2010 NRC RO Exam

Question 1

1. Which one of the following describes the response of a Relief/Safety Valve to a rising RPV pressure if a pneumatic supply is NOT available from any source?

A Relief/Safety Valve will …

A. function in its pressure relief mode and in its safety mode.
B. NOT function in its pressure relief mode but will function in its safety mode.
C. function in its pressure relief mode and NOT function in its safety mode.
D. NOT function in its pressure relief mode and NOT function in its safety mode.

Answer 1

A: Incorrect: A pneumatic supply (accumulator I continuous) is required to lift in the Pressure Relief mode.

B: Correct: A pneumatic supply is not required in the Safety mode. In the safety mode, whenever reactor pressure is greater than spring pressure the valve will open.

Question 2

2. While at rated conditions, a fire in the Relay Room results in entry into N2-S0P-78, Control Room Evacuation. The following occur:

• All SOP-78 Control Room actions are completed before evacuating the Control Room
• Before any action is taken outside the Control Room, the Automatic Depressurization System (ADS) spuriously initiates and all ADS valves open.
• Shortly thereafter all appendix R disconnect switches on panels 2CES*PNL415, 416 and 417 are placed in the actuate position.

Which one of the following describes the response of the ADS valves?

A. All ADS valves will close when the disconnect switches are placed in actuate.
B. All ADS valves remain open and cannot be closed unless solenoid fuses are removed.
C. The four ADS valves controlled from Remote Shutdown Panel, 2CES*PNL405 close. The remaining three will not close until solenoid fuses are removed.
D. The four ADS valves controlled from Remote Shutdown Panel, 2CES*PNL405 will close when their switches are taken out of the NORMAL position. The remaining three will not close until solenoid fuses are removed.

Answer 2

A: Correct: Per N2-0P-78, Remote Shutdown System Attachment 2, disconnect switches SW1-2CESA02 (Div 1) and SW1-2CESB02 (Div 2) isolate solenoids for the 4 ADS valves controlled from the remotes shutdown panel. Switches SW1-2CESA04 (Div 1) and SW1-2CESB04 (Div 2) isolate remainder of ADS circuitry.

B: Incorrect: Per N2-0P-78, Remote Shutdown System Attachment 2, disconnect switches SW1-2CESA02 (Div 1) and SW1-2CESB02 (Div 2) isolate solenoids for the 4 ADS valves controlled from the remotes shutdown panel. Switches SW1-2CESA04 (Div 1) and SW1-2CESB04 (Div 2) isolate remainder of ADS circuitry.

C & D: Incorrect: When the disconnect switches are repositioned, the ADS solenoids lose power and the valves close. See A & B above.

Question 3

3. Which one of the following power supply failures would result in a loss of the RPS “A” Analog Trip Units on panel 2CEC*PNL609?

Loss of …

A. 2VBB-UPS1A.
B. RPS UPS 2VBB-UPS3A.
C. RPS MG set 2RPM-MG1A.
D. Emergency UPS 2VBA*UPS2A.

Answer 3

B: Correct: RPS UPS 2VBB-UPS3A is the power supply

Question 4

4. The plant is in Mode 4 when 600V distribution panel 2NJS-PNL600 is de-energized.
   Assuming no operator actions which one of the following is the affect of this loss on the Source Range Monitors (SRM)?

A. SRM Channels A and C de-energize immediately.
B. SRM Channels A and C de-energize after several hours.
C. SRM Channels Band D de-energize immediately.
D. SRM Channels Band D de-energize after several hours.

Answer 4

D: Correct: 2NJS-PNL600 powers chargers 2BWSCHGR3B1 and 2BWS-CHGR3D1. Although both chargers are lost, the SRMs remain energized via 2BWS-BAT3B/3D. These batteries are sized to maintain voltage for up to 4 hours before voltage begins to decay. If the chargers are not restored, SRMs will eventually de-energize.

Question 5

5. The plant is at rated conditions when a complete loss of Normal 125 VDC Power bus 2BYS-SWG001A occurs.

Which one of the following is a direct plant response to this event?

A. Main turbine immediately trips
B. BOTH Recirc pumps immediately trip
C. Loss of Main Generator protective relays control power
D. Loss of control power to 2NPS-SWG001/2/3 load breakers

Answer 5

C: Correct: The loss of 2BYS-SWG001A causes a loss of Main Generator protective relays control power and results in Annunciator 852604, Generator protective relays control power failure.

Question 6

6. A plant startup in progress with the following:

• Reactor power is 30%
• The Rod Worth Minimizer (RWM) is in the Transition Zone Then, the UPS supply breaker to the RWM trips.

Other than scramming the reactor which one of the following describes the current control rod movement capability?

A. Rods can be withdrawn OR inserted using the reactor manual control system.
B. Rods CANNOT be withdrawn OR inserted using the reactor manual control system.
C. Rods can only be inserted if the Continuous Insert push button is used. Rods CANNOT be withdrawn using any method.
D. Rods can only be inserted if the Continuous Insert push button is used. Rods can only be withdrawn if the Continuous Withdraw mode is used.

Answer 6

B: Correct: Loss of power to RWM removes permissive to move rods.

Question 7

7. The following conditions exist:

• The reactor is in Mode 3.
• RHR Loop “B” has just been placed in Shutdown Cooling
• Reactor pressure has risen to 150 psig

Which one of the following would indicate that a failure had occurred in the automatic response of the RHR system?

A. RHR*P1 B pump is running
B. RHS*MOV2B, PMP 1 B SDC SUCT VLV is open
C. RHS*MOV40B, SDC B RETURN THROTTLE is closed
D. RHS*MOV8B, HEAT EXCHANGER 1B INLET BYPASS VALVE is in mid position (throttled)

Answer 7

A: Correct: The pump should have tripped on loss of suction flowpath when MOV112 and 113 closed.

Question 8

8. The plant is at rated conditions when the following indications are received.

• Annunciator 601706, HPCS System INOP alarms
• Amber System Status Light “HPCS Line Break” illuminates

Which one of the following describes the impact if a Standby Liquid injection is required? If the system is actuated Standby Liquid will inject into the …
A.
core shroud region of the reactor vessel.
B.
downcomer region of the reactor vessel.
C.
drywell and will NOT reach the reactor vessel.
D.
core shroud region but NOT the HPCS spray header.

Answer 8

B: Correct: Alarm annunciates on high pressure when CS injection line pressure rises. This occurs when the CS injection line rises to downcomer pressure when the line breaks outside the shroud but inside the vessel. SLS injection would still be via the downcomer.

Question 9

9. The plant was at rated condition when the following occur:

• A complete loss of off-site power occurred
• Emergency Diesel Generators (EDGs) start and re-power their respective buses.
• Ten minutes later, all Division I 125 VDC is lost.

Which one of the following is the current status of 2EGS*EG1 and its output breaker 101-1?

EGS*EG1 ; BRKR 101-1
A. Tripped ; Open
B. Tripped ; Closed
C. Running ; Open
D. Running ; Closed

Answer 9

B: Correct: Loss of DC control power will de-energize the EDG trip relays tripping the diesel, and breaker 101-1 will not open due to no power to its trip coil.

Question 10

10. A reactor plant startup is being conducted in accordance with N2-0P-101A.

• The reactor is critical and SRM/IRM overlap data has just been completed.
• All SRMs are reading between 5 X 10E4 and 1 X 10E5 cps
• All IRMs are on mid scale on range 1
• The operator has selected both the SRMs and the IRMs for withdraw.

Which one of the following will be the first automatic protective action as the detectors are withdrawn?

A. SRM INOP trip
B. IRM Downscale rod block
C. SRM Downscale rodblock
D. IRM Detector NOT fully inserted rod block

Answer 10

D: Correct: When the withdraw function is selected the first action will be the movement of the SRM and IRM detectors from the core. Immediately the IRMs will be detected to be not fully inserted This rod block is active whenever the mode switch is not in run.

Question 11

11. A plant startup is in progress, with the following:

• Feedwater Pump A is in service
• All three Feedwater Pump Suction MOVs, CNM-MOV84A, B and C, are open
• Low Flow Control Valve 2FWS-LV55A is controlling RPV level in AUTO
• 2FWS-LV55A is currently 40% open

Then, instrument air to ALL Feedwater valves is lost

Which one of the following describes the effect on 2FWS-L V55A position and reactor water level?

2FWS-LV55A Position ; Reactor Water Level
A. Fails closed ; Lowering, below normal level
B. Fails as is ; Constant, normal level
C. Fails as is ; Lowering, below normal level
D. Fails open ; Rising, above normal level

Answer 11

C: Correct: Low Flow Control Valve 2FWS-LV55A fails as is on loss of air. Feed pump min flow valves fail open on loss of air, diverting water from the reactor resulting in a lowering level.

Question 12

12. The plant is at rated conditions when a complete loss of pneumatics to 2CCP-TV108, RB CLOSED LOOP COOLING TEMPERATURE CONTROL VALVE, occurs.

Which one of the following describes the effect of this failure on the CCP heat exchanger?

A. MORE flow is directed to SHELL side of heat exchanger.
B. MORE flow is directed to TUBE side of heat exchanger.
C. LESS flow is directed to SHELL side of heat exchanger.
D. LESS flow is directed to TUBE side of heat exchanger.

Answer 12

A: Correct: On a loss of air, 2CCP-TV1 08 positions itself to provide maximum cooling. The valve in the heat exchanger discharge header will fail open and the valve in the heat exchanger bypass line fails closed, which directs MORE flow through the SHELL side of the heat exchanger.

Question 13

13. Following a small break on a Recirc Loop the following conditions exist:

• Low Pressure Core Spray is the only injection system available
• RPV water level has been restored
• Low Pressure Core Spray Injection Valve, 2CSL *MOV1 04, is throttled and Core Spray Flow is 2000 gpm

Then, the recirc loop under goes a double guillotine shear. In response to the lowering RPV level the operator fully opens 2CSL*MOV104.

Per design, Core Spray injection will:
A. Raise level to the level of the Main Steam Lines
B. Stabilize level at approximately -14 inches (actual level)
C. Stabilize level at approximately -62 inches (actual level)
D. Be insufficient to recover level to above the Bottom Of the Active Fuel

Answer 13

C: Correct: -62 inches is the level of the jet pump suctions. During a design bases accident, one ECCS is expected to re-flood to the top of the Jet pump suctions.

Question 14

14. Following a reactor scram and loss of off-site power, RCIC is manually started to control RPV level. Current conditions are:

• RPV level is 180 inches and slowly rising
• RCIC Flow Controller 2ICS*FC101 is in Auto
• RCIC Flow Controller 2ICS*FC101 setpoint has been reduced to 400 gpm.

In order to control RPV level, ICS*MOV124, Test Bypass to Condensate Storage Tank, is opened and ICS*FV108, Test Bypass to Condensate Storage Tank, is throttled in mid position.

Which one of the following are the responses of RPV injection flowrate and RCIC Flow indication on RCIC Flow Controller 2ICS*FC101, if ICS*FV108 is opened further?

RPV injection flowrate ; RCIC Flow Controller indication
A. Lower ; Remain the same
B. Lower ; Lower
C. Raise ; Remain the same
D. Raise ; Raise

Answer 14

A: Correct: Further opening of the Test Bypass will divert additional flow to the CST. The controller sees the increase in flow and closes down on the turbine, keeping flow at 400 gpm.

Question 15

15. The plant is operating at 100% power with a normal Reactor Building ventilation configuration. Reactor Building differential pressure is -0.6 inches WG when the following sequence of events occur:

1300: Drywell pressure rises to 14 psig
1300: GTS Trains A and B automatically start
1310: GTS Train A is shutdown by placing Train A Initiation control switch in Auto After Stop
1315: GTS Train B Fan trips due to a blown control power fuse

Which one of the following describes the impact of these events on Reactor Building differential pressure (DP) and the actions required, per N2-0P-61 B, to restore RB differential pressure?

Reactor Building DP ; Actions Required
A. Becomes less negative ; GTS Train A must be manually restarted
B. Becomes less negative ; Confirm automatic restart of GTS Train A
C. Becomes more negative ; Defeat high Drywell pressure interlocks and restart HVR
D. Becomes more negative ; Confirm automatic restart of GTS Train A

Answer 15

B: Correct: GTS Train A restarts because the high Drywell pressure (>1.68 psig) initiation signal is still present and with no running GTS train RB differential pressure will degrade toward 0. When pressure reaches -0.25 inches WG, the manually shutdown train will automatically restart.

Question 16

16. Following a loss of coolant accident and loss of off-site power the following conditions exist:

• RHR Loops A and B are injecting
• NO other ECCS injection sources are available
• Reactor pressure is 100 psig and steady
• Reactor level is 50 inches and steady

While in these conditions RHR Pump A trips and cannot be restarted.

Which one of the following describes an Alternate Injection System that N2-EOP-6 Attachment 6 specifies to re-establish level control?

A. Fire water cross tied to RHR Loop B
B. Service Water cross tied to RHR Loop B
C. Fire water cross tied to RHR Loop A
D. Service Water cross tied to RHR Loop A

Answer 16

C: Correct: Attachment 6 provides direction to align firewater provided the loop is currently NOT being utilized. Reactor pressure is also low enough to allow firewater injection.

Question 17

17. With the plant at rated conditions the following occur:

• Drywell pressure rises to 2 psig
• The Standby Diesel generators start and run unloaded.

Two minutes later the following 2EGS*EG1 indications are observed in the control room:

• Engine RPM: 608 RPM
• Generator VARs: 0 VARS
• Generator AC output voltage: 4170 VAC

Given these indications, which one of the following is correct?

A. The Auto Voltage Regulator has failed and the voltage regulator has shifted to manual.
B. The Primary Control Circuit (Control Circuit #1) has failed and the engine is operating in Parallel Mode.
C. The electronic speed control governor has failed and the hydraulic governor is controlling engine speed.
D. A complete failure of the electronic-hydraulic control governor has occurred and engine speed is being controlled by the overspeed governor.

Answer 17

C: Correct: A failure of the electronic speed control governor will shift control to the hydraulic governor which will control the engine at -608 RPM. The frequency indicated is the frequency associated with a 12 pole engine operating at 608 RPM.

Question 18

18. Following a plant transient, High Pressure Core Spray (HPCS) auto initiates and injects. HPCS recovers level and injection flow stops when CSH*MOV107, Pump 1 Injection Valve, closes on RPV high level. Current plant conditions are as follows:

• RPV level is +180 inches and lowering
• Drywell pressure 1.2 psig and rising
• No operator actions have been taken up to this point.

Then Drywell pressure rises above 2 psig.

Given the above, which one of the following is correct regarding the response of CSH*MOV1 07 and the required actions, if ANY, to restore injection at this time?

CSH*MOV107 will… , Required Actions

A. Automatically open, None
B. Remain closed, Depress the Hi WTR Level Seal-In Reset
C. Remain closed, Depress the Manual Initiation Seal-In Reset
D. Remain closed, Place the control switch for CSH*MOV107 to the OPEN position

Answer 18

A: Incorrect: CSH*MOV107 is interlocked closed until level lowers to level 2 or the high water level reset is depressed.
B: Correct: CSH*MOV107 will re-open if high level reset is depressed.
C: Incorrect: The initiation will reset but CSH*MOV107 will not re-open until level lowers to level 2.
D: Incorrect: CSH*MOV107 cannot be opened manually until the high level is reset.

Question 19

19. The reactor is at 100% power when the Key Lock switch on LPRM Chassis 1 is placed in the INOP position.

How will this affect the Input Status lights for APRM 1 on Two-Out-Of-Four Logic Module, MDL 1?

A. All status lights will remain extinguished.
B. The OPRM status light will illuminate. All others will be extinguished.
C. The HIGHIINOP status light AND the OPRM status light will illuminate.
D. The HIGHIINOP status Ught will illuminate. All others will be extinguished.

Answer 19

C: Correct: Per N2-0P-92 precaution 15, Placing an APRM slave (LPRM chassis) keylock switch in the INOP position will be detected by the APRM master and place the APRM channel in INOP Gust as if the APRM keylock switch has been placed in INOP). This will trip the OPRM input to the voter as well.

Question 20

20. A plant Startup is in progress with the following:

• All APRMs are reading 6%
• The Reactor Mode Switch is in STARTUP with final checks in progress for placing the mode switch to RUN.

Which one of the following identifies the effect, if any, of placing the IRM A Mode Switch (S1) in the Standby position on panel 2CEC*PNL606?

A. Nothing happens at this power level.
B. Annunciator IRM UPSCALE / INOPERABLE alarm only.
C. Control Rod block and Annunciator IRM UPSCALE / INOPERABLE alarm only.
D. Half scram, Control Rod block and Annunciator IRM UPSCALE / INOPERABLE alarm.

Answer 20

D: Correct: Taking the mode switch out of operate generates a half scram and rod block

Question 21

21. Primary Containment Isolation Valve testing is being conducted on containment purge AOVs in accordance with N2-OSP-CPS-Q001, Primary Containment Purge System Valve Operability Test.

Which one of the following would be the correct method for determining whether an AOV’s closure time was acceptable for a fully open AOV?

Start the timing when the … Stop the timing …

A. control switch is taken to the close ; when the green closed indication position illuminates
B. green closed indication illuminates ; when the red open indication extinguishes
C. control switch is taken to the close ; when the red open indication extinguishes position
D. green closed indication illuminates ; two seconds after red open indication extinguishes

Answer 21

A: Incorrect: This is the timing closed method for a CPS SOV.

Correct: Sect. 4.2.1 Measuring Valve Stroke times for all valves except Solenoid Operated Valves.
• Measure opening stroke time from the time the control switch is placed to OPEN until the green indicating light de-energizes.
• Measure closing stroke time from the time the control switch is placed to CLOSE until the red indicating light de-energizes.

Question 22

22. Which one of the following describes the electrical interlock with RHS*MOV2A, PMP 1A SDC SUCT VL V, and the reason for that interlock?

If RHS*MOV2A is OPEN then …

A. RHS*MOV24A, LPCI A Injection Valve, will not open to prevent bypassing the recirculation loop.
B. RHS*MOV4A, RHR Pump 1A Minimum Flow Valve, will not open to prevent draining the RPV to the suppression pool.
C. RHS*FV38A, RHR A Return to the Suppression Pool, will not open to prevent draining the RPV to the suppression pool.
D. RHS*MOV24A, LPCI A Injection Valve AND RHS*MOV40A, SDC A Injection valve, cannot both be opened at the same time to prevent running out the RHR pump.

Answer 22

C: Correct: If in SDC none of the suppression pool cooling Ispray valves can be opened.

Question 23

23. Following a Reactor scram the RO places the mode switch to Shutdown and commences inserting all IRMs and SRMs.

Which one of the following SRM alarms will occur?

A. • SRM Downscale
• SRM Detector Position Abnormal
B. • SRM Short Period
• SRM Upscale/Inoperable
C. • SRM Short Period
• SRM Detector Position Abnormal
D. • SRM Downscale
• SRM Upscale/Inoperable

Answer 23

B: Correct: An upscale trip is expected as the detectors are driven into the core and move into high flux regions. A short period is also expected. The detectors interpret the rise in counts as they are inserted as a short period.

Question 24

24. The plant is at rated power when the following occur:

• 2ENS*SWG101 de-energizes when Reserve Transformer 2 RTX-XSR1A locks out
• Diesel Generator 1 (EG 1) automatically starts
• EG 1 output breaker, 101-1, does NOT automatically close

Which one of the following is the potential impact of immediately attempting to manually close EG 1 output breaker, 101-1 and what actions are required by N2-S0P-3, Loss of AC Power, prior to manually closing the breaker?

Impact of Immediately Closing 101-1 ; Required Actions

A. Overloading EG 1 ; Place selected Bus loads in Pull to Lock ONLY
B. Energizing a faulted bus ; Verify there are no faults on the Bus ONLY
C. Overloading EG 1 ; Place selected Bus loads in Pull to Lock AND Place Bus 101 Synchronizing Switch to On
D. Energizing a faulted bus ; Verify there are no faults on the Bus AND Place Bus 101 Synchronizing Switch to On

Answer 24

D: Correct: A potential fault on 2ENS*SWG101 would prevent EG 1 from automatically closing on the bus. SOP-3 requires the operator to verify the bus is not faulted by verifying the bus fault annunciators (852147 and 852148) are clear. With EG 1 already running the procedure directs placing the Synch Switch in on and then manually closing the breaker.

Question 25

25. The reactor is at 100% power and 100% core flow when the following occurs:

• Main Turbine trip.
• RPS trips but NO rod insertion occurs.
• After the initial RPV pressure response, RPV pressure is stabilized between 970 psig to 1000 psig.

Which one of the following describes the amount of reactor power that is being generated 30 to 45 seconds after the transient begins and why power is lower than 100%?

(Assume that all systems, except rods, respond as designed but there continues to be zero rod motion and no operator actions have taken place other than stabilizing RPV pressure).

Approximate Reactor Power ; Basis for Power Lowering

A. Within Bypass Valve capacity ; Recirc pumps are at slow speed Recirc pumps are at slow speed
B. Greater than Bypass Valve capacity ; Recirc pumps are at slow speed Recirc pumps are at slow speed
C. Within Bypass Valve capacity ; Recirc pumps have tripped and Feedwater flow has runback
D. Greater than Bypass Valve capacity ; Recirc pumps have tripped and Feedwater flow has runback

Answer 25

D: Correct: The LFMG sets are tripped and the feed flow runback occurs 25 seconds after the high pressure trip. 30 to 45 seconds after the event occurs, with the recirc pumps tripped, reactor power lowers but remains >25% which is greater than bypass valve capacity.

Question 26

26. The plant is operating at rated conditions when a complete loss of CCP Mini-Loop cooling water to the Instrument Air system occurs.

What is the impact on the Instrument Air Compressors?

The operating Compressor trips on … Backup Compressors …
A. low cooling water flow… are blocked from starting.
B. high outlet air temperature… are blocked from starting.
C. low cooling water flow… start and eventually trip.
D. high outlet air temperature… start and eventually trip.

Answer 26

D: Correct: The compressors trip on high outlet air temperature. There is no interlock to prevent the lagging and backup compressors from starting. The backup air compressors will start and eventually also trip on high outlet air temperature.

Question 27

27. Which of the following components are cooled by the Condensate System?

1 . Offgas Condensers A and B

2. Steam Jet Air Ejector Intercondensers
3. Steam Jet Air Ejector Precoolers
4. Mechanical Vacuum Pump Seal Cooler

A. Components 1 and 3
B. Components 2 and 4
C. Component 1 ONLY
D. Component 2 ONLY

Answer 27

A: Incorrect: Off gas condensers are cooled by TBCCW.
B: Incorrect: Mechanical Vacuum Pump Seal Cooler is cooled by Service Water.
C: Incorrect: Off gas condensers are cooled by TBCCW.
D: Correct: Steam Jet Air Ejector Intercondensers are cooled by the Condensate System.

Question 28

28. Which one of the following describes the potential effect of a loss of RDS Backfill Injection prior to a depressurizing event?

RPV water level indication will… PCIS water level isolations will occur. ..

A. indicate lower than actual… above the actual setpoint.
B. indicate lower than actual… below the actual setpoint.
C. indicate higher than actual… above the actual setpoint.
D. indicate higher than actual… below the actual setpoint.

Answer 28

D: Correct: RDS Backfill keeps the reference leg full of water and prevents the accumulation of dissolved gasses in solution that may come out of solution during sudden depressurization causing a loss of density in the reference leg that results in lowering the dp across the transmitter which correlates to indicated level being higher than actual. When indicated level lowers to 159.3, actual level is lower, therefore isolations that should have occurred at the actual level will occur below the actual water level.

Question 29

29. The plant is at rated conditions when annunciator 603303, ROD DRIVE CONTROL SYSTEM INOPERABLE, alarms.

Which one of the following describes the significance of this annunciator as it relates to the Reactor Manual Control System (RMCS) and the probable cause of this alarm?

Effect on Control Rod motion… Probable cause…

A. Rod motion is NOT impacted… Loss of RMCS permissive signal
B. Rod motion is NOT impacted… Loss of RDS Drive Water DP
C. Rod withdrawal is blocked… Loss of RMCS permissive signal
D. Rod withdrawal is blocked… Loss of RDS Drive Water DP

Answer 29

C: Correct: Per the associated ARP a rod block is generated to prevent rod motion when rod blocks may be disabled. The annunciator is caused by a:

1. Low Voltage
2. Computer Failure
3. Master Test Pushbutton depressed
4. Loss of permissive signal from (fcn. Of Rx Mode Switch):
   a. RPIS (INOP)
   b. NMS
   c. RSCS
   d. RWM
   e. Scram Discharge Volume
   f. Scram Discharge Volume Bypass
   g. Refuel Platform
   h. Service Platform
5. Clock Failure

Question 30

30. Twenty minutes after a fuel failure, reactor scram, and MSIV isolation, leakage past the MSIVs is indicated. Conditions are as follows:
    • N2-EOP-MSL has been entered
    • Main Condenser Vacuum is 0 inches
    • Main Steam Line Radiation and Turbine Building Release limits of N2-EOP-MSL have been exceeded

Given these conditions, which one of the following system lineups is available to limit the escape of radioactive steam?

Place the …
A. Main Steam Line Drains in service.
B. Mechanical Vacuum Pump in service.
C. Bypass Opening Jack momentarily to open.
D. Steam Jet Air Ejectors in service utilizing Aux Boiler Steam.

Answer 30

D: Correct: N2-EOP-MSL directs that the SJAEs be placed in service per N2-EOP-6. EOP-6, attachment 16 lines up the SJAEs utilizing Aux Steam.

Question 31

31. During a reactor plant startup Reactor Water Cleanup (VVCS) is being used to aid in RPV level control by rejecting reactor water to the main condenser.

Which one of the following is correct regarding the WCS flow through the Regenerative Heat Exchanger (RHX) and the Non Regenerative Heat Exchanger (NRHX) when 2WCS-FV135, REJECT FLOW CONTROL MANUAL CONTROL, is opened?

A WCS flow to the NRHX remains the same
B. WCS flow to the NRHX decreases
C. WCS return flow from the filter demineralizers to the RHX increases
D. WCS return flow from the filter demineralizers to the RHX decreases

Answer 31

D: Correct: WCS flow branches downstream of the filter demins. One branch goes to the reject line and the other branch is the return flow to the RHX. Therefore, as reject flow goes up, RHX flow goes down.

Question 32

32. The reactor is at 40% power with the following:

• One of the four Main Steam Line Flow Transmitter inputs to the Feedwater Level Control System failed downscale
• Feedwater Level Control has been placed in Single Element Control.

Following this failure the RWM will be operating …..

A. below the Low Power Set Point, alarms and rod blocks due to control rod mispositionings are enforced.
B. above the Low Power Alarm Point, alarms and rod blocks due to control rod mispositionings are not enforced.
C. in the Transition Zone, alarms are active but rod blocks due to control rod mispositionings are not enforced.
D. in the Transition Zone, alarms are not active and rod blocks due to control rod mispositionings are not enforced.

Answer 32

C: Correct. Transition zone is set when steam flow is between 25% and 40%. With power initially at 40%, a single steam flow indicator failing will result in a total steam flow signal lowering to 30%. At 30% steam flow alarms are active but rod blocks are not enforced.

Question 33

33. Which one of the following is the effect of a Loss of Coolant Accident (LOCA) coincident with a total Loss of Offsite Power (LOOP) on Fuel Pool temperatures and what actions are required to re-establish temperature control?

Fuel Pool Temperature ; To Re-establish Temperature Control

A. Lowers ; Manually close the FilterlDemin Bypass Valve, 2SFC-HIC113 only.
B. Rises ; Wait 70 seconds then restart the Fuel Pool Cooling Pump only.
C. Lowers ; Manually close the Filter/Demin Bypass Valve, 2SFC-HIC113 and place the Filter Demins back in service.
D. Rises ; Wait 70 seconds then restart the Fuel Pool Cooling Pump and line up Service Water Backup Cooling to the SFC heat exchanger.

Answer 33

D: Correct: On a Loss Of Coolant Accident (LOCA), the running Pump will continue to run but a Pump cannot be started or restarted during the first 70 seconds after receipt of the LOCA signal. If a Loss of Offsite Power (LOOP) and a LOCA occur simultaneously, the running Pumps will trip. A Pump may be manually restarted after 60 seconds using N2SOP-03. The Reactor Building Closed Loop Cooling Water supply to the SFC Heat Exchanger is also lost on a LOOP/LOCA and Service Water Backup Cooling to the SFC Heat Exchanger must be manually lined up by the operator.

Question 34

34. The plant is at rated conditions when annunciator 602105, RECIRC FCV A MOTION INHIBIT, alarms due to associated Hydraulic Power Unit failure.

Which one of the following is a concern while troubleshooting and repairing this condition and what action, per N2-0P-29, will address the concern?

Concern ; Action Taken

A. Recirc loop flow mismatch ; Close the Loop A Hydraulic Fluid Outside Isolation valves
B. Power level exceeds 3467 MWt ; Close the Loop A Hydraulic Fluid Outside Isolation valves
C. Recirc loop flow mismatch ; Manually trip the “A” Recirc pump after a scram if the reactor scrams
D. Inadequate Recirc Pump NPSH ; Manually trip the CIA” Recirc pump after a scram if the reactor scrams

Answer 34

A: Correct: Per precaution # 25 in OP-29, flow forces on the valve cause the FCV to drift close. If the Hydraulic Power Supply is to be shutdown at power at the discretion of the SM, close the appropriate loop outside hydraulic isolation valve by taking the LOOP A(B) HYDR FLUID OUTSIDE ISOL switch to the CLOSE position.

Question 35

35. The plant is at 100% power with the following conditions:

• A TIP trace is in progress
• The TIP is being operated in the MANUAL MODE
• The TIP is out of its shield chamber and running into the core but has NOT yet reached the CORE TOP LIMIT

With these initial conditions a feed water transient results in RPV level lowering to 100 inches before recovering. Ten minutes later RPV level is 180 inches and steady.

Which one of the following sets of indications would signify that the TIP system had responded as designed to the transient?
(Assume no operator actions.)

A. Amber Squib Monitor light: OFF Amber Shear Valve Monitor light: OFF Red Ball Valve Open light: ON White In-Core light: ON
B. Amber Shear Valve Monitor light: OFF Red Ball Valve Open light: ON White In-Core light: OFF White In-Shield light: ON
C. Amber Shear Valve Monitor light: OFF Green Ball Valve Closed light: ON White In-Core light: OFF White In-Shield light: ON
D. Amber Squib Monitor light: ON Amber Shear Valve Monitor light: ON White In-Core light: ON White In-Shield light: ON

Answer 35

A: Incorrect: These are indications that the TIP never received an isolation signal. RPV level dropping to 100 inches would actuate the automatic withdraw.
B: Incorrect: Although the TIP withdrew and is in the shield, the ball valve did not close. The automatic withdraw sequence will also close the ball valve.
C: Correct: These are indications that the TIP has withdrawn and is now in the shield. This is an expected response to an isolation signal.
D: Incorrect: These are indications that the shear fired. The shear does not automatically fire.

Question 36

36. Following a scram from full power all control rods indicate fully inserted with the exception of Control Rod 20-37. Additional information is as follows:

• Rod 20-37 indicates Blank on the Four Rod Display
• Blue Scram Light for rod 20-37 IS illuminated
• Full-In Light for Rod 20-37 is NOT illuminated
• OD-7 Control Rod Position Printout displays rod 20-37 as XX.
• No other rod position information is available.

Given the above information which one of the following is correct regarding the position of rod 20-37?

A. The rod’s position cannot be determined.
B. The rod fully inserted but is beyond full in.
C. The rod is stuck in between notch positions.
D. The rod is at 00 but the reed switch for position 00 is stuck open.

Answer 36

A: Correct: A double X (XX) indicates that the RPIS is receiving abnormal data. With these indications, the rod could be anywhere from fully withdrawn to beyond full in.

Question 37

37. In accordance with N2-0P-92, Neutron Monitoring, which one of the following describes:
    (1) How the Rod Block Monitor (RBM) operates as a function of APRM input AND
    (2) How the “nulling” or renormalization of a RBM channel is performed? (Assume a peripheral control rod is NOT currently selected)

A. (1) RBM “A” is provided an input signal from APRM 1 and with an alternate signal from APRM 3 ONLY.
(2) ONLY by first selecting any peripheral control rod and then reselecting any other control rod including the originally selected control rod.
B. (1) RBM “B” is provided an input signal from APRM 2 and with an alternate signal from APRM 3 ONLY.
(2) By first selectill9 another control rod and then reselecting the originally selected or any other control rod.
C. (1) RBM “A” is provided an input signal from APRM 1 and with an alternate signal from APRMs 3 or 4.
(2) ONLY by first selecting any peripheral control rod and then reselecting any other control rod including the originally selected control rod.
D. (1) RBM “B” is provided an input signal from APRM 2 and with an alternate signal from APRM 3 or 4.
(2) By first selecting another control rod and then reselecting the originally selected or any other control rod.

Answer 37

D: Correct -lAW N2-0P-92, Page 7 system description -The APRM provides Simulated Thermal Power to the RBM which uses that power level to determine which RBM power range setpoint is enabled. RBM A receives a signal from APRM 1 (APRM 3 and then 4 are alternates). RBM B receives a signal from APRM 2 (APRM 4 and then 3 are alternates) . The RBM is calibrated to a fixed (constant) reference each time a non-peripheral rod is selected. At the time of rod selection the RBM takes input from the surrounding LPRMs at various core heights and averages these readings. A “nulling” operation is performed which establishes the pre-rod motion value. This value is normalized to 100%. Renormalization is allowed by deselecting and reselecting the rod.

Question 38

38. Which one of the following describes when the Refueling Platform Main Hoist Raise Block is received?
    When the ___

(1) Fuel Hoist Interlock occurs
(2) Slack Cable Interlock occurs
(3) Safety Travel Interlock is received and TRAVEL OVERRIDE Switch is in NORM
(4) Normal Up Limit is reached while HOIST OVERRIDE push button is not depressed

A. (1), (2) AND (3)
B. (2) and (4)
C. (1),(3)AND(4)
D. (1) and (4)

Answer 38

B: Incorrect: Slack Cable interlock (2) is a Hoist Lower Block
C: Correct -lAW N2-0P-39 Section 5.4

Question 39

39. The plant was at full power when a LOCA occurred.
    Which one of the following failures/actions could result in exceeding the Primary Containment Pressure Limit (PCPL) and subsequent containment failure?

A. Initiating Drywell sprays with Suppression Pool level above 217 feet.
B. One set of Suppression Chamber to Drywell vacuum breakers failing closed.
C. Initiating Drywell sprays when outside the limits of the Drywell Spray Initiating Limit.
D. A break in a drywell to Suppression Chamber downcomer above the level in the Suppression Pool.

Answer 39

A: Incorrect -Initiating drywell spray at this level might challenge containment integrity but the failure mechanism would be due to exceeding the negative pressure rating due to the vacuum breakers being covered.

C: Incorrect -Initiating drywell sprays when outside the limits of the Drywell Spray Initiating Limit could result in containment failure but the failure mechanism would be due to exceeding the negative pressure rating.

D: Correct -The PCPL is a high pressure limit based on maintaining containment integrity. The lowest pressure is 45 psig which is also the design pressure of the containment. The UFSAR worst-case steam bypass analysis assumes a total bypass leakage path area of less than 0.1 ft2 (primarily drywell floor seams, downcomer and SRV piping penetrations, and vacuum breakers). The bypass path area of a single downcomer (assuming a double-ended rupture) would be almost 30 times larger. Although no specific data exists, NMPC Engineering is certain that the 45 psig primary containment design pressure would be exceeded.

Question 40

40. The plant has scrammed due to an MSIV closure, with the following:

• A large number of control rods have failed to fully insert
• Reactor power is 6% and steady
• 6 SRVs automatically lifted initially
• RPV pressure is now being maintained with SRVs

Which one of the following describes the status of the Reactor Recirculation Pumps (RCS) and Standby Liquid Control Pumps (SLS) two minutes following the initial event?

RCS Pumps ; SLS Pumps

A. Running ; NOT running
B. Running ; Running
C. NOT running ; Running
D. NOT running ; NOT running

Answer 40

C: Correct-SLC pumps start with the APRM’s >4% and RCS pumps trip on a RRCS reactor high press signal & power >4% after 25 sec

Question 41

41. The plant is operating at 100% power, with the following:

• Annunciator 849105 FIRE DETECTED PNL 127 SW STAIRl237 (for the Control Building EI 237) alarms
• Fire is confirmed
• HVC*ACU1A, CONTROL ROOM AC FAN tripped
• HVC*ACU2A, RELAY ROOM AC FAN tripped

Which one of the following identifies the actions required to be taken for Control Building Ventilation (HVC) and the reason?

A. Defeat cross divisional interlocks to prevent a Control Room evacuation due to smoke infiltration.
B. Actuate Appendix R disconnects to prevent tripping the Division II ACUs due to faulty electrical circuits.
C. Defeat cross divisional interlocks to ensure Control Room Envelope temperature can be maintained 90°F or less.
D. Actuate Appendix R disconnects to place HVC in a lineup that ensures the Control Room Envelope pressure does not become negative.

Answer 41

C: Correct -Per ARP 849105, N2-0P53A off normal section H.14.0 is required to be performed immediately, to defeat the HVC cross divisional interlocks. Implementation of this section directs the starting of Div II ACUs and using the cross divisional interlock key lock override switch to prevent loss of Div II components because of fire affecting Div I components. Note 2 states this is required to maintain Control Room Envelope temperature below 90°F.

Question 42

42. The plant is operating at rated conditions.

Which one of the following describes the effect of a loss of instrument air to the following Reactor Building isolation dampers and the response of the Reactor Building ventilation system?

• 2HVR*AOD1A/B
• 2HVR*AOD9A/B
• 2HVR*AOD10A/B

Dampers fail ; Reactor Building ventilation

A. Open ; No other automatic actions occur
B. Closed ; No other automatic actions occur
C. Open ; Reactor Building supply and exhaust fans trip
D. Closed ; Reactor Building supply and exhaust fans trip

Answer 42

D: Correct -A loss of Instrument Air to the Reactor Building will cause the Reactor Building isolation dampers 2HVR*AOD1A1B, AOD9A1B AND AOD10AlB to close, tripping all supply AND exhaust fans. The resulting low above/below Refuel Floor air flow will auto start the lead emergency recirculation unit cooler, 2HVR*UC413B.

Question 43

43. Emergency Procedure N2-EOP-C5, Failure To Scram, specifies under certain conditions to terminate and prevent injection and lower RPV level to below 100 inches before restoring injection.

Which one of the following is the EOP bases for performing this action?

This action …

A. increases the concentration of boron being injected into the core.
B. allows preheating the feedwater to lower the possibility of thermal hydraulic instabilities.
C. reduces the amount of natural circulation in the core following the tripping of the Recirc Pumps.
D. increases the amount of inlet subcooling to further reduce power via the temperature coefficient.

Answer 43

B: Correct: Per EOP bases, 100 inches is 24 inches below the feedwater spargers. When injection is commenced, feedwater will fall though a steam space and be preheated before entering the core region. This reduction in subcooling reduces the probability of thermal hydraulic oscillations since the step Q5 of C5 tripped the Recirc pumps.

Question 44

44. While executing the EOPs the following conditions exist:

• The Narrow Range RPV Level indicators are steady +152 inches.
• The hottest reactor building temperature is 170 degrees.
• All drywell temperature instruments are now pegged high at > 350 degrees
• RPV pressure is stable at 1000 psig.

Utilizing Table C from N2-EOP-RPV, which one of the following is correct regarding the Narrow Range RPV level indication?

Narrow Range Level indication …
A. CANNOT be used. Flashing of the reference legs may be occurring.
B. CANNOT be used. Actual RPV Level may be below the variable leg tap.
C. CAN be used for trending purposes. Indicated level is lower than actual.
D. CAN be used for trending purposes. Indicated level is higher than actual.

Answer 44

B: Correct: The minimum useable level for the Narrow Range is 150 inches when temperatures are less than 350 degrees and 155 inches when temperature is > 350 degrees in either the drywell or the reactor building. Per EOP Tech Bases document and EOP-RPV control, Caution A, the Narrow Range level indicators cannot be used when level is below the Minimum Indicating level.

Question 45

45. During a loss of shutdown cooling direction is given in N2-S0P-31, Loss of Shutdown Cooling, to raise level if no RHR pump or Recirculation Pump can be started.

Which one of the following identifies the reason for raising RPV water level to 227 to 243 inches?

To flood the …

A. dryer assembly to promote natural circulation.
B. steam separators to promote natural circulation.
C. dryer assembly to provide long term decay heat removal.
D. steam separators to provide long term decay heat removal.

Answer 45

B: Correct -the steam separators are flooded to connect flow from inside to outside the shroud to aid in natural circulation

Question 46

46. Refueling is in progress when an irradiated fuel bundle is dropped in the Spent Fuel Pool causing a refuel floor high radiation alarm and a reactor building isolation.

Which one of the following describes a reason for the reactor building ventilation system response?

This isolation wilL ..

A. reduce refuel floor radiation levels as quickly as possible.
B. ensure the greatest amount of air dilution prior to discharge.
C. prevent the spread of contamination to other parts of the Reactor Building.
D. ensure that air discharged from the refuel floor goes through a filtration system.

Answer 46

D: Correct -In the event of an accident condition, the Standby Gas Treatment System (GTS) prevents leakage of radioactive gases and particulates to the environment by maintaining a negative pressure in the Reactor Building by exhausting air (4000 cfm)

Question 47

47. Procedure N2-EOP-PC, Primary Containment Control, contains the following caution:

CAUTION: Operating ECCS or RCIC with suppression pool water level below EI. 195 ft may cause system damage.

Which one of the following describes the hazard of NOT complying with this caution?

A. RCIC exhaust line pressure oscillations could potentially cause system damage.
B. NPSH and vortex limits may be exceeded if RCIC is aligned to its alternate suction.
C. Reduced cooling water flow to the RCIC turbine bearings may cause system damage.
D. RCIC steam will discharge into the Suppression Chamber air space and overpressurize containment.

Answer 47

B: Correct -lAW the approved RCIC training material; EOPs caution that operating RCIC with suppression pool water level below EI. 195 ft may cause system damage. NPSH and vortex limits for these systems should be observed, if possible, but may be exceeded if necessary to maintain adequate core cooling. EOP-6 Attachment 29 is used when suppression pool level is below elevation 195 feet to determine the actual limits. If
necessary, EOP-6 Attachment 29 refers to use of EOP-6 Attachments 3 and 4 for throttling ECCS and RCIC flows to control within applicable limits.

Question 48

48. Following a grid disturbance the following conditions exist:

• Main Generator load is 800 megawatts electric (MWe)
• Main Generator reactive loading is 200 MVARs to the Generator
• Main Generator voltage regulation is in AUTO

System Power Control now requests that the Main Generator MVAR loading be changed to 100 MVARs lagging.

This will be accomplished by going to:

A. RAISE on the AC Voltage Regulator Control Switch
B. LOWER on the AC Voltage Regulator Control Switch
C. RAISE on the DC Voltage Regulator Control Switch
D. LOWER on the DC Voltage Regulator Control Switch

Answer 48

A: Correct: Voltage regulation is in AUTO which places the AC regulator in control. Initial MVAR loading is in the leading direction. To establish 100 MVARs in the lagging direction must go to raise which will increase excitation.

Question 49

49. Following a plant transient, the following conditions exist:

• Drywell Pressure is 3.8 psig
• Drywell Temperature is 200°F
• Suppression Chamber Pressure is 2.6 psig
• Suppression Pool Temperature is 92°F
• Suppression Pool Level is 200 feet

Which one of the following actions should be taken? Initiate …

A. Suppression Pool Cooling ONLY.
B. Suppression Chamber Sprays AND Drywell Sprays ONLY.
C. Suppression Chamber Sprays AND Suppression Pool Cooling ONLY.
D. Suppression Chamber Sprays AND Suppression Pool Cooling AND Drywell Sprays.

Answer 49

C: Correct -above 90 degrees SP temperature and below 10 psig SP Pressure -both sprays and cooling are put in service per EOP-PC step PCP-2 and SPT -3. Drywell sprays are not initiated based on the DWSIL curve.

Question 50

50. The plant is operating at 100% power with the following:

• Division I Diesel Generator 2EGS*EG1 is operating in parallel with offsite power for monthly surveillance testing
• 2EGS*EG1 is supplying 3960 -4400 KW to bus Then Offsite breaker R-50 trips open

Which one of the following describes the effect on 2EGS*EG1 and the Electrical Distribution circuit breakers?

DIV 1 Diesel 2EGS*EG1 …
A continues to run with its output breaker 101-1 closed. Offsite feeder breaker 101-13 is tripped open.
B. continues to run with its output breaker 101-1 and Offsite feeder breaker 101-13 tripped open.
C. trips on overspeed and its output breaker 101-1 is open. Offsite feeder breaker 101-13 is closed.
D. trips on overspeed and its output breaker 101-1 and Offsite feeder breaker 101-13 are tripped open.

Answer 50

A: Correct -R-50 trip results in a loss of offsite Line 5. Loss of offsite power causes offsite Feeder 101-13 to trip open. Diesel Generator 2EGS*EG1 continues to run with is output breaker 101-1 closed supplying the emergency switchgear.

Question 51

51. The plant is in Single Loop operation, following an unplanned Recirc Pump trip. The following conditions exist:

• Indicated drive flow in the operating Recirculation loop is 21,000 gpm
• Indicated total core flow on Control Room Panel P603 is 22 Mlbm/hr

Which one of the following identifies the current relationship between indicated total core flow on P603 flow recorder and actual total Core Flow and the reason for the difference?

Indicated total core flow is …

A. higher than actual due to idle loop flow being added to the total core flow summing network.
B. higher than actual due to idle jet pump flow being added to the total core flow summing network.
C. lower than actual due to idle loop flow being subtracted from the total core flow summing network.
D. lower than actual due to idle jet pump reverse flow being subtracted from the total core flow summing network.

Answer 51

C: Correct -NOTE 2 following Step H.6.0 of N2-0P-29, Reactor Recirculation System, states ‘When calculating total core flow in single loop operation and the operating loop drive flow is less than 22,000 gpm, Loop Flows should be added instead of subtracted”. In addition, Step 6.2 & 6.3 of N2-RESP-07, When one recirculation pump is not running, the forward/reverse flow logic network automatically subtracts the measured flow in the idle loop from the measured flow in the active loop and displays the true core flow on the P603 chart recorder (B22-R613) and computer point NSSFA101 (total core flow), and NSSFA01S (total core flow -smooth). If in single loop operation with less than 22,000 GPM flow, flow in the shutdown loop will be positive instead of negative as assumed by the summing network (Step 6.2). Therefore in this situation, a calculated total core flow must be substituted into computer point NSSFA101 and NSSFA01S prior to demanding a core monitoring case.

Question 52

52. A small LOCA has occurred with the following conditions:

• Drywell pressure is 2.1 psig
• Drywell temperature is 170°F
• EDGs have started and are running unloaded
• EOP-RPV and EOP-PC have been entered

Which one of the following describes the status of CCP flow to the drywell unit coolers?

CCP is …

A. NOT isolated and will only isolate if off site power is also lost.
B. NOT isolated and will only isolate if RPV level lowers below 108.8 inches.
C. isolated, but can be procedurally restored using LOCA OVERRIDE switches.
D. isolated and cannot be procedurally restored until drywell pressure lowers below 1.68 psig.

Answer 52

C: Correct –A CCP isolation to the drywell unit coolers occurs on a group 8 isolation signal (1.68 psig Drywell pressure or 108.8 RPV level). EOP-6, ATTACHMENT 24 allows restoration of drywell cooling as long as drywell temperature is below 250°F.

This is only allowed due to the fact that Drywell Temp is above 150°F, If not in the Drywell Temp leg of EOP-PC, then D would be correct per SOP-60:

5.2 If system isolation has occurred due to a valid signal, the problem must be determined and corrected prior to resetting or bypassing the isolation signal, unless otherwise directed by the EOPs.

Question 53

53. The plant was operating at 22% power when grid instabilities caused a Main Generator Lockout Relay Trip.

Which one of the following describes the turbine valve status?

A. The Turbine Stop Valves, Control Valves, and Combined Intermediate Valves are closed.
B. The Turbine Stop Valves and Control Valves are closed. The Combined Intermediate Valves are open.
C. The Turbine Stop Valves are closed. The Control Valves and Combined Intermediate Valves are open.
D. The Turbine Stop Valves and Combined Intermediate Valves are closed. The Control Valves are open.

Answer 53

A: Correct -All valves close on a turbine trip

Question 54

54. The plant is operating at rated conditions, when a loss of 125VDC Bus 2BYS-SWG001 C occurs.

Which one of the following identifies the Uninterruptible Power Supplies (UPS’s) that will lose DC power?

A. 2A and 2B
B. 1D and 3B
C. 1A, 1C and 1G
D. 1B, 1G and 3A

Answer 54

D: Correct -UPS 1 B, 1 G AND 3A receive backup DC power from 2BYS-SWG001 C

Question 55

55. Which one of the following is the bases for Technical Specifications Iodine 131 limit?

Limits the maximum amount of …

A. TEDE dose received by an individual at the exclusion area boundary.
B. CEDE dose received by an individual at the exclusion area boundary.
C. TEDE dose received by an individual at the protected area boundary.
D. CEDE dose received by an individual at the protected area boundary.

Answer 55

Correct -Per TS 3.4.8 bases -Limits on the maximum allowable level of radioactivity in the reactor coolant are established to ensure, in the event of a release of any radioactive material to the environment during a DBA, radiation doses are maintained within 10% of the limits of 10 CFR 50.67 Per 10 CFR 50.67 -An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 0.25 Sv (25 remf total effective dose equivalent (TEDE). 10% of this limit would therefore be 2.5 Rem.

Question 56

56. The plant was operating at rated conditions with APRM 1 inoperable and bypassed. when the “A” Reactor Recirc pump trips. Following the trip, the following conditions exist:
    • Core flow 40%
    • Reactor power 60%
    • Annunciator 603218, OPRM TRIP ENABLED, alarmed
    • OPRM Trip Enabled status lights are illuminated for OPRM 2 and 3
    • The remaining OPRM status lights are extinguished.
    • CRS has determined that OPRM #4 is INOP.

When the APRM Numac Displays on P608 are checked the message “OPRM TRIP ENABLE” is displayed on OPRM 2 and 3 only. NO other OPRM messages are displayed.

Which one of the following is required by station procedures?

A. Insert the first 4 CRAM rods
B. Place the reactor mode switch to Shutdown
C. Raise core flow with the “B” Recirculation Pump to exit the Exit Region
D. Shift APRM recorders to fast speed AND monitor for core oscillations

Answer 56

B: Correct: Per ARP 603218, OPRMs should have enabled. Only 2 of the three available did so. Per SOP-29, if operating in the OPRM Dependent Stability Region and less than 3 are available, the reactor is to be scrammed.

Question 57

57. The following plant conditions exist:

• A fire required entry into N2-S0P-78, Control Room Evacuation
• All Offsite power is lost
• The Reactor is shutdown
• Reactor pressure is 200 psig
• 2EGS*EG1 and 2EGS*EG3 started and powered the emergency buses
• Plant is being controlled from the Remote Shutdown Panel

Which one of the following strategies is to be used, per N2-0P-78, Remote Shutdown System and N2-S0P-78, Control Room Evacuation?

A. Open four (4) ADS Valves then place Alternate Shutdown Cooling in service.
B. Open four (4) ADS Valves then place Normal Shutdown Cooling in service.
C. Continue operating RCIC then place Alternate Shutdown Cooling in service.
D. Continue operating RCIC then place Normal Shutdown Cooling in service.

Answer 57

C: Correct: Per N2-0P-78, Remote Shutdown System, Alternate Shutdown Cooling is used if offsite power is lost, because Normal Shutdown Cooling cannot be placed in service. RCIC is still available to reduce pressure and provide makeup to the RPV.
D:

Question 58

58. The following conditions exist:

• A small break LOCA is in progress.
• Appropriate EOPs have been entered.
• No injection sources are available.
• RPV level is approaching the TAF.

Which one of the following describes why RPV Blowdown is delayed until water level has lowered to the Minimum Zero-Injection RPV Water Level per N2-EOP-C3?

A. RPV Blowdown before this level reduces the time the core remains adequately cooled during these conditions.
B. Steam Cooling will NOT maintain fuel cladding temperature below 1500°F if blowdown is performed with water level above TAF.
C. RPV Blowdown before this level produces insufficient steam mass removal rate for adequate core cooling.
D. Steam cooling will NOT maintain fuel cladding temperature below 1800°F if blowdown is performed with water level above TAF.

Answer 58

A: Correct -Per EOP-C3, (Page 10-14 of EOP Bases Document), it states “opening the SRVs before RPV water level reaches the MZIRWL would reduce the time over which the core remains adequately cooled with no injection”.

Question 59

59. Given the following:

• The reactor is critical with a heatup in progress.
• Reactor pressure is 550 psig.
• “A” CRD pump is out of service for maintenance.
• “B” CRD pump tripped and cannot be restarted.
• 5 different CRD accumulator trouble alarms were received due to low accumulator pressure. The associated control rods are fully withdrawn.
o 18-27 @ 920 psig
o 26-51 @ 1020 psig
o 38-43 @ 1000 psig
o 46-27 @ 960 psig
o 54-23 @ 900 psig

Which one of the following describes the current manual control rod insertion capability using the Reactor Manual Control System (RMCS) and the status of the control rod scram function per Tech Specs and N2-S0P-30?

Manual insertion capability ; Scram function

A. NOT available ; Not Degraded
B. Available ; Degraded
C. NOT Available ; Degraded
D. Available ; Not Degraded

Answer 59

C: Correct -Per TS Bases 3.1.4 Background -“If the reactor pressure is low, such as during startup, the accumulator will fully insert the control rod within the required time without assistance from reactor pressure.” At least 2 withdrawn control rods have low accumulator pressures rendering them inoperable «940 psig). Rods mayor may not fully insert or they may not insert fast enough under the given conditions, so the scram function may be degraded.
With both CRD pumps off, there is no drive pressure, so manual insertion using RMCS is not possible. Furthermore, CRD charging water pressure is also lost. In addition, reactor pressure is less than 900 psig, so reactor pressure alone may not complete the scram.

Question 60

60. The plant was operating at 100% power. The following conditions exist:

At Time = 0 (T=O)
• A failure to scram occurred
• Reactor power is at 50%.
• Reactor Pressure has remained at 1000 psig during the event
• Main Turbine is still on line
• Reactor Water level is 106 inches and slowly lowering

Which of the following identifies the status of the Recirc Pumps at T= +15 seconds?

A. NOT running
B. Running at Low Speed but will trip in 10 seconds
C. Running at High Speed but will shift to Low Speed in 10 seconds
D. Running at Low Speed and will remain running until tripped manually

Answer 60

A: Correct -the recirc pumps trip at RPV level 2 (108.8”)

Question 61

61. The plant was operating at rated conditions when a spurious Group 8 isolation occurs.

Which one of the following describes the response, if any, of the Drywell Unit Coolers and the effect, if any, on Drywell pressure?

Drywell Unit Cooler Fans ; Drywell pressure
A. Trip ; Rises
B. Trip ; Remains the same
C. Operating ; Rises
D. Operating ; Remains the same

Answer 61

A: Correct -The Level 2 signal will cause a group 8 isolation which isolates CCP flow to the drywell and trips the drywell fans. Drywell pressure will rise.

Question 62

62. A LOCA has occurred with the following plant conditions:

• RPV Level is 20 inches and rising slowly
• Reactor Pressure is 40 psig and steady
• LPCS is injecting at design flow rate
• No other sources of injection are available
• The Reactor Building sump reaches the High-High setpoint
• The LPCS Pump is the source of the leak

Which one of the following is the required action regarding the LPCS Pump?

A. Continue to inject with the pump
B. Isolate the pump when annunciator 601411, LPCS PUMP ROOM FLOODING, alarms
C. Isolate the pump when LPCS area water level exceeds the Max Normal Operating Value
D. Isolate the pump when two area water levels exceed the Max Normal Operating Values

Answer 62

A: Correct -N2-EOP-SC specifies that a system is not to be isolated if needed for EOP actions. LPCS is maintaining adequate core cooling.

Question 63

63. The plant was operating at 99% power when a main turbine trip occurred but the reactor did not scram.

Direction is given in the EOPs that if SRVs are cycling, to manually open SRVs until pressure drops to 970 psig.

Which of the following describes the basis for this direction?

A. Maintains adequate core cooling in the RPV.
B. Maximizes the amount of steam condensed in the suppression pool.
C. Maximizes the amount of energy directed to the main condenser.
D. Maintains stable reactor parameters to prevent the loss of high pressure feed systems.

Answer 63

C: Correct -Pressure is stabilized at a value below 1052 psig, the high RPV pressure scram setpoint, to avoid SRV actuation and to permit the scram logic to be reset (if no other scram signal exists). SRVs are opened and pressure is reduced to 970 psig to stop them from cycling because this is the pressure at which steam flow through the TBVs is at 100%. This minimizes reactor power swings and maximizes the amount of energy going to the main condenser. Lowering pressure lower than this risks the TBVs going closed. This could then cause the SRVs to reopen and put more energy in the suppression pool.

Question 64

64. The plant is operating at 50% power with the following:

• Annunciator 851358, TURBINE CNSR A/B/C VACUUM LOW alarmed
• Main Condenser Vacuum is 24.4 inches Hg vac. and slowly degrading

Which one of the following actions is FIRST required by N2-S0P-09, Loss of Vacuum?

A. Scram the reactor lAW N2-S0P-101C, Reactor Scram.
B. Immediately trip the Main Turbine per N2-S0P-21, Turbine Trip.
C. Reduce reactor power lAW N2-S0P-101D, Rapid Power Reduction
D. Verify the Main Turbine has tripped and enter N2-S0P-101 C, Reactor Scram.

Answer 64

A: Incorrect -A scram would be required if offgas inlet pressure could not be restored and maintained <19 psi a and the reactor was still critical.
B: Incorrect -A turbine trip would be required if vacuum was <24.6” Hg and the turbine was loaded to <30% (363 MWe).
C: Correct -Per N2-S0P-9, First reduce power to attempt to stabilize vacuum.
D: Incorrect -The Main Turbine will not auto trip until vacuum lowers to 22.1 inches.

Question 65

65. Following a loss of coolant accident the following conditions exist:

• Reactor pressure is 800 psig and lowering
• Suppression Pool level is 212 feet and rising

Which one of the following is the limiting component for these conditions per N2-EOP-PC Bases?

A. SRV Tailpipes
B. Vacuum Breakers
C. Drywell Spray Ring
D. Suppression Chamber Spray Sparger

Answer 65

A: Correct -Per EOP-PC Bases Rev 5 page 5-21, at that SP level you have reached the SRV tailpipe level limit as given in EOP-PC Table N. This level is the highest suppression pool water level at which opening an SRV will not result in exceeding the code allowable stresses in the tail pipe, tail pipe supports, quencher, or quencher supports.
B: Incorrect -The bottom of the vacuum breaker openings is at 227.25 feet and is not the concern per EOP Bases at this SP level (Per EOP bases page 5-28).
C: Incorrect -The drywell spray ring is >240 feet and is not the concern per EOP Bases at this SP level.
D: Incorrect -the SP spray sparger is at elev. 231 ft and is not the concern per EOP Bases at this SP level (Per EOP bases page 5-27)

Question 66

66. During a Refuel Outage, with the Reactor Mode Switch in Refuel, the following conditions exist:

• Annunciator 603442, Control Rod Out Block, alarms
• One Control Rod is selected
• All Control Rods are inserted to position 00

Which one of the following is a possible cause of the annunciator?

The Refuel Bridge is …
A. NOT over the core with the Grapple full up and loaded.
B. NOT over the core with the Grapple full down and loaded.
C. over the core with the Grapple full up and the Trolley Hoist is loaded.
D. over the core with the Grapple full down and the Trolley Hoist is NOT loaded.

Answer 66

C: Correct: With the Refuel Bridge over the core with the Grapple up, but an auxiliary hoist loaded a Control Rod Block will be initiated.

Question 67

67. The plant is in Mode 5 with core alterations in progress. Which one of the following conditions REQUIRES that Secondary Containment be OPERABLE?

A. Any core alteration regardless of the time since plant shutdown.
B. Any core alteration if performed within 1 day of initial plant shutdown.
C. Movement of irradiated fuel regardless of the time since plant shutdown.
D. Movement of recently irradiated fuel if performed within 1 day of initial plant shutdown.

Answer 67

D: Correct: Tech Spec 3.6.4.1 requires secondary containment to be operable during movement of any recently irradiated fuel assembly. Tech Spec bases defines recently irradiated as the bundle having been part of a critical core within the past 24 hours.

Question 68

68. A plant startup and power ascension is in progress. The following conditions exist:

• Reactor Power is 26%
• Generator output is 250 MWe
• Load Limit setpoint is 250 MWe because of a generator winding issue

Then, annunciator 851150, TURBINE BYPASS VALVE OPEN alarms and #1 Bypass Valve opens.

Which one of the following actions is required to close the #1 Bypass Valve?

A. Insert control rods.
B. Raise Pressure Set.
C. Lower Pressure Set.
D. Lower Bypass Opening Jack setpoint.

Answer 68

A. Correct -With the Load Limit set at 250 MWe the turbine cannot accept any more load. Since the EHC systems wants to open the control valves, but can’t because of the load limit the bypass valve is opening to control pressure. The only way to close the bypass valves is to lower reactor power and hence pressure by inserting the control rods.

Question 69

69. Which one of the following is an example of a “Group Tagging” situation and how would protected personnel sign on to this type of clearance per CNG-OP-1. 01-1007, Clearance and Safety Tagging?

Covered under Group Tagging ; Sign onto

A: Entry into the suppression chamber to conduct inspections ; Specific Radiation Work Permit
B: Reactor disassembly on the Refueling Floor ; Specific Radiation Work Permit
C: Entry into the suppression chamber to conduct inspections ; Operating Permit for Personnel Protection
D: Reactor disassembly on the Refueling Floor ; Operating Permit for Personnel Protection

Answer 69

A: Correct: An individual that requires protection of a Tagout for general area entries of the drywell and Suppression ChamberlTorus. The use of the Specific Radiation Work Permit for these areas provides this protection.

B: Incorrect: A group tag out is only used for entry into the drywell and Suppression ChamberlT orus.

Question 70

70. Core Alterations are in progress with the following:

• An irradiated fuel bundle is being moved from the reactor cavity to Spent Fuel Pool
• The bundle becomes ungrappled and falls into the reactor vessel downcomer area. (Between the vessel wall and the shroud)
• The bundle integrity is maintained

Which one of the following workers is at greatest risk of radiation overexposure?

A. I&C Tech at SLS Tank
B. Refuel SRO on the Bridge
C. Mechanic working on the installed SRVs
D. RP Tech at Refuel Floor Access Point

Answer 70

C: Correct: Worker closest to the bundle with the least amount of shielding will be at greatest risk. SRVs are in the Drywell at the approximate elevation of the downcomer.

Question 71

71. The plant is in a refuel outage, with the following:

• An accident on the refueling floor caused a serious injury to a worker
• The worker is on the Refueling Platform
• Radiation levels in the area of the injured operator are 2500 mRem/hr
• Emergency exposure limit for life saving operations has been authorized

In accordance with EPIP-EPP-15, Emergency Health Physics Procedure, which one of the following is the maximum stay time for the individual providing life saving assistance?

A. 2 hours
B. 4 hours
C. 6 hours
D. 10 hours

Answer 71

D: Correct. Based on correct limit of 25 Rem limit (life saving). 25R12.5R per hr =10 hours

Question 72

72. The plant is operating at 100% power, with the following:

• RCIC Full Flow Test surveillance is in progress
• Suppression Pool temperature rises to 91F due to RCIC operation

Which one of the following identifies the proper EOP implementation requirements? N2-EOP-PC …

A. entry is NOT required because no emergency condition exists.
B. entry is NOT required because no entry conditions have been met.
C. must be entered but can be exited since no emergency condition exists.
D. must be entered and can only be exited when entry conditions are not met.

Answer 72

C: Correct: “An EOP may be exited if the exit conditions of the procedure are met or any time it is determined that an emergency no longer exists.” The determination is not dependent upon the status of entry conditions …”. EOP-PC entry conditions exists, so EOP entry IS required. The EOP can be exited when it is determined that an emergency does not exist.

*** This question may be dated, we are allowed to go up to 105F during surveillance testing***

Question 73

73. With the plant at 100% power a fire occurs in the Control Room and the CRS announces that they have entered SOP-78, Control Room Evacuation.

Which one of the following are the required OATC operator actions before leaving the Control Room?

Shutdown the Reactor, …
A. close the MSIVs, and trip the feedwater pumps.
B. trip one feedwater pump, and verify 2FWS-LV1 Os in Auto.
C. verify emergency diesel generators running unloaded, and main turbine tripped.
D. main turbine tripped, house loads transferred and turbine bypass valves controlling pressure.

Answer 73

A: Correct: lAW SOP-7B, the OATC operator must perform the following:

1. Mode switch to SHUTDOWN.
2. Confirm ALL rods full in.
3. Close the MSIVs.
4. Trip feedwater pumps.
5. Verify 2FWS-LV1 Os in MANUAL AND closed.
6. Verify Main Turbine tripped.
7. Confirm House Loads transferred OR Diesels energizing buses.

Question 74

74. The plant is operating at 100% power. What is the maximum Technical Specifications amount of Reactor Coolant System Leakage (gpm) allowed for continued plant operation?

Unidentified ; Identified

A. 2 ; 20
B. 2 ; 25
C. 5 ; 20
D. 5 ; 25

Answer 74

D: Correct: 5 gpm is the maximum unidentified leakage in 24 hours and 25 is the maximum identified within 24 hours.

Question 75

75. A Site Area Emergency has been declared at NMP2.

In accordance with the Emergency Plan, which Emergency Response Facility provides the location for the overall response to an emergency, once all the required Emergency Response Facilities are activated?

A. Control Room
B. Technical Support Center
C. Operations Support Center
D. Emergency Operations Facility

Answer 75

A: Incorrect: -The Control room manages the plant the EOF provides the overall response.
B: Incorrect: -The TSC provides technical support to the EOF
C: Incorrect: -The OSC provides support to the EOF.
D: Correct: The TSC, OSC and EOF are activated during an Alert, Site Area Emergency or General Emergency, or when directed by the SM/ED or ED/RM. The Emergency Director/Recovery Manager (ED/RM) is responsible for managing all aspects of the NMP response to an emergency at NMPNS. The ED/RM operates from the EOF.