure signal is cleared.phat needs to be completed to reset Drywell

';/

Cooling isolation logic?

4 A.

Reset Div I & II DRWL CLG logic on 1 C601 Reset Div I & II RBCW IS0 VALVE POS logic on IC681 Go to open for A & B Drywell Cooler inboard and outboard isolation valves B.

Go to close for A & B Drywell Cooler inboard and outboard isolation valves Reset Div I & I I DRWL CLG logic on IC601 Reset Div I & II RBCW IS0 VALVE POS logic on 1 C681 C.

Reset MN STM LINE DIV I & II logic on IC601 Go to close for A & B Drywell Cooler inboard and outboard isolation valves Reset Div I & II RBCW IS0 VALVE POS logic on IC681 D.

Reset Div I & II DRWL CLG logic on IC601 Reset Div I & II RBCW IS0 VALVE POS logic on IC681 Verify A & B Drywell Cooler inboard and outboard isolation valves open Question Data Answer: D Reset Div I & II DRWL CLG logic on IC601 Reset Div I & II RBCW IS0 VALVE POS logic on IC681 Verify A & B Drywell Cooler inboard and outboard isolation valves open ExplanationlJustification:

A.

B.

C.

D.

Correct answer No need to open valves No need to go to close.

No need to go to close Sys#

System Category KA Statement 290001 Secondary Containment Ability to manually operate andlor monitor in the System reset: PlantSpecific WA#

290001.~4.11 WA Importance 3.4 Exam Level RO control room:

References provided to Candidate None Technical

References:

ON-159402 Question Source:

Modified Hope Creek Unit 1,1998 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam L

66 The US, PCO, and I&C have discussed the performance of a Surveillance Procedure on the ADS system. The annunciators expected to be received during the surveillance have been identified and reviewed.

Which of the following describes the required actions when one of the expect is received?

The operator shall acknowledge the alarm and: /

A.

is NOT required to report the alarm to the US. The op ator does NOT have to the associated alarm response procedure.

is NOT required to report the alarm to the US. The associated alarm respon procedure shall be checked.

the annunciator shall be reported to the US. The operator does associated alarm response procedure.

B.

C.

D.

the annunciator shall be reported to the US. The associated alarm response procedure shall be checked.

Question Data Answer: A is NOT required to report the alarm to the US. The operator does NOT have to refer to the associated alarm response I

procedure.

ExplanationlJustification:

A.

B.

C.

D.

correct answer, The operator shall acknowledge the alarm and is NOT required to report the alarm to the US. The operator does NOT have to refer to the associated alarm response procedure.

The Alarm Procedure does not have to be referred to under these conditions No report is required to the US No report is required to the US Sys#

System Category KA Statement Conduct of Operations Knowledge of operator responsibilities during all modes of plant operation.

WA#

2.1.2 WA Importance 3.0 Exam Level RO References provided to Candidate None Technical

References:

OP-AD-004 (Sect 11.2.c)

Question Source:

Modified Cooper I, 1999 Level Of Difficulty: (1-5) 2 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC I 9 NRC Exam L-67 In accordance with STANDARDS OF SHIFT OPERATION, OP-AD-002, which of the following is the PREFERRED method to perform a required verification on a throttled valve set at two turns open?

A.

B.

C.

D.

Independent visual check of required valve position.

Perform a second valve operation to verify the position.

Observe flow indication through the throttled valve's system during Observe the initial operator's action in positioning the throttled

'c J

Question Data Answer: D Observe the initial operator's action in positioning the throttled valve.

ExplanationlJustification:

A.

B.

C.

D.

throttle valves are aligned using concurrent verification throttle valves are aligned using concurrent verification throttle valves are aligned using concurrent verification correct answer, throttle valves are aligned using concurrent verification Conduct of Operations Ability to perform specific system and integrated plant procedures during different modes of plant operation.

WA#

2.1.23 WA Importance 3.9 Exam Level RO References provided to Candidate None Technical

References:

OP-AD-002 (7.3.5)

Question Source:

Modified Quad Cities 1,1998 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExarnPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam

-v.

68 During a startup on Unit 2 reactor, The Plant Control Operator withdraws control rod 26-27 from notch 32 to notch 34, reactor period changes from 200 seconds to a stable 50 second period.

Which of the following identifies the required action to be taken?

A.

Re-insert control rods as necessary to achieve sub-criticality.

B.

Shutdown the reactor until a thorough assessment has been performed.

C.

Re-insert control rod 26-27 to obtain a stable period indication of greater than 100 seconds.

D.

Do not move any additional rods until a Core Monitor is run.

Question Data Answer: C Re-insert control rod 26-27 to obtain a stable period indication of greater than 100 seconds.

ExplanatiodJustification:

A.

B.

obtain a stable period indication of greater than 100 seconds.

obtain a stable period indication of greater than 100 seconds.

obtain a stable period indication of greater than 100 seconds.

L-'

C.

correct answer D.

Sys#

System Category Conduct of Operations wA## 2.1.7 WA Importance 3.7 References provided to Candidate None Question Source:

Modified Browns Ferry 2,2001 Question Cognitive Level:

KA Statement Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

Exam Level RO Technical

References:

GO-200-002,6.23 Level Of Difficulty: (1-5) 3 10 CFR Part 55 Content:

55.41 NRC 2003 Rev I H:\\NRCExamPrep\\Rich\\NRCAForrn.doc Printed on 06/19/03

SSES LOC I 9 NRC Exam v

69 Given the following conditions:

- Reactor power is 40%

- ALL Turbine Bypass Valves fail OPEN.

- The MSlVs FAIL to automatically close.

- The MSlVs are closed manually

~

Prior to MSlV closure, which of the following combinations of reactor power and reactor pressure would indicate a safety limit violation had occurred?

B A.

Reactor power is 30% and RPV pressure is 810 psig &

B.

Reactor power is 30% and RPV pressure is 775 psig C.

Reactor power is 20% and RPV pressure is 795 psig D.

Reactor power is 10% and RPV pressure is 810 psig Question Data Answer: B Reactor power is 30% and RPV pressure is 775 psig ExplanationlJustification:

k B.

C.

D.

Pressure and flow within Safety Limit correct answer, <785 psig, 4 0 E 6 Ibm core flow, thermal power must be < 25%.

Power within Safety Limit of 25%

Pressure, power and flow within Safety Limit.

u Sys#

System Category KA Statement Equipment Control Knowledge of limiting conditions for operations and safety limits.

WA#

2.2.22 WA Importance 3.4 Exam Level RO References provided to Candidate. None Technical

References:

T.S. 2.1 Question Source:

Modified Dresden 2,1998 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC I 9 NRC Exam 70 A Procedure Change for immediate use is required for RClC Quarterly Flow Verification, SO-150-002. The HV-149-F022 TEST LINE IS0 TO C 6 n e s d s to be set to a position other than 40% OPEN as required by The Proce ure Change because of a pump impeller modification.

steps to be completed prior to use of the procedure. -4' A.

Obtain PORC committee review, Log PCAF in the control room Procedure PCAF Log, Stamp PCAF placed in controlled manuals as CONTROLLED, Deliver the original PCAF to training.

B.

Obtain responsible Functional Unit Manager approval, Stamp PCAF placed in controlled \\._

w manuals as CONTROLLED, QC approve insertion into appropriate manuals.

C.

Log PCAF in the control room Procedure PCAF Log, Stamp PCAF placed in cont"rolte6 manuals as CONTROLLED, Deliver the original PCAF to DCS, Obtain P m

ittee review,.

D.

Obtain responsible Functional Unit Manager approval, Stamp PCAF placed in controlled manuals as CONTROLLED, Deliver the original PCAF to DCS.

Question Data Answer: C

--.~

Log PCAF in the control room Procedure PCAF Log, Stamp PCAF placed in controlled manuals as CONTROLLED, Deliver the original PCAF to DCS, Obtain PORC committee review,.

Expfanation/Justification:

A.

B.

C.

Correct answer.

D.

PCAF requires PORC approval.

PCAF is delivered to DCS.

QC does not approve PCAFs Sys#

System Category KA Statement Equipment Control KIA#

2.2.11 KIA Importance 2.5 Exam Level RO Knowledge of the process for controlling temporary changes.

References provided to Candidate None Technical

References:

NDAP-QA-0002 (8.3)

Question Source:

Modified Quad Cities I, 1998 Level Of Difficulty: (I

-5) 3 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

55.41 NRC 2003 Rev I H:\\NRCExarnPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam L-71 The plant is in MODE 4 preparing for a refueling outage.

What is/are the MINIMUM action(s) that must be performed to enter into MODE 5?

A.

De-tension one reactor vessel head closure bolt.

B.

Place the Reactor Mode Switch in the REFUEL position.

C. Place the Reactor Mode Switch in the REFUEL position and de-tension one reactor vessel head closure bolt.

D.

Place the Reactor Mode Switch in the REFUEL position and de-tension all the reactor vessel head closure bolts.

Question Data De-tension one reactor vessel head closure bolt.

ExplanatiodJustification:

A.

B.

C.

D.

correct answer, when detensioning of head begins, plant is in Mode 5 by definition of Tech Specs.

By procedure the mode switch should be in refuel, but head stud detensioning is accpted definition of Refuel Mode.

By procedure the mode switch should be in refuel, but head stud detensioning is accpted definition of Refuel Mode.

By procedure the mode switch should be in refuel, but head stud detensioning is accpted definition of Refuel Mode.

Sys#

System Category KA Statement Equipment Control Knowledge of refueling administrative requirements.

KIA#

2.2.26 WA Importance 2.5 Exam Level RO References provided to Candidate None Technical

References:

Tech Spec Definitions 1.01 Question Source:

Modified Perry 1,2001 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

55.41 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam 72 A 22 year old operator is working in a radiation field under the following conditions:

The operators cumulative dose for the year is 940 mrem.

The job is in a 20 mrem/hr radiation area.

No dose extension has been or will be authorized.

Select the number of hours the operator may work in the radiation area administrative limit for the year?

A.

26 hrs

9. 53 hrs C.

203 hrs D.

153 hrs Question Data Answer: B 53 hrs Explanation/Justification:

A.

8.

C.

0.

1000 Admin limit (minus) 940 cum dose = 60 dose available (divided by) 20 dose rate = 26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br /> correct answer, 2000 Admin limit (minus) 940 cum dose = 1060 dose available (divided by) 20 dose rate = 53 hours6.134259e-4 days <br />0.0147 hours <br />8.763227e-5 weeks <br />2.01665e-5 months <br /> 5000 NRC* (minus) 940 cum dose = 4060 dose available (divided by) 20 dose rate = 203 hours0.00235 days <br />0.0564 hours <br />3.356481e-4 weeks <br />7.72415e-5 months <br /> 4000 Admin limit (minus) 940 cum dose = 1060 dose available (divided by) 20 dose rate = 153 hours0.00177 days <br />0.0425 hours <br />2.529762e-4 weeks <br />5.82165e-5 months <br /> L,

Sys#

System Category Radiological Controls KA Statement Knowledge of 10 CFR 20 and related facility radiation control requirements.

WA#

2.3.1 WA Importance.s Exam Level RO References provided to Candidate None Technical

References:

NDAP-QA-0625 (6.2)

Question Source:

Modified Clinton 1,2001 Level Of Difficulty: (1-5) 3 Question Cognitive Level :

Comprehension 10 CFR Part 55 Content:

55.41 NRC 2003 Rev I H:\\NRCExamPrep\\Rich\\NRCAForm

.doc Printed on 06/19/03

SSES LOC 19 NRC Exam

.-..-/-.

73

- Unit 1 is in Mode 3.

- It is desired to de-inert the Unit 1 Primary Containment as soon as possible to permit containment access for maintenance.

What flowpath is permited for de-inerting of the Unit 1 Suppression Chamber?

A.

Both Standby Gas Treatment trains in service, vent Suppression Pool via SUPP CHMBR VENT IB IS0 HV-15703 and SUPP CHMBR VENT OB IS0 HV-15704.

B.

One Standby Gas Treatment train in service, the other Standby Gas Treatment Train operable, vent Suppression Pool via SUPP CHMBR VENT IB IS0 HV-15703 and SUPP CHMBR VENT OB IS0 HV-15704.

C.

Both Standby Gas Treatment trains in service, vent Suppression Pool via SUPP CHMBR VENT IB IS0 HV-15703, SUPP CHMBR VENT OB IS0 HV-15704, and SUPP CHMBR VENT BYPS OB IS0 HV-15705.

D.

One Standby Gas Treatment train in service, the other Standby Gas Treatment Train operable, vent Suppression Pool via SUPP CHMBR VENT IB IS0 HV-15703 and SUPP CHMBR VENT OB IS0 HV-15704, and SUPP CHMBR VENT BYPS OB IS0 HV-15705.

Question Data Answer: B One Standby Gas Treatment train in service, the other Standby Gas Treatment Train operable, vent Suppression Pool via SUPP CHMBR VENT IB IS0 HV-15703 and SUPP CHMBR VENT OB IS0 HV-15704.

ExplanationlJustification:

A.

B.

Correct answer.

C.

D.

Only one SBGT train I/S at a time.

Only one SBGT train 11s at a time, and either the 04 or 05 valve open, not both.

Either the 04 or 05 valve open, not both.

d Sys#

System Category c/

KA Statement Radiological Controls Knowledge of the process for performing a containment purge.

KIA#

2.3.9 WA Importance 2.5 Exam Level RO References provided to Candidate None Technical

References:

OP-173401 Question Source:

Modified Quad Cities 1,2001 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Comprehension 10 CFR Part 55 Content:

55.41 NRC 2003 Rev I H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam L

74 EO-I 00-1 02, "RPV Control", directs the operator to reset the main generator lockout if RPV level can be maintained > -1 29 inches.

SELECT the correct ba@ for this step from the reasons listed below.

A.

To allow the Stator Water Cooling Pumps to be restarted to provide cooling to the main generator B.

To allow the Recirc Pumps to be restarted to establish forced reactor coolant circulation C.

D.

To prevent a trip of the Stator Water Cooling Pump To prevent a plant Auxiliary Bus load shed Question Data Answer: D To prevent a plant Auxiliary Bus load shed ExplanatiodJustification:

A.

B.

C.

D.

correct answer No half scram signal received No RRP runback signal exists No Rod Block signal received Sys#

System Category KA Statement Emergency Procedures and Plan Knowledge of the specific bases for EOPs.

WA## 2.4.18 WA Importance 2.7 Exam Level RO Technical

References:

AR-103-001 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

55.41 J

References provided to Candidate None Question Source:

Modified NRC 2003 Rev I H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 0611 9/03

SSES LOC 19 NRC Exam 75 The reactor is shutdown with one loop of shutdown cooling in use and NO Recirculation Pumps running.

How would the Shutdown Cooling System respond if Reactor Vessel Level decreased from

+50 inches to +4 inches?

A.

The shutdown cooling suction inboard and outboard isolation valves isolate, the operating RHR pump trips.

B.

The shutdown cooling suction inboard and outboard isolation valves isolate, the operating RHR pump remains running.

C.

Shutdown cooling continues unaffected.

D.

The shutdown cooling suction inboard and outboard isolation valves isolate, the operating RHR pump trips, HV-151-F)15A opens, remaining RHR pumps auto start..

Question Data Answer: A The shutdown cooling suction inboard and outboard isolation valves isolate, the operating RHR pump trips.

ExplanatiodJustification:

A.

6.

C.

D.

correct answer, running RHR pump will trip due to loss of suction path.

No suction path running pump will trip.

No suction path, running pump will trip.

F015 valve auto opens and RHR pumps auto start at -129.

Sys#

System KA Statement Knowledge of low power I shutdown implications in accident (e.g. LOCA or loss of RHR) mitigation strategies.

55.41 WA#

2.4.9 WA Importance 3.3 Exam Level RO References provided to Candidate None Technical

References:

TM-OP-049 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

Question Source:

New Level Of Difficulty: (1-5) 3 NRC 2003 Rev 1 H :\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 0611 9/03

SSES LOC 19 NRC Exam v

76 Unit 2 is operating at power when a recirculation flow reduction event results in entry into Region 2 of the Power to Flow Map. Plant conditions PRIOR to the event were as follows:

- Reactor power was 90%.

- APRMs indicated 90%

- All LPRMs were above downscale alarms and below upscale alarms.

- Period meters indicated infinity.

After the flow reduction event and core flow first reaches its lowest flow rate, which of the following instrumentation responses would you use as the Unit Supervisor to justify entry into the core oscillation off normal procedure?

A.

Peak to peak oscillations on RBM 10% and growing larger.

B.

Period meters are oscillating and short period alarms are received on a I O to 20 second frequency.

C.

Peak to peak oscillations on APRM's 5% to 6% and their magnitude is growing larger./

D.

Total Steam flow oscillations 10 to 12% and their magnitude is growing larger

\\-

Question Data Answer: C Peak to peak oscillations on APRM's 5% to 6% and their magnitude is growing larger.

ExplanationNustification:

A.

B.

C.

correct answer D.

RBM not referenced in the off normal procedure.

No reference to period indication in the off normal procedure.

Steam flow not referenced in the off normal procedure.

~~

Category KA Statement they apply to PARTIAL OR COMPLETE LOSS OF Sys#

System 295001 Partial or Complete Loss Ability to determine andlor interpret the following as Neutron mpRit6ring of Forced Core Flow Circulation FORCED CORE FLOW CIRCULATIOW:

WA#

Z ~ ~ O O I. A A Z. O ~

WA Importance 3.2 Exam Level SRO References provided to Candidate None Technical

References:

O N - 2 7 6 Question Source:

Modified Peach Level Of Difficulty: (1-5) 3 Question Cognitive Level:

10 CFR Part 55 Content:

55.43 NRC 2003 Rev 1 H:\\NRCExarnPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC I 9 NRC Exam 9

77 Unit I is operating at 100% power w h e n b t r u m e n t air line in the Turbine Building rupt-2 The air compressors are unable to keep up'with the loss of air and instrument airpPessur e IS lowering.

What will the overall Reactor Pressure Vessel level control and pressure control strategy be for the loss of instrument air?

____I A.

Pressure Control using Main Steam Line drains, condensate pumps for level control.

B.

SRVs for pressure control, HPCVRCIC for level control C.

Maximize CRD, RClC for level control, SRVs for pressure control D.

HPCVRCIC for level control, SRVs and Main Steam line drains for pressure control.

Question Data Answer: B SRVs for pressure control, HPCI/RCIC for level control ExpIanationlJustification:

A.

6.

C.

Condenser is not available and no condensate line up is possible due to level control valves fail closed on a loss of air.

correct answer, Outboard MSlVs will go closed on a loss of air, therefore no steam for feedpumps or use of the main condenser for decay heat. Condensate will be unavailable due to no feedpath on a loss of air.

CRD flow control valves fail closed on a loss of air, D.

Condenser is not available for pressure control Exam Level SRO Technical

References:

ON-118-001, Eo-100-1 02 Level Of Difficulty: ( l r 3

10 CFR Part 55 Content:

of Instrument Air INSTRUMENT AIR:

WA#

295019.~~2.02 WA Importance g References provided to andidate None Question Cognitive Level:

55.43 Question Source:

Yf NRC 2003 Rev I H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

\\-,

78 Which of the following describes the consequences if a gh Exhaust Duct rad - monito Any release as a result of this act@ will:

A.

not be processed by SGTS and may result whole body and an Emergency Plan entry at the General Emergency level.

B.

not be processed but will be monitored by Zone 1 & Zone 2 rad monitors for conditions requiring entry into EO-I 00-1 05, Radioactivity Release Control and an Emergency Plan entry at the Site Area Emergency level.

C.

will be processed by Zone 1 and Zone 2 HVAC but still may result in a site boundary dose in excess of 25 rem whole body and an Emergency Plan entry at the General Emergency level.

D.

will be processed by Zone 1 and Zone 2 HVAC but still may result conditions requiring entry into EO-I 00-1 05, Radioactivity Release Control and an Emergency Plan entry at the Site Area Emergency level.

Question Data Answer: A

/

not be processed by SGTS and may result in a site boundary dose in excess of 25 rem whole body and an Emergency Plan entry at the General Emergency level.

ExplanationNustification:

A.

B.

C.

D.

Correct, inop rad monitors, SGTS wont start, release wont be processed, and release may exceed 10CFR100 limits at site boundary which is 25 Rem whole body and 300 Thyroid.

Incorrect, Zone 3 independent of 1 and 2, no monitoring done anywhere Incorrect, Zone 3 independent of 1 and 2, no impact on site boundary limits by IOCFR I00 Incorrect, Zone 3 independent of 1 and 2, no processing done anywhere.

Sys#

System Category KA Statement 295023 Refueling Accidents Ability to determine andlor interpret the following as Entry conditions of emergency plan W M 295023.~~2.05 WA Importance Exam Level SRO they apply to REFUELING ACCIDENTS:

References provided to Candidate None Technical

References:

Tech Spec 83.3.6.2 Question Source:

New Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Corn prehension 10 CFR Part 55 Content:

55.43 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAFom.doc Printed on 06/19/03

SSES LOC 19 NRC Exam 79 A loss of drywell cooling occurs on Unit 1. Drywell pressure increases to 2.9 psig. Drywell

-I temperature increases to 155 degrees F.

Which portions of the Emergency Operating Procedures 7

u be working for the above plant conditions?.

A.

All legs of EO-I 00-1 02, RPV Control and all legs of EO-I 00-1 03, Primary Containment Control.

B.

All legs of EO-I 00-1 03, Primary Containment Control. No entry to EO-1 00-1 02, RPV Control is required.

C.

Primary Containment Pressure (PC/P) and Drywell Temperature (DWn) legs of EO-100-1 03, Primary Containment Control. No entry to EO-I 00-1 02, RPV Control is required.

D.

Primary Containment Pressure (PC/P) and Drywell Temperature (DW/T) legs of EO-100-1 03, Primary Containment Control and all sections of EO-I 00-1 02, RPV Control.

Question Data Answer: A All legs of EO-100-102, RPV Control and all legs of EO-100-103, Primary Containment Control.

ExplanatiodJustification:

A.

B.

C.

correct answer, high drywell pressure would be a scram requiring entry into 102 and 103.

Must enter 102 for scram.

Must enter 102 for scram.

KJA#

295024.~~2.02 KIA Importance 4.0 Exam Level References provided to Candidate Emergency Operating Technical

References:

EO-000-102 or 103 Procedures Question Source:

Modified Susquehanna, 1996 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

55.43 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam

-.--A 80 Unit 1 at 100% power when a MSIV isolation occurs

- Partial scram with one qu

- Reactor pressure peaked What automatic actions shoul procedur8)wII be used for reactor vessel pressure control?

A.

B.

C.

D.

ARI, SRV operation. EO-I 00-1 02, EO-I 00-1 13.

n t r - o t fully inserted.

curred to control the pressure transient and what Reactor Recirc Pump trip, SRV operation. EO-I 00-1 13 Main Turbine Bypass valves, SRV operation. EO-I 00-1 02, EO-I 00-1 13.

Reactor Recirc Pump trip, Main Turbine Bypass valves. EO-I 00-1 02

,PA Question Data Reactor Recirc Pump trip, SRV operation. EO-I 00-1 13 ExplanationNustification:

A.

B.

C.

D.

correct answer, Recirc pump ATWS trip at 1123, SRV ops at 1126. EO-100-1 13 for ATWS.

Bypass valves not available with MSlVs closed. EO-100-102 entered but exited due to more than one rod out, thus not used for pressure control.

Bypass valves not available with MSlVs closed. EO-100-102 entered but exited due to more than one rod out, thus not used for pressure control.

EO-100-102 entered but exited due to more than one rod out, thus not used for pressure control.

L Sys#

System Category KA Statement WA#

295025.2.1.23 WA Importance 4.0 Exam Level SRO Procedures References provided to Candidate Emergency Operating Technical

References:

AR-103-001 Question Source:

New Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

55.43 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam v-81 Unit 1 received a 100% power load reject and a complete loss of offsite power:

- Reactor scram 2

3 e

n

- RPV Level

-10 inches

- Drywell Temperature 255 F-/

- Suppression Pool Level

- Suppression Chamber Pressure

- Suppression Pool Temperature

- RPV Pressure 3 4 0 h

- Drywell Pressure

+m$g 1 16 feet What systems are available for Reactor Vessel Level control?

A. M S L C and CRD RHR Service Water, Core Spray, CRD Condensate pumps, RHR, SLC

/-

Core Spray, WtC,

• \\

CRD Question Data Answer: A HPCI, SLC and CRD ExplanatiodJustification:

A.

B.

correct answer, core spray if flow less than 5200 gpm RHR SW doesn't have the required dischargepressure at this time n at >I40 deg F due to lube oil c Sys#

System Category KA Statement KIA#

295026.2.4.11 WA Importance 3.6 References provided to Candidate None Question Source:

New Question Cognitive Level:

Analysis Exam Level SRO Technical

References:

ON-100-009 Level Of Difficulty: (1-5) 3 10 CFR Part 55 Content:

55.43 NRC 2003 Rev 1 H:\\NRCExarnPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam k-1 82 A-LOCA has occurred and level has trended down and the following conditions exist in the p l a n t 7 Reactor Pressure Wide Range Level Fuel Zone Level Upset Range Level Shutdown Range Level Narrow Range Level Drvwell Pressure Drvwell Temperature 50 psig J

- 150 inches

- 190 inches 0 inches 0 inches 0 inches

+ 5.2 psig 200 degrees F

/

J J

As the Unit Supervisor mI,,ch of the following Reactor Level Instruments would you instruct the operators to use as water level lowers?

A.

ShutdownRange B.

WideRange C.

Upset Range (fh ?P D.

Fuel Zone Range Question Data Answer: D Fuel Zone Range ExplanatiorVJustification:

A 0-500, calibrated cold

6. range to -150, calibrated for 1035 reactor pressure C.

0 - 180 calibrated hot.

D.

correct answer, calibrated for existing conditions, on scale Sys#

System Category KA Statement 295028 High Drywell Temperature Ability to determine and/or interpret the following as Reactor water level WA#

295028.~~2.03 WA Importance 3.9 Exam Level SRO References provided to Candidate None Technical

References:

they apply to HIGH DRYWELL TEMPERATURE:

Question Source:

Modified Grand Gulf 1,2000 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Comprehension 10 CFR Part 55 Content:

55.43 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam 83 An ATWS from 100% power has occurred on Unit 1. The following conditions now exist:

RPV level band:

Reactor Power:

SLC Tank level 5%

-161 inches to -60 inches 25 on IRMs range 6 Based on the above conditions, what will be the next directions given to the shift crew?

A.

restore and maintain RPV level in the normal band, +I3 to +54 inches B.

C.

D.

commence a controlled cooldown of the reactor vessel per EO-I 00-1 02 exit the EO-I 00-1 13 flowchart and control level per EO-I 00-1 02.

restore and maintain RPV level, -60 to -1 10 inches restore and maintain RPV level in the normal band, +I3 to +54 inches el should be restored to ExplanatiodJustification:

+I 3-54 B.

D.

commence a controled cooldown as per EO-100-113, not 102 board.

with HSBW added. level is restored to +I3 to +54 inches.

'L-C.

remain on the EO-100-113 board for level control.

Sys#

System Category KA Statement KIA#

295031.2.4.22 KIA Importance 4.0 Exam Level SRO Procedures References provided to Candidate Emergency Operating Technical

References:

EO-loo-113 Question Source:

New Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Comprehension 10 CFR Part 55 Content:

55.43 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC I 9 NRC Exam d

84 Unit 1 operating at 100% power. B Circ Water (CW) pump motor winding temperature indicates a rising trend on the PlCSY computer display and CW pump discharge pressure PI1 151 1A on IC668 indicates a lowering trend.

As the Unit Supervisor which of the following directions to the Plant Control Operators would be appropriate for the given plant conditions.

A.

Monitor CW discharge pressure, at 5 psig prior to discharge pressure alarm setpoint -

trip CW pump, Monitor Main Condenser vacuum, reduce power per ON-143-001 as required.

B.

Monitor CW pump motor temperatures, Monitor Main Condenser vacuum for degradation, monitor Reactor power for a lowering trend, reduce power per GO-1.00-012 as required.

JP*.v C.

Monitor CW motor temperature, at 10 degrees before computer temperature alarm - trip 1 dp

. I CW pump, monitor Reactor power for a rising trend, reduce power per CRC Book as I

required.

D.

Monitor Main Condenser vacuum for degradation, monitor condensate temperature for a lowering trend, reduce power per ON-143-001 as required.

L Question Data Answer: B Monitor CW pump motor temperatures, Monitor Main Condenser vacuum for degradation, monitor Reactor power for a lowering trend, reduce power per GO-100-012 as required.

Explanation/Justifiction:

A.

no discharge pressure alarm on CW system, a trip of the CW pump will cause a reurc runback which is an undesired automatic function slight reduction in reactor power.

reactor power will lower not rise due to vacuum and condensate temperature.

Condensate temperature will rise instead of lower.

C.

D.

Vacuum they apply to LOSS OF MAN CONDENSER VACUUM:

WA#

295002.~~2.02 WA Importance 3.3 Exam Level SRO Technical

References:

10 CFR Part 55 Content:

55.43 References provided to Candidate None Question Source:

New Level Of Difficulty: (1-5)

Question Cognitive Level:

Analysis k

NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam 85 Given the following conditions on Unit I

- Reactor power has been lowered to 95% in preparation for Turbine Control Valve testing

- Prior to starting the test, the PCOM reports APRM reactor power is rising

- Power peaks and stabilizes at 100%

- No alarms are received

- After investigation, the PCOP discovers HPCl is running and injecting

- All other equipment and instruments are operating as designed HPCl initiation caused by relay room cabinet door jarring relays.

Which of the following is the required Unit Supervisor direction regarding reactor power and the reportablity requirements for these conditions?

The US shall direct a Recirc Flow reduction to:

A.

less than 75% power and ensure an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> reportability call is made.

B.

less than or equal to 95% power and !E4w reportability.

C.

less than 75% power and ensure a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> reportability call is made.

L d

D.

less than or equal to 95% power and ensure a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> reportability call is made 4M Question Data Answer: B less than or equal to 95% power and r q reportability.

ExplanationlJustification:

A.

power reduction to 75% is required for a loss of feedwater heating.

signal.

power reduction to 75% is required for a loss of feedwater heating. Incorrect reportability time.

C.

D.

Incorrect reportability time.

References provided to Candidate NDAP-QA-0720 Technical

References:

ON-156-001 & NDAP-QA-Attachment G 0720 Question Source:

New Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Comprehension 10 CFR Part 55 Content:

55.43 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForrn.doc Printed on 06/19/03

SSES LOC 19 NRC Exam

\\-'

86 On Unit 1 a scram has occurred. The following plant conditions exist:

- CIG/MSIV interlocks have been bypassed

- Instrument Air supplying CIG

- The SDV is full

- The scram pilot valve air header is "0" psig

- Numerous control rods are NOT fully in rted

- Reactor power is stable at 12%

As the Unit Supervisor reviewing the in 'cations of the failure to scram transient, you direct the PCOM to:

k A.

Vent the scram airheader. / f i A L I

B.

Initiate ARi. &"J$&

C.

Reset the scram to drain the SDV and insert a manual scram.

gbJ@

D.

De-energize the scram solenoids.

Question Data Answer: C Explanation/Justification:

A.

6.

C.

Correct answer D.

Reset the scram to drain the SDV and insert a manual scram.

'v-Venting the scram air header would not serve any additional purpose since air header is 0.

ARI would not serve any additional purpose since air header is 0.

de-energizing the scram solinoids would not serve any additional purpose since air header is 0.

Sys#

System Category KA Statement WA#

295015.2.1.7 WA Importance 4.4 Exam Level SRO References provided to Candidate Emergency Operating Technical

References:

EO-000-113 Question Source:

Modified Level Of Difficulty: (1-5) 10 CFR Part 55 Content:

55.43 Question Cognitive NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC I 9 NRC Exam 87 A loss of both CRD pumps has occurred during a reactor startup with reactor pressure at 500 psig.

During the crew brief to discuss the loss of both CRD pumps, an NPO verifies a low pressure accumulator alarm that was received on a rod at position 12%'

As the Unit Supervisor which of the following actions are required?

L.

A.

Continue the brief, direct the PCOM to check for the second accumulator alarm, then insert manual scram.

B.

Suspend the brief, direct the PCOM to insert the rod with the accumulator alarm one notch.

C.

Continue the brief, within 20 min. direct the PCOM to restart a pump and insert one rod one notch D.

Suspend the brief, direct the PCOM to immediately insert a manual scram Question Data Answer: D Suspend the brief, direct the PCOM to immediately insert a manual scram

\\--'

ExplanationlJustification:

A.

B.

C.

D.

No CRD pumps, Rx pressure less than 900, accumulator alarm, Mode Switch to SID.

No CRD pumps, Rx pressure less than 900, accumulator alarm, Mode Switch to SID.

No CRD pumps, Rx pressure less than 900, accumulator alarm, Mode Switch to S/D.

correct answer, If reactor steam dome pressure c 900 psig and one or more scram accumulators are inoperable, PLACE Reactor Mode Switch in SHUTDOWN position.

Sys#

System Category KA Statement 295022 Loss of CRD Pumps Ability to determine andlor interpret the following as Accumulator pressure they apply to LOSS OF CRD PUMPS:

WA#

295022.~~2.01 WA Importance 3.6 Exam Level SRO References provided to Candidate None Technical

References:

ON-I 55-007 Question Source:

Modified Nine Mile Point 1,1998 Level Of Difficulty: (1 -5) 3 Question Cognitive Level:

Comprehension 10 CFR Part 55 Content:

55.43 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam 88 Given the following conditions with Unit I in Mode 4 and Unit 2 at 100% power LOOP/LOCA testing is in progress on Unit 1.

As part of the test, all four Diesel Generators start and load to their respective Unit 1 busses.

ECCS responses are as follows:

- 1A RHR Pump Does NOT,start

- 1 B and I C RHR Pumps Start adseconds

- I D RHR Pump Starts at 10 seconds

- All 4 Unit 1 Core Spray Pumps Which of the following Tech Spec actions should be taken?

Start at seconds la A.

Declare 1A RHR Pump Inoperable, Unit I enters Tech Spec 3.5.2 Declare D DG Inoperable, Unit 1 and 2 enter Tech spec 3.8.2 B.

Declare 1A RHR Pump Inoperable, no Tech Spec entry required Declare D DG Inoperable, Unit 2 enters Tech Spec 3.8.1 D DG can be returned to Operable if 1A RHR Pump breaker DC Knife Switch opened.

C.

Declare 1A RHR Pump Inoperable, no Tech Spec entry required Declare 1 D RHR Pump Inoperable, no Tech Spec entry requried D DG remains Operable D.

Declare 1A and I D RHR Pumps Inoperable, Unit I enters Tech Spec 3.5.2 Declare D DG Inoperable, Unit 2 enters Tech Spec 3.8.1 Question Data Answer: B Declare 1A RHR Pump Inoperable, no Tech Spec entry required Declare D DG Inoperable, Unit 2 enters Tech Spec 3.8.1 D DG can be returned to Operable if 1A RHR Pump breaker DC Knife Switch opened.

ExplanationlJustification:

A.

B.

3.5.2 U-I SID ECCS is not applicable with one pump out of service. 3.8.2 SID Electrical is not applicable for unit 2.

correct answer, With Unit 1 in Mode 4 the RHR pump can be declared out of with the other pump in the loop still operable and no entry into Tech Specs. Surveillance requirements for Unit 2 DIG operability in Mode 1 requires DG auto-starts from standby condition and: energizes permanently connected loads in < 10 seconds, energizes auto-connected shutdown loads through individual load timers. With the RHR pump DC knife switch open the pump not loading does not inop the DIG for unit 2.

d RHR pump is not inoperable.

3.5.2 U-I S/D ECCS is not applicable with one pump out of service.

C.

D.

Sys#

System Category KA Statement 203000 RHWLPCI: Injection Mode (Plant Specific)

Ability to (a) predict the impacts of the following on the RHWLPCI: INJECTION MODE; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Emergency generator failure KIA#

203000.~2.06 WA Importance 3.9 Exam Level SRO References provided to Candidate None Technical

References:

TM-OP-049 Question Source:

New Level Of Difficulty: (1-5) 4 L,

Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

55.43 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam 89 Given the following conditions:

- Unit 1 is in Mode 4 with Shutdown Cooling in service utilizing the A Residual Heat Removal

- Electrical Maintenance has reported that one of the B RPS MG Set EPA Breakers is running

- They recommend B RPS be transferred to the Alternate Power Supply (RHR) Pump.

hotter than acceptable Which of the following will prevent a loss of Shutdown Cooling (SDC) and addresses the operability issues of opening breakers?

prevented by:

A.

op SDC Outboard Isolation Valve (HV-151-FOO8)

Inboard and Outboard Isolation Valves 3.6.1.3, Condition B.

an Inboard and Outboard Isolation Valves Condition B.

</-"

D.

opening the breaker supplying power to SDC Outboard Isolation Valve (HV-151-F008) and NO LCO is required due to the plant being in Mode 4.

Question Data Answer: 6 opening the breakers supplying power to SDC Inboard and Outboard Isolation Valves (HV-151-F009 and F008) and taking LCO 3.6.1.3, Condition B:

ExplanatiorVJustification:

A.

6. Correct, allowed by procedure C.

D.

Incorrect, loss of B RPS cause both F008 and FOO9 to close.

Incorrect, not an instrumentation LCO ncorrect, loss of B RPS cases both F008 and FOO9 to close, must take LCO Sys #

System Category KA Statement 205000 Shutdown Cooling System Ability to monitor automatic operations of the Pump trips (RHR Shutdown Cooling SHUTDOWN COOLING SYSTEMlMODE including:

Mode)

KIA#

205000.~3.02 KIA Importan-7 Exam Level References provided to Candidate Technical

References:

OP-149-002 Question Source:

New Level Of Difficulty: (1-5) 4 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 0611 9/03

SSES LOC 19 NRC Exam

\\..-

90 Given the following conditions with Unit I at 100% power:

- Core Spray Loop A Header Break Detection High Differential Pressure alarm has j

- The Reactor Building NPO reports PDIS-E21-1 N004A is reading -3.4 psid on Pan received Which of the following actions are required for these conditions?

A.

Declare Core Spray A header d/p instrumentation channel inope Operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

ko*

B.

Declare Core Spray Loop A Inoperable. Restore to Operable status within7hys.

C.

Write an AR to document the out-of-specification differential pressure condition.

D.

Write a tracking LCO to document the systems ability to inject inside the vessel but not spray above the core.

Question Data Answer: A Declare Core Spray A header dIp instrumentation channel inoperable. Restore to Operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Explanation/Justification:

A.

Correct.

6.

C.

D.

.L-Incorrect. CS still operable, dIp still good, bad alarm Incorrect. DIP still good, AR should be for incorrect alarm received Incorrect. CS still capable of performing its intended function Sys#

System Category KA Statement 209001 Low Pressure Core Spray System Ability to (a) predict the impacts of the following on the LOW PRESSURE CORE SPRAY SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Core spray line break WA#

2o9001.~2.05 WA Importance 3.6 Exam Level SRO Technical

References:

TRO 3.5.2, Pages 3.5-3 through 3.5-5 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Comprehension 10 CFR Part 55 Content:

55.43 References provided to Candidate Question Source:

New TRO 3.5 NRC 2003 Rev I H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC I 9 NRC Exam I.

91 Given the following conditions with Unit 2 in Mode 1 :

- Surveillance SO-024-001, Monthly Diesel Generator Operability Test, is being performed for

- The 2A201 bus is currently loaded at 800 W

- The A DG is operating at the SO hold point of 1000 KW.

- The Supply Breaker to 2A201 from Transformer OX201 trips due to a breaker problem.

the A DG to the 2A201 bus.

Which of the following describes the expected electric plant response and the Unit Supervisor directed actions for these conditions?

The A Diesel Generator Output Breaker:

A.

trips and the US should direct resetting the 2A201 bus lockouts and power restoration from its alternate power source.

B.

trips and the US should direct verification of 2A201 automatically re-energizing from its alternate power source.

C.

does NOT trip and the US should direct restoration of normal bus voltage and frequency parameters.

L.

D.

does NOT trip and the US should direct an immediate trip of the Diesel Generator Question Data Answer: C does NOT trip and the US should direct restoration of normal bus voltage and frequency parameters.

ExplanationlJustification:

A.

B.

C.

D.

Incorrect, breaker does not trip, no lockouts trip, bus remains energized Incorrect, breaker does not trip, bus remains energized Correct, a loss of 200 KW will show up as changes in frequency and voltage, return parameters to normal bands.

Incorrect, DG can handle this transient without breaker or engine tripping, no reason to direct trip

~

~

Sys#

System Category KA Statement 264000 Emergency Generators (DiesellJet)

Ability to (a) predict the impacts of the following on the EMERGENCY GENERATORS (DIESEUJET); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Loss of AC. power WA#

264000.A2.09 WA Importance 4.1 Exam Level SRO References provided to Candidate None Technical

References:

Electrical Theory and Question Source:

New Level Of Difficulty: (1-5) 4 application Question Cognitive Level:

Comprehension 10 CFR Part 55 Content:

55.43 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam I

92 Unit 1 is at rated conditions with no LCOs entered.

Electrical Maintenance has submitted a work package to overhaul the HV-151 -F006A Shutdown Cooling Suction Valve motor actuator.

The work package requires the HV-151-FO04A to be closed to allow manual operation of the HV-151-FO06A to set the valve limit switches.

What Limiting Conditions of Operation will be in effect?

B.

LCO 3.4.8, LCO 3.4.9, LCO 3.5.1, LCO 3.6.1.3 D. %v9, LCO 3.6.2.3, LCO 3.6.2.4 Question Data Answer: C LCO 3.5.1 ExplanationlJustification:

A.

B.

C.

D.

SDC not required at rated conditions.

SDC not required at rated conditions.

correct answer, to open the 06, the 04 suction valve must be closed, taking 'A' RHR pump out of service.

ECCS LCO required for RHR pp out of service.

-6 Sys#

System Category KA Statement WA#

219000.2.2.24 KIA Importance 3.8 Exam Level SRO Question Source:

New Level Of Difficulty: (1-5) 4 References provided to Candidate TS 3.4,3.5 & 3.6 Technical

References:

Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

55.41 NRC 2003 Rev I H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam 93 With the plant in Mode 5 and the Refueling Platform approaching the reactor vessel, which of the following will initiate a rod withdraw block for any selected rod?

A.

Grapple NOT engaged.

B.

Any Hoist extended.

C.

D.

Any Hoist Loaded.

Any rod NOT fully inserted.

Question Data Answer: D Any Hoist Loaded.

ExplanatiodJustification:

A.

B.

C.

D.

Correct answer, grapple not engaged will not cause a rod block.

no interlock for hoist extended, everything uses load cells.

1 rod allowed full out.

Sys#

System Category KA Statement 234000 Fuel Handling Equipment Ability to monitor automatic operations of the FUEL Interlock operation HANDLING EQUIPMENT including:

KIA##

234000.~3.02 WA importance Exam Level SRO Question Source:

Modified Nine Mile Point 1,1996 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

55.43 L--

References provided to Candidate None Technical

References:

TM-OP-56 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForrn.doc Printed on 06/19/03

SSES LOC I 9 NRC Exam

• \\lkV 94 Unit 2 is operating at 100% power when fuel damage occurs, the manual scram fails. %' &@

The following plant conditions exist:

u

- Reactor power 18%

- Reactor pressure 940 psig

- RPV water level

- Main Steam Line 6

- Main Turbine Tripped

- Site boundry release (Adult Thyroid)

-100 inches Inboard & Outboard MSlVs failed open 4.90 Rem(rising)

Given the above conditions, which of the following actions are required?

A.

Use the SRVs to maintain reactor pressure less than 965 psig.

B.

Use Main Turbine BPVs to commence a reactor cool down at less than a 9OF/Hr rate.

C.

Use Main Turbine BPVs to commence a reactor cool down at greater than a 9OF/Hr rate.

D.

Perform Emergency RPV Depressurization.

U'

~

~

~~~~

Question Data Answer: D Perform Emergency RPV Depressurization.

ExplanationlJustification:

A.

B.

C.

D.

EO-200-1 05 rad release EO, leads to EO-200-1 02 RPV level control but back out to the level/pwer control EO therefore there is no cooldown allowed.

EO-200-105 rad release EO, leads to EO-200-102 RPV level control but back out to the levellpwer control EO therefore there is no cooldown allowed.

EO-200-105 rad release EO, leads to EO-200-102 RPV level control but back out to the Ievellpwer control EO therefore there is no cooldown allowed.

correct answer, MSlVs failing to close is a "Primary system" discharging. The ATWS is being controlled by EO-200-1 13 with level being maintained at -100" as required by the procedure.

Sys#

System Category KA Statement Equipment Control '-7 Knowledge of refueling administrative requirements.

WA Importance 4.2 Exam Level SRO Technical

References:

EOP Flow charts Fermi 2 2,2001 Level Of Difficulty: (1-5) 3 Analysis 10 CFR Part 55 Content:

55.43 to Candidate None Question Source:

Modified Question Cognitive Level:

NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForrn.doc Printed on OW1 9/03

SSES LOC I 9 NRC Exam L

95 In accordance with Unit 1 Tech Specs and ON-I 83-001, Stuck 06 Safety Relief Valve, the Reactor Mode Switch was placed in Shutdown at 108 degrees F due to a stuck open SRV.

Post-scram, the Suppression Pool reached a peak of 114 degrees F before Suppression Pool Cooling was able to begin removing heat. The reactor was NOT required to be placed in Mode

4.

s v

Which of the following describes the restrictions on the ensuing reactor startup? Assume the SRV has been repaired.

Suppression Pool temperature must be less than or equal to:

A.

105 degrees F prior to placing the Reactor Mode Switch in Startup/Standby.

B.

90 degrees F prior to placing the Reactor Mode Switch in Startup/Standby.

C.

90 degees F prior to exceeding 1 % power.

D.

105 degrees F prior to exceeding 1 % power.

Question Data Answer: C ExplanationlJustification:

A.

B.

C.

D.

90 degees F prior to exceeding 1% power.

~ - '

Incorrect, 105 limit only applies >I%

power with testing in progress Incorrect, no requirement for this since the reactor remains in one of the three modes for which the LCOs apply at all times Correct, in Modes 1, 2 and 3 this limit applies once >I%

power.

Incorrect, 105 limit only applies >I%

power with testing in progress.

Sys#

System Category KA Statement Conduct of Operations Knowledge of less than one hour technical specification action statements for systems.

Exam Level SRO References provided to Technical

References:

TS 3.6.2.1 K/A#

2.1.11 Question Source:

New Question Cognitive Level:

Level Of Difficulty: (1-5) 3 10 CFR Part 55 Content:

55.43

\\-

NRC 2003 Rev I H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 0611 9/03

SSES LOC 19 NRC Exam

\\.--

96 Which of the following is/are considered an Unreviewed Safety Question?

1. Emergency actions that depart from T.S. are needed to protect the public health and safety.

2. The possibility of an accident exists that has not been evaluated by the FSAR

3. The consequence of a malfunction of equipment evaluated by the FSAR is increased.

4. The margin of safety as defined in T.S. is reduced.

5. An emergency event that can not be classified by the Emergency plan.

A.

2, 3, 4 B.

1, 2, 3

c.

3,4,5 D.

2, 3, 5 Question Data Answer: A 2,3,4 ExplanatiodJustification:

A.

correct answer, Not USQ - Emergency actions that depart from T.S. are needed to protect the public health and safety.

Is USQ - The possibility of an accident exists that has not been evaluated by the FSAR Is USQ - The consequence of a malfunction of equipment evaluated by the FSAR is increased.

Is USQ - The margin of safety as defined in T.S. is reduced.

Not USQ -An emergency event that can not be classified by the Emergency plan.

v B.

1, not USQ C.

5, not USQ D.

5, not USQ WA#

References 2.2.8 provided to Candidate/

WA Importance 3.3 Question Source:

Modified a Gen Sta 2, I999 Question Cognitive Level:

Compreh sion Sys#

System Category KA Statement Equipment Control Knowledge of the process for determining if the proposed change, test, or experiment involves an unreviewed safety question.

Exam Level SRO Technical

References:

Level Of Difficulty: (1-5) 3 10 CFR Part 55 Content:

55.43 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC I 9 NRC Exam 97 Which of the following is the bases for the Technical Specifications, Minimum Suppression Chamber Water Volume in Operation Conditions 1, 2, and 3?

A.

Ensures a sufficient supply of water is available, with the Minimum CST Volume in the event of a LOCA to permit recirculation cooling flow to the core.

B.

Ensures a sufficient amount of water would be available to adequately condense the steam from the SRV tailpipe break above the suppression pool level.

C.

Ensures a sufficient amount of water would be available to adequately condense the steam from the SRV discharges, downcomers, or HPCl and RClC turbine exhaust lines and provide emergency make-up to the reactor vessel.

D.

Provides sufficient supply of water that, with the Minimum CST Volume, Long Term Cooling is available for the design basis accident.

Question Data Answer: C Ensures a sufficient amount of water would be available to adequately condense the steam from the SRV discharges, downcomers, or HPCl and RClC turbine exhaust lines and provide emergency make-up to the reactor vessel.

ExplanationNustification:

A.

6.

C.

CST is not part of basis.

HPCl & RClC are included in the basis.

correct answer, If the suppression pool water level is too low, an insufficient amount of water would be available to adequately condense the steam from the SRV discharges, downcorners, or HPCl and RClC turbine exhaust IinesWLow suppression pool water level could also result in an inadequate emergency makeup water source to the Emergency Core Cooling SystemWThe lower volume would also absorb less steam energy before heating up excessively. Therefore, a minimum suppression pool water level is specified.

CST is not part of basis.

D.

~

~

~

WStatemZt

~

Sys#

System Category Equipment Control Knowledge of bases in technical specifications for limiting conditions for operations and safety limits.

WA#

2.2.25 WA Importance 3.7 Exam Level SRO References provided to Candidate None Technical

References:

TS Basis - B 3.6.2.2 Question Source:

Modified Fermi 2 2,1998 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Fundamental 10 CFR Part 55 Content:

55.43 NRC 2003 Rev I H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC I 9 NRC Exam 98 Maintenance must be performed in the Unit 1 Reactor Water Cleanup (RWCU) Backwash k Room (Room 1-509) to support a modification.

Per NDAP-QA-0323, Standard Blocking Practices, and NDAP-QA-0626, Radiologically Controlled Area Access and Radiation Work permit (RWP) System, which of the following actions is required BEFORE the Unit Supervisor may allow work to begin in the room to comply with the ALARA BLOCKING principle?

A.

Flush, then drain the Backwash Receiving Tank and maintain the tank empty.

B.

Backwash the RWCU filters, then drain the filters and maintain the filters empty.

C.

Flush and drain, then fill the Backwash Receiving Tank and maintain the tank full.

D.

Backwash the RWCU filters, then fill the tank and maintain the tank full.

Question Data Answer: C Flush and drain, then fill the Backwash Receiving Tank and maintain the tank full.

ExplanatiodJustification:

A.

B.

C.

D.

The tanks must also be filled with water to act as shielding.

This would not help the radiation levels in the Backwash Receiving Tank Rooms and may make it worse since the filters are backwashed to the room.

correct answer, NDAP-QA-0323, Standard Blocking Practices requires the tanks be flushed drained and filled before entry into the rooms.

This would not help the radiation levels in the Backwash Receiving Tank Rooms and may make it worse since the filters are backwashed to the room.

-x-Sys#

System Category K/A#

2.3.10.

WA Importance 3.3 Exam Level SRO References provided to Candidate None Technical

References:

NDAP-QA-0626, Sect. 6.2.5 Question Source:

Modified Susquehanna, 2001 Level Of Difficulty: (1-5) 3 Question Cognitive Level:

Memory 10 CFR Part 55 Content:

55.43 NRC 2003 Rev I H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam

J 99 Maintenance has just reported to the Control Room that a 55 gallon drum of lube oil leaked into the River Water Makeup intake bay overnight.

Using the attachments from NDAP-QA-0720, STATION REPORTABILITY EVALUATION GUIDANCE, RT MATRIX AND the offsite agencies that must be notified after PA DEP is notified.

A.

LCEMA, Coast Guard, NRC B.

CCEMA, Coast Guard, NRC C.

LCEMA, Coast Guard, PEMA D.

LCEMA, CCEMA, PEMA Question Data Answer: A LCEMA, Coast Guard, NRC Explanation/Justification:

A.

B.

NRC must be notified.

C.

D.

correct answer, PA DEP, LCEMA, and Coast Guard must be notified due to a petroleum product being discharged to a waterway.

The NRC must be notified anytime an offsite agency is notified.

Correct for a Comprehensive Environmental Response, Compensation, and Liability Act release which this spill is NOT.

Columbia county not notified for this spill not PA Emergency Management Agency.

u

~-

Sys#

System Category KA Statement Knowledge of which events related to system operationslstatus should be reported to outside agencies.

Emergency Procedures and Plan W M 2A.30 KIA Importance 3.6 Exam Level SRO References provided to Candidate NDAP-QA-0720 Att Q &

Technical

References:

NDAP-QA-0720 T

Question Source:

New Level Of Difficulty: (1 -5) 4 Question Cognitive Level:

Comprehension 10 CFR Part 55 Content:

55.43 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03

SSES LOC 19 NRC Exam

• c 100 The Mode Switch has been placed in Shutdown.

Reactor Power is 15%.

Reactor Water Level is lowering at 1 inch per minute.

Drywell pressure is 0.75 psig and slowly rising.

The Unit Supervisor is about to give direction from EO-I 00-1 13, LEVEUPOWER CONTROL to inhibit ADS and bypass MSlV and CIG interlocks, why?

A.

Current level control range is near the auto initiation of ADS which could allow uncontrolled low pressure injection and main condenser D s e d as a heat sink for as long as possible.

B.

Initiation of ADS and/or closure of the MSlVs would cause large pressure changes. GE calculations indicate swings in pressure cause large power oscillations.

C.

Initiation of ADS would cause large pressure changes with resulting power oscillations.

MSlVs must remain open for use of the main condenser as a heat sink.

D.

MSlVs must remain open for use of the main condenser as a heat sink. CIG must remain available for later use of ADS valves.

Question Data Answer: A Current level control range is near the auto initiation of ADS which could allow uncontrolled low pressure injection and main condenser is used as a heat sink for as long as possible.

ExplanationlJustification:

A.

B.

C.

D.

correct answer, Prevent low pressure unborated injection and prevent MSlVs from clossing due to a loss of pneumatic supply.

ADS would depressurize the RPV.

ADS would depressurize the RPV.

CIG not required, 2200# bottles supply header.

Sys#

System Category Emergency Procedures and Plan KA Statement Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.

UJA#

2.4.49 WA Importance 4.0 Exam Level SRO References provided to Candidate None Technical

References:

EO-000-113 Question Source:

New Level Of Difficulty: (1 -5) 4 Question Cognitive Level:

Analysis 10 CFR Part 55 Content:

55.43 NRC 2003 Rev 1 H:\\NRCExamPrep\\Rich\\NRCAForm.doc Printed on 06/19/03